Appendix CRF - Part 3 - Northamptonshire County Council

AUGEAN PLC EAST NORTHANTS RESOURCE 

MANAGEMENT FACILITY 

AU/KCE/MM/1561/01PEI 

April 2011 

pl14591 

APPENDIX D 

APPLICATION FOR DISPOSAL OF LLW INCLUDING HV-LLW UNDER THE 

RADIOACTIVE SUBSTANCES ACT 1993 FOR THE EAST NORTHANTS RESOURCE 

MANAGEMENT FACILITY SUPPORTING INFORMATION. JULY 2009 

WS010001/ENRMF/CONSAPPCRF 319

Application for disposal 

of LLW including HV-VLLW 

Under the Radioactive Substances Act 

1993, for the East Northants Resource 

Management Facility 

Supporting Information 

July 2009 


Preface 

This authorisation application was prepared by Augean plc with support from: 

Decommissioning and waste specialists from the United Kingdom Atomic Energy 

Authority (UKAEA) Harwell site (recently renamed RSRL). 

Technical assessments were provided by Galson Sciences Ltd using a 

framework developed by the Scotland & Northern Ireland Forum for 

Environmental Research. 

Supplementary technical studies were provided by UKAEA Ltd. Technical 

Services Group. 

Background materials concerning radioactivity for those unfamiliar with the 

subject were obtained from the International Atomic Energy Authority (IAEA). 

Occupational radiation protection advice was provided by the Health Protection 

Agency (HPA). 

Capability statements for the professional team are given in Annex J. 

Application for disposal of LLW including HV-VLLW under RSA 1993, 

for the East Northants Resource Management Facility: 


July 2009 

2 


Contents 

Summary 

Supporting Information for the Application 

1.0 Introduction 

2.0 Authorisation 

2.1 Background 

2.2 What is LLW? 

2.3 Strategic Need 

3.0 Policy and Regulatory Background 

3.1 Radioactive Substances Regulation 

3.2 The Radioactive Substances Act 

3.3 Risk 

3.4 UK Government Policy 

3.5 Basic Safety Standards 

3.6 Environmental Permitting Regulations 

3.7 Conservation Regulations 

3.8 Ionising Radiations Regulations 

3.9 Nuclear Industry LLW Strategy 

3.10 Other 

4.0 Site Background Information 

4.1 Site Description and Local Environment 

4.2 Business Plans and Site Development Plans 

4.3 Existing Permits 

5.0 Radioactive Waste Disposal Proposal 

5.1 Principles and Dose Criteria 

5.2 Sources of Waste 

5.3 Road Transport 

5.4 Receipt and Assay 

5.5 Accumulation and Quarantine 

5.6 Disposal, Waste Emplacement, Compaction, Cover and Handling 

5.7 Worker Radiation Protection 

5.8 Environmental Radioactivity Monitoring 

6.0 Waste Disposal History 

7.0 Proposals for Liquid and Gaseous Discharges 

8.0 Radioactive Waste Disposal Consequence Assessment and Radiological Capacity 

8.1 Pre-Closure – expected to occur 

Direct Radiation Exposure from Waste Handling and Emplacement 


Exposure from Gas Generation from the Landfill 

8.3 Pre-Closure – not expected to occur 

Dropped Load of Waste 





July 2009 

3 


Wound Exposure 


Exposure from Fire 

8.6 Pre Closure and Aftercare Period – expected to occur 

Exposure from Leachate Processing Offsite – Sewage Works 

8.7 Pre Closure and Aftercare Period – not certain to occur 

Exposure from Leachate - Spillage 


Exposure from Aerosols 

8.9 Post-Closure – expected to occur 

Exposure by Using Groundwater at Nearest Abstraction Point 




Exposure to Wildlife from all sources 


External dose from emplaced wastes 

8.13 Post –Closure not expected to occur 

Exposure by Using Groundwater from a Borehole Constructed at the Boundary of 

the Landfill 


Exposure by Intrusion into the Emplaced Waste Post Closure of the Landfill 

8.15 Results of the Assessment 

8.16 Landfill Radiological Capacity 

9.0 Radioactive Waste Disposal Proposed Authorisation Conditions and Waste Acceptance 

Criteria 

9.1 Potential Conditions Arising - Standard RSA Authorisation Template 

9.2 Potential Conditions Arising - Existing Landfill Permit & the Landfill Regulations 

9.3 Conditions Arising from the Site Specific Risk Assessment and Industry Practice 

10.0 BPEO Assessment for LLW Disposal of Waste from Nuclear Sites 

10.1 BPEO 

11.0 BPM and ALARA Assessment for the Proposed Radioactive Waste Disposal 

11.1 ALARA 

11.2 BPM 

12.0 Landfill Engineering and BAT Features of the Existing Landfill 

13.0 Waste Hierarchy and Waste Minimisation at Source 

14.0 Summary of the Existing Environmental Statement for the Site and Impacts of the 

Proposal 

15.0 Outline of Management and Operating Arrangements 

16.0 Stakeholder Consultation 

17.0 The Applications Forms 

17.1 Waste Disposal 

18.0 Conclusion 




July 2009 

4 


References 

Figures 

Figure 1 Site Location 

Figure 2 Site Layout 

Glossary 

Annexes 

A Radiation, People and the Environment (IAEA, 2004) 

B Suitability Assessment – Galson Sciences 

C ENRMF, IRRs 1999, Radiation Risk Assessment for Low Level Waste Disposal, HPA 

D Dose Rate calculations in support of Low Level waste disposal authorisation, 

TSG(09)0487 

E SNIFFER Methodology Information 

F Copy of Application Form 

G Example Capacity Calculation Layout 

H Calculation of dose rate at landfill, TSG(09)0488 

I Baseline Groundwater and Leachate Sample Results 

J Capability Statements 




July 2009 

5 


Summary 

Introduction 

S1 This document provides supporting information for an application for 

authorisation under the Radioactive Substances Act 1993, for disposal of solid 

Low Level Radioactive waste (LLW) of up to 200 Bq/g, including High Volume 

Very Low Level Waste (HV-VLLW), at the East Northants Resource Management 

Facility (ENRMF), operated by Augean plc. 

S2 The waste has a very low radioactivity content which this application 

demonstrates would present a very low risk if disposed. 

S3 This document provides information to the Environment Agency, as regulator, in 

order that they can consider the application for authorisation. This document is 

also a public document. 

S4 This document contains specialist terms which are required to communicate the 

information to the regulator and it is recognised that this may make the document 

less accessible to a wider audience. The main document contains a 

comprehensive Glossary of technical terms used. A more detailed booklet on 

radiation, people and the environment (published by the International Atomic 

Energy Agency) is referenced in Annex A to the main document. These 

information sources may be of further use to readers new to the subject matter. 

S5 This application for an authorisation contains proposed arrangements and 

conditions which are subject to regulatory approval and changes. If the 

application is granted, the conditions that apply will be those established by the 

authorisation and by detailed supporting operational documentation prepared to 

address the authorisation. 




July 2009 

6 


Background 

S6 The use of landfill is an established approach to the disposal of LLW with low 

specific activity and is supported by Government policy (ref 3). Disposal of LLW 

to landfill is authorised under the Radioactive Substances Act 1993 (ref 4) using 

permits/authorisations issued by the Environment Agency in England. The 

permitting arrangements are currently under review to incorporate the approach 

within the Environmental Permitting Regulations 2010 (ref 19). 

S7 Disposal routes in the UK for LLW are limited and often the only option available 

is disposal to the LLW repository near to the village of Drigg in Cumbria. The 

LLW repository does not have capacity for the volumes of the full range of LLW 

(up to 4000 Bq/g alpha and 12,000 Bq/g beta/gamma) that will be generated from 

broad decommissioning of the nuclear industry. The disposal of LLW at the 

lower end of the range of specific activity is not thought to be a sustainable use of 

the repository, which has been designed and engineered to a standard suitable 

for materials with a radioactive content at the higher end of the range for LLW. 

The strategic need for alternative fit for purpose disposal routes is established 

and detailed within the UK nuclear industry LLW strategy (consultation) (ref 20) 

and for the non-nuclear industry in UK government policy (ref 3). 

S8 The proposed LLW contains very small amounts of radioactivity; less than or 

equal to 200 Bq/g. The waste can be handled safely by humans in direct contact 

with the material in a manner similar to other low hazard wastes. The material is 

a radioactive waste in accordance with legal definitions but, in the case of this 

application, it contains radioactivity of less than the bottom 5% of the range of low 

level radioactive wastes. The waste does not need special security measures. 

S9 The LLW that will be disposed of arises from the decommissioning and clean-up 

of nuclear industry sites and from non-nuclear industry sources, such as 

hospitals. 

S10 Typically the waste is rubble, soils, crushed concrete, bricks and metals that arise 

from demolition of buildings that were previously used for nuclear research or 

power generation. A large programme of work to decommission the nuclear 

legacies created since the 1940’s is currently underway in the UK that will 

generate significant volumes of LLW. The UK Nuclear Industry LLW strategy and 

supporting inventories (ref 20) provide detailed information on the potential types 

and nature of the wastes. 

S11 During decommissioning, the highest radioactive hazards are removed prior to 

demolition of structures. What remains after decommissioning is a mixture of 

construction materials/soils either that can be proven clean or which sometimes 

contain trace levels of radioactivity. Efforts are made to separate out 

radioactivity, to sort wastes, to recycle materials and to reuse materials. The 

wastes that remain with trace levels of radioactivity after this process are typical 

of the wastes proposed for disposal at ENRMF. 




July 2009 

7 


Radioactivity & Risk 

S12 This summary contains a short Annex which briefly explains radioactivity. 

Humans are exposed to ionising radiation every day. This exposure comes from 

background radiation. 

S13 Humans can have additional exposure from other sources; for example having an 

x-ray or flying in a commercial aeroplane results in additional exposure to ionising 

radiation. It is necessary to limit exposure because there is a possibility of 

adverse health effects. 

S14 In the UK there is a consensus that for exposure to radioactivity the risk of a 

fatality of 1 in 1,000,000 (one in a million) per year can be regarded by society as 

a level of risk to a member of the public beyond which further reduction may not 

be justified. 

S15 A risk of one in a million is very low. For comparison the average annual risk of 

death for the following is approximately: 

Smoking 10 cigarettes per day 1 in 200 

Natural causes for someone aged 40 1 in 700 

Accidents in the home 1 in 10,000 

Lighting strike 1 in 10,000,000 (1 in 10 million) 

S16 The waste disposal process proposed has been designed such that the risk to 

the public in the long term is broadly less than one in a million per year. This is 

consistent with the risk guidance level set by regulatory guidance (ref 18) and is 

better than the risk constraint established by the HPA guidance (ref 14) of 1 in 

100,000 per year. This risk guidance level is achieved by limiting the radioactive 

content and amount of the waste that the landfill can receive. Hence long term 

public safety is an inherent feature of the proposal and does not depend on future 

human actions. 

S17 For unlikely intrusion events, including for example, if the landfill is excavated by 

future society and the land reused for residential properties, a dose guidance 

level has been used which is the lower end of the range indicated by regulatory 

guidance (refs 18, 14). 

S18 Over time radioactivity decreases because radioactive decay is a process that 

eventually leads to the original wastes becoming non-radioactive. For some 

types (radionuclides), this takes so long that it can be ignored, but for others the 

effect is relatively quick and after only a few years or decades the waste 

becomes essentially non-radioactive. 




July 2009 

8 


The East Northants Resource Management Facility 

S19 The landfill site lies approximately 2.5km north of the village of King’s Cliffe in the 

East Northamptonshire District of the County of Northamptonshire (Figure 1). 

The closest village to the site is Duddington, approximately 2.2km to the North 

West. The setting is generally rural with a majority of the land surrounding the 

landfill site comprising open farmland or woodland. 

S20 Landfilling operations at East Northants Resource Management Facility 

commenced in 2002. The site has been previously known as the Kingscliffe 

landfill site and as the Slipe Clay Pit. 

S21 The facility is the subject of a Permit and operates as a hazardous waste landfill 

with a number of ancillary waste activities and treatment processes on the site. 

The landfill site is engineered to the highest standards consistent with hazardous 

landfills. It is lined with a composite barrier of high density polyethylene and 

1.5m thickness of clay. The disposal rate of the engineered landfill cells with 

hazardous and inert (the inert material is used for cover and construction of 

access tracks) waste is permitted at a maximum of 249,999 tonnes/year (Figure 

2). 

S22 It is currently envisaged that landfill operations will continue until approximately 

2013, dependant on the actual importation rate. The site will be progressively 

restored and once complete will undergo a defined scheme of capping and final 

restoration. The afteruse of the site will be principally grassland and wildflower 

meadows for ecological and agricultural purposes. 

S23 The proposal for LLW disposal at the site will not change the annual tonnage, the 

total capacity of the site or the physical features that contributed to the original 

landfill permitting decision. 




July 2009 

9 


The LLW Disposal Proposal 

S24 The application document provides an outline of the proposed arrangements for 

the LLW disposal process. After granting of the authorisation this outline would 

be developed into detailed operating arrangements in accordance with the 

authorisation conditions. 

S25 The process has the following key features: 

The wastes will be transported to the site in accordance with relevant 

transport regulations that apply to radioactive wastes. The regulations are 

established to control the risks from, for example, transport accidents that 

result in waste spillage. The waste would typically be contained in double 

sealed bulk bags or 200 litre metal drums. 

Loose or exposed LLW waste will not be transported to the landfill site or 

handled at the site. 

Wastes arriving at the landfill under the authorisation will be pre-notified both 

for transport purposes and for acceptability against the waste acceptance 

criteria. Prior to physical receipt of the waste a package of information 

concerning the characteristics of the waste will be submitted by the sender for 

acceptance by the landfill. Augean’s Technical Assessment team will check 

the characterisation information to ensure that the waste is adequately 

described and that the waste meets the waste acceptance criteria. These 

checks may involve quality assurance analysis that is independent of the 

sender. 

Wastes arriving at the landfill will be subject to physical inspection to check 

the integrity of the waste packages and to check the external radiation dose. 

If a waste consignment fails to be acceptable upon receipt at the site 

entrance and can safely be returned to the sender, it will be refused entry to 

the site. 

The stringent pre-acceptance measures will ensure that only acceptable 

wastes are received at the landfill. In the very unlikely case that a waste 

consignment fails to be acceptable upon receipt and may not be safe to 

return to the sender (for example, if a package has been damaged) the 

landfill site operator will quarantine the waste in a safe area set aside for that 

purpose. The response plan for such cases would utilise the resources of the 

consignor and would involve the regulatory authorities. 

Acceptable wastes will be disposed to the landfill void after receipt. The 

waste will be moved to the landfill working face along roads made of suitable 

hardcore materials. 

The waste packages will be lifted using mechanical equipment and placed 

into the landfill at the base of the waste face. Waste packages will not be 

tumble tipped. 




July 2009 

10 


Immediately after completion of a phase of waste emplacement, the waste 

will be covered with at least a 300mm thickness of suitable cover on all 

exposed surfaces and sufficient to ensure that the dose rate at a height of 1 

metre is less than 2 microSv/hr. 

A record will be kept of the waste disposal location. 

S26 Workplace and environmental monitoring will be carried out including, 

groundwater monitoring, leachate monitoring, surface water monitoring and 

monitoring of the working areas. 




July 2009 

11 


Key Facts Summary 

Operational 

Legal Dose Limit for Workers 20 mSv/yr ++ 

Legal Dose Limit for the Public 1 mSv/yr 

Dose Criteria for Workers for this application

Summary of the Environmental Impacts of the Proposal 

S27 The risk from radioactivity has been assessed using a conservative predictive 

model. The model has been used to calculate the capacity of the landfill for 

radioactivity that would result in a risk of less than one in a million per year to the 

public in the long term and the capacity that meets the dose criteria set for the 

unlikely circumstance of future human intrusion into the waste. 

S28 In practice the risk will be even further reduced through good operating practices 

and future management arrangements, but these are not assumed by the model. 

S29 An assessment has been carried out of the exposure of the landfill workers which 

shows that exposures can be maintained below a dose criterion of 1 mSv/yr (the 

dose limit for workers is 20 mSv/yr). This has been confirmed by advice 

received from the Radiological Protection Advisor for the site, the HPA (ref 16). 

S30 The following list summarises the overall impact of the LLW proposal on the 

existing environmental impact statement for the landfill site. 

Impact to Groundwater An insignificant risk from 

pollution 

Impact to Surface Water An insignificant risk from 

pollution 

Impact to Landscape No change in the landform 

Traffic Impact No additional traffic, very low 

risks from traffic incident 

Impact from Noise No additional noise 

Impact to Ecology No additional landtake, 

insignificant risk to animals. 

Impact to Air Quality An insignificant risk from 

release to atmosphere 

Impact to Human Health Insignificant risk in the long 

and short term 

Archaeology No impacts identified 

Proposed Authorisation Conditions 

S31 The application contains a series of proposed authorisation conditions which are 

subject to agreement with the Environment Agency. A proposal is made to 

reflect certain conditions from the existing landfill permit/risk assessment into the 

LLW authorisation in order to ensure consistency with current limits and 

standards. 




July 2009 

13 


S32 The actual capacity of the landfill for LLW depends on the mixture of different 

radionuclides disposed. This is due to the fact that some radionuclides present 

more risk under certain scenarios than others. 

S33 The exact mixture of radionuclides that will be sent to the landfill is not known 

prior to the process commencing because in many cases the wastes have not yet 

been generated by the senders through completion of their decommissioning 

works. The mixture of nuclides in any particular consignment would always be 

known and approved by Augean prior to receipt. 

S34 The proposal is that the capacity of the landfill is subject to a total capacity limit 

combined with a series of other conditions. The total capacity limit would apply 

from the date of issue until closure of the landfill or until the capacity is reached. 

The landfill would receive no more LLW wastes under the permit once the 

capacity limit is reached. The capacity limit cannot be expressed as a single 

number because it depends on the mixture received up to any point in time, so 

the proposal is for a continuously revised capacity limit based on individual 

nuclides (including appropriate daughter chains). The total capacity limit would 

be established using an authorised spreadsheet model agreed with the regulator. 

The spreadsheet model would represent the most restrictive case from the risk 

assessment and would produce as an output the remaining capacity of the landfill 

on an individual nuclide basis given the exact wastes received to that point in 

time. Prior to accepting any further waste the model would be used by the landfill 

operator to determine that the consignment would not lead to a breach of the 

total capacity limit. 

S35 This disposal is not in addition to the existing 249,999 tonne per year landfill 

disposal rate but is part of it and hence no additional traffic results. 

Example Waste Stream 

S36 The application contains information on an example waste stream from the 

Harwell site. It is proposed that the authorisation at ENRMF facilitates the 

reception of LLW with a specific activity of less than or equal to 200 Bq/g from 

any consignor in the UK where the consignor demonstrates to their regulator that 

disposal to ENRMF is the best practicable environmental option. The Harwell 

site is typical of the decommissioning sites and information from the site has 

been included for purposes of illustration only. 




July 2009 

14 


Questions & Answers 

S37 Why has the East Northants Resource Management Facility been proposed for 

LLW disposal? 

The East Northants Resource Management Facility is a modern landfill site 

constructed to the high quality standards required for hazardous waste disposal 

and is hence technically suitable for LLW disposal. 

There are few such well engineered sites in the UK. The UK Nuclear Industry 

Strategy (ref 20) notes that whilst transport and proximity are important 

considerations, when considered on a national level the issue is not a strong 

differentiator between options because the additional impact to transport 

infrastructure or carbon emissions is low. The proposal in this case would not 

result in a net increase in traffic to the site because the annual tonnage capacity 

limit is unchanged. 

S38 Why is it proposed to use a hazardous waste site for LLW? 

Hazardous waste sites are constructed using high standards of environmental 

protection engineering and are subject to rigorous regulation throughout their 

operating period and post-closure. The site has also an established preacceptance, 

technical assessment, consignment and acceptance regime unlike 

non-hazardous landfills. Regulations require that LLW is managed using the 

Best Practicable Means and this is achieved by using a hazardous waste site. 

S39 What is the worst case impact that could happen as a result of the disposal of 

LLW? 

The worst case events are considered in detail in the authorisation 

application. 

The worst case during the waste disposal phase is that a waste package is 

dropped and the contents spilled. The consequences of such an unlikely 

occurrence are minor. 

The worst case after closure of the landfill site is an occurrence in which the 

waste is excavated without knowledge of the contents and then the waste is 

used, for example, as the surface material for new housing development. 

This is a very unlikely occurrence and the consequences of it happening are 

that members of the public would be exposed to low levels of radioactivity, 

below the regulatory dose criteria. 

The radioactive content of the LLW (

S40 What are the impacts of transporting the LLW to the site? 

The safety of the transport of radioactive materials (including LLW with very 

low radioactivity content) is governed by UK dangerous goods transport 

regulations. These require the wastes to be contained in packages 

appropriate for the level of radioactivity bearing in mind what would happen in 

a transport accident. 

If the waste were involved in an accident during transport to the site an 

established response arrangement involving the emergency services 

augmented by suitably qualified and experienced advisors and monitoring 

specialists would be enacted. If waste were spilled it would be a simple 

matter of recovering the spilt materials and sweeping the road. The levels of 

radioactivity involved would not require extensive arrangements during the 

recovery operation and the risk to members of the public from exposure to 

radioactivity would be very low. 

The number of vehicle movements and the environmental impact of transport 

are not increased by this authorisation application because the total capacity 

of the landfill is unchanged by the proposal. 

S41 Have other alternatives to landfill disposal been considered? 

Every site in the nuclear industry (which includes the power stations and 

research sites) wishing to consign LLW to the landfill will first have to 

demonstrate to the regulatory authorities that landfill of their waste is the Best 

Practicable Environmental Option (BPEO). This is a requirement for the 

consigning sites to be granted a transfer authorisation under the Radioactive 

Substances Act which they will require to send wastes. They will also have to 

demonstrate that they have complied with the waste hierarchy and have 

therefore exhausted options to Reduce, Recycle and Reuse the materials. 

Consideration of the BPEO requires all alternatives to be considered. In 

some cases, for example, LLW can be treated, incinerated or recovered 

through smelting. Landfill disposal will be considered as a last resort after 

these other approaches have been considered. 

It is likely the majority of LLW will arise from historically contaminated land 

and buildings for which the other waste management options are generally 

less applicable. 

S42 What controls would be in place to ensure the LLW waste can be disposed of 

safely? 

The waste will be subject to pre-acceptance tests to ensure it is acceptable. 

The waste will be checked upon arrival at the site. Radioactivity can be 

easily and immediately measured using simple instruments to ensure waste 

acceptability. 




July 2009 

16 


The waste will be handled in enclosed containers. 

The waste will be disposed immediately after receipt and covered with 

material layers. 

The landfill is experienced with handling hazardous wastes, has well 

developed procedures/arrangements and has a good safety culture. 

Specific procedures and a radiation protection plan will be established for the 

LLW operations. 

Monitoring will be carried out to ensure protection of the workers and the 

public. 

The operations are subject to specific authorisation, regulation and inspection 

by the EA and works will be carried out in accordance with the authorisation 

conditions. 

S43 Who regulates the disposal of LLW? 

Conclusion 

The principal regulator in England is the Environment Agency and the 

disposal is authorised under the Radioactive Substances Act. The 

occupational safety of workers and the public is regulated by the Heath & 

Safety Executive principally under the Ionising Radiations Regulations. The 

road transport of radioactive materials is regulated by the Department for 

Transport. 

S44 There is a strategic need for landfill waste disposal routes for materials 

containing very low levels of radioactivity that arise from decommissioning the 

nuclear industry sites and from other sources. Several such routes are likely to 

be required in the UK and the proposal to use the East Northants Resource 

Management Facility would provide a significant capability. 

S45 The amount and concentration of radioactivity in the waste is very low and 

presents a very low risk. A risk assessment has been carried out and the 

capacity of the landfill to receive the waste has been estimated using a 

conservative method. 

S46 The proposal for disposal of LLW is subject to agreement of, and the issue of an 

authorisation by the Environment Agency. 




July 2009 

17 


Annex to the Summary: What is Radioactivity? 

Introduction 

Humans are exposed to ionising radiation every day. This exposure comes from background 

radiation. 

Humans can have additional exposure from other sources; for example having an x-ray or flying 

in a commercial aeroplane results in additional exposure to ionising radiation. It is necessary to 

limit exposure because there is a possibility of adverse health effects. 

Annex A to the application document gives full background information on radiation. 

What is Ionising Radiation? 

All matter is made up of atoms consisting of a nucleus surrounded by negatively charged 

electrons, similar to the sun surrounded by the planets. The nucleus consists of neutrons and 

positively charged protons. 

Atoms containing the same number of protons have identical chemical properties and are known 

as elements. Elements with a different number of neutrons are known as isotopes. There are 88 

naturally occurring elements some examples of which are oxygen, iron, sulphur, uranium and 

radon gas. 

Some atoms are radioactive (they are called radionuclides) and the nucleus of such atoms can 

change structure (lose energy); in so doing the energy is emitted as radiation in three main forms: 

alpha rays, 

beta rays and 

gamma rays. 

This process is termed radioactive decay and the resulting daughter product, a new element, is 

formed as a result. These radiations can interact with surrounding matter to produce positively 

and negatively charged particles (a type of electricity). This process is called ionisation, hence 

the term “ionising radiation”. X-rays are also known as ionising radiation and they are identical to 

gamma rays except they are emitted by electrons, not by the nucleus. 




July 2009 

18 


What are the Properties of Ionising Radiation? 

Alpha rays and beta rays are sub-atomic particles that travel at close to the speed of light 

(300,000,000 metres per second). Alpha rays can be stopped (energy absorbed) by a piece of 

paper, while beta rays can be stopped by one or two centimetres of human tissue. 

Gamma rays and X-rays are waves of energy similar to visible light, except they have more 

energy and are invisible. They travel at the speed of light and penetrate matter more easily than 

the particulate radiations. 

What Units are used to Measure Radioactivity? 

Radiation is measured in decays (disintegrations) per second which corresponds to the number 

of nuclei losing energy each second. One Becquerel (abbreviation Bq) is equal to one decay per 

second: one megabequerel is equal to one million disintegrations per second. The human body 

is naturally radioactive due to the presence of radioactive potassium: A 70 kilogram person would 

contain about 3500 Bq. 

How Does Radiation Interact with Matter? 

When the energy from radiation is absorbed by matter, chemical changes occur at the atomic 

level. If the exposure is large enough these changes can be readily observed. For example, if 

glass is heavily irradiated it changes colour. Some precious stones are coloured for commercial 

purposes using this method. When the body is subjected to a medical X-ray the bones absorb 

most of the energy and a photographic film can then give an image of the skeleton. The amount 

of radiation absorbed per gram of matter is called the “absorbed dose”. 

What Units Are Used To Measure Absorbed Dose? 

Absorbed dose is measured in grays (abbreviation Gy). One gray corresponds to one joule of 

radiation energy deposited in one kilogram of matter. (Note: It would require 320,000 joules of 

energy to boil one kilogram (one litre) of water). This is a large unit and the milligray (mGy), 

which is one thousandth of a gray, is more commonly used. 

When radiation interacts with living tissue the effect it has varies with the type of radiation. Alpha 

rays are 20 times more effective than beta and gamma rays at causing tissue damage. To allow 

for this, the dose in grays is multiplied by an effectiveness factor and the new units are called 

sieverts (abbreviation Sv) and the dose is called the “equivalent dose”. A one milligray dose of 

alpha rays is equal to 20 mSv (millisieverts) of equivalent dose. A one milligray dose of beta rays 

is equal to 1 mSv equivalent dose because the effectiveness factor is 1 for beta rays. In most 

cases the effectiveness factor is unity and the dose in grays is equal to the dose in sieverts. 




July 2009 

19 


How Does Radiation Interact with the Human Body? 

When radiation is absorbed in the body it causes chemical reactions to occur which can alter the 

normal functions of the body. At high doses (above 1 sievert) this can result in massive cell death, 

organ damage and possibly death to the individual. At low doses (less than 50 mSv) the situation 

is more complex. 

The body is made up of different cells. For example we have brain cells, muscle cells, blood cells 

etc. It is the genes within a cell that determine how a cell functions. If damage occurs to the 

genes then it is possible for a cancer to occur. This means the cell has lost the ability to control 

the rate at which it reproduces. 

Radiation can cause this effect and at low doses it is the only known deleterious health effect. 

This type of event is very unlikely to occur, and an estimate of its frequency can only be obtained 

by measuring the effect at higher doses and calculating the probability at low doses. 

Annex A of the application document gives more detail of the health effects of radiation. 

The Natural Background 

The effect of radiation on health must be discussed within the context of the natural background. 

Background radiation consists of cosmic rays from space and radiation present in the earth from 

when it was formed. Cosmic radiation increases with altitude and so airline pilots receive a 

higher exposure from this source; the dose rate at 12,000 metres being about 150 times the sea 

level dose. The terrestrial radiation comes from naturally occurring radioisotopes of potassium 

and rubidium and from decay products of uranium and thorium. On average two thirds of the 

dose people receive comes from terrestrial sources. Most of this dose comes from the gas, 

radon, which is a decay product of uranium and thorium. Radon emanates from the soil and 

tends to concentrate in buildings. 

Source Of Exposure Exposure 

Total Natural Radiation (Average UK) 2.2 mSv per year 

Seven Hour Aeroplane Flight 0.02 to 0.07 mSv 

Chest X-Ray 0.04 mSv 

Cosmic Radiation Exposure of Domestic Airline Pilot 2 mSv per year 

Examples of Exposure to Ionising Radiation 




July 2009 

20 


Exposure Limits 

The International Commission on Radiological Protection (ICRP) has set the following limits on 

exposure to ionising radiation: 

The general public shall not be exposed to more than 1 mSv per annum (over and above 

natural background). 

Occupational exposure shall not exceed 20 mSv per annum. 

These limits exclude exposure due to background and medical radiation. 

More restrictive targets than these limits are proposed by the authorisation application. 

In this application the dose criteria for workers is

Supporting Information for the Application 




July 2009 

22 



1.0.1 This document provides supporting information for an application for 

authorisation under the Radioactive Substances Act (RSA) 1993, for disposal of 

solid Low Level Radioactive waste (LLW) with a specific activity of less than or 

equal to 200 Bq/g, including High Volume Very Low Level Waste (HV-VLLW), at 

the East Northants Resource Management Facility, operated by Augean plc. 

1.0.2 The waste has a very low radioactivity content which this application 

demonstrates would present a very low risk if disposed. 

1.0.3 This document provides information to the Environment Agency, as regulator, in 

order that they can consider the application for authorisation. This document is 

also a public document. 

1.0.4 The key sections of the document are: 

The summary. 

The main body of text provides information relating to the application in 

accordance with the guidance on contents issued by the Environment Agency 

(ref 1). 

A glossary with explanations of special terms. 

An Annex containing further introductory information on radiation and 

radioactivity. 

Annexes containing risk assessments for the application which examine the 

safety of the proposed waste disposals. 

An Annex containing background information on the risk assessment 

methodology. 

A copy of the application form for the disposal authorisation (under sect 13 of 

the RSA). 

1.0.5 This document contains proposed arrangements and conditions which are 

subject to regulatory approval and changes. If the application is granted the 

conditions that apply will be those established by the authorisation and by 

detailed supporting operational documentation prepared to address the 

authorisation. 




July 2009 

23 


2.0 Authorisation 


2.1.1 Landfill disposal is considered a valid option (ref 3, 18, 19, 20) for the disposal of 

LLW with a low specific activity from the nuclear industry and from other nonnuclear 

industry sources such as hospitals. 

2.1.2 The disposal will be authorised under the Radioactive Substances Act 1993 (ref 

4). The permitting arrangements are currently under review to incorporate the 

approach within the Environmental Permitting Regulations 2010 (ref 19). 

2.1.3 The proposed LLW (

The proposed LLW waste will have a radioactivity content of less than or 

equal to 200 Bq/g. Where Bq/g is Becquerel per gram, a Becquerel is a 

measure of radioactivity equivalent to 1 disintegration per second and hence 

Bq/g is a measure of the “concentration” of radioactivity, also called specific 

activity. 

The lower limit of LLW for man made substances is currently 0.4 Bq/g below 

which the material is not subject to specific regulatory control. Other 

exemption/exclusion levels may apply to particular nuclides/radioelements. 

For this authorisation application, the waste is a LLW in the range: 

- a specific activity greater than an applicable exemption/exclusion level 

and up to 200 Bq/g total specific activity. 

If wastes of less than the exemption/exclusion level are mixed in with the 

LLW as an inevitable result of their production, in a manner that makes 

separation impracticable, then these would also be treated as LLW. 

The total specific activity would be averaged appropriately in order to be 

representative of the individual waste package and in any case over not more 

than 4 tonnes. 

The LLW may contain waste which were it not classified as a radioactive 

waste would be classified as Inert, Non-Hazardous or Hazardous. 

The LLW may contain waste which would be defined as High Volume – Very 

Low Level Waste (HV-VLLW) in accordance with policy (ref 3), but is not 

limited to that definition. 

2.3 Strategic Need 

2.3.1 Disposal routes in the UK for LLW are limited and often the only option available 

is disposal to the LLW repository near to the village of Drigg in Cumbria. The 

LLW repository does not have capacity for the volumes of LLW that will be 

generated from broad decommissioning of the nuclear industry and it is not 

thought to be a sustainable use of the repository, which has been designed and 

engineered for materials with radioactivity content in the higher range of activity 

of LLW. The strategic need for alternative fit for purpose disposal routes is 

established and detailed within the UK nuclear industry LLW strategy (ref 20) and 

for non-nuclear industry users in UK government policy (ref 3). 

2.3.2 The strategic drivers for new LLW disposal routes are: 

Decommissioning: A disposal route for LLW with low specific activity will 

make it possible to decommission many nuclear industry and non-nuclear 

industry legacies across the UK. The lack of such a route may hold-up 

decommissioning and increase costs for the taxpayer. 

Sustainability: It is government policy that LLW management solutions 

should be provided earlier rather than later. The provision of a new LLW 




July 2009 

25 


management option will allow current stockpiles of waste to be disposed 

safely and hence not leave the issue to future generations. 

Technical Benefit: Some of the waste that will be labelled as LLW from 

the nuclear industry will be essentially clean demolition materials that for 

technical reasons cannot be proven to be clean for free release. Such 

projects are currently difficult to undertake and the provision of a new 

LLW management option will enable such projects to proceed. 

Regional Scale: One approach for the provision of new LLW 

management options is to provide specialised landfill facilities at the site 

of origin. However, this could result in many relatively small landfills 

being constructed across the UK at the existing nuclear sites, many of 

which are not in favourable geological settings for such uses. As is the 

case for conventional wastes, it is reasonable to propose that regional 

solutions which balance transport distance with economies of scale are 

worth consideration. The transport of LLW with low specific activity does 

not present challenging hazards because of the very low levels of 

radioactivity involved. The amounts are small compared to conventional 

wastes and will be generated slowly over several decades. A LLW 

disposal route at the East Northants Resource Management Facility site 

could serve multiple nuclear industry sites. There are few such well 

engineered sites in the UK. The UK Nuclear Industry Strategy (ref 20) 

notes that whilst transport and proximity are important considerations 

when considered on a national level the issue is not a strong differentiator 

between options because the additional impact to transport infrastructure 

or carbon emissions is low. The proposal in this case would not result in 

a net increase in traffic to the site because the annual tonnage capacity 

limit is unchanged. 

International Experience: Other countries that have progressed with the 

clean-up of their nuclear legacies have found great benefit from having 

waste routes for LLW disposal to landfill. There are examples from the 

USA and both Spain and France have recently opened such a route. 

UK Government Policy: Advisory committees in the UK have examined 

the case for provision of such waste routes (ref 5) and concluded that 

government policy should be supportive. Government policy in this area 

has recently been revised and enables the provision of landfill waste 

routes for LLW under appropriate circumstances (ref 3). The UK Nuclear 

Industry LLW Strategy is supportive of the option (ref 20). 

Low Level Waste Repository (LLWR) Acceptance Criteria: The LLW 

Repository Ltd criteria for waste acceptance at the disposal facility near 

the village of Drigg, Cumbria states that: “Waste shall not be Consigned 

for disposal if reasonably practicable measures could be adopted to 

segregate it from other arisings such that disposal is possible as any of 

the following: very low level waste, low level waste in domestic refuse or 

as a special precautions disposal at suitable landfill sites”. This is 

consistent with government policy. 




July 2009 

26 


3.0 Policy and Regulatory Background 

3.0.1 This section is not intended to be a comprehensive review and reference should 

be made to the source documents for further detail. Points of particular 

relevance to the application for authorisation are made. 

3.1 Radioactive Substances Regulation 

3.1.1 Regulation of LLW is summarised by the Environment Agency in “Considerations 

for Radioactive Substances Regulation under the RSA 1993 at Nuclear Sites...” 

(ref 2). Although the East Northants Resource Management Facility is not and 

will not be a nuclear licensed site, the provisions apply to wastes that arise from 

the nuclear industry who operate on nuclear licensed sites. 

The Environment Agency is responsible under the Radioactive 

Substances Act 1993 for regulating all disposals of LLW from nuclear 

sites in England and Wales. 

The Environment Agency issues authorisations (permits) which include 

limits and conditions. 

The Environment Agency regulates the “source” site which generates or 

transfers the LLW and the “destination” disposal site which receives the 

waste in the case of solid waste disposal. 

In addition to issuing or varying authorisations, the Environment Agency 

periodically reviews authorisations, carries out inspections, investigates 

incidents, assesses public exposure and has powers of enforcement. 

3.1.2 The guidance on requirements for authorisation (ref 18) for near-surface disposal 

facilities for solid radioactive wastes has recently been issued. It is proposed that 

this application falls outside of the scope of that guidance because the 

application does not involve a facility solely for the disposal of solid radioactive 

waste and can be assessed using simple conservative approaches. However, 

the application has been prepared to be consistent with the guidance. The 

following list describes key relevant points from the guidance and where they are 

addressed by the application: 




July 2009 

27 


Objective/Principle/Requirement from Guidance on Requirements 

for Authorisation 

The fundamental protection objective is to ensure that all disposals of 

solid radioactive waste to facilities on land are made in a way that 

safeguards the interests of people and the environment, now and in 

the future, commands public confidence and is cost-effective. 

Principle1: Solid radioactive waste shall be disposed of in such a way 

that the assessed radiological risks to people and the environment in 

the future are no greater than the risks that would be acceptable at the 

time of disposal. 

Principle 2: Both at the time of disposal and in the future, the 

radiological risks to people and the environment from a disposal of 

solid radioactive waste shall be as low as reasonable achievable 

under the circumstances prevailing at the time of disposal, taking into 

account economic and societal factors and the need to manage any 

non-radiological hazards. 

Principle 3: Both at the time of disposal and in the future, the standard 

of protection to people and the environment against radiological 

hazards from a disposal of solid radioactive waste shall be no less 

stringent than the nationally acceptable standard at the time of the 

disposal. 

Principle 4: The level of protection to people and the environment 

against any non-radiological hazards associated with disposing of 

solid radioactive waste shall be no less stringent than that provided by 

the nationally acceptable standard for such hazards from the disposal 

of any other waste at the time of disposal for wastes that present a 

non-radiological but not a radiological hazard. 

Principle 5: Both at the time of disposal and in the future, 

unreasonable reliance shall not be placed on human action to protect 

the public and the environment against radiological and any nonradiological 

hazards from a disposal of solid radioactive waste. 

R1 and R2 n/a 

R3: The developer should take the lead on dialogue with the potential 

host community, other interested parties and the general public. 

R4: An application under RSA 93 relating to a proposed disposal of 

solid radioactive waste should be supported by an environmental 

safety case. 

R5: The developer/operator of a disposal facility for solid radioactive 

waste should foster and nurture a positive environmental safety culture 

at all times and should have a management system, organisational 

structure and resources sufficient to provide the following functions: (a) 

planning and control of work; (b) the application of sound science and 

good engineering practice; (c) provision of information; (d) 

documentation and record keeping; (e) quality management. 




Addressed by… 

Section 5.1 and 

8.0 

Section 8.0 

Section 11.1 


8.0 


12.0 

Section 8.0 

Section 16.0 

Section 8.0 

Section 15.0 

R6: During the period of authorisation of a disposal facility for solid Section 8.0 

July 2009 

28 


Objective/Principle/Requirement from Guidance on Requirements 

for Authorisation 

radioactive waste, the effective dose from the facility to a 

representative member of the critical group should not exceed a 

source-related dose constraint of 0.3 mSv/year. 

R7: After the period of authorisation, the assessed radiological risk 

from a disposal facility to a person representative of those at greatest 

risk should be consistent with a risk guidance level of 10 -6 per year (i.e. 

1 in a million per year). 

R8: The developer/operator of a near-surface disposal facility should 

assess the potential consequences of human intrusion into the facility 

after the period of authorisation on the basis that it is likely to occur. 

The developer/operator should, however, consider and implement any 

practical measures that might reduce the chance of its happening. 

The assessed effective dose to any person during and after the 

assumed intrusion should be consistent with a dose guidance level in 

the range of around 3 mSv/year to around 20 mSv/year. 

R9: The choice of waste acceptance criteria, how the selected site is 

used and the design, construction, operation, closure and post-closure 

management of the disposal facility should ensure that radiological 

risks to members of the public and to the environment, both during the 

period of authorisation and afterwards, are as low as reasonably 

achievable (ALARA), taking into account economic and social factors. 

R10: The developer/operator should carry out an assessment to show 

that the radiological effects of a disposal facility on the accessible 

environment are acceptably low, both during the period of 

authorisation and afterwards. 


waste should demonstrate that the disposal system provides adequate 

protection against non-radiological hazards. 

R12: n/a 


waste should make sure that the site is used and the facility is 

designed, constructed, operated and capable of closure so as to avoid 

unacceptable effects on the performance of the disposal system. 


waste should establish waste acceptance criteria consistent with the 

assumptions made in the environmental safety case and with the 

requirements for transport and handling, and demonstrate that these 

can be applied during operations at the facility. 

R15: In support of the environmental safety case, the 

developer/operator of a disposal facility for solid radioactive waste 

should carry out a programme to monitor for changes caused by 

construction, operation and closure of the facility. 




Addressed by… 

Section 8.0 

Section 8.0 


9.0 


14.0 


12.0 


9.0 

Section 9.0 

Section 5.8 

July 2009 

29 


3.1.3 The HPA have recently issued their advice on Radiological Protection Objectives 

for the Land-Based Disposal of Solid Radioactive Wastes (ref 14) which provides 

overlapping guidance to the guidance on requirements for authorisation. Where 

there is a conflict between these guidance documents the guidance on 

requirements for authorisation has been used. 

3.2 The Radioactive Substances Act 1993 

3.2.1 The Radioactive Substances Act (RSA) 1993 defines what a radioactive waste is, 

establishes that users of radioactive material/waste must be registered (section 6 

and 7) and establishes arrangements for the authorisation of disposal and 

accumulation of radioactive waste (section 13). The Environment Agency 

provides regulation of RSA 1993 in England and Wales. 

3.2.2 This application for authorisation of disposal and for registration as a user of 

radioactive materials is made under the RSA 1993. 

3.2.3 Note that the East Northants Resource Management Facility is an existing 

permitted hazardous waste landfill under the Landfill Regulations 2002. The LLW 

authorisation, should one be granted, is additional to the existing permits and 

may be considered a separate but overlapping regime. The total tonnage 

capacity and the broad environmental impact of the landfill will be unaffected by 

the permitting of this type of waste. 

3.2.4 Where wastes are currently acceptable under the existing permits it is proposed 

that these arrangements will not be altered by any new authorisation granted 

under RSA 1993. The LLW authorised under RSA 1993 will have to comply with 

both the existing risk assessments (which underpin the existing permit conditions, 

as far as they can be applied and referred to in the new permit) and the RSA 


3.2.5 LLW is radioactive waste and is therefore not a “controlled” waste in England and 

Wales (ref 7 and 11). The regime of regulation that applies to conventional 

landfill sites encompasses controlled wastes and does not cover LLW. The 

Hazardous Waste Regulations exclude radioactive waste except in the special 

case where the waste is exempt from the requirements of the RSA 1993 for 

disposal but is still a radioactive waste. This can occur for some so-called 

“exempt” wastes that are

waste properties the risk assessment underpinning the existing permits is 

conservative because LLW with non hazardous or inert properties will have a 

lower non-radiological risk than the existing risk assessment allows for. 

3.2.7 An approach to regulation could be for the RSA authorisation to replicate the 

relevant conditions of the existing permit/risk assessments in addition to further 

conditions specific to the LLW. This could be achieved by reference to the 

existing risk assessments which underpin the current landfill permit and this 

approach is assumed in the remainder of this document. 

Existing Permit under the 

Landfill Regulations 2002 




New Authorisation 

under RSA 1993 for 

LLW disposal which 

also refers to existing 

risk assessment 

constraints 

A Hazardous Waste stream Applies Does not apply because 

the waste is not a LLW 

LLW that does not also Does not apply because the 

Applies 

possess Hazardous Waste 

properties 

waste is not a controlled waste 

LLW that does also possess Does not apply because the 

Applies 

Hazardous Waste properties waste is not a controlled waste 

The parallel regimes of the Landfill Regulations and the Radioactive Substances Act 

3.2.8 The majority of the wastes will be LLW with other properties that are nonhazardous 

or inert. This raises the question of why a hazardous waste site such 

as the East Northants Resource Management Facility should be authorised as 

opposed to a non-hazardous site. The reasons for proposing this arrangement 

are: 

Use of a hazardous site will allow the small amounts of LLW that are also 

landfillable hazardous waste to be consigned (for example, asbestos 

gaskets in radioactively contaminated waste ventilation ducts) and this 

helps prevent such wastes becoming orphaned. 

Hazardous waste sites have to be engineered using Best Available 

Techniques (BAT) that are to the highest standard for landfill and 

therefore represent a good choice for LLW which have to be disposed of 

using Best Practicable Means (BPM) that also meet a high standard. 

Hazardous landfills have well developed operational procedures for the 

acceptance and handling of difficult wastes. For example, the East 

Northants Resource Management Facility has a comprehensive 

laboratory with qualified chemists who will be able to manage the 

acceptance and disposal of the wastes. The facility utilises waste 

handling methods that are identical to those required for emplacing LLW. 

July 2009 

31 


Hazardous waste landfills receive wastes which are essentially 

incombustible and which pass a flammability waste acceptance test. This 

reduces the probability of a landfill fire to a very low level. 

Hazardous wastes sites are not common in the UK and are therefore 

much less likely to be “forgotten” by future societies. This is useful in 

minimising the risk that future generations will be exposed to the waste. 

(Note that the risk assessment assumes that the site is forgotten as a 

worst case assumption and demonstrates that the disposal is safe 

regardless of this.) 

The probability of inadvertent human intrusion is reduced by the following 

features of the East Northants Resource Management Facility : 

Depth of the disposal horizon. In the majority of cases the LLW will be 

buried several metres into the hazardous waste disposals which will act 

as a visual indicator for future generations that the area is a waste site. 

The site is in an area of low mineral resource potential (the clay has 

been extracted) which makes it less likely that future generations will 

wish to dig deep into the waste. 

The nature of the site as a hazardous site makes it more likely that 

records and knowledge of the site will be retained by future generations. 

The engineered cover design provided for hazardous waste sites makes 

penetration into the waste less likely. 

The landform design, whilst being sympathetic with the surroundings, is 

nonetheless relatively obvious as a non natural feature. 

3.2.9 For the purposes of this authorisation the applicant wishes the site to be treated 

as a “non-nuclear premises” and for the authorisation to be held by the applicant 

for the site (ref 18, 9.2.16). The applicant does not wish for the disposal 

authorisation to be held by the consignor(s). 

3.3 Risk 

3.3.1 Radiation protection and regulation is concerned with ensuring the protection of 

humans from the risks presented by ionising radiation whilst taking into account 

the benefits offered by a particular process. 

3.3.2 In the case of the East Northants Resource Management Facility authorisation 

the focus is to assess the risks presented by the disposal of LLW to the workers 

and the public. The potential benefits of such a waste disposal route have been 

outlined above. The overall regulatory assessment is a balancing of risks and 

benefits. 




July 2009 

32 


3.3.3 Any nuclear site wishing to send waste to such a disposal route will have to carry 

out an assessment of risks and benefits using a framework called “Best 

Practicable Environmental Option”. 

3.3.4 The risk presented by ionising radiation is directly related to the amount of 

radiation to which a person is exposed. The radioactivity exposure is called 

“dose”. Internationally accepted systems of relating risk to dose have been 

established and are used by the UK (refs 12, 14, 16). 

3.3.5 Risk guidance levels/criteria are established by UK policy and internationally 

accepted good practice. The amount of LLW that can be disposed in a landfill is 

limited by the risk guidance levels/criteria and subject to further optimisation 

through consideration of the process as a whole. 

3.3.6 The risk arising from a waste disposal can be calculated for the short and long 

term to both workers and the public. This can then be used to establish the 

amount of radioactivity that can be disposed in the landfill (which is called the 

radiological capacity) without exceeding the risk criteria. 

3.3.7 The authorised radiological capacity is set to ensure that the resulting dose 

presents a very low risk. This is achieved through prospective dose 

estimation/risk assessment calculations. Once the disposal process is operating, 

workplace and environmental monitoring can be used to check the actual 

exposure of humans to radioactivity. 

3.3.8 It is not necessary to calculate the exposure of all humans exposed to the 

radioactivity as long as the humans that might be most exposed are assessed. 

This is called the “critical group” methodology. In practice, a few groups may be 

assessed to ensure coverage of workers and the public both in the long and short 

timescales. 

3.4 UK Government Policy 

3.4.1 Policy on radioactive waste is set out in the White Paper, Review of Radioactive 

Waste Management Policy, Cm 2919. In respect of LLW, the policy has been 

amended by the Policy for the Long Term Management of Solid Low Level 

Radioactive Waste (ref 3). 

3.4.2 The policy allows for disposal of LLW at specified landfill sites, provided that this 

meets regulatory requirements. The maximum specific activity of the LLW 

(

easonable attempts have been made to avoid, reduce, recycle and reuse the 

material. 

3.4.4 The government policy advises that a risk target of 10 -6 /year (one in a million) of 

developing a fatal cancer or serious hereditary defect should be used as an 

objective in the design process. Where the regulators are satisfied that best 

practicable means have been adopted by the operator to limit risks and the 

estimated risks to the public (now and in the future) are below this target, then no 

further reductions in risk should be sought. The guidance has been developed 

further in guidance of requirements for authorisation (ref 18) discussed above. 

3.5 Basic Safety Standards Directive 1996 (BSSD) and The 

Radioactive Substances Direction 2000 

3.5.1 European law influences the regulation of LLW disposal as follows: 

A requirement is established to achieve the ALARA (as low as reasonably 

achievable) principle. All exposures to ionising radiation of any member 

of the public and of the population as a whole resulting from the disposal 

of LLW are kept as low as reasonably achievable, economic and social 

factors being taken into account (ALARA). 

To achieve ALARA the stated maximum doses to individuals resulting 

from a defined source are 0.3 mSv/year from any single new source (mSv 

is milliSievert a measure of radiation dose; for comparison the 

background natural radiation dose in the UK is 2.2 mSv/year or more). 

This applies to current discharges that could be altered by changes to 

current operating arrangements. Government policy in Cm 2919 proposes 

a threshold of 0.02 mSv/year as equivalent to an annual risk of death of 

around 1 in a million/year (10 -6 /yr). 

The range 0.3 mSv/yr to 0.02 mSv/yr is the dose range over which a 

process can be optimised in accordance with the ALARA principle. If 

doses are below 0.02 mSv/yr, the regulators should not seek to secure 

further reductions provided they are satisfied that the operator is using 

best practicable means to limit discharges. 

Regardless of the above targets there is a UK legal limit derived from 

European law for the exposure of nuclear workers and the public from all 

man-made sources of radioactivity (other than medical exposure). The 

dose limit for members of the public is 1 mSv/yr and the dose limit for 

workers is 20mSv/yr. In the case of the ENRMF the landfill and other 

workers would be managed to a dose limit of 1 mSv/yr as specified by the 

site’s radiation protection plan. 

3.5.2 BSSD defines the optimisation principle. In current practice a concept called the 

use of Best Practicable Means or BPM is used to demonstrate in part that 

optimisation has been addressed. The EA is currently consulting on whether 




July 2009 

34 


adioactive substances regulation should be brought under the Environmental 

Permitting regime. In the latter case the Best Available Techniques or BAT 

concept would be applied to the LLW management process. 

3.5.3 BSSD sets out requirements on the EA in relation to permitting which have been 

addressed by this application including: dose limits, dose constraints, 

authorisation conditions designed to be protective of human health, authorisation 

limits which have in-built safety factors, flexible authorisation limits and proposals 

for environmental monitoring. 

3.5.4 BSSD sets out the requirements that management of radioactive waste disposal 

should be undertaken following consultation by an operator with a Qualified 

Expert. For this application qualified experts are provided under contract to the 

site operator by the Health Protection Agency and the site operator has used 

suitably qualified and experienced advisors to prepare the application obtained 

from Galson Sciences and UKAEA. 

3.6 Environmental Permitting Regulations 2007 

3.6.1 The existing East Northants Resource Management Facility is permitted under 

the Environmental Permitting(England and Wales) Regulations 2007. These 

regulations set out a pollution control regime for landfills. 

3.6.2 To operate the landfill, a permit was issued under the Pollution Prevention and 

Control (PPC) Regulations 2000. These regulations have just been rationalised 

into the Environmental Permitting Regulations 2007. A new Environmental 

Permit was issued for the site in March 2009. PPC and EPR seeks to improve 

environmental protection by introducing measures to reduce or prevent 

emissions to air, land and water. 

3.6.3 The existing landfill has been built and is operated within these regulations and 

hence the pollution prevention measures which exist are of direct use in 

minimising pollution from the LLW. 

3.6.4 The EPR regime does not currently incorporate radioactive materials because 

they are not controlled wastes. However an activity may be controlled by both 

EPR and the Radioactive Substances Act 1993 and that regulators will ensure 

that the two regimes do not impose conflicting obligations on the same matter. 

3.6.5 The existing landfill is permitted under these regulations and that permit applies 

conditions. In order for LLW to be permitted for disposal in the landfill they 

require to be permitted under the Radioactive Substances Act and it is proposed 

that the risk assessment constraints that apply to the existing landfill should be 

replicated within the RSA authorisation where applicable and in a non conflicting 

manner. 




July 2009 

35 


3.6.6 This application document does not include detailed information concerning the 

existing permit and the underpinning risk assessments because these documents 

have been previously submitted through the regulator. 

3.7 Conservation (Natural Habitats and Conservation) Regulations 

1994 

3.7.1 The Habitats Regulations require the Environment Agency to be satisfied that the 

integrity of designated “European sites” (sites with certain ecological value) will 

not be adversely affected by relevant permissions issued by the Agency. These 

regulations have been addressed by the existing permit for the landfill. 

3.8 Ionising Radiations Regulations 1999 

3.8.1 Ionising radiations occur as either electromagnetic rays (such as X-rays and 

gamma rays) or particles (such as alpha and beta particles). Radiation occurs 

naturally (e.g. from the radioactive decay of natural radioactive substances such 

as radon gas and its decay products) but can also be produced artificially. People 

can be exposed externally, to radiation from a radioactive material or a generator 

such as an X-ray set, or internally, by inhaling or ingesting radioactive 

substances. Wounds that become contaminated by radioactive material can also 

cause radioactive exposure. 

3.8.2 Everyone receives some exposure to natural background radiation and much of 

the population also has the occasional medical or dental X-ray. The Health and 

Safety Executive (HSE) is concerned with the control of exposure to radiation 

arising from the use of radioactive materials and radiation generators in work 

activities in the nuclear industry; waste; medical and dental practice; 

manufacturing; construction; engineering; paper; offshore drilling; education 

(colleges, schools) and non-destructive testing industries. 

3.8.3 The main legal requirements enforced by HSE are detailed in the Work with 

ionising radiation: Ionising Radiations Regulations 1999 Approved code of 

practice and guidance. 

3.8.4 The regulations will apply to the workers at the landfill and visitors to the site. 

The regulations require risk assessment and specialist advice from a Radiation 

Protection Adviser to be enacted prior to work with ionising radiations. The 

advice received will result in a safe system of work at the site to limit, control and 

measure exposure. 

3.8.5 The very low amounts of radioactivity in the LLW mean that relatively simple 

arrangements will be sufficient and that the workers will be treated as members 

of the public for purposes of dose limitation. 




July 2009 

36 


3.9 Nuclear Industry LLW Strategy 

3.9.1 The UK Nuclear Industry Strategy has been published for consultation by the 

Nuclear Decommissioning Authority (NDA). The strategy underpins the strategic 

need for new waste management options for LLW disposal. A key theme of the 

strategy is “development and use of new fit for purpose management and 

disposal routes, so waste producers have more choice in determining and 

implementing waste management routes”. 

3.10 Other 

3.10.1 Article 37 of the Euratom Treaty requires member states to the European 

Commission to provide sufficient information about plans to dispose of 

radioactive waste to allow the Commission to decide whether the plans could 

cause radioactive contamination of the water, soil or airspace of another Member 

State. It is assumed by this application that Article 37 submissions, where 

required, are implemented by the consigning nuclear industry sites. 

3.10.2 The 1992 OSPAR convention and related national strategies seeks progressive 

and substantial reductions of discharges, emissions and losses of radioactive 

substances to the marine environment. The strategies use a dose guidance level 

for the public of 0.02 mSv/yr which is consistent with that used in this application. 

The use of BAT methods is proposed and is consistent with the approach used 

for this application. 

3.10.3 The UK ratified Joint Convention on the Safety of Spent Fuel Management and 

on the Safety of Radioactive Waste Management (IAEA 1997) sets out a 

framework for radioactive waste management. This application has been made 

in a manner consistent with the objectives of this convention. 

3.10.4 The IAEA publish good practice guidance in relation to radioactive waste 

management. This application is consistent with IAEA guidance. 

3.10.5 The ICRP is an independent advisory body that provides recommendations on 

radiation protection. In the UK the HPA provide advice on the recommendations 

of the ICRP. The HPA’s advice concerning the most recent publication ICRP 103 

changes the previous advice to use a single source dose constraint of 0.15 mSv. 

Whilst this application uses the 0.3 mSv figure which is current to UK policy, it is 

not used for calculating the radiological capacity of the landfill and adoption of the 

0.15 mSv constraint would make no impact on the key dimensions of the 

proposal. 

3.10.6 The Town and Country Planning (Environmental Impact Assessment) 

Regulations 1999 may require the environmental effects of development of 

radioactive waste facilities to be assessed as part of the planning approval 

process under the Town and Country Planning Act 1990. A separate planning 

approval process is being pursued for the ENRMF in order for it to receive LLW. 




July 2009 

37 


3.10.7 The new regulations to transpose the 2006 Groundwater Directive expected in 

2009 may apply to radioactive substances. 

3.10.8 The EA has previously issued guidance concerning BPM and BPEO in relation to 

waste management options. The current move is to replace these concepts with 

BAT, this being underpinned by developing Radioactive Substances Regulation 

Environmental Principles. This application has been prepared taking into 

account these developments in so far as practicable. 

3.10.9 The Nuclear Safeguards Act 2000 establishes arrangements through which the 

UK accounts for and protects nuclear materials (plutonium, uranium and 

thorium). The consigning nuclear sites would provide accountancy for such 

materials up to the point of disposal. 

3.10.10 Radioactive waste is transported to a disposal facility under strict controls in 

accordance with Dangerous Goods Regulations which are based upon the 

transport regulations issued by the IAEA. For this application the site will not 

accept unpackaged wastes even where such wastes are compliant with the 

transport regulations. The Dangerous Goods Safety Advisor (DGSA) for the 

consigning sites ensures compliance with the regulations and the contract 

conditions. For this application the receiving site will also have an appropriately 

qualified DGSA to provide advice on auditing, checking and receiving packages. 

3.10.11 The Environmental Permitting Regulations 2010 may be applied to Radioactive 

Substances Regulation and are currently under consultation. 

3.10.12 An assessment of the radiological impact on species other than humans may 

be required to address the Environment Act 1995 and the Conservation 

Regulations 1994. This application has incorporated such an assessment. 




July 2009 

38 


4.0 Site Background Information 

4.1 Site Description and Local Environment 

4.1.1 The landfill site lies approximately 2.5km north of the village of King’s Cliffe in the 

East Northamptonshire District of the County of Northamptonshire at National 

Grid Reference TF010 001 (ref 15) (Figure 1). The setting is generally rural with 

a majority of the land surrounding the landfill site comprising open farmland or 

woodland. The only properties in the immediate vicinity of the landfill comprise a 

terrace of three houses (Westhay Cottages) and Westhay Farm with associated 

agricultural and commercial buildings. These properties are all located to the 

east of the eastern boundary of the landfill site on the other side of the site 

access road. 

4.1.2 Landfilling operations at East Northants Resource Management Facility 

commenced in 2002. 

4.1.3 The closest village to the site is Duddington, approximately 2.2km to the 

northwest and King’s Cliffe village which lies approximately 2.5km to the south 

(Figure 1). Collyweston village is approximately 3.3 km to the north of the site. 

The only other development within the vicinity of the site is the RAF airfield at 

Wittering which lies approximately 800m to the northeast. This is an operational 

airfield used for pilot training. 

4.1.4 The landfill lies within the Rockingham Forest/Lower Nene Valley Special 

Landscape Area, a local designation adopted by the County Council in 1974. 

This is an area of relatively level to gently undulating land at an elevation of 

approximately 85m above ordnance datum. The predominant land uses within 

the immediate area of the site are agriculture and woodland. 

4.1.5 Part of the Collyweston Great Woods to the north of the site is designated a Site 

of Special Scientific Interest (SSSI) and a National Nature Reserve (NNR). 

4.1.6 The A47 road lies approximately 1 km to the north of the site, with the Stamford 

Road, an unclassified road linking the A47 to the village of King’s Cliffe, passing 

along the eastern boundary of the site (Figure 1). 

4.1.7 The geology of the site consists of an upper clay strata formed from glacial till 

and estuarine mudstone to a depth of approximately 11.5m, underlain by 

Jurassic limestone. The clay strata have been partially worked as part of the 

quarrying operations on site, with overburden materials stockpiled to the western 

end of the site and when required silica clay is exported off site. 

4.1.8 There are no main water courses in proximity to the landfill site, the closest being 

Willow Brook (3km south) and the River Welland (2.5km west), the landfill site 

being approximately on the watershed between these two. 




July 2009 

39 


4.2 Business Plans and Site Development Plans 

4.2.1 The landfill site is operated as a hazardous waste landfill with a number of 

ancillary and related waste activities on the site. The disposal rate of the 

engineered landfill cells with hazardous and inert (for cover etc.) waste is 

permitted at a maximum rate of 249,999 tonnes/year (Figure 2). 

4.2.2 It is envisaged that landfill operations will continue until approximately 2013, 

dependant on the actual importation rate. The site will be progressively restored 

and once complete will undergo a defined scheme of capping and restoration. In 

accordance with the extant planning permission the landfill site will be restored 

principally to grasslands for ecological and agricultural afteruse. 

4.2.3 The proposal for LLW disposal at the site would not be envisaged to change the 

total capacity of the site or the physical features that contributed to the original 

landfill permitting decision. 

4.2.4 Operating details for the site are not presented here and are available in the 

supporting documentation for the existing permitted operations (ref 15). The 

operating arrangements and culture at the site are consistent with the 

arrangements proposed for LLW disposal in this application. 

4.3 Existing Permits 

4.3.1 The East Northants Resource Management Facility landfill is operating under an 

Environmental Permit (TP 3430GW) issued May 2009, for the disposal of 

hazardous waste. The site commenced operations in 2002 under a PPC Permit 

and was originally a co-disposal site for the disposal of non-hazardous and 

hazardous wastes. Since the beginning of 2004, the site has received 

predominantly hazardous waste and the practice of co-disposal has ceased. The 

site is therefore now a hazardous only site apart from the need for suitable cover 

materials. The permit boundary covers an area of 17.27 hectares with some 

14.36 hectares for the maximum extent of the cells. 

4.3.2 The wastes accepted at East Northants Resource Management Facility cover a 

broad spectrum of those defined as hazardous under the European Waste 

Catalogue subject to the hazardous waste acceptance criteria. These criteria in 

particular exclude explosive, flammable, corrosive and infectious materials. 




July 2009 

40 


5.0 Radioactive Waste Disposal Proposal 

5.0.1 This section provides an outline of the proposed arrangements for the LLW 

disposal process. After granting of the authorisation this outline would be 

developed into detailed operational written safe systems of work in accordance 

with the authorisation conditions. 

5.0.2 It is useful to consider four distinct phases for the timeline of the facility: 

The operational phase when the facility is receiving waste. 

The active institutional control (aftercare) phase which covers the time 

from closure of the facility to the time when provisions for active aftercare 

ceases (60 years or greater in the case of the East Northants Resource 

Management Facility in accordance with the existing permit). 

The passive control period over which records are expected to inform 

future generations of the presence of radioactive waste. 

The uncontrolled phase when all records might be expected to have been 

lost. 

5.1 Principles and Dose Criteria 

5.1.1 Dose criteria for LLW disposal are well established in the UK (ref 18) and 

considerable good practice guidance exists. In order of decreasing dose, the 

criteria can be “limits” which are legally established, “constraints” which are levels 

established by approved practice or “criteria”/“targets” /”guidance levels” which 

are good practices established by guidance. The dose criteria used for this 

authorisation application are: 

For workers the legal dose limit is 20 mSv/year, and the criterion used for 

this application is 1 mSv/year, which is the same as the current legal limit 

for the public. This is an operational criterion and is not used to set the 

radiological capacity of the landfill because the exposure arises in a 

manner unrelated to the total capacity of the site. This criterion does 

affect some of the authorisation conditions, in particular external dose 

limits on packages and the limit on specific activity. This criterion will be 

used for radiation protection purposes during operation of the facility. 

For the public a legal dose limit of 1 mSv/year and a dose constraint of 

0.3 mSv/year would be used during the operational phase. The aim 

would be to ensure through radiation protection measures and monitoring 

that no person received more than the dose constraint during the 

operational phase. This is a constraint and is not used to set the 

radiological capacity of the landfill because this is considered to be an 




July 2009 

41 


upper bound on the region of dose optimisation. This constraint will be 

used for radiation protection purposes during operation of the facility. 

For all persons during the post-closure phases, for natural processes, the 

dose guidance level used to set the radiological capacity is 0.02 mSv/yr 

which corresponds to a risk of 10 -6 /yr or 1 in a million per year. This is 

used to set the radiological capacity of the landfill as a whole and not for 

occupational radiation dose protection. The same criterion is also used 

as a design target for the operational phase for public exposure. 

Inadvertent intrusion into the site in the future is not certain to occur and 

therefore this event has a low probability of occurrence. The dose 

guidance level used is 3 mSv/year, which is the lower end of the range 

indicated by the guidance on requirements for authorisation and HPA 

guidance (ref 14,18). This is used for direct physical intrusion scenarios 

and for intrusion by extraction borehole at the site boundary. 

It is assumed that following closure of the landfill and the end of the 

aftercare period the continued ability of the design to meet the risk target 

does not depend on actions of future generations to maintain integrity of 

the disposal system. 

Following closure of the landfill, it is assumed for the purposes of 

conservative risk assessment that society prevents intrusion into the 

waste form for at least 60 years after closure which is consistent with the 

current financial provision for the long term aftercare of the landfill. In 

practice the site will be under the control of the existing Environmental 

Permit until the Environment Agency is satisfied that the site no longer 

represents a significant risk of harm to human health and pollution of the 

environment. This period will almost certainly be considerably longer than 

60 years. Beyond the 60 year period, it is assumed conservatively and 

for the purpose of risk assessment that humans may penetrate into the 

landfill in a manner that results in continuous exposure without realisation 

of the hazards present. Note that this risk assessment applies only to the 

LLW component of the waste and not to other hazardous wastes that may 

be present in the landfill at this time and which have already been 

permitted for the site on the basis of other risk assessments. 60 years is 

considered reasonable given that the landfill in question is an extensively 

engineered and capped hazardous waste permitted site. The assumption 

of inadvertent intrusion which goes unrealised and which results in 

humans living on the exposed waste form is considered conservative. 

The risk assessment is based upon the principles and scenarios in the 

“SNIFFER SPB model” (Annex E) as adjusted for this site specific case 

(Annex B) and as corrected to address developments required to the 

original SNIFFER model. The SNIFFER model was developed in 

conjunction with the UK regulatory authorities. 

The use of a high quality modern hazardous waste engineered landfill and 

the stated operational arrangements are designed to ensure that the risk 

of radiological exposure to members of the public is as low as reasonably 




July 2009 

42 


Operational 

achievable. This is achieved through the application of best practicable 

means. 

The facility design is already established and is that provided already for 

the landfill to operate as a permitted hazardous waste disposal facility. 

Operational arrangements specific to the LLW are established to augment 

existing arrangements (Sections 5.4 to 5.6). 

The summarised dose criteria are: 

Legal Dose Limit for Workers 20 mSv/yr 

Legal Dose Limit for the Public 1 mSv/yr 

Dose Criterion for Workers for this application

proposed that if the waste has an underpinning justification for disposal 

established by the consigning site and it meets the waste acceptance criteria and 

the waste acceptance criteria of the existing permit as reflected in the RSA 

authorisation, then the waste is acceptable. This would include wastes that if 

they were not radioactive would be classified as Inert, Non-Hazardous or 

Hazardous. 

5.3 Road Transport 

5.3.1 The following outline arrangements are proposed and will be detailed in the 

operating arrangements for the process which will be developed if the 

authorisation is approved. 

5.3.2 The main legislation covering the safe transport of the LLW material is The 

Carriage of Dangerous Goods…Regulations 2007 (ref 13). The emphasis of the 

regulations is for the safe management of each stage of the transport chain. 

Annex A of the ADR contains a section specific to package design to provide the 

main element of safety in normal and accident conditions. 

5.3.3 The onus is on the consignor and carrier of the waste from the source site to 

ensure that it is transported in accordance with the transport regulations. 

Specialist advice must be sought from an appropriately trained person holding 

certification as a Dangerous Goods Safety Advisor (DGSA). Part of the waste 

acceptance arrangements at the landfill will be checks to ensure that the records 

and physical condition of the packages meet the transport regulations upon 

arrival at the landfill site. This is standard practice at the landfill for all waste 

accepted. Nuclear industry sites are experienced in these arrangements and 

have developed practices to ensure they are implemented. 

5.3.4 LLW contains low amounts and concentrations of radioactivity which mean that 

they can be transported safely in standard packages used in the transportation of 

dangerous substances. In some cases the amount of radioactivity will be so low 

that the packages will be exempt from the regulations. Some of the lower activity 

wastes will be transportable in “excepted” packages as defined under the 

regulations and the remainder will be transportable in “industrial” packages. 

Even where wastes could be transported unpackaged as low specific activity 

materials in accordance with the regulations, all wastes will be contained in 

sealed packages. 

5.3.5 Typical packages will be either: 

Flexible Intermediate Bulk Containers. These are usually called “bulk 

bags”. The bags would be transported singly stacked on an enclosed 

freight vehicle and would be handled using pallets or integral lifting loops. 

It is normal to use double sealed bags. The bags would be placed into 

the disposal void using mechanical handling equipment. 




July 2009 

44 


Non reusable type approved waste transport drums (200 litre nominal 

capacity). The drums would be handled on pallets or using drum handling 

equipment. The drums would be placed into the disposal void using 

mechanical handling equipment. 

Single items may be wrapped and sealed. 

5.3.6 Loose or exposed LLW will not be accepted at the landfill site. 

5.3.7 Under the transport regulations the consignee (the receiving landfill site) has 

duties. The landfill site must have a staff training plan in place and a quality 

assurance programme (operating arrangements) to ensure that regulations are 

being adhered to. This would entail checks by the landfill operator on receipt of a 

shipment that the records are correct (all shipments would be pre-notified and 

pre-accepted for shipment), that the shipment is in accordance with the 

regulations, that the packages are in good order and that the external dose rate 

is in accordance with regulations (in addition to the waste acceptance criteria for 

dose rate). A quarantine arrangement would be implemented for non-compliant 

consignments. 

5.3.8 The consignee must maintain a radiation protection programme in accordance 

with the Ionising Radiations Regulations. 

5.3.9 The consignee must ensure appropriate segregation of the packages, upon 

receipt, from persons and other dangerous goods in accordance with the 

regulations. 

5.3.10 The consignee must maintain an operating arrangement for emergencies and 

spillages and provide information to affected parties. 

5.4 Pre-acceptance and Assay 


operating arrangements for the process which will be developed if authorisation 

is approved. 

5.4.2 Wastes that will be delivered at the landfill under the authorisation will be prenotified 

both for radioactive transport purposes and for waste acceptability 

against the waste acceptance criteria. Prior to dispatch of the waste from the 

consignor a package of information concerning the characteristics of the waste 

will be submitted by the consignor for acceptance by the landfill. Augean will 

check the characterisation information to ensure that the waste is adequately 

described and that the waste meets the waste acceptance criteria and that the 

landfill has adequate radiological capacity to receive the waste. 

5.4.3 The waste will be characterised so as to facilitate their subsequent management, 

including waste disposal. Arrangements for characterisation would be regulated 




July 2009 

45 


y the authorisations for transfer issued to the source sites and established by 

the contract between the consignor and the disposal facility. Nuclear industry 

sites operate waste characterisation methodologies in accordance with industry 

good practice guidance. 

5.4.4 Characterisation will include relevant physical, chemical and radiological 

properties. 

5.4.5 Wastes generated within a well-defined process or which can be demonstrated to 

have self-similar characteristics may be characterised as a waste stream. This 

may mean that reduced characterisation of individual packages is required. 

However, the radioactive composition and specific activity of each individual 

waste package would always be reported and averaged over the waste package 

(or 4 tonnes whichever is the smaller). 

5.4.6 Certain characterisation is required by the existing permits for the landfill in 

relation to the receipt of hazardous wastes. Such characterisation will be 

provided for LLW in so far as the conditions of existing permits are referred to by 

the RSA authorisation for LLW disposal. 

5.4.7 Radioactivity related characterisation information for each individual package will 

include: 

An assessment of the amount, concentration and isotopic composition of 

the radioactivity in each individual package. This could be obtained, for 

example, by radiochemical analysis and gamma spectrometry of a 

representative sample, using “fingerprinting” where applicable or using 

radiochemical analysis and bulk gamma spectrometry. The history and 

nature of occurrence of the waste will be taken into account by the source 

site when designing characterisation approaches. The isotopic 

composition must be adequate to characterise the waste in accordance 

with the list of nuclides (heads of chains) used in the risk assessment 

(Annex B). 

The waste characterisation methodology used to obtain the 

measurements and a justification that this meets a best practicable 

means approach. The landfill site operator will review the 

characterisation methodology as part of the waste acceptance process. 

The quality assurance methodology used by the consignor to validate the 

radioactivity measurements. 

The external radiation dose required by the transport regulations and at 1 

metre from the package on all sides and the top. The justification that this 

meets the waste acceptance criteria and transport regulations. 

The surface contamination clearance records. 

Unique identification labelling of each waste package as required under 

the transport regulations. 




July 2009 

46 


5.4.8 Other characterisation information will include: 

The generic information required for basic characterisation by the existing 

permit for the hazardous waste stream: 

Source and origin of the waste 

The process producing the waste 

Waste treatments applied to the waste 

The composition of the waste and an assessment against relevant 

limit values 

The appearance of the waste 

Any equivalent codes applicable to the waste (EWC etc.), 

although this would not apply legally to LLW 

Any hazardous properties according to Schedule 3 of the 

Hazardous Waste Regulations. 

The relevant properties which make the waste hazardous should 

that apply 

A demonstration that the waste is not prohibited 

For waste streams where each package is not characterised and it 

is argued that some of the characteristics are the same across the 

waste stream; the compositional range for the individual 

packages. 

A measurement of the weight of each package. 

The acceptability and characterisation of the waste against the hazardous 

waste landfill acceptance criteria under existing permits where these are 

referred to by the RSA authorisation or other non-radiological criteria 

established by the RSA authorisation. For a hazardous waste landfill site 

this might generically include: 

compliance with the banned substance list, 

compliance with the waste acceptance limits for hazardous waste 

disposal, 

the designated leach testing of the waste/waste streams, where 

applicable, 

Details of any pre-conditioning/treatment of the wastes that has 

been utilised and details of compliance with the “three-point test” 

for pre-treatment. 

The amount of voidage in the waste package (where relevant, for 

example in the case of wrapped items). 

Justification that the waste hierarchy of avoid, reduce, recycle and reuse 

has been applied to the waste. 

Information relating to the safe transport of the waste as required under 

the transport regulations. 




July 2009 

47 


5.4.9 Each package or self-similar group of packages will have a representative 

sample taken at the time of packing. The sample will be retained by the source 

site and be identified uniquely as linked to the package. The nature of LLW 

means that sampling and analysis has to be carried out using specialist 

laboratories and that often the source site will be better equipped to manage this 

than the landfill. Samples may be subject to transport regulations. 

5.4.10 The provision of sealed samples is designed to enable the landfill operator or 

regulators to request check analysis (compliance testing) without the requirement 

to open the main package, thereby avoiding double handling and unnecessary 

exposure of loose waste at the landfill. 

5.4.11 Samples will be retained by the source site for 1 year after disposal of the 

package and then disposed to the landfill. 

5.4.12 Wastes arriving at the landfill will be subject to the following on site verification: 

The shipment will be checked while still on the vehicle against the prenotified 

characterisation information for consistency. 

The external dose rate at 1 metre will be checked. 

The packages will be visually checked for integrity. 

The transport documentation will be checked for compliance with the 

transport regulations. 

The characterisation documentation will be checked to ensure the waste 

has been pre-accepted and is compliant. 

Receipt records will be generated. 

The waste packages will not be opened or sampled at the landfill in order 

to minimise unnecessary exposure. 

5.5 Accumulation and Quarantine 


operating procedures and instructions for the process which will be prepared if 

authorisation is approved. It is noted that receipt of unacceptable waste 

packages is very unlikely because of the stringent pre-acceptance and transport 

arrangements applied to the wastes. The arrangements for quarantine are 

provided as a contingency measure. 

If a waste consignment fails to be acceptable upon receipt at the site 

entrance and can safely be returned to the consignor, it will be refused 

entry to the site. 




July 2009 

48 


If a waste consignment fails to be acceptable upon receipt and may not 

be safe to return to the consignor (for example a package has been 

damaged, or a dose rate exceeded) the landfill site operator will: 

Consult the consignor and available information to enable a safe 

response plan to be generated. 

Enact the contingency arrangements where required as outlined 

below. 

In cases where safe to do so, move the consignment and offload 

to a designated quarantine area for LLW. 

Inform the Environment Agency. 

For the unlikely case that these contingency measures are executed, the 

disposal contract between the landfill and the consignor will establish 

responsibility for remedial action. The consignor will take responsibility 

for remedial action in co-ordination with the landfill operator. The 

consignor will complete any necessary regulatory notifications, 

investigations, remedial action planning and remedial works. Such 

actions will be subject to specific safety planning prior to execution such 

that no significant risks are imposed on the landfill operators or public. 

The consignors of significant volumes of LLW are operators of nuclear 

licensed sites and will have considerable resources that they can bring 

into play in order to ensure effective remedial action. 

A designated quarantine area will be provided that is marked, physically 

demarcated off from the rest of the site and well segregated by distance 

from persons on the site. The design of the quarantine area will be part of 

the radiation protection plan for the site established to meet the 

requirements of the Ionising Radiations Regulations. For illustration, the 

quarantine area might consist of a lockable steel HISO freight container 

set to one side of the main traffic routes and suitably labelled. The 

quarantine area shall be so designed to ensure that the dose at the 

perimeter does not exceed 2 microSv/h. 

LLW will not be intentionally accumulated. Wastes received to the site 

will be placed in the landfill void and covered before the end of the 

working day. During the working day any accumulations of LLW required 

for operational reasons will be kept together in designated and marked 

locations that avoid human exposure through distance. If, for any 

exceptional reason, LLW cannot be disposed and covered on the day of 

arrival they will be stored in a designated and segregated area and the 

Environment Agency informed. The storage area shall be so designed to 

ensure that the dose at the perimeter does not exceed 2 microSv/h. 




July 2009 

49 


5.6 Disposal, Waste Emplacement, Compaction, Cover and 

Handling 



is approved. 

Waste will be disposed to the landfill void as soon as practicable after 

receipt. The waste will be moved to the landfill working face along roads 

made of suitable hardcore materials. 

The waste packages will be lifted using mechanical equipment with air 

conditioned cabins (P3 filter equivalent) and placed into the landfill at the 

base of the working face. 

Immediately after placement of the load the waste will be covered with at 

least a 300mmm thickness of suitable cover on all exposed surfaces. If 

the doserate at 1 metre above the emplaced and covered waste is greater 

than 2 microSv/hr further cover will be added until this doserate is 

achieved. 

A bowser to dowse the waste with water will be on standby wherever 

waste is unloaded in case of spillage. 

A record will be kept of the waste disposal location. 

Waste will be disposed of in the current working cell or cells and will be 

spread throughout the landfill void of that cell without deliberate 

concentration into one location. 

Waste will not be co-located with incompatible other wastes; for example, 

other wastes that could damage the package during emplacement. 

Traffic over existing wastes will be on suitable cover tracks in order to 

avoid vehicle penetration to the waste layer. 

The most likely point at which a load could be dropped or damaged would 

be during emplacement. Emergency procedures will be enacted to deal 

with a dropped load situation should this occur. 

No loose, unpackaged or exposed wastes will be handled under normal 

operating conditions. 

5.7 Worker Radiation Protection 



is approved. 




July 2009 

50 


5.7.2 The landfill operator will develop a written safe system of work (radiation 

protection plan) to implement controls and arrangements in accordance with the 

Ionising Radiations Regulations to ensure worker radiation protection. A plan 

has been prepared by the appointed Radiation Protection Advisor, the HPA 

(Annex C). 

5.7.3 The system of work will include: 

Arrangements for radiation protection for all operations that take place 

within the boundaries of the landfill site including transport, receipt, 

quarantine, accumulation, disposal and post-disposal operations. 

The system of work will aim to ensure that dose to individuals is optimised 

and below the dose limit and dose constraint. 

The system of work will be based on the principle that landfill operators 

are not specialist nuclear workers and that they should be protected to 

limits that would apply to members of the public during the operational 

phase. 

The system of work shall aim to achieve the ALARA principle regardless 

of limits and constraints that apply. 

Actual radiation exposures from direct radiation will be monitored and the 

systems subject to annual review and improvement. 

The system of work will be based upon a prior risk and dose assessment 

of the potential exposures which will consider routine and accident 

conditions. 

The system of work will document roles and responsibilities, dose 

assessment and optimisation, surface contamination assessment, 

segregation, other protective measures, emergency response 

arrangements, training provision, information provision, competency 

assessment and quality assurance. 

Arrangements for employing the services of a qualified Radiation 

Protection Adviser to assist with the establishment and operation of the 

system of work. 

Arrangements for monitoring packages and conveyances for radiation 

dose and arrangements for ensuring surface contamination clearance. 

Arrangements for workplace monitoring, where recommended by the 

radiological protection advisor. 

Arrangements for worker monitoring. 

Arrangements for recording and reporting exposures. 

Arrangements for limiting the external dose received by landfill site 

workers through primary limitations of the radiation dose waste 

acceptance criteria, coupled with time of exposure, distance from 




July 2009 

51 


package and additional shielding (for example soil cover on disposed 

wastes). 

5.7.4 The main source of radiation exposure to the landfill workers under normal 

conditions will be external radiation from the packages during handling and 

emplacement. The controls are: 

Regardless of the requirements of the transport regulations the maximum 

dose rate at a specified distance from the package will be limited to a 

value that ensures that given the likely number of shipments per year the 

dose constraint of 1 mSv/year will not be exceeded under routine 

operational conditions. 

The maximum concentration of radioactivity in the package is limited by 

the waste acceptance criteria and conditions of authorisation. 

The package is sealed. 

Packages are designed to withstand specified drop tests to withstand 

accidents during emplacement or are robust to the conditions of handling 

at the landfill. 

Incident procedures will be enacted to minimise exposure for accident 

conditions, such as dropped loads. 

The packages will be handled at a distance using mechanical equipment 

with air conditioned cabins (P3 filter equivalent). 

Accumulated packages will be set aside in a quarantined area that is 

segregated by distance and barriers. 

Packages will normally be emplaced in the disposal void immediately 

upon receipt. 

The working zone and face of the disposal area will be covered with an 

adequate covering of cover material (soils) after each emplacement 

operation or at the end of the working day in order to reduce external 

radiation dose to trivial levels. 

A bowser to dowse the waste with water will be on standby wherever 

waste is unloaded in case of spillage. 

There will be no loose handling or tipping of wastes. 

Packages will be placed in the disposal void and will not be tipped. 

Monitoring of the package, workplace and worker for external radiation 

will be carried out in accordance with an established plan commensurate 

with risks. 

Monitoring of working areas for surface contamination including 

occasional reassurance monitoring of, for example, wheel wash, traffic 

routes, gateways and change rooms will be undertaken in accordance 

with an established plan commensurate with risks. 




July 2009 

52 


Designation of areas (if applicable) under the Ionising Radiations 

Regulations. 

5.8 Environmental Radioactivity Monitoring 



is approved. The monitoring under the Environmental Permit is described in 

detail in a series of Monitoring and Action Plans (MAPs). The MAPs set out the 

parameters, frequencies, methodologies and reporting of monitoring for the 

landfill. Contingency action plans are included in the event that a limit specified 

in the Permit is exceeded. It is anticipated that radiological monitoring will be 

added to the MAPs. 

5.8.2 The following environmental radiological monitoring is proposed: 

Annual radiochemical analysis of groundwater to current monitoring 

schedules as described under the existing permit for several existing 

boreholes close to the site. Analysis would be for gamma spectrometry, 

gross alpha / beta in waters and 3 H in aqueous samples. 

Annual radiochemical analysis of leachate. Analysis would be for gamma 

spectrometry, gross alpha / beta in waters and 3 H in aqueous samples. 

Quarterly radiochemical analysis of surface water discharge. Analysis 

would be for gamma spectrometry, gross alpha/beta in waters and 3 H in 

aqueous samples. 

Annual radiochemical analysis of the landfill gas flare stack emission for 

the radioactive gases identified in the risk assessment. 

Quarterly radiochemical analysis for dust deposited on a powered static 

air sampler paper at one predominantly downwind location on the site 

boundary to include gamma spectrometry and gross alpha/beta. (This 

would be baselined against equivalent samples taken in the period prior 

to first waste receipt). 

Annual analysis of randomly selected surface soils from four points 

around the site boundary to include gamma spectrometry and gross 

alpha/beta. (This would be baselined against equivalent samples taken in 

the period prior to first waste receipt). 




July 2009 

53 


6.0 Waste Disposal History 

6.0.1 The original Permit for the site was issued in 2002. The Permit was reviewed in 

early 2009 and a new Permit issued in May 2009. Landfilling commenced in 

2002 in a series of 5 Phases (Figure 2). The site was originally operated as a codisposal 

facility in which Special Wastes were disposed of with non-hazardous 

biodegradable wastes. This principle of co-disposal relies on the biological 

activity in the biodegradable wastes to ameliorate the Special Wastes. Phases 1 

and 2 of the site were operated by co-disposal. The use of co-disposal ended 

with the implementation of the Landfill Regulations in 2004. Under the 2004 

Regulations the facility became a landfill site for the acceptance of hazardous 

wastes. Therefore Phases 3 onwards have received hazardous and inert wastes 

only. 

6.0.2 To date Phases 1, 2 and 3 have been completed capped and restored. Phase 4 

is operational and Phase 5 needs to be excavated and engineered (Figure 2). 

Approximately 650,000m 3 of waste has been disposed of and there is 

approximately 700,000m 3 of void remaining. 

6.0.3 Baseline samples of leachate and groundwater have been analysed for 

radioactivity and results are given in Annex I. The baseline sampling showed no 

enhancements of radioactivity in the groundwater samples. The leachate 

samples showed a slight enhancement for Tritium, which is commonly found in 

leachate from landfills across the UK. The levels of Tritium were lower than 

those usually found in landfill leachates (Annex I). 




July 2009 

54 


7.0 Proposals for Liquid and Gaseous Discharges 

7.0.1 The proposal involves no specific authorised liquid or gaseous discharge routes. 

Inadvertent discharge to the air from gas generation from the waste form has 

been included in the risk assessment which shows that the radioactive emissions 

will be negligible (Annex B). Discharge to groundwater has been included in the 

risk assessment which shows that the potential emissions are very low and will 

not result in exceedance of relevant constraint limits (Annex B). 

7.0.2 A specific risk assessment has been provided for any radioactive exposure or 

environmental impact arising from leachate management practices (Annex B). 

Leachate is currently treated offsite. It is not reasonable to model the impact 

from leachate using the same conservative assumptions concerning leachate 

activity levels that are used for the groundwater modelling. The optimal approach 

for controlling any impact that may arise from leachate management is to place 

an authorisation limit on leachate activity levels such that the consequent impact 

is insignificant. Should the leachate ultimately have higher activity levels than the 

authorisation limit, then an alternative treatment would be considered which could 

include a detailed risk assessment and revised limit (see 8.6). 




July 2009 

55 


8.0 Radioactive Waste Disposal Consequence 

Assessment and Radiological Capacity 

8.0.1 To determine the amount of radiological material that can be disposed of within 

the site without exceeding the proposed dose criteria a series of consequence 

and risk assessments have been undertaken. The consequence and risk 

assessment for the disposal of LLW at the East Northants Resource 

Management Facility is included in Annex B. Supplementary information on the 

assessment methodology is included in Annex E. 

8.0.2 An additional assessment of direct radiation dose to workers is included in 

Annexes D and H. 

8.0.3 The primary assessment has been carried out using the SNIFFER 2006 model 

for assessing the suitability of controlled landfills to accept disposals of solid low 

level radioactive waste. A range of 43 nuclides has been considered with their 

associated daughters. The methodology for the other supplementary 

assessments is described in the Annexes. 

8.0.4 A summary of the scenarios considered follows: 

Scenario Annual Dose Criteria 

Used for Assessment 

Pre-Closure – expected to occur 

Direct Radiation 20 mSv/yr Worker 

Exposure from (Ionising Radiation 

Waste Handling and Radiations) 

Emplacement 1 mSv/yr Worker 

(Operational Criterion) 

0.02 mSv/yr Public 

(GRA)- Lower Bound 

0.3 mSv/yr Public (GRA) 

– Upper Bound 

Exposure from Gas 20 mSv/yr Worker 

Generation from the (Ionising Radiation 

Landfill 

Radiations) 

1 mSv/yr Worker 






Pre-Closure – not expected to occur 

Dropped Load of 20 mSv/yr Worker 

Waste (and 

(Ionising Radiation 

hypothetical aircraft Radiations) 

impact ) 







Public Worker Assessment Capacity 

Constraint? 

8.1 

Annex D and H 

8.2 

Annex B (5.5) 

8.3 

Annex C 

Not used to 

define landfill 

capacity 

Considered as a 

constraint to 

landfill capacity 

Not used to 


capacity 

July 2009 

56 







3 mSv/yr Public for 

aircraft intrusion 

Wound Exposure 20 mSv/yr Worker 


Radiations) 



Exposure from Fire See discussion at 8.5 

Pre Closure and Aftercare Period – expected to occur 

Exposure from 

Leachate Processing 

Offsite – Sewage 

Works 



Radiations) 







Pre Closure and Aftercare Period – not certain to occur 

Exposure from 

Leachate - Spillage 

Exposure from 

Leachate - Aerosols 



Radiations) 









Radiations) 







Post-Closure – expected to occur 

Exposure by Using 

Groundwater at 

Nearest Abstraction 

Point 




Public Worker Assessment Capacity 

Constraint? 

x 8.4 

Annex C 

Not assessed 

Discussed in 

Annex B 

8.6 

Annex B 

8.7 

Annex B 

8.8 

Annex B 

0.02 mSv/yr Public (GRA) x 8.9 

Annex B 

Not used to 


capacity 

Not used to 


capacity 

Not Considered 

as a constraint 

to landfill 

capacity 

Used to set 

authorisation 

conditions for 

leachate 

discharge 

Not used as a 

constraint on 


because worst 

case constraint 

is larger than the 

physical landfill 

Not used as a 

constraint on 


or leachate 

discharge 

concentration 

because the 

case in section 

8.6 is more 

constraining 


potential 

constraint to 


July 2009 

57 


Scenario Annual Dose Criteria Public Worker Assessment Capacity 


Constraint? 

Exposure from Gas 0.02 mSv/yr Public (GRA) x 8.10 


Generation from the 

Annex B potential 

Landfill 

constraint to 


Exposure to Wildlife 10 microgray/hr x x 8.11 

Not used to 

from all sources 

Annex B define landfill 

capacity 

External dose from 0.02 mSv Public (GRA) x 8.12 

Not considered 

emplaced wastes 

Annex B as a constraint 

to landfill 

capacity 

because the 

resulting doses 

are trivial 

Post –Closure not expected to occur 

Exposure by Using 3 mSv Public (GRA and x 8.13 

Has the potential 

Groundwater from a HPA) 

Annex B to be a 

Borehole 

constraint to 

Constructed at the 

Boundary of the 

Landfill 


Exposure by 3 mSv Public or Worker 8.14 

Has the potential 

Intrusion into the (GRA and HPA) 

Annex B to be a 

Emplaced Waste 

constraint to 

Post Closure of the 

Landfill 


Other potential scenarios are discussed in Annex B and have not been assessed for documented reasons. 

8.0.5 Some other potential scenarios are discussed in Annex B and have not been 

assessed for documented reasons. Additionally: 

- The aircraft crash scenario is discussed with the dropped load scenario. 

- A scenario involving drilling into the waste form for construction of new 

sampling or leachate wells is not discussed because this would be 

executed with knowledge under appropriate regulations with appropriate 

precautions as necessary. 

- The effects of very long term climate change are not assessed because 

the site is already permitted as a hazardous site and LLW disposal gives 

rise to no additional considerations in respect of flooding, coastal erosion 

or sea level rises. Future glaciation would have similar or lesser effects 

than the “residential intrusion scenario” considered in 8.14. 

- The effects of seismic events. The engineered containment structures at 

the site are not formed of brittle materials such as concrete that may 

fracture as a result of a severe earthquake. The HDPE and clay lining 

materials have a high shear strength and have the flexibility to withstand 




July 2009 

58 


the stresses which would be imposed during the types of earthquake 

which occur in the UK. 

- Transport accident scenarios are not discussed because transport is 

outside of the scope of the authorisation and is regulated under an 

existing regime of Dangerous Goods Regulations. Transport accidents 

on the site are considered as part of the dropped load scenario and a 

transport accident involving leachate is specifically considered. 

8.06 The full set of results is given in the Annexes and the following conclusions are 

reached from the assessment: 


Direct Radiation Exposure from Waste Handling and 

Emplacement 

8.1.1 The dose criteria are the legal limit to workers of 20 mSv/yr, the site criterion of 1 

mSv/yr for workers, the dose guidance level of 0.02 mSv/yr for the public and the 

dose constraint for the public of 0.3 mSv/yr. 

8.1.2 The impacted group is landfill workers and the public near to the site. The 

emplaced waste can only affect the landfill workers because there is no line of 

sight for direct radiation from landfill void. 

8.1.3 The assessment is contained within Annexes D and H. 

8.1.4 Waste Emplacement: The scenario is the external radiation exposure of workers 

in the vicinity of the waste emplaced in the landfill after it has been covered. 

8.1.5 The assessment is contained within Annex H. 

8.1.6 Annex H illustrates the dose rate for varying cover thicknesses using two 

illustrative cases, one of which is a worst case. The advice of the radiation 

protection advisor is that the maximum radiation dose 1 m above the covered 

waste should be less than 2 microSv/hr in order to ensure the occupational dose 

is considerably less than the dose criterion of 1 mSv/yr. 

8.1.7 Annex H demonstrates that for most cases a 300mm thick cover layer will more 

than achieve the dose rate. For the worst case of waste containing Co-60 at 200 

Bq/g a cover layer of 700mm would be required to achieve the dose rate, but this 

is exceptional. 

8.1.8 The proposed authorisation condition is that a minimum cover layer of 300mm be 

utilised and that if the dose rate 1 m above the waste is still greater than 2 

microSv/hr then further cover will be added in order to achieve the dose rate. 

The minimum cover layer of 300mm is adequate to ensure daily physical 

protection of the waste. 




July 2009 

59 


8.1.9 Additional ALARA precautions are that all wastes are handled by machines and 

operatives generally do not enter the operational area on foot. On most days the 

only reason to enter the operational area on foot is for final inspection at the end 

of the day and health physics monitoring. Workplace monitoring will confirm 

actual doses and enable dose limitation to be managed. 

8.1.10 The original SNIFFER model uses occupational external dose as a constraint to 

set the radiological capacity of the landfill but since this dose is specific to 

workers during the operational phase and can be managed through occupational 

radiation dose protection practices this is not considered necessary. Hence the 

external dose assessment for waste emplacement has not been used to 

constrain the overall radiological capacity. 

8.1.11 Waste Handling: The scenario is the external radiation exposure to workers from 

their occupancy near to a waste package prior to disposal. 

8.1.12 The SNIFFER model does not include this scenario and it has been assessed in 

Annex D. Annex D considers the external radiation dose for a series of cases 

and package types. The hypothetical worst case is identified to be a waste 

flexible type container with 200 Bq/g of Co-60. A flexible container carrying Co- 

60 at 200 Bq/g is an unlikely case and another case is included in Annex D to 

illustrate more typical exposures. 

8.1.13 The hypothetical worst case dose identified in Annex D is 14.5 microSv/hr at 1 

metre from the package face. However the radiation protection advisor has 

advised that the maximum dose at 1 metre from a package should be less than 

10 microSv/hr in order to ensure the occupational dose is considerably less than 

the dose criterion of 1 mSv/yr. Thus 10 microSv/hr will be used as an 

acceptance criterion and constrains the contents of the package to this limit. 

8.1.14 The proposed authorisation condition is that the dose at 1 metre from the 

package face must be less than 10 microSv/hr. This would be measured by the 

consignor prior to sending the package and would be checked by the consignee 

upon arrival of the package. 

8.1.15 Additional ALARA precautions are that dose can be measured directly and 

managed actively to prevent unnecessary exposure. As illustrated in Annex D the 

field dose drops quickly with distance from the package and hence the simple 

precaution of managing occupancy time and distance is practicable. 

8.1.16 This dose is specific to workers during the operational phase and can be 

managed through occupational radiation dose protection practices, hence it is not 

used to constrain overall radiological capacity. 

8.1.17 There is an additional scenario that a member of the public stands at a distance 

in direct line of sight of a waste package/shipment and hence receives direct 

radiation exposure. This can be estimated by considering the waste as a single 




July 2009 

60 


point source with a 10 microSv/hr doserate at 1 metre, assuming that the 

member of the public is located 50 metres from the waste. The doserate at 50 

metres can be estimated from: 

D1=D2 (X2 2 /X1 2 ) 

Where, D1 and D2 = doserate at positions 1 and 2 

X1 and X2 = the distance from the source at positions 1 and 2 

This gives an estimated maximum doserate at 50 metres of 4E-3 microSv/hr. If 

the person stands in that location for 8 hours per day and there is waste at the 

maximum level in that location every day then the person would receive 12 

microSv per year. This is based on conservative assumptions and is within the 

20 microSv per year lower bound dose criterion. 

Under the same assumptions but with a 100 metre distance to the person, the 

maximum estimated dose would be 3 microSv/yr. 

These calculations do not take into account the significant shielding afforded by 

the soil screen bund at the boundary of the site. 






8.2.2 The impacted groups during the pre-closure phase are the public and workers. 

8.2.3 The assessment is contained within Annex B, section 5.5. 

8.2.4 The scenario is radioactive gas release from the landfill in the pre-closure phase. 

Annex B indicates that the worst case is for Ra-226 in both the case of the 

worker and the public. For this worst case and using a worker dose criterion of 1 

mSv/yr the capacity of the landfill to take only Ra-226 would be 8.4E9 MBq (42 

million tonnes at 200 Bq/g). For this worst case and using a public dose criterion 

of 0.02 mSv/yr the capacity of the landfill to take only Ra-226 would be 81E6 

MBq (403 thousand tonnes at 200 Bq/g). 

8.2.5 This scenario has the potential to constrain landfill capacity for the public 

exposure case but not for the worker case (because the landfill is physically 

smaller than the radiological capacity). 




July 2009 

61 



Dropped Load of Waste 

Dropped Load 





8.3.3 The scenario is not contained within the SNIFFER model and has been 

separately addressed in Annex C, which is a radiological risk assessment for 

occupational exposure completed by the HPA. 

8.3.4 The conclusion is that with appropriate precautions the worker exposure can be 

kept with the site criterion under the unlikely circumstance of a dropped container 

which gave rise to a release. 

8.3.4 This scenario is not used to constrain landfill capacity. 

8.3.5 To augment the calculations in Annex C the following table gives exposure to 

both workers and the public under the following assumptions using the UKAEA 

dropped load methodology from the safety assessment handbook (ref 22). 

8.3.6 The assumptions are: 

- A one cubic metre flexible container of wastes is dropped and spills 10% 

of its contents through broken seams. 

- The bag is filled with a dry solid. 

- The bag contains the maximum concentration of a single nuclide at 200 

Bg/g. 

- The bag weighs 1 tonne. 

- The distance to the nearest public is 50m and the event duration is 30 

minutes. 

- The worker remains very close to the dropped waste without taking 

precautions or retreating for at least 30 minutes. 

- The atmospheric conditions are worst case, still conditions. 




July 2009 

62 


8.3.7 Dose from inhaling material discharged from a dropped container is given by: 

Dose 

inh, 

bag 

I RF1 

RF2 

C B 

D 

 

DF 

where I is the inventory of radionuclide Rn releasable (Bq) 

RF1 is the release fraction (-) 

RF2 is the respirable fraction (-) 

C is the dispersion coefficient (s m -3 ). 

B is the breathing rate (m 3 s -1 ) 

DF is the decontamination factor (-) 

Dinh is the dose coefficient for inhalation of radionuclide, 

Rn (Sv Bq -1 ). 

Parameter Description Value Units 

inventory of radionuclide in the bag 200E6 Bq 

I inventory of radionuclide Rn 

releasable 

20E6 Bq 

RF1 release fraction 1E-3 - 

RF2 respirable fraction 0.1 - 

C dispersion Worker 5 

s m 

coefficient Public 1.7E-2 

-3 

B breathing rate 3.3E-4 m 3 s -1 

DF decontamination factor 1 - 




Rn 

inh 

July 2009 

63 


Doses from Dropped Load Scenario 

Inhalation 

Dose Coefficent Worker Dose 

Radionuclide (Sv/Bq) 

(microSv) Public Dose (microSv) 

H-3 2.60E-10 8.58E-04 2.92E-06 

C-14 5.80E-09 1.91E-02 6.51E-05 

Cl-36 7.30E-09 2.41E-02 8.19E-05 

Fe-55 7.70E-10 2.54E-03 8.64E-06 

Co-60 3.10E-08 1.02E-01 3.48E-04 

Ni-63 4.80E-10 1.58E-03 5.39E-06 

Sr-90 1.62E-07 5.35E-01 1.82E-03 

Nb-94 1.10E-08 3.63E-02 1.23E-04 

Tc-99 1.30E-08 4.29E-02 1.46E-04 

Ru-106 6.60E-08 2.18E-01 7.41E-04 

Ag-108m 3.70E-08 1.22E-01 4.15E-04 

Sb-125 5.46E-09 1.80E-02 6.13E-05 

Sn-126 3.12E-08 1.03E-01 3.50E-04 

I-129 3.60E-08 1.19E-01 4.04E-04 

Ba-133 3.10E-09 1.02E-02 3.48E-05 

Cs-134 6.80E-09 2.24E-02 7.63E-05 

Cs-137 3.90E-08 1.29E-01 4.38E-04 

Pm-147 5.00E-09 1.65E-02 5.61E-05 

Eu-152 4.20E-08 1.39E-01 4.71E-04 

Eu-154 5.30E-08 1.75E-01 5.95E-04 

Eu-155 6.90E-09 2.28E-02 7.74E-05 

Pb-210 9.99E-06 3.30E+01 1.12E-01 

Ra-226 1.95E-05 6.44E+01 2.19E-01 

Ac-227 5.69E-04 1.88E+03 6.38E+00 

Th-229 2.56E-04 8.45E+02 2.87E+00 

Th-230 1.00E-04 3.30E+02 1.12E+00 

Th-232 1.70E-04 5.61E+02 1.91E+00 

Pa-231 1.40E-04 4.62E+02 1.57E+00 

U-232 4.69E-05 1.55E+02 5.26E-01 

U-233 9.60E-06 3.17E+01 1.08E-01 

U-234 9.40E-06 3.10E+01 1.05E-01 

U-235 8.50E-06 2.81E+01 9.54E-02 

U-236 3.20E-06 1.06E+01 3.59E-02 

U-238 8.01E-06 2.64E+01 8.99E-02 

Np-237 5.00E-05 1.65E+02 5.61E-01 

Pu-238 1.10E-04 3.63E+02 1.23E+00 

Pu-239 1.20E-04 3.96E+02 1.35E+00 

Pu-240 1.20E-04 3.96E+02 1.35E+00 

Pu-241 2.30E-06 7.59E+00 2.58E-02 

Pu-242 1.10E-04 3.63E+02 1.23E+00 

Am-241 9.60E-05 3.17E+02 1.08E+00 

Cm-243 3.11E-05 1.03E+02 3.49E-01 

Cm-244 2.71E-05 8.94E+01 3.04E-01 

8.3.8 Only in the case of Ac-227 does the assessment fail to meet the site criterion for 

workers. Ac-227 is very unlikely to be present at 200 Bg/g given the low 




July 2009 

64 


occurrence of this nuclide. The above calculations assume that the bag is filled 

with a loose dry material that disperses readily, that the package fails and that 

the worker does not respond correctly. These are highly conservative 

assumptions, especially given the operational precautions proposed in Section 

5.6. 

8.3.9 A key measure to mitigate dropped load dispersion events will be to engineer the 

waste containers such that they withstand or substantially withstand accidental 

drops during handling. Where drums are used these will be rated under existing 

dangerous good transport regulations for radioactive material to withstand a drop 

test. Flexible containers may only be used where this is acceptable under 

dangerous goods transport regulations and these regulations specify isotope 

specific limits designed to ensure public safety. 

8.3.10 The dropped bag scenario is not used to establish radiological capacity of the 

landfill because it is independent of the total tonnage received. 

Aircraft Impact 

8.3.11 The event could be considered an intrusion in which case the 3-20 mSv/yr dose 

criteria would apply. 


The event is assessed for the pre-closure phase but could also apply to the postclosure 

phase for the public if the landfill closure cap (at least 1.5m thick) did not 

provide full protection from the impact. 


separately addressed below. 

8.3.14 This scenario is not used to constrain landfill capacity because it is independent 

of tonnage received. The scenario has a very low probability of occurrence. 

8.3.15 The following gives exposure to both workers and the public under the following 

assumptions using the UKAEA release methodology from the safety assessment 

handbook (ref 22). The approach used is to assume an amount of material is 

physically displaced by crater formation through impact of a high velocity military 

aircraft. This is considered a reasonable scenario given the presence of an RAF 

base close to the landfill when compared to much less likely scenarios involving 

heavy civilian aircraft. 

8.3.16 Due to the complexity of such an event this assessment can only be considered 

as a scoping calculation based on conservative assumptions. 

8.3.17 The assumptions are: 

- The aircraft hits an area of exposed waste and forms a crater. 




July 2009 

65 


- The crater size can be estimated from theoretical models for estimated 

impact parameters such as densities, impact velocity, impact angle, 

missile dimensions and target density/type (Ref 21). Scoping calculations 

indicate that crater sizes of 300 cum are conceivable. Actual crater sizes 

from impacts due to Harrier jets (the type of aircraft currently based at 

RAF Wittering) reveal a wide variation from virtually no displacement to 

significant craters dependent on the nature of the event. A record (ref 23) 

notes a Harrier jet impact forming a crater of approximately 300 cum. For 

comparison, the Lockerbie B747 impact formed a crater of 560 cum (ref 

24). 

- The displaced waste contains the maximum concentration of a single 

nuclide at 200 Bg/g. The chosen nuclide is Pu-239 which has a 

conservative inhalation dose coefficient and yet is a credible nuclide to 

occur at the concentration assumed. 

- The density of the displaced waste is 1.5 t/cum. 300 cum or 450 tonnes 

are displaced. Giving rise to displacement of 90,000 MBq. 

- The distance to the nearest public is 200m and the event has 30 minute 

release duration. This is on the basis that immediate evacuation of the 

near zone would occur from such an extreme event and within the very 

near zone immediate fatality due to impact would be likely. 

- The effect of fire on dispersal is not included (refer Section 8.5).. 

- The worker exposure is the same as the public exposure because 

workers would evacuate quickly to the same distance. 

- The atmospheric conditions are worst case still conditions and mixing is 

not assumed to be enhanced by fire. 

8.3.18 Dose from inhaling material discharged from displaced material: 

Dose 

inh, 

bag 

I RF1 

RF2 

C 

B 

D 

 

DF 

where I is the inventory of radionuclide Rn releasable (Bq) 

RF1 is the release fraction (-) 

RF2 is the respirable fraction (-) 

C is the dispersion coefficient (s m -3 ). 

B is the breathing rate (m 3 s -1 ) 

DF is the decontamination factor (-) 

Dinh is the dose coefficient for inhalation of radionuclide, 

Rn (Sv Bq -1 ). 




Rn 

inh 

July 2009 

66 



inventory of radionuclide 90,000E6 Bq 

RF1 release fraction 1E-3 - 

RF2 respirable fraction 0.1 - 

C dispersion Worker 1.5 E-3 s m -3 

coefficient Public 1.5 E-3 

B breathing rate 3.3E-4 m 3 s -1 

DF decontamination factor 1 - 

8.3.19 The resulting dose would be approximately 0.5 mSv. Such a calculation could 

have a relatively wide range of uncertainty, but this conservative scoping 

estimate indicates that public and worker legal dose limits would not be exceeded 

and the 3 mSv/yr intrusion dose limit would not be exceeded by this low 

probability extreme event. 


Wound Exposure 

8.4.1 The dose criteria are the legal limit to workers of 20 mSv/yr and the site criterion 

of 1 mSv/yr for workers. 

8.4.2 The impacted groups during the pre-closure phase are the landfill site workers. 


separately addressed in Annex C, which is a radiological risk assessment for 

occupational exposure completed by the HPA. 

8.4.4 The conclusion is that wound exposures are unlikely and can be further reduced 

in likelihood and impact through simple precautions. It is very likely this will be 

effective in maintaining individual exposures within the site criterion. 

8.4.5 This scenario is not used to constrain landfill capacity 




July 2009 

67 



Exposure from Fire 

8.5.1 The dose criteria are not defined within the guidance on requirements for 

authorisation for this scenario but could be taken to be the worker dose target of 

1 mSv/yr and the public dose constraint of 0.3 mSv/yr. 


8.5.3 The scenario is transient and for practical purposes can only occur whilst the 

wastes are not covered with the final capping layer. The lack of biodegradable 

wastes makes fires very unlikely after the cap is in place. 

8.5.4 Furthermore, some of the exposure pathways considered in the SNIFFER fire 

model would not arise in practice because intervention would occur. So for 

example exposures from any subsequent deposition would be controlled by 

remedial activities. 

8.5.5 Although an aircraft crash could lead to a fire, the fire would mostly consume 

aircraft fuel and wreckage. The main feature of an aircraft impact which could 

lead to exposure would be the physical displacement of material and this is 

considered in Section 8.3. 

8.5.6 The waste in the landfill, the cover materials and the LLW are essentially 

incombustible. The current waste acceptance criterion for the landfill largely 

excludes organic material and includes a flammability test. As such it is difficult 

to conceive that the fire scenario included in the SNIFFER model can occur for 

this type of landfill and it has not been utilised to constrain landfill capacity. 

8.6 Pre Closure and Aftercare Period – expected to occur 

Exposure from Leachate Processing Offsite – Sewage Works 




8.6.2 The impacted groups are sewage workers and the public impacted by the 

sewage works. 

8.6.3 The scenario is addressed in Annex B. 

8.6.4 Leachate levels at the ENRMF are maintained by pumping excess leachate to 

tankers and transporting this leachate to a water treatment plant at Avonmouth. 

An initial assessment of the potential impacts from routine, off-site leachate 

management has been made using the Environment Agency’s methodology and 




July 2009 

68 


the assumption that doses from water treatment would be similar to doses from 

sewage treatment. The Environment Agency’s methodology allows for a range of 

exposure groups affected by releases to a public sewer, depending on the 

discharge route for treated effluent. For this assessment, only the groups 

associated directly with operation of the treatment plant, farming of land 

conditioned by sludge or using the estuary are considered. These groups and 

the relevant exposure pathways are: 

Sewage treatment workers (adults only) 

External irradiation from radionuclides in raw sewage and sludge 

Inadvertent inhalation and ingestion of raw sewage and sludge containing 

radionuclides 

Farming family living on land conditioned with sewage sludge 

Consumption of food produced on land conditioned with sludge and 

incorporating radionuclides 

External irradiation from radionuclides in sludge conditioned soil 

Inadvertent inhalation and ingestion of sludge conditioned soil 

Fisherman’s family (estuary/coastal water receives treated effluent from sewage 

works, typically via a river) 

External irradiation from radionuclides deposited in sediments 

Consumption of fish incorporating radionuclides 

8.6.5 The results in Annex B are expressed as specific dose per MBq/year of activity in 

the leachate treated. The groundwater assessment uses a very pessimistic 

assumption regarding leachate concentrations which it is not appropriate to use 

in this assessment. There is no empirical evidence on which to base leachate 

activity concentrations. 

8.6.6 The worst case result is for Th-232 with the “farming family” public exposure 

group. If we assume a dose constraint of 0.3 mSv/yr for the public during the 

operational phase, the results indicate that the maximum allowable leachate 

discharge per year is 216 MBq if the leachate only comprised Th-232. 

8.6.7 The proposed approach is that this scenario is not used to constrain radiological 

capacity, but that it is used to derive authorisation discharge limits for the 

leachate which can then be subsequently refined when empirical monitoring 

results become available. Based upon the above approach the authorisation 

limits for individual nuclides if the leachate comprised 100% of that nuclide, with a 

0.3 mSv/y dose criterion are given in the following table. In practice authorisation 

discharge limits will be set after discussion with the Environment Agency and 

then optimized through operational experience such that restrictive dose criterion 

are met. 




July 2009 

69 


Nuclide 

Maximum Leachate Authorisation Limits for Individual Nuclides 

Farming family Specific dose 

(microSv / y per MBq / y) 




Leachate Authorisation Limit 

MBq/yr 

H-3 2.83E-06 106E+6 

C-14 4.72E-03 63E+3 

Cl-36 7.78E-02 3.9E+3 

Fe-55 1.33E-03 225E+3 

Co-60 7.78E-01 385 

Sr-90 2.17E-02 14E+3 

Tc-99 2.83E-01 1060 

Ru-106 3.06E-03 98E+3 

I-129 6.11E-02 4909 

Cs-134 1.17E-01 2564 

Cs-137 1.00E-01 3000 

Pm-147 1.67E-05 18E+6 

Eu-152 2.67E-01 1123 

Eu-154 2.72E-01 1102 

Eu-155 5.06E-03 59E+3 

Pb-210 5.33E-01 562 

Ra-226 5.56E-01 540 

Th-230 1.28E-02 23E+3 

Th-232 1.39E+00 215 

U-234 1.17E-03 256E+3 

U-235 7.78E-03 38E+3 

U-238 2.06E-03 145E+3 

Np-237 7.22E-02 4155 

Pu-238 2.00E-02 15E+3 

Pu-239 2.28E-02 13E+3 

Pu-240 2.28E-02 13E+3 

Pu-241 3.39E-04 885E+3 

Pu-242 2.22E-02 13E+3 

Am-241 3.94E-02 7614 

Cm-243 6.67E-02 4497 

Cm-244 1.78E-02 16E+3 

July 2009 

70 



Exposure from Leachate - Spillage 




8.7.2 Notwithstanding any radioactive components, landfill leachate poses a hazard to 

the environment if spilt and any road accident involving loss of an entire load 

would be subject to mitigation measures. Leachate that did enter water 

resources would also become diluted. For this assessment, it is conservatively 

assumed that an entire tanker load of leachate (30 m 3 of leachate) reaches a 

small reservoir (2 x 10 6 m 3 ) that is used for drinking water, irrigation and fishing. 

8.7.3 The scenario is addressed in Annex B, table 5.5. 

8.7.4 The worst case is if the leachate comprises only Ra-226. The public dose 

constraint of 0.3 mSv can be used because this event is low probability and 

clean-up actions would in reality be taken to largely mitigate the event altogether. 

The resulting radiological capacity for the crops exposure case is then 71E6 MBq 

or 356,000 tonnes at 200 Bq/g. Given the mitigation measures noted above this 

scenario does not constrain radiological capacity. 


Exposure from Aerosols 




8.8.2 There is potential, during leachate management or spillage, for the production of 

aerosols which could lead to doses via the inhalation pathway. 

8.8.3 The assessment is presented in Annex B, table 5.6. The results are presented in 

terms of specific dose (μSv y -1 per MBq per hour). The worst case result is if the 

leachate aerosol comprises only Ac-227, for public exposure. Assuming 1600 

hours exposure per year and a dose guidance level of 0.02 mSv/yr, the resulting 

leachate discharge limit would be 7002 MBq/yr. This case results in levels which 

are orders of magnitude higher than the case considered in section 8.6 above 

and hence this scenario is not a constraint on either radiological capacity or 

leachate discharge concentration. 




July 2009 

71 



Exposure by Using Groundwater at Nearest Abstraction Point 

8.9.1 The dose criterion is the dose guidance level of 0.02 mSv/yr for the public. 

8.9.2 The scenario only impacts members of the public. 

8.9.3 The scenario is assessed in Annex B, table 5.1 (1500m irrigation case). 

8.9.4 If a well or river is used for irrigation, then doses can result from ingestion of 

foodstuffs raised on contaminated soil, inhalation of dust from the soil, and 

external exposure to the soil. Drinking of contaminated water from a well or river 

is also a potential exposure pathway. If contaminated groundwater discharges to 

surface water (spring, river, sea), then ingestion of foodstuffs from the surface 

water is a potential exposure pathway. This scenario considers such exposure 

from the nearest abstraction point. 

8.9.5 The following table shows the results of the assessment on the assumption of 

using the public dose guidance level of 0.02 mSv/yr and the tonnage capacities 

assuming 200 Bq/g of each nuclide. These capacities are for the case where all 

of the waste in the landfill is comprised of the single nuclide at the maximum 

concentration. The information has been sorted is order of ascending capacity 

with the most restricted capacity nuclides at the top of the table. 

8.9.6 With the exception of I-129 and Np-237 the tonnage capacities are larger than 

the physical size of the landfill void. This scenario has the potential to constrain 

landfill capacity. 

Groundwater Pathway, Borehole, 1500m Irrigation Case (Table 5.1 Annex B) 

Radionuclide 

Specific Dose 

(microSv/yr per 

MBq) 




Radiological Capacity (MBq) 

(0.02 mSv/yr dose criterion) 

Tonnage Capacity 

(tonnes) 

(@200 Bq/g 

concentration) 

I-129 1.38E-05 1.45E+06 7.25E+03 

Np-237 1.22E-06 1.64E+07 8.20E+04 

Cl-36 6.52E-08 3.07E+08 1.53E+06 

Th-232 4.04E-08 4.95E+08 2.48E+06 

Pa-231 3.60E-08 5.56E+08 2.78E+06 

Pu-242 1.69E-08 1.18E+09 5.92E+06 

Ra-226 1.56E-08 1.28E+09 6.41E+06 

Pu-239 1.54E-08 1.30E+09 6.49E+06 

Th-229 1.44E-08 1.39E+09 6.94E+06 

Sn-126 1.20E-08 1.67E+09 8.33E+06 

July 2009 

72 


Groundwater Pathway, Borehole, 1500m Irrigation Case (Table 5.1 Annex B) 

Radionuclide 

Specific Dose 

(microSv/yr per 

MBq) 




Radiological Capacity (MBq) 

(0.02 mSv/yr dose criterion) 


(tonnes) 

(@200 Bq/g 


Pu-240 1.04E-08 1.92E+09 9.62E+06 

Th-230 8.24E-09 2.43E+09 1.21E+07 

U-233 4.62E-09 4.33E+09 2.16E+07 

U-234 4.35E-09 4.60E+09 2.30E+07 

U-235 4.35E-09 4.60E+09 2.30E+07 

U-238 4.35E-09 4.60E+09 2.30E+07 

U-236 4.22E-09 4.74E+09 2.37E+07 

Tc-99 2.12E-09 9.43E+09 4.72E+07 

C-14 2.06E-09 9.71E+09 4.85E+07 

Cm-244 1.29E-09 1.55E+10 7.75E+07 

Nb-94 3.76E-10 5.32E+10 2.66E+08 

Cm-243 1.82E-10 1.10E+11 5.49E+08 

Pu-241 1.59E-10 1.26E+11 6.29E+08 

Am-241 5.37E-11 3.72E+11 1.86E+09 

Pu-238 1.86E-12 1.08E+13 5.38E+10 

Ag-108m 1.68E-17 1.19E+18 5.95E+15 

U-232 5.07E-20 3.94E+20 1.97E+18 

Ni-63 7.94E-21 2.52E+21 1.26E+19 

Sr-90 3.04E-24 6.58E+24 3.29E+22 

Cs-137 1.28E-25 1.56E+26 7.81E+23 

Pb-210 1.25E-25 1.60E+26 8.00E+23 

Ac-227 5.66E-26 3.53E+26 1.77E+24 

H-3 3.66E-30 5.46E+30 2.73E+28 

Ba-133 1.44E-31 1.39E+32 6.94E+29 

Eu-152 2.77E-32 7.22E+32 3.61E+30 

Eu-154 3.28E-35 6.10E+35 3.05E+33 

Co-60 1.21E-39 1.65E+40 8.26E+37 

Eu-155 3.63E-41 5.51E+41 2.75E+39 

Sb-125 4.81E-42 4.16E+42 2.08E+40 

Cs-134 1.82E-43 1.10E+44 5.49E+41 

Fe-55 1.04E-43 1.92E+44 9.62E+41 

Pm-147 3.41E-44 5.87E+44 2.93E+42 

Ru-106 3.44E-46 5.81E+46 2.91E+44 

July 2009 

73 





8.10.2 The scenario is relevant to members of the public (post closure workers are 

treated as members of the public in this case). 

8.10.3 The scenario is assessed in Annex B, table 5.7, the “resident after closure case”. 

8.10.4 The scenario is the release of radioactive gas in the post-closure phase. The 

results indicate that the worst case is for Ra-226 which using the 0.02 mSv/yr 

dose criterion gives a capacity of 7E6 MBq or 35,000 tonnes at 200 Bq/g. 

8.10.5 This scenario has the potential to constrain landfill capacity. 


Exposure to Wildlife from all sources 

8.11.1 A dose criterion for screening purposes of 10 microGy/hr has been used in 

accordance with the ERICA tool. 

8.11.2 The scenario relates to a representative range of organisms and wildlife groups. 

8.11.3 The scenario is the release of radionuclides into the environment. A set of 

pessimistic assumptions have been used for a hypothetical release. 

8.11.4 The scenario is described and assessed in Annex B, section 5.7. Tables 5.10, 

5.11 and 5.12 give the results. 

8.11.5 The conclusion is that the exposure to the range of organisms and wildlife groups 

is below the screening dose criteria and therefore does not need further 

assessment. This scenario does not constrain landfill capacity 


External dose from emplaced wastes 


8.12.2 The scenario applies only to members of the public post-closure. 

8.12.3 The scenario is that persons walking on the closed waste site will experience 

direct radiation exposure through the cover materials. The scenario is included in 

the SNIFFER model and is assessed in Annex B, table 5.2, “Public dose 1.5m 




July 2009 

74 


cap”. The assumption is that the waste is shielded by a 1.5m thick cap, although 

in practice this is likely to be a conservative assumption. 

8.12.4 The results shows that the resulting doses from a worst case where the landfill is 

filled with the worst case nuclide at the maximum concentration, are far below the 

dose criterion. This case does not constrain landfill capacity. 


Exposure by Using Groundwater from a Borehole Constructed 

at the Boundary of the Landfill 

8.13.1 The dose criterion is the lower dose guidance level of 3 mSv/yr for the public for 

an intrusion scenario. 

8.13.2 The scenario applies only to members of the public post-closure and is unlikely to 

occur given the presence of the hazardous landfill. 

8.13.3 The scenario is that a new groundwater abstraction point is licensed at the 

boundary of the landfill site. The scenario is assessed in Annex B, table 5.1”Site 

boundary drinking”. 

8.13.4 The resulting radiological capacities and tonnage capacities (based on the 

maximum concentration of 200 Bq/g) are shown below in order of ascending 

capacity with the most restricted capacity nuclides at the top of the table. 

8.13.5 With the exception of I-129 and Np-237 the tonnage capacities are larger than the 

physical size of the landfill void. This scenario has the potential to constrain 

landfill capacity. 




July 2009 

75 


Groundwater Pathway, Site Boundary Drinking Case, (Table 5.1 Annex B) 

Radionuclide 

Specific Dose 

(microSv/yr 

per MBq) 

Radiological Capacity 

(MBq) 

(3 mSv/yr dose 

criterion) 




Tonnage Capacity (tonnes) 

(@200 Bq/g concentration) 

I-129 3.02E-04 9.93E+06 4.97E+04 

Np-237 9.52E-05 3.15E+07 1.58E+05 

Th-232 3.59E-06 8.36E+08 4.18E+06 

Pa-231 3.08E-06 9.74E+08 4.87E+06 

Cl-36 1.69E-06 1.78E+09 8.88E+06 

Pu-242 1.51E-06 1.99E+09 9.93E+06 

Pu-239 1.37E-06 2.19E+09 1.09E+07 

Ra-226 1.36E-06 2.21E+09 1.10E+07 

Th-229 1.29E-06 2.33E+09 1.16E+07 

Pu-240 9.33E-07 3.22E+09 1.61E+07 

Th-230 7.35E-07 4.08E+09 2.04E+07 

Sn-126 4.87E-07 6.16E+09 3.08E+07 

U-233 4.13E-07 7.26E+09 3.63E+07 

U-234 3.89E-07 7.71E+09 3.86E+07 

U-238 3.89E-07 7.71E+09 3.86E+07 

U-235 3.87E-07 7.75E+09 3.88E+07 

U-236 3.78E-07 7.94E+09 3.97E+07 

Tc-99 1.52E-07 1.97E+10 9.87E+07 

C-14 1.39E-07 2.16E+10 1.08E+08 

Cm-244 1.15E-07 2.61E+10 1.30E+08 

Pu-241 3.62E-08 8.29E+10 4.14E+08 

Cm-243 1.63E-08 1.84E+11 9.20E+08 

Am-241 1.08E-08 2.78E+11 1.39E+09 

Nb-94 9.07E-09 3.31E+11 1.65E+09 

Pu-238 1.67E-10 1.80E+13 8.98E+10 

Ag-108m 1.51E-12 1.99E+15 9.93E+12 

U-232 6.19E-14 4.85E+16 2.42E+14 

Ni-63 3.47E-15 8.65E+17 4.32E+15 

Sr-90 2.19E-17 1.37E+20 6.85E+17 

Pb-210 1.43E-18 2.10E+21 1.05E+19 

Cs-137 9.32E-19 3.22E+21 1.61E+19 

Ac-227 6.70E-19 4.48E+21 2.24E+19 

July 2009 

76 


Groundwater Pathway, Site Boundary Drinking Case, (Table 5.1 Annex B) 

Radionuclide 

Specific Dose 

(microSv/yr 

per MBq) 


(MBq) 


criterion) 




Tonnage Capacity (tonnes) 

(@200 Bq/g concentration) 

H-3 6.89E-23 4.35E+25 2.18E+23 

Ba-133 3.25E-24 9.23E+26 4.62E+24 

Eu-152 4.84E-25 6.20E+27 3.10E+25 

Eu-154 7.98E-28 3.76E+30 1.88E+28 

Co-60 3.79E-32 7.92E+34 3.96E+32 

Eu-155 1.29E-33 2.33E+36 1.16E+34 

Sb-125 2.03E-34 1.48E+37 7.39E+34 

Cs-134 8.12E-36 3.69E+38 1.85E+36 

Fe-55 4.45E-36 6.74E+38 3.37E+36 

Pm-147 1.47E-36 2.04E+39 1.02E+37 

Ru-106 1.67E-38 1.80E+41 8.98E+38 


Exposure by Intrusion into the Emplaced Waste Post Closure of 

the Landfill 

8.14.1 The dose criterion is the lower dose guidance level of 3 mSv/yr for the public and 

workers for an intrusion scenario. 

8.14.2 The scenario applies to members of the public and workers post-closure and is 

uncertain to occur. 

8.14.3 The scenario is assessed in Annex B, table 5.3 in the “Intruder, 60 years case 

and the Resident 60 years case”. The scenario is that either workers or 

members of the public intrude into the waste. For the public case the scenario 

includes residence on the waste material after intrusion. 

8.14.4 The resulting radiological capacities and tonnage capacities (based on the 

maximum concentration of 200 Bq/g) are shown below in order of ascending 

capacity with the most restricted capacity nuclides at the top of the table. 

8.14.5 These scenarios have the potential to constrain landfill capacity. 

July 2009 

77 


Intrusion (Table 5.3, Annex B) "Intruder 60 year Case" 

Radionuclide 

Specific Dose 

(microSv/yr 

per MBq) 


(MBq) 

(3 mSv/yr dose criterion) 





(tonnes) 

(@200 Bq/g 


Th-232 1.87E-04 1.60E+07 8.02E+04 

Sn-126 1.34E-04 2.24E+07 1.12E+05 

Ra-226 1.08E-04 2.78E+07 1.39E+05 

Th-229 9.48E-05 3.16E+07 1.58E+05 

Nb-94 7.52E-05 3.99E+07 1.99E+05 

Pa-231 6.82E-05 4.40E+07 2.20E+05 

Ag-108m 5.13E-05 5.85E+07 2.92E+05 

Pu-239 3.84E-05 7.81E+07 3.91E+05 

Pu-240 3.82E-05 7.85E+07 3.93E+05 

Pu-242 3.54E-05 8.47E+07 4.24E+05 

Th-230 3.49E-05 8.60E+07 4.30E+05 

Am-241 2.79E-05 1.08E+08 5.38E+05 

Np-237 2.47E-05 1.21E+08 6.07E+05 

Ac-227 2.14E-05 1.40E+08 7.01E+05 

Pu-238 2.03E-05 1.48E+08 7.39E+05 

Pu-241 1.47E-05 2.04E+08 1.02E+06 

U-232 8.99E-06 3.34E+08 1.67E+06 

U-235 8.96E-06 3.35E+08 1.67E+06 

Cs-137 5.60E-06 5.36E+08 2.68E+06 

U-233 3.89E-06 7.71E+08 3.86E+06 

U-238 3.88E-06 7.73E+08 3.87E+06 

U-234 3.29E-06 9.12E+08 4.56E+06 

Cm-243 3.03E-06 9.90E+08 4.95E+06 

Pb-210 2.19E-06 1.37E+09 6.85E+06 

Cm-244 1.57E-06 1.91E+09 9.55E+06 

Eu-152 1.42E-06 2.11E+09 1.06E+07 

U-236 1.38E-06 2.17E+09 1.09E+07 

I-129 1.06E-06 2.83E+09 1.42E+07 

Eu-154 2.41E-07 1.24E+10 6.22E+07 

Ba-133 1.65E-07 1.82E+10 9.09E+07 

Sr-90 9.59E-08 3.13E+10 1.56E+08 

July 2009 

78 


Intrusion (Table 5.3, Annex B) "Intruder 60 year Case" 

Radionuclide 

Specific Dose 

(microSv/yr 

per MBq) 


(MBq) 

(3 mSv/yr dose criterion) 





(tonnes) 

(@200 Bq/g 


Cl-36 3.02E-08 9.93E+10 4.97E+08 

Co-60 1.27E-08 2.36E+11 1.18E+09 

Tc-99 1.09E-08 2.75E+11 1.38E+09 

C-14 7.05E-09 4.26E+11 2.13E+09 

Ni-63 8.68E-10 3.46E+12 1.73E+10 

Eu-155 8.04E-11 3.73E+13 1.87E+11 

H-3 4.52E-12 6.64E+14 3.32E+12 

Sb-125 5.68E-13 5.28E+15 2.64E+13 

Cs-134 6.83E-15 4.39E+17 2.20E+15 

Fe-55 4.85E-17 6.19E+19 3.09E+17 

Pm-147 3.54E-17 8.47E+19 4.24E+17 

Ru-106 1.39E-26 2.16E+29 1.08E+27 

July 2009 

79 


Intrusion (Table 5.3, Annex B) "Resident 60 year Case" 

Radionuclide 

Specific Dose 

(microSv/yr 

per MBq) 


(MBq) 


criterion) 





(tonnes) 

(@200 Bq/g 


Ra-226 5.40E-06 5.56E+08 2.78E+06 

Pa-231 1.66E-06 1.81E+09 9.04E+06 

I-129 1.36E-06 2.21E+09 1.10E+07 

Sn-126 4.90E-07 6.12E+09 3.06E+07 

Th-232 4.75E-07 6.32E+09 3.16E+07 

Cl-36 4.28E-07 7.01E+09 3.50E+07 

Tc-99 3.68E-07 8.15E+09 4.08E+07 

Sr-90 3.31E-07 9.06E+09 4.53E+07 

Nb-94 2.44E-07 1.23E+10 6.15E+07 

Ag-108m 1.68E-07 1.79E+10 8.93E+07 

Th-230 1.56E-07 1.92E+10 9.62E+07 

Pb-210 1.28E-07 2.34E+10 1.17E+08 

Th-229 8.84E-08 3.39E+10 1.70E+08 

Np-237 4.65E-08 6.45E+10 3.23E+08 

Cs-137 2.61E-08 1.15E+11 5.75E+08 

U-235 2.47E-08 1.21E+11 6.07E+08 

Pu-239 1.86E-08 1.61E+11 8.06E+08 

Pu-240 1.85E-08 1.62E+11 8.11E+08 

Ac-227 1.81E-08 1.66E+11 8.29E+08 

Pu-242 1.76E-08 1.70E+11 8.52E+08 

U-232 1.57E-08 1.91E+11 9.55E+08 

Am-241 1.57E-08 1.91E+11 9.55E+08 

Pu-238 9.86E-09 3.04E+11 1.52E+09 

C-14 8.68E-09 3.46E+11 1.73E+09 

Pu-241 8.29E-09 3.62E+11 1.81E+09 

U-238 6.87E-09 4.37E+11 2.18E+09 

Eu-152 4.63E-09 6.48E+11 3.24E+09 

Cm-243 4.49E-09 6.68E+11 3.34E+09 

U-233 4.32E-09 6.94E+11 3.47E+09 

U-234 3.71E-09 8.09E+11 4.04E+09 

U-236 3.14E-09 9.55E+11 4.78E+09 

Cm-244 9.45E-10 3.17E+12 1.59E+10 

July 2009 

80 


Intrusion (Table 5.3, Annex B) "Resident 60 year Case" 

Radionuclide 

Specific Dose 

(microSv/yr 

per MBq) 


(MBq) 


criterion) 





(tonnes) 

(@200 Bq/g 


Eu-154 7.87E-10 3.81E+12 1.91E+10 

Ba-133 5.63E-10 5.33E+12 2.66E+10 

Ni-63 2.60E-10 1.15E+13 5.77E+10 

H-3 1.28E-10 2.34E+13 1.17E+11 

Co-60 4.19E-11 7.16E+13 3.58E+11 

Eu-155 2.74E-13 1.09E+16 5.47E+13 

Sb-125 1.89E-15 1.59E+18 7.94E+15 

Cs-134 2.76E-17 1.09E+20 5.43E+17 

Fe-55 8.13E-18 3.69E+20 1.85E+18 

Pm-147 4.55E-19 6.59E+21 3.30E+19 

Ru-106 7.44E-29 4.03E+31 2.02E+29 

July 2009 

81 


8.15 Results of the Assessment 



Direct Radiation 20 mSv/yr Worker 

Exposure from (Ionising Radiation 

Waste Handling and Radiations) 

Emplacement 1 mSv/yr Worker 






Exposure from Gas 20 mSv/yr Worker 

Generation from the (Ionising Radiation 

Landfill – Pre Radiations) 

Closure 







Dropped Load of 20 mSv/yr Worker 

Waste (and 


hypothetical aircraft Radiations) 

impact ) 







Wound Exposure 20 mSv/yr Worker 


Radiations) 



Exposure from Fire See discussion at 8.5 

Exposure from 

Leachate Processing 

Offsite – Sewage 

Works 

Exposure from 

Leachate - Spillage 



Radiations) 









Radiations) 




Assessment Results 

8.1 

Annex D and H 

8.2 

Annex B (5.5) 

8.3 

Annex C 

8.4 

Annex C 

Exposure from the emplaced 

wastes is constrained by a site rule 

limiting dose rate. 

Exposure from handling waste 

packages is constrained by a site 

rule limiting dose rate. 

This scenario may limit radiological 

capacity for certain cases. 

Worker exposure and public 

exposure is within dose targets. 

Worker exposure is within dose 

targets. 

Annex B Not used to limit radiological 

capacity because the waste is 

8.6 

Annex B 

8.7 

Annex B 

essentially incombustible. 

Will be used to establish leachate 

discharge concentration limits. 

Does not restrict landfill capacity 

because the landfill is smaller than 

the most restrictive case. 

July 2009 

82 










Assessment Results 

Exposure from 20 mSv/yr Worker 8.8 

This scenario is less restrictive that 

Leachate - Aerosols (Ionising Radiation Annex B the case at 8.6 which will result in 

Radiations) 

more restrictive leachate discharge 







limits. 

Exposure by Using 0.02 mSv/yr Public (GRA) 8.9 


Groundwater at 

Nearest Abstraction 

Point 

Annex B capacity for certain cases. 

Exposure from Gas 0.02 mSv/yr Public (GRA) 8.10 


Generation from the 

Landfill – post 

closure 


Exposure to Wildlife 10 microgray/hr 8.11 

Not restrictive to landfill capacity. 

from all sources 

Annex B 

External dose from 0.02 mSv Public (GRA) 8.12 

Not restrictive to landfill capacity. 

emplaced wastes 

Annex B 

Exposure by Using 3 mSv Public (GRA and 8.13 


Groundwater from a 

Borehole 

Constructed at the 

Boundary of the 

Landfill 

HPA) 


Exposure by 3 mSv Public or Worker 8.14 


Intrusion into the 

Emplaced Waste 

Post Closure of the 

Landfill 

(GRA and HPA) 


8.15.1 A number of sensitivity studies have been undertaken (Annex B) to examine 

uncertainties associated with the assessment results. 




July 2009 

83 


8.16 Landfill Radiological Capacity 

8.16.1 The actual radiological capacity depends on the proportions of the different 

isotopes in the mixture of the entire waste that is disposed. 

8.16.2 Section 6 of Annex B gives the radiological capacity for a typical and well-defined 

waste stream from the Harwell site, for illustration, on the assumption this was 

the only waste stream sent to the site and presents nuclide specific capacities for 

the various scenarios. 

8.16.3 As discussed above the scenarios which could constrain landfill capacity are: 

- Exposure from Gas Generation from the Landfill 

- Exposure by Using Groundwater at Nearest Abstraction Point 

- Exposure by Using Groundwater from a Borehole Constructed at the 

Boundary of the Landfill 

- Exposure by Intrusion into the Emplaced Waste Post Closure of the 

Landfill 

8.16.4 The table below identifies the specific doses for each of these cases as derived 

from the results in Annex B. The table identifies the most restrictive specific dose 

in each case (shown in yellow highlight) and gives the radiological capacity for 

that case for that nuclide only. Those shown in red outlined border are the 

nuclides which could have capacities that are smaller than the remaining physical 

capacity of the landfill (~1 million tonnes) and hence may be restrictive. 




July 2009 

84 


Radiological 

Capacity 

(MBq) 

(0.02 mSv/yr 

dose 

criterion) 

Specific 

Dose 

(microSv/yr 

per MBq) 

Gas 

Generation 

from the 

Landfill 

Combined 

Pre- 

Closure 

and Post 

Closure 

Public 

Radiological 

Capacity 

(MBq) 

(3 mSv/yr 

dose 

criterion) 

Specific 

Dose 

(microSv/yr 

per MBq) 

Intrusion 

(Table 5.3, 

Annex B) 

"Resident 

60 year 

Case" 

Radiological 

Capacity 

(MBq) 

(3 mSv/yr 

dose 

criterion) 

Specific 

Dose 

(microSv/yr 

per MBq) 

Intrusion 

(Table 5.3, 

Annex B) 

"Intruder 60 

year Case" 

Radiological 

Capacity 

(MBq) 

(3 mSv/yr 

dose 

criterion) 

Specific 

Dose 

(microSv/yr 

per MBq) 

Groundwater 

Pathway, 

Site 

Boundary 

Drinking 

Case, 

(Table 5.1 

Annex B) 

Radiological 

Capacity 

(MBq) 

(0.02 mSv/yr 

dose 

criterion) 

Specific Dose 

(microSv/yr 

per MBq) 

Groundwater 

Pathway, 

Borehole, 

1500m 

Irrigation 

Case (Table 

5.1 Annex B) 

Radionuclide 

H-3 3.66E-30 5.46E+30 6.89E-23 4.35E+25 4.52E-12 6.64E+14 1.28E-10 2.34E+13 1.13E-07 1.77E+08 

C-14 2.06E-09 9.71E+09 1.39E-07 2.16E+10 7.05E-09 4.26E+11 8.68E-09 3.46E+11 

Cl-36 6.52E-08 3.07E+08 1.69E-06 1.78E+09 3.02E-08 9.93E+10 4.28E-07 7.01E+09 

Fe-55 1.04E-43 1.92E+44 4.45E-36 6.74E+38 4.85E-17 6.19E+19 8.13E-18 3.69E+20 

Co-60 1.21E-39 1.65E+40 3.79E-32 7.92E+34 1.27E-08 2.36E+11 4.19E-11 7.16E+13 

Ni-63 7.94E-21 2.52E+21 3.47E-15 8.65E+17 8.68E-10 3.46E+12 2.60E-10 1.15E+13 

Sr-90 3.04E-24 6.58E+24 2.19E-17 1.37E+20 9.59E-08 3.13E+10 3.31E-07 9.06E+09 

Nb-94 3.76E-10 5.32E+10 9.07E-09 3.31E+11 7.52E-05 3.99E+07 2.44E-07 1.23E+10 

Tc-99 2.12E-09 9.43E+09 1.52E-07 1.97E+10 1.09E-08 2.75E+11 3.68E-07 8.15E+09 

Ru-106 3.44E-46 5.81E+46 1.67E-38 1.80E+41 1.39E-26 2.16E+29 7.44E-29 4.03E+31 

Ag-108m 1.68E-17 1.19E+18 1.51E-12 1.99E+15 5.13E-05 5.85E+07 1.68E-07 1.79E+10 

Sb-125 4.81E-42 4.16E+42 2.03E-34 1.48E+37 5.68E-13 5.28E+15 1.89E-15 1.59E+18 

Sn-126 1.20E-08 1.67E+09 4.87E-07 6.16E+09 1.34E-04 2.24E+07 4.90E-07 6.12E+09 

I-129 1.38E-05 1.45E+06 3.02E-04 9.93E+06 1.06E-06 2.83E+09 1.36E-06 2.21E+09 

Ba-133 1.44E-31 1.39E+32 3.25E-24 9.23E+26 1.65E-07 1.82E+10 5.63E-10 5.33E+12 

Cs-134 1.82E-43 1.10E+44 8.12E-36 3.69E+38 6.83E-15 4.39E+17 2.76E-17 1.09E+20 

Cs-137 1.28E-25 1.56E+26 9.32E-19 3.22E+21 5.60E-06 5.36E+08 2.61E-08 1.15E+11 

Pm-147 3.41E-44 5.87E+44 1.47E-36 2.04E+39 3.54E-17 8.47E+19 4.55E-19 6.59E+21 

July 2009 

85 





Radiological 

Capacity 

(MBq) 

(0.02 mSv/yr 

dose 

criterion) 

Specific 

Dose 

(microSv/yr 

per MBq) 

Gas 

Generation 

from the 

Landfill 

Combined 

Pre- 

Closure 

and Post 

Closure 

Public 

Specific 

Dose 

Specific Dose 

(microSv/yr 

Specific 

(microSv/yr 

per MBq) 

Specific 

Dose 

per MBq) 

Groundwater 

Dose 

(microSv/yr 

Groundwater 

Pathway, 

(microSv/yr 

per MBq) 

Pathway, Radiological Site 

Radiological per MBq) Radiological Intrusion Radiological 

Borehole, Capacity Boundary Capacity Intrusion Capacity (Table 5.3, Capacity 

1500m 

(MBq) 

Drinking (MBq) 

(Table 5.3, (MBq) 

Annex B) (MBq) 

Irrigation (0.02 mSv/yr Case, 

(3 mSv/yr Annex B) (3 mSv/yr "Resident (3 mSv/yr 

Case (Table dose 

(Table 5.1 dose 

"Intruder 60 dose 

60 year dose 

Radionuclide 5.1 Annex B) criterion) Annex B) criterion) year Case" criterion) Case" criterion) 

Eu-152 2.77E-32 7.22E+32 4.84E-25 6.20E+27 1.42E-06 2.11E+09 4.63E-09 6.48E+11 

Eu-154 3.28E-35 6.10E+35 7.98E-28 3.76E+30 2.41E-07 1.24E+10 7.87E-10 3.81E+12 

Eu-155 3.63E-41 5.51E+41 1.29E-33 2.33E+36 8.04E-11 3.73E+13 2.74E-13 1.09E+16 

Pb-210 1.25E-25 1.60E+26 1.43E-18 2.10E+21 2.19E-06 1.37E+09 1.28E-07 2.34E+10 

Ra-226 1.56E-08 1.28E+09 1.36E-06 2.21E+09 1.08E-04 2.78E+07 5.40E-06 5.56E+08 2.85E-06 7.02E+06 

Ac-227 5.66E-26 3.53E+26 6.70E-19 4.48E+21 2.14E-05 1.40E+08 1.81E-08 1.66E+11 

Th-229 1.44E-08 1.39E+09 1.29E-06 2.33E+09 9.48E-05 3.16E+07 8.84E-08 3.39E+10 

Th-230 8.24E-09 2.43E+09 7.35E-07 4.08E+09 3.49E-05 8.60E+07 1.56E-07 1.92E+10 7.30E-08 2.74E+08 

Th-232 4.04E-08 4.95E+08 3.59E-06 8.36E+08 1.87E-04 1.60E+07 4.75E-07 6.32E+09 

Pa-231 3.60E-08 5.56E+08 3.08E-06 9.74E+08 6.82E-05 4.40E+07 1.66E-06 1.81E+09 

U-232 5.07E-20 3.94E+20 6.19E-14 4.85E+16 8.99E-06 3.34E+08 1.57E-08 1.91E+11 

U-233 4.62E-09 4.33E+09 4.13E-07 7.26E+09 3.89E-06 7.71E+08 4.32E-09 6.94E+11 

U-234 4.35E-09 4.60E+09 3.89E-07 7.71E+09 3.29E-06 9.12E+08 3.71E-09 8.09E+11 2.02E-11 9.90E+11 

U-235 4.35E-09 4.60E+09 3.87E-07 7.75E+09 8.96E-06 3.35E+08 2.47E-08 1.21E+11 

U-236 4.22E-09 4.74E+09 3.78E-07 7.94E+09 1.38E-06 2.17E+09 3.14E-09 9.55E+11 

U-238 4.35E-09 4.60E+09 3.89E-07 7.71E+09 3.88E-06 7.73E+08 6.87E-09 4.37E+11 2.57E-12 7.78E+12 

Np-237 1.22E-06 1.64E+07 9.52E-05 3.15E+07 2.47E-05 1.21E+08 4.65E-08 6.45E+10 

Pu-238 1.86E-12 1.08E+13 1.67E-10 1.80E+13 2.03E-05 1.48E+08 9.86E-09 3.04E+11 1.64E-15 1.22E+16 

Pu-239 1.54E-08 1.30E+09 1.37E-06 2.19E+09 3.84E-05 7.81E+07 1.86E-08 1.61E+11 

July 2009 

86 





Radiological 

Capacity 

(MBq) 

(0.02 mSv/yr 

dose 

criterion) 

Specific 

Dose 

(microSv/yr 

per MBq) 

Gas 

Generation 

from the 

Landfill 

Combined 

Pre- 

Closure 

and Post 

Closure 

Public 

Specific 

Dose 

Specific Dose 

(microSv/yr 

Specific 

(microSv/yr 

per MBq) 

Specific 

Dose 

per MBq) 

Groundwater 

Dose 

(microSv/yr 

Groundwater 

Pathway, 

(microSv/yr 

per MBq) 

Pathway, Radiological Site 

Radiological per MBq) Radiological Intrusion Radiological 

Borehole, Capacity Boundary Capacity Intrusion Capacity (Table 5.3, Capacity 

1500m 

(MBq) 

Drinking (MBq) 

(Table 5.3, (MBq) 

Annex B) (MBq) 

Irrigation (0.02 mSv/yr Case, 

(3 mSv/yr Annex B) (3 mSv/yr "Resident (3 mSv/yr 

Case (Table dose 

(Table 5.1 dose 

"Intruder 60 dose 

60 year dose 

Radionuclide 5.1 Annex B) criterion) Annex B) criterion) year Case" criterion) Case" criterion) 

Pu-240 1.04E-08 1.92E+09 9.33E-07 3.22E+09 3.82E-05 7.85E+07 1.85E-08 1.62E+11 

Pu-241 1.59E-10 1.26E+11 3.62E-08 8.29E+10 1.47E-05 2.04E+08 8.29E-09 3.62E+11 

Pu-242 1.69E-08 1.18E+09 1.51E-06 1.99E+09 3.54E-05 8.47E+07 1.76E-08 1.70E+11 6.46E-21 3.10E+21 

Am-241 5.37E-11 3.72E+11 1.08E-08 2.78E+11 2.79E-05 1.08E+08 1.57E-08 1.91E+11 

Cm-243 1.82E-10 1.10E+11 1.63E-08 1.84E+11 3.03E-06 9.90E+08 4.49E-09 6.68E+11 

Cm-244 1.29E-09 1.55E+10 1.15E-07 2.61E+10 1.57E-06 1.91E+09 9.45E-10 3.17E+12 

Restrictive Cases for Radiological Capacity on an Individual Nuclide Basis 

July 2009 

87 





8.16.5 The likely conclusion is that the landfill can receive unlimited amounts of these 

nuclides at up to 200 Bq/g without exceeding the radiological capacity before 

reaching the physical capacity. For those highlighted nuclides their effect in 

practice will be reduced by being parts of mixtures with other nuclides and at 

average concentrations less than 200 Bq/g. Several of the most restrictive 

nuclides are short half-life and will generally not feature in decommissioning 

wastes in significant concentrations. 

8.16.6 Notwithstanding the overall conclusion that capacity is not particularly restricted 

in this case, the proposal is that the capacity of the landfill is subject to a total 

capacity limit combined with a series of other conditions. The total capacity limit 

would apply from the date of issue until closure of the landfill or until the capacity 

is reached. The landfill would receive no more LLW under the permit once the 

capacity limit is reached. The capacity limit cannot be expressed as a single 

number because it depends on the mixture received up to any point in time, so 

the proposal is for a continuously revised capacity limit based on individual 

nuclides (including appropriate daughter chains). The total capacity limit would 

be established using an authorised spreadsheet model agreed with the regulator. 

The spreadsheet model would represent the most restrictive case from the risk 

assessment and would produce as an output the remaining capacity of the landfill 

on an individual nuclide basis given the exact wastes received to that point in 

time. Prior to accepting any further waste the model would be used by the landfill 

operator to determine that the consignment would not lead to a breach of the 

total capacity limit. This approach has a number of features: 

The approach requires a comprehensive level of waste characterisation by the 

consignors, but this is considered practicable and is optimal for ensuring public 

health is not impacted by imprecise waste assay. This is also sustainable 

because future generations will receive comprehensive information on the 

disposed nuclides enabling them to make informed decisions. 

The approach cannot be expressed as a simple number and hence may be less 

transparent to the public, but the approach is highly transparent and detailed to 

the regulator. 

The approach is “modern” in the sense that it aligns the authorisation with the risk 

assessment. This is in line with proposals to use risk based approaches for RSA 

exemptions orders, waste definitions, clearance definitions and nuclear site 

delicensing conditions. Hence the criteria used to produce the waste, to 

categorise the waste and to dispose of the waste are based on a consistent risk 

based approach that can be expressed in common terms of risk and dose. 

The approach is based on the total life cycle of the facility. This addresses a 

potential public concern that the authorised capacity may “creep” upwards at the 

point of annual reviews. Authorisation creep of this type was identified as a 

concern from the pre-application stakeholder workshops. 




July 2009 

88 


The approach drives the correct behaviour down to the consignor in respect of 

the waste hierarchy because the approach enables remaining nuclide capacity 

(hence price) to be directly related to the environmental risk of that nuclide. 

The approach addresses the optimisation principle because the remaining 

capacity of the landfill is continuously optimised in a manner that ensures the 

overall risk guidance levels are not exceeded and the model will enable the 

operator and regulator to make informed choices. 

The capacity of the landfill is administered in a manner that ensures dose limits 

and constraints will not be exceeded with a significant safety factor to account for 

uncertainties. 

The approach maintains flexibility to account for the uncertain overall inventory of 

decommissioning wastes to be produced as required by regulatory guidance (ref 

19). 

8.16.7 For the purposes of authorisation this approach can be described as maintaining 

a condition in which the remaining radiological capacity for a particular nuclide or 

wastestream, which is proposed for receipt, is greater than zero, taking into 

account the wastestream received to that point in time. Where: 

Radiological capacity is the amount of radioactive material that can be consigned 

to a site without any of the potentially exposed groups considered receiving a 

dose above a specified criterion, for the specific scenario. 

For a single radionuclide, the radiological capacity (in Bq) can be easily 

calculated by dividing the dose criterion (expressed in Sv) by the maximum 

specific dose for that radionuclide (expressed in Sv/Bq). For mixtures of 

individual radionuclides, the capacity can be simply apportioned (e.g., half of the 

overall capacity to each of two radionuclides). In the case of waste streams, 

however, in which the proportions of different radionuclides are fixed, the 

calculation of capacity must consider both the specific dose and the activity 

ratios. 

The radiological capacity for radionuclide Rni in a waste stream (RCi) is given by: 

RC 

i 

where: 

 

 

DC 

i 

SD 

f 




f 

i 

i 

fi is the fraction of the overall activity arising from Rni (such that fi=1) 

SDi is the specific dose from Rni 

DC is the dose constraint 

Furthermore: 

Total activity limit for 

each radionuclide (Bq) = Dose Limit (Sv/y).Waste Activity (Bq) 

Dose Estimate (Sv/y) 

July 2009 

89 


With the additional constraint that the total dose from all of the radionuclides must 

not exceed the relevant dose limit: 

i Qi / Qi,l

9.0 Radioactive Waste Disposal Proposed Authorisation 

Conditions and Waste Acceptance Criteria 

9.0.1 The following are proposed authorisation conditions and waste acceptance 

criteria subject to development during the authorisation process: 

9.1 Potential Conditions Arising from the Standard RSA 

Authorisation Template 

The use of Best Practicable Means for operation, management and 

maintenance. 

Maintaining equipment and systems provided for the waste disposal process 

in good repair. 

The requirement for a management system, organisational structure and 

sufficient resources to achieve compliance. 

The requirement for training of staff in respect of the conditions of the 

authorisation, the operating techniques and emergency action plans. 

The requirement for written operating arrangements. 

The requirement for audit and review of arrangements. 

The requirement for sampling, testing, calculations and analysis to determine 

compliance. 

The requirement to keep records. 

The requirement to inform the EA of certain matters of compliance. 

9.2 Potential Conditions Arising from the Existing Landfill Permit 

and the Landfill Regulations 2002 

Inclusion of the underpinning limits established by existing risk assessments 

for the existing Landfill Permit for hazardous waste disposal, where 

applicable. Including: 

Compliance with appropriate waste acceptance criteria for hazardous waste 

disposal, including compliance with the prohibited substance list, compliance 

with the waste acceptance limits for hazardous waste disposal established by 

underpinning risk assessments and for the appropriate designated leach 

testing of waste packages or self-similar waste streams. 

Prohibited wastes include: liquids, explosive, corrosive, oxidising, 

flammable or highly flammable wastes, infectious clinical wastes, 

unknown chemical substances. 




July 2009 

91 


Limit values are prescribed for certain leach parameters and total organic 

carbon. The limit values vary between granular or monolithic waste forms 

(LLW could be either type). The limit values appropriate for the LLW 

would be established through reference to the underpinning risk 

assessments for the landfill design. 

Special arrangements apply to handling of asbestos bearing wastes. 

Compliance with site specific authorisation conditions. 

Compliance with the total tonnage limits placed on the landfill by the existing 

permit (It is important in order to maintain the integrity of the existing 

environmental impact assessment that the total physical capacity of the 

landfill is unchanged). 

Prohibition of deliberate dilution or mixing to achieve waste acceptance 

criteria. 

9.3 Conditions Arising from the Site Specific Risk Assessment and 

Industry Practice 

LLW will not be loose handled or tipped at the site. 

LLW will be transported to the site in radioactive materials compliant sealed 

packages and in packages suitable for handling at the landfill site. 

Waste will undergo an agreed pre-acceptance, pre-notification, receipt and 

disposal process in accordance with operating arrangements. 

The paper work for each consignment and the waste load will be inspected to 

confirm that they are consistent with the waste booked into the site. 

Wastes will be placed on the same day as receipt, or will be suitably 

quarantined where this is not practicable. 

Wastes will be covered by at least 300m thickness of suitable cover after 

each emplacement campaign or at the end of the working day such that there 

is no exposed face. Sufficient cover will be used to ensure the doserate at 1 

metre above the waste is less than 2 microSv/hr. 

The waste will not be trafficked or compacted without a covering protective 

layer of suitable cover adequate to protect the waste from exposure to the 

surface. 

A radiation protection plan and scheme of environmental monitoring will be 

operated in accordance with agreed operating arrangements. 

The maximum concentration of radioactivity in any package will be 200 Bq/g 

averaged over the package and in any case not averaged over more than 4 

tonnes. 

The minimum concentration of radioactivity averaged in any package shall be 

such that the waste is defined as low level or very low level radioactive waste. 




July 2009 

92 


Exempt and excluded wastes are not a relevant waste to the permit. It is 

proposed that if wastes of less than a relevant exemption or exclusion order 

are mixed in with the LLW as an inevitable result of their production then 

these would also be treated as LLW. Should the RSA exemption orders 

which define the boundary of Exempt and LLW wastes be revised, then the 

authorisation would automatically incorporate such changes. 

The waste will be a solid waste as defined under the Landfill Regulations. 

Liquid wastes and slurries etc. are prohibited. 

The radiological capacity will be monitored and complied with. 

The waste will consist of waste material that is deemed to be contaminated 

and not be primary contaminant material (such material would be normally 

recoverable and hence not be a waste). 

Notwithstanding the requirements of the existing permit under the Landfill 

Regulations which concern acceptability of chemical hazards in respect of 

hazardous waste - the waste will not be capable of generating toxic or 

explosive gases, vapours or fumes that would be harmful to persons involved 

in the waste process. 

Notwithstanding the requirements of the existing permit under the Landfill 

Regulations which concern acceptability of chemical hazards in respect of 

hazardous waste – the waste will not contain pressurised gas receptacles as 

defined within the Carriage of Dangerous Goods…Regulations 2004 (or as 

amended). 

LLW containing putrescible materials (materials liable to be readily 

decomposed by micro-organisms, excluding wood and paper) will be 

excluded in so far as is reasonably practicable. 

Conditions for acceptance will dictate that the consignor ensures that external 

non-fixed contamination levels on waste packages will be as low as 

reasonably practicable throughout the process and in any case not more than 

4 Bq/cm 2 beta/gamma and 0.4 Bq/cm 2 alpha averaged over an area of 

300cm 2 (as derived from normal industry practice). 

External dose rates throughout the process will be as low as reasonably 

practicable, shall be in accordance with the transport regulations and shall 

not exceed 0.01 mSv/hr (10 microSv/hr) at 1m from the waste package. 

LLW with hazardous properties that would mean it would be a hazardous 

waste if it were not radioactive, will comply with all the relevant conditions of 

the RSA authorisation in respect of non-radiological hazards. 

The capacity of the landfill is subject to a total capacity limit combined with a 

series of other conditions. The total capacity limit would apply from the date 

of issue until closure of the landfill or until the capacity is reached. The 

landfill would receive no more LLW under the permit once the capacity limit is 

reached. The capacity limit cannot be expressed as a single number 

because it depends on the mixture received up to any point in time, so the 

proposal is for a continuously revised capacity limit based on individual 




July 2009 

93 


nuclides (including appropriate daughter chains). The total capacity limit 

would be established using an authorised spreadsheet model agreed with the 

regulator. The spreadsheet model would represent the most restrictive case 

from the risk assessment and would produce as an output the remaining 

capacity of the landfill on an individual nuclide basis given the exact wastes 

received to that point in time. Prior to accepting any further waste the model 

would be used by the landfill operator to determine that the consignment 

would not lead to a breach of the total capacity limit. An example 

spreadsheet to illustrate the model is included in Annex G. It is considered 

that a condition as proposed below will provide robust and enforceable 

means of regulating the site capacity: 

Radiological capacity is the amount of radioactive material that can 

be consigned to the site without any of the potentially exposed 

groups considered receiving a dose above a specified criterion, for 

the specific scenario. 

The radiological capacity for radionuclide Rni in a waste stream 

(RCi) is given by: 

RC 

DC 

i 

SD 

f 




i 

where: 

Furthermore: 

 

 

f 

i 

i 

fi is the fraction of the overall activity arising from Rni 

(such that fi=1) 

SDi is the specific dose from Rni 


Total activity limit for 

each radionuclide (Bq) = Dose Limit (Sv/y).Waste Activity (Bq) 

Dose Estimate (Sv/y) 

With the additional constraint that the total dose from all of the 

radionuclides must not exceed the relevant dose limit: 

i Qi / Qi,l

Any isotopes not modelled in the risk assessment will be modelled prior to 

acceptance and incorporated into the approved spreadsheet model for 

radiological capacity. 

The conditions for the aftercare period and revocation of the authorisation, 

including the provisions for closure of the authorisation at the time the 

disposals cease and any provisions for monitoring during the aftercare 

period. 




July 2009 

95 


10.0 BPEO Assessment for Disposal of LLW from 

Nuclear Sites 

10.1 BPEO (Best Practicable Environmental Option) 

10.1.1 The Royal Commission on Environmental Pollution defined the Best Practicable 

Environmental Option (BPEO) as: 

“The outcome of a systematic and consultative decision-making procedure which 

emphasises the protection and conservation of the environment across land, air 

and water. 

The BPEO procedure establishes, for a given set of objectives, the option that 

provides the most benefits or the least damage to the environment as a whole, at 

acceptable cost, in the long term as well as the short term.” 

10.1.2 The Environment Agency requires the use of the BPEO methodology by nuclear 

industry sites in order to underpin their waste management choices and 

strategies. 

10.1.3 A BPEO study involves a rational consideration of all the options against a series 

of criteria. Importantly it involves extensive consultation. 

10.1.4 Any nuclear industry site that wished to send waste to the East Northants 

Resource Management Facility under an authorisation would be required to have 

an underpinning BPEO study that justified this approach before they would be 

granted a transfer authorisation under the RSA 1993 by the EA. 

10.1.5 The consigning site would in any case have to apply the waste management 

hierarchy of avoid – minimise – recycle – reuse to any waste stream prior to 

consideration of disposal and would have to demonstrate the use of best 

practicable means in respect of waste generating activities. So, for example, a 

nuclear industry site is required to extensively sort wastes prior to dispatch in 

order to avoid unnecessary disposals. 

10.1.6 The proposed approach is that the BPEO methodology would not be applied to 

this authorisation application, but would be applied to each individual transfer 

authorisation application from nuclear industry sites. 

10.1.7 Non-nuclear industry sites are not required to prepare BPEO assessments for 

their waste streams, although these would normally be small volumes in 

comparison to the potential higher volumes from the nuclear industry. 




July 2009 

96 


11.0 BPM and ALARA Assessment for the Proposed 

Radioactive Waste Disposal 

11.1 ALARA 

11.1.1 The “As Low as Reasonably Achievable” (ALARA) principle is concerned with 

optimising radiation doses to humans. 

11.1.2 “In relation to any particular source within a practice, the magnitude of individual 

doses, the number of people exposed, and the likelihood of incurring exposures 

where these are not certain to be received should all be kept as low as 

reasonably achievable, economic and social factors being taken into account.” 

11.1.3 Conservative radiological assessments for workers and the public in the 

operational and post-closure stages are presented in this report and demonstrate 

that it is likely that dose constraints, dose limits, design risk targets and design 

dose targets will be achieved. The design targets are set at levels beyond which 

further measures should only be considered necessary if they do not involve 

disproportionate costs. 

11.1.4 Operational optimisation measures are described in this report and would be 

developed as part of the radiation protection plan for the site. Feedback from 

workplace and environmental monitoring would be used to implement further 

optimisation measures if required in order to achieve actual exposures which are 

a fraction of the constraints and limits. 

11.2 BPM 

11.2.1 The Best Practicable Means (BPM) principle is essentially a consideration of 

whether an adequate argument has been made that further measures to reduce 

risk are not needed because the measures cannot be implemented at a 

reasonable cost given economic and social factors. 

11.2.2 Having carried out a BPEO study to consider what the right option to pursue is, 

BPM is concerned with executing that option in the right way. 

11.2.3 Whereas ALARA applies to dose optimisation, BPM applies to optimise 

radioactive waste management. 

11.2.4 Within a particular waste option, the BPM is that level of management and 

engineering control that minimises, as far as practicable, the release of 

radioactivity to the environment whilst taking account of a wider range of factors, 

including cost-effectiveness, technological status, operational safety and 

social/environment factors. 

11.2.5 To some extent the BPM concept overlaps with the BAT (Best Available 

Techniques) concept that underpins the provision of new landfill designs under 




July 2009 

97 


the Environmental Permitting Regulations (formerly the Landfill Regulations) and 

which applies to non-radioactive pollutants. Indeed, there is a view that that BAT 

and BPM are synonymous. Many BAT features of new landfill sites for 

hazardous waste such as the East Northants Resource Management Facility are 

effectively prescribed (for example, the permeability performance of barrier 

layers), whereas BPM features are not so specifically prescribed. 

11.2.6 The BPEO study undertaken for the example Harwell waste stream is likely to be 

typical for all decommissioning waste arising. That study indicates that shallow 

disposal in an engineered facility is likely to be the BPEO for most low level 

decommissioning wastes of the type proposed for ENRMF. Hence, BPM for 

such an option focuses on the design of the facility, whether it meets modern 

standards and whether any further straightforward improvements are feasible. 

11.2.7 The current state of the wastes on the various nuclear sites, whilst adequately 

controlled, is unarguably less satisfactory and less sustainable than final 

disposal. If the waste is not disposed to engineered facilities it will remain in 

above ground stores or in contaminated land areas and will present a higher risk 

to future generations. The proposed option represents a net reduction in risk 

from the current situation. 

11.2.8 It is submitted that use of a modern standard hazardous waste landfill that has 

been designed and implemented using BAT under recent legislative guidance 

represents BPM for the disposal of LLW of the type proposed for ENRMF. The 

reasoning is that the LLW has the same chemical properties whilst being no more 

mobile and are generally less reactive than the hazardous wastes for which the 

landfill was designed and that the landfill was designed in such a way as to 

prevent harm to humans and the environment. If a new specialist landfill were 

designed for the LLW of the type proposed for ENRMF it is unlikely to use 

engineering features and standards beyond those currently used to define BAT 

for modern hazardous waste landfills. 

11.2.9 The risk assessments in this application support the case that the existing landfill 

design will prevent harm arising from the LLW to an appropriate risk standard. 

11.2.10Further limitations have been proposed on the disposal that are additional to the 

BAT features of the existing landfill and these are described throughout this 

application and in particular in section 5, including : 

Wastes will only be accepted for disposal if the source site (in the case of 

a nuclear industry site) demonstrates that the option is BPEO and that 

BPM has been used to apply the waste hierarchy and to characterise the 

waste. 

The radiological capacity of the landfill has been back calculated to give a 

design risk target under the most restrictive future scenarios of

The maximum concentration of radioactivity has been limited through 

proposed waste acceptance criteria to limit the effects of routine and 

accidental exposures from transport and emplacement operations such 

that dose constraints are achieved and improved upon. 

Additional waste acceptance criteria have been proposed to further limit 

exposure. For example, a constraint on the external dose on the transport 

package has been proposed which is more constraining than required by 

transport regulations. 

Operational arrangements have been proposed to further reduce 

exposure, including for example, no loose handling of materials, the use 

of suitable cover materials, the use of segregation arrangements, the use 

of contamination clearance and control arrangements, the use of 

personnel, workplace and environmental monitoring and the use of 

emergency arrangements. 

11.2.11The likely result of these additional measures will be that the risk presented by 

the waste disposal will be less than the design risk target over the long term and 

that occupational dose to workers will be well within the dose constraints. 

11.2.12Can further measures be implemented? 

The two principal possible further restrictions are to further limit the 

radiological capacity or to further limit the radioactivity concentration. 

The long term risks are driven by the total activity disposed. Further 

reductions in the capacity are not reasonably practicable because the 

capacity has been designed to a basic risk target that meets current 

guidance and in practice the further optimisation measures described 

above will reduce the risk still further. The capacities that result are of a 

size to be useful and economic for the decommissioning industry and 

reductions would make the waste route considerably less able to meet 

regional demand. 

The short term risks are driven by the concentration of activity and 

resulting doserate in any one package of waste for any given isotope 

type. The concentrations and doserates have been optimised through 

application of restrictive waste acceptance criteria and further restrictions 

are not required in order to achieve the occupational dose constraints for 

workers (based on those workers using the constraints applicable to the 

public). In practice the ALARA and BPM arrangements described above 

will further reduce exposures. The proposed radioactivity concentrations 

and doserate criteria are of practical use for the decommissioning industry 

and further reductions would severely limit the applicability of the waste 

route to solve the strategic drivers described above. 




July 2009 

99 


12.0 Landfill Engineering and BAT Features of the 

Existing Landfill 

12.0.1 The existing hazardous waste landfill at the East Northants Resource 

Management Facility is authorised under the Landfill Regulations and the 

Pollution Prevention and Control Regulations. 

12.0.2 The landfill is designed in accordance with these regulations and utilises Best 

Available Techniques (BAT). These BAT design features and arrangements 

would also be utilised by the LLW and contribute to the BPM case above. 

Recently created hazardous waste landfill sites, such as the East Northants 

Resource Management Facility, have the highest level of BAT features 

(compared to non-hazardous or inert landfill sites). 

12.0.3 The details of the BAT design features of the existing landfill are contained in ref 

15 which gives hydrogeological, stability, landfill gas, environmental impact and 

nuisance risk assessments. The arrangements for construction design, waste 

acceptance, groundwater protection, landfill gas management, leachate 

management, landfill stabilisation, pollution prevention, nuisance prevention and 

quality assurance, construction quality assurance, maintenance, landfill capping, 

site restoration, operations, waste handling/placement, security, use of raw 

materials, secondary wastes, accident arrangements, monitoring, closure, 

aftercare and surrender are described in existing documentation for the landfill 

site. 

12.0.4 These BAT features represent a solid foundation for the management of the LLW 

and have been taken into account in the risk assessment for LLW disposal to the 

extent detailed in this document. The features are not described in detail in this 

document. An outline of the key landfill engineering features follows: 

A full containment landfill engineering system designed to meet the 

requirements of the Landfill Regulations 2002 and 2004 (as amended). 

This requires a basal lining system with, or equivalent to having, a 

permeability of 1 x 10 -9 m/s or lower and a thickness of no less than 5m or 

equivalent. For the basal liner the landfill incorporates a 1.5m thick layer 

of reworked clay with a maximum permeability of 3x10 -10 m/s and a 2mm 

high density polyethylene synthetic liner. The sidewalls are formed from 

the in-situ clay materials with the liner placed over these. 

The surface capping system comprises from the waste surface upwards: 

300mm regulating layer 

geosynthetic clay liner 

1mm welded geomembrane 

protector geotextile or drainage geocomposite 

1m of restoration soils 




July 2009 

100 


A leachate collection system 

A gas collection system 

Ancillary systems such as vehicle cleaning equipment 

A surface water, groundwater and environmental monitoring system 

Restoration of the site to grassland including wildflower meadow and 

agricultural grassland 

Operational arrangements for site construction, operation, closure, 

restoration and aftercare. 

12.0.5 The proposal for LLW disposal does not change the existing arrangements and 

augments the arrangements for the LLW component. 




July 2009 

101 


13.0 Waste Hierarchy and Waste Minimisation at Source 

13.0.1 Producing sites will be required to demonstrate that the waste hierarchy has 

been applied to the waste prior to acceptance by the landfill. 

Avoid – Wastes are not generated if this is feasible, for example, maintaining 

separation of clean and radioactive materials in a building rather than 

deliberately mixing the wastes to produce a larger volume of lower 

concentration material. 

Segregate – For example through application of BPM methods to 

characterise waste into exempt and LLW waste streams rather than mixing. 

Prevent Spread – For example, preventing the spread of contamination to 

clean materials. 

Recycle/Reuse – For example, reuse of contaminated or activated concrete 

as a construction material within the nuclear industry. 

Clearance, Exemption and Exclusion – For example, use of the good 

practice to sort materials into classifications that can be defined as nonradioactive 

waste for the purposes of disposal. 

Volume Reduction – For example, compacting wastes which are 

compactable to reduce disposal volume. 

Disposal 




July 2009 

102 


14.0 Summary of the Existing Environmental Statement 

for the Site and Impacts of the Proposal 

14.0.1 The environmental impact statement for the existing landfill (ref 15) describes the 

impacts from hazardous waste disposal operations. 

14.0.2 It is submitted that the addition of a LLW stream to the inventory of waste 

acceptable at the site makes no significant change to the existing environmental 

impact assessment or current/future use of the site. 

14.0.3 The following table is a summarised version of table 15.1 from ref 15 which 

summarises the existing environmental impact assessment and to which has 

been added comments on the changes introduced by the LLW stream. 

Feature and Interest Existing Impact and 

Changes Resulting from 

GROUNDWATER 

Mitigation 

LLW Disposal Authorisation 

Aquifer – flow characteristics Change in recharge and flow None – the design of the 

considered insignificant. existing landfill is unchanged 

Aquifer – groundwater quality Residual impact from leachate Insignificant additional risk as 

considered to be insignificant demonstrated by the risk 

SURFACE WATER 

due to engineered mitigation 

features 

assessment 

Steams - flow Potential change in flow due to No additional change in 

new landform, considered to 

be insignificant after impact 

landform. 

Steams - quality Potential impact from leachate Insignificant additional risk as 

and runoff, considered to be demonstrated by the risk 

LANDSCAPE 

insignificant after impact. assessment 

Westhay Cottages – visual Design and screening to No additional change in 

receptor 

mitigate impact. 

landform. 

Landscape Character Area – Design and screening to No additional change in 

landscape receptor 

TRAFFIC 

mitigate impact. 

landform. 

Traffic – Residents Change in traffic insignificant No additional change in traffic 

quantity. 

Traffic - Motorists Roads have adequate No additional change in traffic 

capacity 

quantity. 

Traffic - Safety Motorists Mud on road mitigated through Additional monitoring to check 

NOISE 

decontamination measures for contamination spread 

Noise – Local community Some noise mitigation 

measures applied to mitigate 

noise from landfilling 

operations 

No additional noise 




July 2009 

103 


ECOLOGY and NATURE CONSERVATION 

Woods, hedgerows, scrub, 

grassland, standing water, 

badgers, birds, amphibians, 

reptiles, bats, rare plant 

species. 

AIR QUALITY 

Local property and global 

climate 

Some preventative measures, 

surveys and habitat creation 

schemes used to mitigate 

potential impacts. 

Mitigation measures to control 

gas generation and 

odour/dust. 

HEALTH 

Health – people Risk of exposure to dusts, 

aerosols, gas, contaminants, 

leachate, vermin, and litter 

mitigated through engineering 

design features and 

operational arrangements. 




No additional land usage. 

The risk assessment shows 

that the impact will be 

insignificant. 

Risk from additional gas 

discharges assessed as 

insignificant. No handing of 

loose wastes. Emergency 

measures to address spillages 

and dropped loads. 

Health impact from LLW 

assessed to be insignificant 

during the operational and 

post-closure phases. 

ARCHAEOLOGY 

Archaeology No known impacts. No additional impacts. 

RADIOACTIVITY (added to this table for this report) 

Health – people in the long 

and short term – workers and 

Risk of exposure to radiation. 

the public 

The risk to workers has been reduced to low levels which are 

better than occupational dose targets and constraints through 

limits on waste acceptance and through operational 

arrangements. 

The risk to the public in the short and long term has been 

reduced to insignificant levels through limits on waste 

acceptance and through operational arrangements. 

July 2009 

104 


15.0 Outline of Management and Operating Arrangements 

15.0.1 Augean formed in 2004 is the UK’s market leader specialising in the 

management of hazardous waste. The Company provides a complete solution 

for the management of hazardous wastes and works in partnership with 

producers to provide long-term answers to the treatment and disposal of our 

more difficult to manage wastes. Augean operates proactively to ensure that 

regulatory standards are met and often exceeded. Best practice is considered 

normal practice. The Company currently owns more than 8 million cubic metres 

of void space, five treatment centres and employ over 150 people across 10 

sites. 

15.0.2 The locations of Augean facilities are shown below: 

15.1 Augean Corporate Social Responsibility 

15.1.1 Augean is committed to Corporate Social Responsibility as demonstrated through 

the publication annually of a Corporate Social Responsibility report which 

measures our performance in respect of business, health and safety, our 

employees, our neighbours and the environment. An essential element of our 

approach to our business is our core business values. 




July 2009 

105 


15.2 Augean’s Core Business Values: 

Transparency we are open and transparent in all that we do 

Integrity we are trustworthy and honest in all that we say and do 

and take responsibility for our own actions 

Social and 

community 

responsibility 

Environmental 

responsibility 




we recognise that our actions have a material impact on 

the communities in which we operate and take that 

responsibility extremely seriously 

we respect the environment and invest time and 

resource in protecting it 

Technical excellence we employ skilled staff and use up-to-date techniques 

and equipment 

Professionalism we are reliable and consistent and deliver 

excellent service 

Respect we are friendly and courteous to colleagues, clients and 

suppliers 

Passion we are proud of our company and dedicated to its 

purpose. We are enthusiastic, enjoy challenges and are 

eager for success 

15.3 Management systems 

15.3.1 Operational performance is maintained through a certified Integrated 

Management System (IMS) delivering protection of health and safety, both 

internally and externally, and the management, protection and improvement of 

the environment for nature and our local communities. The IMS is certified by the 

British Standards Institute to the following standards: 

IS0 9001 Quality management system 

ISO 14001 Environmental management system 

OHSAS 18001 Health and safety management system 

PAS 99 Integrated management system 

15.3.2 Central to the Integrated Management System is the IMS Policy statement. 

July 2009 

106 


15.4 Augean’s IMS Policy: 

Augean is committed to conducting its business operations in a responsible manner 

and we recognise the need to continually improve our operations where practical to 

do so in order to reduce our effects on the environment, ensure the safety and 

welfare of our personnel and neighbours, and ensure client satisfaction through 

service excellence. 

We seek to exceed legal obligations and be among the leading exponents of good 

practice and technological development within the waste management industry. 

At no time shall we provide services that fall short of the professional integrity and 

objectivity that we understand our clients and stakeholders will require and every 

effort shall be sustained to ensure the accuracy, probity and surety of the services 

that we provide. 

To achieve this and remain competitive, we pursue a programme of continuous 

improvement in all aspects of our business. To assist in achieving this high level of 

regulatory compliance, client satisfaction and operational improvement, corporate 

objectives shall be set on an annual basis. Realisation of set objectives is 

continuously monitored, reviewed and communicated throughout the company. 

To ensure a high standard of awareness within the company we provide our employees 

with continuous training to improve their skills and competencies. To maintain external 

awareness and good perception that the company actively liaises with regulatory 

bodies, environmental organisations, stakeholders, the local community and all other 

interested parties. 

The company shall encourage our supply chain and contractors to improve 

business standards through continual assessment. 

It is the policy of the company that the documented Business Management 

System detailed in the Business Manual and supporting administrative 

procedures are the normal basis of working and will be applied to all relevant 

work. 

Paul Blackler, Chief Executive 




July 2009 

107 


15.5 Augean Organisation 

Group Technical 

Director 

Management 

Board 

Site Manager Site Manager 

Technical support 

Technical Manager 

Monitoring Manager 

Site Chemists and Laboratory 

Site Supervisor 

15.5.1 Site Managers are responsible for the quality, health and safety and 

environmental performance of their sites. The Site Manager reports directly to 

the Management Board which is ultimately responsible for performance. The 

Site Manager at East Northants Resource Management Facility is a holder of a 

Certificate of Technical Competence for the management of a hazardous landfill. 

Technical support and expertise is provided by the Technical Team specifically 

the Technical Manager who deals with Authorisation issues and legislative 

compliance, the monitoring team that monitors the environmental impact of the 

site in all media and the site chemists who provide laboratory facilities and 

determine the suitability of waste for acceptance at the site. The Technical Team 

reports to the Group Technical Director who is a member of the Management 

Board and advises the Board on health and safety and environment issues. 

15.5.2 Augean employs a range of highly qualified professionals with expertise in 

environmental and health and safety legislation, environmental management, 

chemistry, ecology, planning, engineering and waste management. As 




Operatives 

Site Manager 

July 2009 

108 


necessary expertise is outsourced from external consultants. The Company 

maintains a list of approved consultants who are selected on the basis of 

qualification and experience and whose place on the list is dependent on good 

service. 

15.6 Operational control 

15.6.1 Through the IMS the aspects and impacts of the business have been 

established. Risk assessments have been conducted for all operational activities 

and where necessary to ensure adequate operational control procedures have 

been developed and implemented. The Table below lists the main operational 

procedures relevant to this application. Additional procedures specific to disposal 

of LLW would be developed as required. 

Reference 

No 

Title 

BM01 Business Manual 

BMS02 Customer Care Policy 

BMS05 Group Environmental Aspects Register 

BMS05 Environmental Aspects - Kings Cliffe 

BMS07 Group Register of Environmental Regulations 

BMS14 COSHH Register and forms - Kings Cliffe 

BMS18 Eye and Eye Sight Test Policy 

BMS19 Group Health and Safety Regulatory Register 

CBP01 Document Control 

CBP03 Training 

CBP04 Communication 

CBP08 Regulatory Compliance 

CBP09 Assessment of Enviornmental Effects 

CBP10 Supplier Evaluation 

CBP12 Control of Contractors 

CBP13 PPC Emergency Preparedness and Response 

CBP15 PPC Handling environmental and safety 

complaints 

CBP16 PPC non-conformance identification, investigation 

and implementation of corrective and preventative 

actions 

CBP17 Monitoring and reporting 

CBP18 Internal Auditing 

CBP19 Management Review 

CBP27 Plant maintenance 

CBP40 Data Back up 

CBP41 Management of change 

CBP43 Permit to work instruction 

CPR 01 Sampling of hazardous waste 

CPR 02 Emergency preparedness and response 

CPR 03 Collection of Windblown Litter Risk Assessment 




July 2009 

109 


CPR 04 Off loading of palletised waste Risk Assessment 

CPR 05 Environmental Monitoring Risk Assessment 

CQP01 Customer Complaints 

CQP02 Customer Feedback 

CQP03 Telephone Contact 

CSP01 COSHH 

CSP02 Risk Assessment 

CSP03 Health Surveillance 

CSP04 Fork Lift Trucks 

CSP05 Welding and flame cutting 

CSP06 Electrical safety 

CSP07 Manual Handling 

CSP08 Noise Control 

CSP09 Personal Protective Equipment 

CSP10 First Aid 

CSP11 Fire Safety 

CSP12 Lifting Operations and lifting equipment 

CSP13 New and expectant mothers 

CSP14 Young persons 

CSP15 Transport Safety 

CSP16 Display Screen Workstations 

CSP17 Violence at Works 

CSP18 Visitors 

CSP19 Housekeeping 

CSP20 Statutory Inspections 

CSP21 Health and Safety information 

CSP23 Consultation with employees 

CSP24 Disabled Persons 

CSP25 Pressure systems 

CSP26 Lone Working 

CSP27 Accident Investigation 

CSP28 Stress at Work 

CSP29 Carriage of Samples 

CSS01 Tipping Artic Systems 

CSS02 Disposal of hazardous waste 

CSS04 Water bowser (filling, use etc) 

CSS05 Unloading palletised loads 

CSS08 Wheel wash maintenance 

CSS10 Use of strimmers 

CSS11 Refuelling of plant and equipment 

CSS15 Towing vehicles 

CSS19 Errection and dismantling of litter fencing 

CSS22 Tipping artics containing asbestos waste 

CSS23 Tipping artics containing non-hazardous gypsum 

wastes 

IG01 Sampling of Waste 




July 2009 

110 


KC01 Site Rules - Drivers 

KC01 Site Rules - Visitors and Contractors 

KC06 Emergency Plan 

LF001 Offices Risk Assessment 

LF002 Operational Areas Risk Assessment 

LF003 Plant Risk Assessment 

LF004 Site Traffic Risk Assessment 

LF005 Wheel cleaning facilities Risk Assessment 

LF006 Tipping Artics Risk Assessment 

MHR1 Rock Salt Bags Risk Assessment 

MHR11 Waste inspection and sampling Risk Assessment 

MHR2 Handling of Drummed Oils and Greases Risk Assessment 

MHR3 Handling and Installation of Leachate Pipes Risk Assessment 

MHR4 Unloading Palletised Waste Risk Assessment 

MHR5 Handling of Deliveries Risk Assessment 

MHR6 Water Bowser Risk Assessment 

MHR7 Erection and Dismantling of Litter Fencing Risk Assessment 

MHR8 Collection of Windblown Litter Risk Assessment 

PPC LF 02 Acceptance of hazardous waste to landfill 

PPC LF 10 Non-conforming waste loads 

PPC LF 11 Quarantined Waste Loads 

PPC LF 12 Cover control 

PPC LF 13 Pest control 

PPC LF 14 Litter control 

PPC LF 15 Noise control 

PPC LF 16 Odour control 

PPC LF 17 Control of dust and particulates 

PPC LF 18 Classification, assessment and accpetance of inert 

wastes 

PPC LF 19 Security Procedures 

SR1 Work on or above water Risk Assessment 

SR2 Tipping Artics Risk Assessment 

SR3 Steam Cleaning Plant and Equipment Risk Assessment 

SR4 Disposal of Asbestos Risk Assessment 

SR6 Leptospirosis Risk Assessment 

SR7 Tetanus Risk Assessment 

Disability Risk Assessment Risk Assessment 

Fire Risk Assessment Risk Assessment 




July 2009 

111 


15.7 Arrangements Specific to LLW Disposal Operations 

15.7.1 The following arrangements will be incorporated into the management system 

specific to LLW disposal operations: 

- A radiation protection plan and risk assessment as required by the 

Ionising Radiations Regulations, prepared by the site Radiological 

Protection Advisor and Qualified Expert. ENRMF, IRRs 1999, Radiation 

Risk Assessment for LLW, HPA (ref 16) 

- An amendment to the site emergency plan to include response 

arrangements to identified fault scenarios including: 

Dropped load 

Contamination discovery 

Non-compliant load 

Dose above threshold discovery 

Potentially contaminated person or wound 

- A procedure for the receipt of waste, assay, quarantine, waste 

emplacement, coverage, record keeping and general LLW disposal 

operations 

- A procedure for routine and periodic health surveillance monitoring for 

contamination and exposure 

- Procedures for environmental monitoring incorporated into the MAPs 

- A procedure for the pre-acceptance of waste including the conditions for 

acceptance for LLW for use in contractual arrangements with consignors 

- Amendments to existing roles and responsibilities to add the roles: 

Radiation Protection Advisor (Qualified Expert), 

Radiation Protection Supervisor(s), 

Dangerous Goods Safety Advisor (Class 7) 

15.7.2 Augean have retained the service of the Health Protection Agency as 

Radiological Protection Supervisor and to act as Qualified Expert in respect of 

such matters as training, advice, emergency response, provision and calibration 

of instrumentation, provision of health physics services, environmental and 

workplace monitoring, analysis, interpretation of specialist information etc. 




July 2009 

112 


16.0 Stakeholder Consultation 

16.0.1 Working in close co-operation with potential consignors of LLW and the 

Environment Agency, Augean plc has undertaken a public consultation 

programme in support of this application for authorisation in accordance with 

Government and Local Planning Authority best practice guidance on this aspect 

of development. 

16.0.2 The specialist development and planning communications company, Jennings 

Nicholson Associates, have assisted Augean in this task. 

16.0.3 The purpose of the communications programme has been to inform and educate 

those affected by this application, to reassure the local communities of the nonthreatening 

nature of what is proposed and to ensure that all the key 

stakeholders have an opportunity to voice their comments and concerns so that 

the company can address them during the authorisation process. 

16.0.4 Augean has operated in the area of its East Northants Resource Management 

Facility (formerly Kings Cliffe Landfill) since 2004. In that time it has built up a 

good working relationship and enhanced its corporate reputation with the local 

communities and those elected to represent them, as well as the statutory and 

non-statutory consultees. It will build on this foundation to engage all the key 

target audiences during the consultation process associated with this application. 

16.0.5 The programme has been set out in a Communications Plan, which established 

clear communications objectives, set out a carefully-timed phased programme to 

reflect milestones in the determination process and identified a variety of proven 

and effective mechanisms to promote the scheme and it key messages. 

16.0.6 The programme has included: meetings of the local liaison committee, a one day 

public surgery, meetings with official bodies and regulators and meetings with 

county, district and parish councils. The results of the feedback from these 

events has been analysed and used to refine this authorisation application. 

16.0.7 The LLW disposal process is subject to a planning approval under Town and 

Country Planning regulations which will involve further stakeholder dialogue. 




July 2009 

113 


17.0 The Application Forms 

17.1 Waste Disposal 

17.1.1 A copy of the application form for a disposal authorisation (under sect 13 of the 

Radioactive Substances Act 1993 (RSA) is included in Annex F. 




July 2009 

114 


18.0 Conclusion 

18.0.1 A proposed set of outline arrangements, waste acceptance criteria and potential 

authorisation conditions has been described for a process to dispose of solid 

LLW wastes to East Northants Resource Management Facility. 

18.0.2 A consequence and risk assessment has been carried out for the public and 

workers in the long and short term. The radiological capacity of the landfill site 

has been back calculated in order to meet defined risk and dose targets. In 

addition, operational arrangements and constraints have been proposed using 

best practicable means to further reduce risk and optimise exposures. 

18.0.3 The proposal is that the capacity of the landfill is subject to a total capacity limit 

combined with a series of other conditions. The total capacity limit would apply 

from the date of issue until closure of the landfill or until the capacity is reached. 

The landfill would receive no more LLW under the permit once the capacity limit 

is reached. The capacity limit cannot be expressed as a single number because 

it depends on the mixture received up to any point in time, so the proposal is for a 

continuously revised capacity limit based on individual nuclides (including 

appropriate daughter chains). The total capacity limit would be established using 

an authorised spreadsheet model agreed with the regulator. The spreadsheet 

model would represent the most restrictive case from the risk assessment and 

would produce as an output the remaining capacity of the landfill on an individual 

nuclide basis given the exact wastes received to that point in time. Prior to 

accepting any further waste the model would be used by the landfill operator to 

determine that the consignment would not lead to a breach of the total capacity 

limit. 

18.0.4 It is submitted that the proposal for disposal of LLW is justified. 




July 2009 

115 


References 

1 Process and Information Document for: Applications for New 

Authorisations;.....issued under the Radioactive Substances Acts 1993 to 

Nuclear Sites in England and Wales, EA, 16/12/05, 

Version 1 

2 Considerations for Radioactive Substances Regulation under the RSA 

1993 at Nuclear Sites in England and Wales, 16/12/05, EA 

3 Policy for the Long Term Management of Solid Low Level Radioactive 

Waste in the UK, March 2007, DEFRA 

4 Radioactive Substances Act 1993 

5 RWMAC, Advice to Ministers on Management of Low Activity Solid 

Radioactive Wastes, 2003 

6 Radioactive Substances Act 1960, A guide to administration of the Act 

7 Environment Act 1990 

8 Hazardous Waste Regulations 2005. 

9 The Pollution, Prevention and Control Regulations 2000 

10 The Landfill Regulations 2002 (as amended 2004 and 2005) 

11 Implications of European Directives for the Disposal of Radioactive 

Wastes, DEFRA, October 2005 

12 Radiological Assessment of Disposal of Large Quantities of Very Low 

Level Waste in Landfill Site, Chen, Kowe, Mobbs and Jones, HPA-RPD 

and Atkins, HPA-RPD-020, March 2007 

13 The Carriage of Dangerous Goods and Use of Transportable Pressure 

Equipment Regulations 2007, No 1573 

14 Documents of the HPA: Radiation Protection Objectives for the Land- 

Based Disposal of Solid Radioactive Wastes, RCE-8, February 2009. 

15 Augean South Ltd., East Northants Resource Management Facility, 

Environmental Statement, Bullen Consultants, June 2005 

16 ENRMF, IRRs 1991, Radiation Risk Assessment for LLW, HPA March 

2009. 

17 SNIFFER, UKRSR05: BPM for the Management of Radioactive Waste, 

2005 

18 Near-surface Disposal Facilities on Land for Solid Radioactive Wastes, 

Guidance on Requirements for Authorisation, February 2009 




July 2009 

116 


19 Environmental Permitting Guidance Radioactive Substances Regulation 

(RSR), Draft Guidance for Consultation, May 2009. 

20 UK Strategy for the Management of Solid Low Level Radioactive Waste 

from the Nuclear Industry: UK Nuclear Industry LLW Strategy, 

Consultation Document, June 2009, Nuclear Decommissioning Authority. 

21 Impact Cratering: A Geologic Process, H.J.Melosh 

22 UKAEA Safety Assessment Handbook 

23 Report into Wisbech Air Crash, 1979, Hansard 

24 Aircraft Accident Report No 2/90 (EW/C1094), Report on the accident to 

B747-121, N739PA, Lockerbie, 1988. 




July 2009 

117 


Figures 




July 2009 

118 




Glossary 

In the context of this Glossary, the term ‘waste’ refers, in general, to radioactive waste unless 

otherwise specified. 

absorbed dose. See dose, absorbed. 

activation. The process of inducing radioactivity. Most commonly used to refer to the induction of 

radioactivity in moderators, coolants, and structural and shielding materials, caused by irradiation 

with neutrons. 

activation product. A radionuclide produced by activation. Often used in distinction from fission 

products. For example, in decommissioning waste comprising structural materials from a nuclear 

facility, activation products might typically be found primarily within the matrix of the material, 

whereas fission products are more likely to be present in the form of contamination on surfaces. 

activity. The quantity A for an amount of radionuclide in a given energy state at a given time. 

The SI unit of activity is the reciprocal second (s–1), termed the Becquerel (Bq). Formerly 

expressed in curie (Ci), which is still sometimes used. 

activity, specific. Of a radionuclide, the activity per unit mass of that nuclide. Of a material, the 

activity per unit mass or volume of the material in which the radionuclides are essentially 

uniformly distributed. 

ALARP & ALARA. As low as reasonably practicable. As low as reasonably achievable. ALARP 

& ALARA describe approaches to optimisation. The optimisation principle states “in relation to 

any particular source within a practice, the magnitude of individual doses, the number of people 

exposed, and the likelihood of incurring exposures where these are not certain to be received 

should all be kept as low as reasonably achievable (ALARA), economic and social factors being 

taken into account…” ALARA is incorporated in UK law via RSA 1993 (BSS) Direction 2000. 

ALARA & ALARP focus on impacts to people. 

alpha bearing waste. See waste, alpha bearing. 

analysis. Often used interchangeably with assessment, especially in more specific terms such as 

safety analysis. In general, however, analysis suggests a more narrowly technical process than 

assessment, aimed at understanding the subject of the analysis rather than determining whether 

or not it is acceptable. Analysis is also often associated with the use of a specific technique. 

Hence, one or more forms of analysis may be used in assessment. 

analysis, consequence. A safety analysis that estimates potential individual or collective 

radiation doses to humans on the basis of radionuclide releases and transport from a nuclear 

facility (e.g. a waste storage facility or disposal site) to the human environment as defined by 

hypothetical release and transport scenarios. 

analysis, deterministic. A simulation of the behaviour of a system utilizing one set of 

parameters, events and features. See also analysis, probabilistic. 

analysis, probabilistic. A simulation of the behaviour of a system defined by parameters, events 

and features whose values are represented by a statistical distribution. The analysis gives a 

corresponding distribution of results. See also analysis, deterministic. 

analysis, risk. An analysis of possible events and their probabilities of occurrence together with 

their potential consequences. 

analysis, safety. An evaluation of the potential hazards associated with the implementation of a 

proposed activity. 




July 2009 

121 


analysis, sensitivity. A quantitative examination of how the behaviour of a simulated system 

(e.g. a computer model) varies with change, usually in the values of its parameters. Two common 

approaches used are: parameter variation, in which the variation of the results is investigated for 

changes in one or more input parameter values within a range around selected reference or 

mean values, and perturbation analysis, in which the variations of the results with respect to 

changes in all the input parameter values are obtained by applying differential, integral or 

probabilistic analysis. 

analysis, uncertainty. An analysis of the amount of variation in the results of assessments or 

analyses due to incomplete knowledge about the current and future states of a system. 

aquifer. A water bearing formation below the surface of the earth that can furnish an appreciable 

supply of water for a well or spring. 

area, controlled. A defined area in which specific protection measures and safety provisions are 

or could be required for controlling normal exposures or preventing the spread of contamination 

during normal working conditions, and preventing or limiting the extent of potential exposures. 

argillaceous. The term applied to all rocks and substances composed of clay or having a notable 

proportion of clay in their composition. 

assessment. The process, and the result, of analysing systematically the hazards associated 

with sources and practices, and associated protection and safety measures, aimed at quantifying 

performance measures for comparison with criteria. Assessment should be distinguished from 

analysis. Assessment is aimed at providing information that forms the basis of a decision whether 

something is satisfactory or not. Various kinds of analysis may be used as tools in doing this. 

Hence an assessment may include a number of analyses. 

assessment, consequence. An assessment of the radiological consequences (e.g. doses and 

activity concentrations) of normal operation and possible accidents associated with a proposed or 

authorized facility or part thereof. This differs from risk assessment in that probabilities are not 

included in the assessment. 

assessment, environmental (impact). An evaluation of radiological and nonradiological impacts 

of a proposed activity, where the performance measure is overall environmental impact, including 

radiological and other global measures of impact on safety and environment. 

assessment, performance. An assessment of the performance of a system or subsystem and 

its implications for protection and safety at a planned or an authorized facility. This differs from 

safety assessment in that it can be applied to parts of a facility, and does not necessarily require 

assessment of radiological impacts. 

assessment, risk. An assessment of the radiological risks associated with normal operation and 

potential accidents involving a source or practice. This will normally include consequence 

assessment and associated probabilities. 

assessment, safety. An analysis to evaluate the performance of an overall system and its 

impact, where the performance measure is radiological impact or some other global measure of 

impact on safety. See also assessment, performance. 

attribute. In the context of multi attribute decision aiding, attributes are features that the options 

possess which can be used to distinguish between the options in terms of advantages and 

disadvantages. For example, when choosing between types of lawnmower attributes might be; 

price, colour, weight, power source, fineness of cut, safety etc. 

audit. A documented activity performed to determine by investigation, examination and 

evaluation of objective evidence the adequacy of, and adherence to, established procedures, 

instructions, specifications, codes, standards, administrative or operational programmes and 

other applicable documents, and the effectiveness of implementation. 




July 2009 

122 


authorization. The granting by a regulatory body or other governmental body of written 

permission for an operator to perform specified activities. Authorization could include, for 

example, licensing, certification and registration. See also licence. 

backfill. The material used to refill excavated portions of a repository (drifts, disposal rooms or 

boreholes) during and after waste has been emplaced. 

background (radiation). The dose, dose rate or an observed measure related to the dose or 

dose rate, attributable to all sources other than the one(s) specified. 

barrier. A physical obstruction that prevents or delays the movement of radionuclides or other 

material between components in a system, for example a waste repository. In general, a barrier 

can be an engineered barrier which is constructed or a natural (or geological) barrier. 

barrier, intrusion. The components of a repository designed to prevent inadvertent access to the 

waste by humans, animals and plants. 

barriers, multiple. Two or more natural or engineered barriers used to isolate radioactive waste 

in, and prevent radionuclide migration from, a repository. See also barrier. 

borehole. A cylindrical excavation, made by a drilling device. Boreholes are drilled during site 

investigation and testing and are also used for waste emplacement in repositories and 

monitoring. 

BPEO. Best Practicable Environmental Option. The outcome of a systematic and consultative 

decision-making procedure which emphasises the protection and conservation of the 

environment across land, air and water. The BPEO procedure establishes, for a given set of 

objectives, the option that provides the most benefits or the least damage to the environment as a 

whole, at acceptable cost, in the long term as well as the short term. 

Bq/g A Becquerel (abbreviated as Bq) is the International System (SI) unit for the activity of 

radioactive material. One Bq of radioactive material is that amount of material in which one atom 

is transformed or undergoes one disintegration every second. A Gram (abbreviated as g) is a 

unit of mass. A Becquerel per Gram (abbreviated Bq/g) is therefore a measure of the 

concentration of radioactivity in a material. 

characterization, site. Detailed surface and subsurface investigations and activities at candidate 

disposal sites to obtain information to determine the suitability of the site for a repository and to 

evaluate the long term performance of a repository at the site. 

characterization, waste. Determination of the physical, chemical and radiological properties of 

the waste to establish the need for further adjustment, treatment, conditioning, or its suitability for 

further handling, processing, storage or disposal. 

clay. Minerals that are essentially hydrated aluminium silicates or occasionally hydrated 

magnesium silicates, with sodium, calcium, potassium and magnesium cations. Also denotes a 

natural material with plastic properties which is essentially a composition of fine to very fine clay 

particles. Clays differ greatly mineralogically and chemically and consequently in their physical 

properties. Because of their large surface areas, most of them have good sorption characteristics. 

cleanup. Any measures that may be carried out to reduce the radiation exposure from existing 

contamination through actions applied to the contamination itself (the source) or to the exposure 

pathways to humans. In a radioactive waste management context, cleanup has essentially the 

same meaning as rehabilitation, remediation and restoration. 

clearance. Removal of radioactive materials or radioactive objects within authorized practices 

from any further regulatory control by the regulatory body. 

clearance level. See level, clearance. 




July 2009 

123 


closure. Administrative and technical actions directed at a repository at the end of its operating 

lifetime — for example covering the disposed waste (for a near surface repository) or backfilling 

and/or sealing (for a geological repository and the passages leading to it) — and termination and 

completion of activities in any associated structures. 

commissioning. The process during which systems and components of facilities and activities, 

having been constructed, are made operational and verified to be in accordance with design 

specifications and to have met the required performance criteria. Commissioning may include 

both non-radioactive and radioactive testing. 

compaction. A treatment method where the bulk volume of a compressible material is reduced 

by application of external pressure — hence an increase in its density (mass per unit volume). 

conditioning. Those operations that produce a waste package suitable for handling, transport, 

storage and/or disposal. Conditioning may include the conversion of the waste to a solid waste 

form, enclosure of the waste in containers, and, if necessary, providing an overpack. See also 

immobilization. 

conductivity, hydraulic, K. Ratio of flow rate n to driving force dh/dl (the change of hydraulic 

head with distance) for viscous flow of a fluid in a porous medium. This is the so-called constant 

of proportionality K in Darcy’s law and depends on both the porous medium and the fluid 

properties. See also permeability. 

container, waste. The vessel into which the waste form is placed for handling, transport, storage 

and/or eventual disposal; also the outer barrier protecting the waste from external intrusions. The 

waste container is a component of the waste package. See also barrier; cask; waste package. 

containment. Methods or physical structures designed to prevent the dispersion of radioactive 

substances. 

contamination. (1) Radioactive substances on surfaces, or within solids, liquids or gases 

(including the human body), where their presence is unintended or undesirable, (2) the presence 

of such substances in such places or (3) the process giving rise to their presence in such places. 

contamination, fixed. Contamination other than non-fixed contamination. 

contamination, non-fixed. Contamination that can be removed from a surface during any 

handling activities, including routine conditions of transport. 

control, institutional. Control of a waste site by an authority or institution designated under the 

laws of a country. This control may be active (monitoring, surveillance and remedial work) or 

passive (land use control) and may be a factor in the design of a nuclear facility (e.g. a near 

surface repository). 

control, regulatory. Any form of control applied to facilities or activities by a regulatory body for 

reasons related to protection or safety. 

controlled area. See area, controlled. 

cover. A layer of material or materials placed over the waste packages or physical structures in a 

near surface repository. The main purpose of covers is to prevent ingress of surface water into 

the repositories and to reduce the likelihood of intrusion. 

criteria. Conditions on which a decision or judgement can be based. They may be qualitative or 

quantitative and should result from established principles and standards. See also requirement; 

specifications. 

critical group. A group of members of the public which is reasonably homogeneous with respect 

to its exposure for a given radiation source and given exposure pathway and is typical of 




July 2009 

124 


individuals receiving the highest effective dose or equivalent dose (as applicable) by the given 

exposure pathway from the given source. 

critical pathway. The dominant environmental route by which members of the critical group are 

exposed to radiation. For example, the critical pathway for iodine discharged with gaseous 

effluents is from pasture to cows and then to milk. Consumption of the milk by individuals gives 

rise to exposure to radiation. 

decommissioning. Administrative and technical actions taken to allow the removal of some or all 

of the regulatory controls from a facility. This does not apply to a repository or to certain nuclear 

facilities used for mining and milling of radioactive materials, for which closure is used. 

decontamination. The complete or partial removal of contamination by a deliberate physical, 

chemical or biological process. 

depleted uranium. See uranium, depleted. 

design. The process and result of developing a concept, detailed plans, supporting calculations 

and specifications for a facility and its parts. 

desorption. See sorption. 

deterministic analysis. See analysis, deterministic. 

diffusion. The movement of atoms or molecules from a region of higher concentration of the 

diffusing species to regions of lower concentration, due to a concentration gradient. 

discharge. A planned and controlled release of (usually gaseous or liquid) radioactive material to 

the environment. 

discharge, authorized. A discharge in accordance with an authorization. See limit, authorized. 

discharges, radioactive. Radioactive substances arising from a source within a practice which 

are discharged to the environment, generally with the purpose of dilution and dispersion. 

disintegration per second. See also Bq/g. A Disintegration is any nuclear transformation that 

emits radiation. Radiation is energy in transit in the form of high speed particles and 

electromagnetic waves. We encounter electromagnetic waves every day. They make up our 

visible light, radio and television waves, ultra violet (UV), and microwaves with a large spectrum 

of energies. These examples of electromagnetic waves do not cause ionizations of atoms 

because they do not carry enough energy to separate molecules or remove electrons from atoms. 

LLW is a radioactive waste because it can emit ionizing radiation. Ionizing radiation is radiation 

with enough energy so that during an interaction with an atom, it can remove tightly bound 

electrons from their orbits, causing the atom to become charged or ionized. Examples are gamma 

rays and neutrons. 

disposal. Emplacement of waste in an appropriate facility without the intention of retrieval. Some 

countries use the term disposal to include discharges of effluents to the environment. 

disposal, near surface. See repository, near surface. 

disposal, on-site. Disposal of the nuclear facility or portions thereof within the nuclear site 

boundary. It includes in situ disposal (entombment) where the nuclear facility is disposed wholly 

or partly at its existing location; or on-site transfer and disposal where the nuclear facility or 

portions thereof are moved to a repository at an adjacent location on the site. 

disposal facility. Synonymous with repository. 

distribution coefficient, Kd. The ratio of the amount of substance sorbed on a unit mass of dry 

solid to the concentration of the substance in a solution in contact with the solid, assuming 

equilibrium conditions. The SI units are: m3/kg. 




July 2009 

125 


dose. A measure of the energy deposited by radiation in a target. Absorbed dose, committed 

equivalent dose, committed effective dose, effective dose, equivalent dose or organ dose, 

depending on the context. All these quantities have the dimensions of energy divided by mass. 

dose, absorbed, D. The fundamental dosimetric quantity D. The unit is J/kg, termed the gray 

(Gy). 

dose constraint. A prospective and source related restriction on the individual dose from a 

source, which provides a basic level of protection for the most highly exposed individuals from a 

source and serves as an upper bound on the dose in optimization of protection for that source. 

The UK government has set a maximum dose constraint value of 0.3 mSv/year when determining 

applications for discharge authorization from a single new source. 

dose, effective, E. A summation of the tissue equivalent doses, each multiplied by the 

appropriate tissue weighting factor: The unit of effective dose is J/kg, with the special name 

sievert (Sv). The committed effective dose is the effective dose that will be received by the 

person over their lifetime as a result of radionuclides taken into the body e.g. by ingestion or 

inhalation. 

dose, equivalent, HT. The radiation-weighted dose in a tissue or organ. This takes account of 

the different amounts of damage caused by different types of radiation eg alpha particles, gamma 

radiation. The unit of equivalent dose is J/kg, termed sievert (Sv). 

dose limit. See limit, dose. The value of the effective dose or the equivalent dose to individuals 

from planned exposure situations that shall not be exceeded. For the purposes of discharge 

authorizations, the UK has (since 1986) applied a dose limit of 1 mSv/year to members of the 

public from all man-made sources of radioactivity (other than from medical applications). 

effluent. Gaseous or liquid radioactive materials which are discharged to the environment. See 

also discharge, authorized. 

emanation. Generation of radioactive gas by the decay of a radioactive solid. 

engineered barrier. See barrier. 

environmental (impact) assessment. See assessment, environmental (impact). 

environmental impact statement. A set of documents recording the results of an evaluation of 

the physical, ecological, cultural and socioeconomic effects of a planned facility (e.g. a repository) 

or of a new technology. 

environmental monitoring. See monitoring, environmental. 

equivalent dose. See dose, equivalent. 

exempt waste. See waste, exempt. 

exemption. The determination by a regulatory body that a source or practice need not be subject 

to some or all aspects of regulatory control on the basis that the exposure (including potential 

exposure) due to the source or practice is too small to warrant the application of those aspects. 

See also level, clearance. 

exemption & exclusion. A number of exemption orders have been made under RSA 1993 

which specify the conditions under which materials or wastes, which are defined as radioactive 

under the Act, can be made Exempt or excluded from some or all provisions of the Act. An 

important exemption order is the Substances of Low Activity (SoLA) Exemption Order. SoLA 

establishes a limit of 0.4 Bq/g for certain radioactive wastes that in effect is the limit below which 

wastes are not treated specifically as a radioactive waste for purposes of disposal. The 

Phosphatic Substances, Rare Earths etc. Exemption Order and the Uranium and Thorium 

Exemption Order are also used to set the practical lower boundaries of what becomes LLW. 




July 2009 

126 


exposure. The act or condition of being subject to irradiation. Exposure can either be external 

exposure due to sources outside the body or internal exposure due to sources inside the body. 

exposure, normal. Exposure which is expected to occur under the normal operating conditions 

of a facility or activity, including possible minor mishaps that can be kept under control, i.e. during 

normal operation and anticipated operational occurrences. 

exposure, potential. Exposure that is not expected to occur with certainty but that may result 

from an accident at a source or owing to an event or sequence of events of a probabilistic nature, 

including equipment failures and operating errors. 

exposure pathway. A route by which radiation or radionuclides can reach humans and cause 

exposure. An exposure pathway may be very simple, for example external exposure from 

airborne radionuclides, or involve a more complex chain, for example internal exposure from 

drinking milk from cows that ate grass contaminated with deposited radionuclides. 

facility. See nuclear facility. 

fissile material. Uranium-233, uranium-235, plutonium-239, plutonium-241, or any combination 

of these radionuclides. Excepted from this definition is: (a) natural uranium or depleted uranium 

which is unirradiated, (b) natural uranium or depleted uranium which has been irradiated in 

thermal reactors only. 

fission product. A radionuclide produced by nuclear fission. 

fixed contamination. See contamination, fixed. 

flow, unsaturated. The flow of water in unsaturated soil by capillary action and gravity. 

fracture. A general term for any breaks in rock whether or not it causes displacement. 

fuel, nuclear. Fissionable and fertile material used in a nuclear reactor for the purpose of 

generating energy. 

geological barrier. See barrier. 

gradient, hydraulic. The change in total hydraulic head per unit distance of flow in a given 

direction. 

groundwater. Water that is held in rocks and soil beneath the surface of the earth. 

half-life, T1/2. The time taken for the quantity of a specified material (e.g. a radionuclide) in a 

specified place to decrease by half as a result of any specified process or processes that follow 

similar exponential patterns to radioactive decay. 

half-life, effective, Teff. The time taken for the activity of a radionuclide in a specified place to 

halve as a result of all relevant processes. 

half-life, radioactive. For a radionuclide, the time required for the activity to decrease, by a 

radioactive decay process, by half. 

Harwell. The UKAEA Harwell site in Oxfordshire is an ex-RAF WWII airbase that has been used 

since 1946 for nuclear research, mainly in support of civilian power generation. The site is now 

well advanced with decommissioning. The aim is to return the site to a delicensed status by 

2025. 

HVLA Waste. High Volume Very Low Level Activity Waste. See main text. 

HV-VLLW. High volume very low level waste. A sub-category of LLW as defined in “Policy for the 

Long Term Management of Solid Low Level Radioactive Waste in the United Kingdom” (DEFRA, 

2007). 




July 2009 

127 


HPA (NRPB) The Health Protection Agency (HPA) is an independent body that protects the 

health and well-being of the population. The HPA includes the ex-National Radiological 

Protection Board (NRPB). 

HSE. Britain's Health and Safety Commission (HSC) and the Health and Safety Executive (HSE) 

are responsible for the regulation of almost all the risks to health and safety arising from work 

activity in Britain. 

hydraulic conductivity, K. See conductivity, hydraulic. 

hydraulic gradient. See gradient, hydraulic. 

hydraulic transmissivity. See transmissivity, hydraulic. 

inadvertent human intrusion. Accidental intrusion into a disposal facility without prior 

knowledge of the presence of the facility or accidental intrusion, without prior knowledge, into an 

area adjacent to the facility in such a way that it degrades the environmental safety performance 

of the facility. 

immobilization. Conversion of waste into a waste form by solidification, embedding or 

encapsulation. The aim is to reduce the potential for migration or dispersion of radionuclides 

during handling, transport, storage and/or disposal. See also conditioning. 

inert waste. Material which does not undergo any significant physical, chemical or biological 

transformations; does not dissolve, burn or otherwise physically or chemically react, biodegrade 

or adversely affect other matter with which it comes into contact in a way likely to give rise to 

environmental pollution or harm to human health; and whose total leachability and pollutant 

content and the ecotoxicity of its leachate are insignificant and in particular do not endanger the 

quality of any surface water or groundwater. This is defined by UK waste legislation for non 

radioactive wastes. 

in situ disposal. See disposal, on-site. 

infiltration. The downward entry of water through the ground surface into soil or rock. 

institutional control. See control, institutional. 

intervention. Any action intended to reduce or avert exposure or the likelihood of exposure to 

sources which are not part of a controlled practice or which are out of control as a consequence 

of an accident. 

leach rate. The rate of dissolution or erosion of material or the release by diffusion from a solid, 

this is hence a measure of how rapidly radionuclides may be released from that material. The 

term usually refers to the durability of a solid waste form but also describes the removal of sorbed 

material from the surface of a solid or porous bed. 

leach test. A test conducted to determine the leach rate of a waste form. The test results may be 

used for judging and comparing different types of waste forms, or may serve as input data for a 

long term safety assessment of a repository. Many different test parameters have to be taken into 

account, for example water composition and temperature. 

leachate. A solution that has been in contact with waste form and, as a result, may contain 

radionuclides. 

level, clearance. A value, established by a regulatory body and expressed in terms of activity 

concentration and/or total activity, at or below which a source of radiation may be released from 

regulatory control. See also clearance. 

level, exemption. A value, established by a regulatory body and expressed in terms of activity 

concentration and/or total activity, at or below which a source of radiation may be granted 

exemption from regulatory control without further consideration. 




July 2009 

128 


licence. A legal document issued by the regulatory body granting authorization to perform 

specified activities related to a facility or activity. The holder of a current licence is termed a 

licensee. A licence is a product of the authorization process, although the term licensing process 

is sometimes used. 

limit, authorized. A limit on a measurable quantity, established or formally accepted by a 

regulatory body. Authorized limit has been commonly used particularly in the context of limits on 

discharges. See also discharge, authorized. 

limit, dose. The value of the effective dose or the equivalent dose to individuals from controlled 

practices that shall not be exceeded. 

liner. (1) A layer of material placed between a waste form and a container to resist corrosion or 

any other degradation of a waste package. (2) A layer of clay, plaster, asphalt or other 

impermeable material placed around or beneath a repository or tailings impoundment to prevent 

leakage and/or erosion. (3) A structural component (made, for example, of concrete or steel) on 

the surface of a tunnel or shaft in a repository. 

LLW. See waste, low and intermediate level. Low Level Radioactive Waste. With certain specific 

exceptions, LLW is defined as waste which has an activity concentration in the range 0.4 – 4,000 

Bq/g for alpha emitters and 12,000 Bq/g for beta-gamma emitters. Where Bq/g is Becquerel per 

gram, a measure of activity within the SI system equivalent to 1 disintegration per second. Where 

an alpha emitter is a form of radioactive decay involving emission of alpha particles (a helium 

nucleus). Where beta decay is a type of radioactive decay involving the emission of electrons or 

positrons. 

Low Level Waste Repository LLWR (Drigg LLW facility). The Drigg site, located 6 km southeast 

of Sellafield, has operated safely for over 40 years disposing of Low Level Radioactive 

Wastes (LLW) from the nuclear and general industries, universities and hospitals. Drigg is 

operated by BN-GS (ex.British Nuclear Fuels Limited (BNFL)). 

long lived waste. See waste, long lived. 

long term. In radioactive waste disposal, refers to periods of time which exceed the time during 

which active institutional control can be expected to last. 

long term stewardship. Conducting, supervising, or managing something entrusted to one's 

care. In the context of nuclear waste sites the phrase encompasses the activities undertaken 

after closure of the site to maintain and monitor the wastes in the long term. 

low level waste (LLW). See waste, low and intermediate level. 

LSG. Local Stakeholder Group. A group of stakeholders that meet regularly in relation to a 

nuclear licensed site. 

Isotope. Different forms of atoms of the same element that have different numbers of neutrons in 

their nuclei. An element may have a number of isotopes. For example, the three isotopes of 

hydrogen are protium, deuterium, and tritium. All three have one proton in their nuclei, but 

deuterium also has one neutron, and tritium has two neutrons. Different isotopes can have 

different radioactive properties and present different risks. 

migration. The movement of radionuclides in the environment as a result of natural processes. 

minimization, waste. The process of reducing the amount and activity of radioactive waste to a 

level as low as reasonably achievable, at all stages from the design of a facility or activity to 

decommissioning, by reducing waste generation and by means such as recycling and reuse, and 

treatment, with due consideration for secondary as well as primary waste. See also pretreatment; 

treatment; volume reduction. 




July 2009 

129 


model. A representation of a system and the ways in which phenomena occur within that system, 

used to simulate or assess the behaviour of the system for a defined purpose. 

model, computational. A calculation tool that implements a mathematical model. 

model, conceptual. A set of qualitative assumptions used to describe a system. 

model, mathematical. A set of mathematical equations designed to represent a conceptual 

model. 

model, pathways. A mathematical representation used to simulate the transport of radionuclides 

from a source to a receptor. 

model, transport. A mathematical representation of mechanisms controlling the movement of 

finely dispersed or dissolved substances in fluids. 

monitoring. Continuous or periodic measurement of radiological and other parameters or 

determination of the status of a system. 

monitoring, environmental. The measurement and evaluation of external dose rates due to 

sources in the environment or of radionuclide concentrations in the environmental media. 

naturally occurring radioactive material (NORM). Material containing no significant amounts of 

radionuclides other than naturally occurring radionuclides. The exact definition of ‘significant 

amounts’ would be a regulatory decision. Materials in which the activity concentrations of the 

naturally occurring radionuclides have been changed by human made processes are included. 

These are sometimes referred to as technically enhanced NORM or TENORM. 

naturally occurring radionuclides. Radionuclides that occur naturally in significant quantities on 

earth. The term is usually used to refer to the primordial radionuclides potassium-40, uranium- 

235, uranium-238 and thorium-232 (the decay product of primordial uranium-236), their 

radioactive decay products, and tritium and carbon-14 generated by natural activation processes. 

NDA. Nuclear Decommissioning Authority. A public body that oversees nuclear 

decommissioning in the UK on designated sites such as Harwell. 

near surface disposal. See repository, near surface. 

nuclear facility. A facility and its associated land, buildings and equipment in which radioactive 

materials are produced, processed, used, handled, stored or disposed of on such a scale that 

consideration of safety is required. 

nuclear material. Plutonium except that with isotopic concentration exceeding 80% in plutonium- 

238; uranium-233; uranium enriched in the isotope 235 or 233; uranium containing the mixture of 

isotopes occurring in nature other than in the form of ore or ore residue; any material containing 

one or more of the foregoing. 

nuclear waste. See waste, radioactive. 

NII. Nuclear Installations Inspectorate. Under UK law (the Health and Safety at Work etc. Act 

1974) employers are responsible for ensuring the safety of their workers and the public, and this 

is just as true for a nuclear site as for any other. This responsibility is reinforced for nuclear 

installations by the Nuclear Installations Act 1965 (NIA), as amended. Under the relevant 

statutory provisions of the NIA, a site cannot have nuclear plant on it unless the user has been 

granted a site licence by the Health and Safety Executive (HSE). This licensing function is 

administered on HSE's behalf by its Nuclear Safety Directorate (NSD). 

nuclear site licence. A licence issued under the Nuclear Installations Act (see NII). 

off-site. Outside the physical boundary of a site. 

on-site. Within the physical boundary of a site. 




July 2009 

130 


on-site disposal. See disposal, on-site. 

operation. All the activities performed to achieve the purpose for which a facility was constructed. 

operational period. The period during which a nuclear facility (e.g. a repository) is being used for 

its intended purpose until it is decommissioned or is submitted for permanent closure. 

optimization. The process of determining what level of protection and safety makes exposures, 

and the probability and magnitude of potential exposures, ‘as low as reasonably achievable, 

economic and social factors being taken into account’ (ALARA). 

overpack. A secondary (or additional) outer container for one or more waste packages, used for 

handling, transport, storage or disposal. 

package, waste. The product of conditioning that includes the waste form and any container(s) 

and internal barriers (e.g. absorbing materials and liners), prepared in accordance with the 

requirements for handling, transport, storage and/or disposal. 

permeability, k. The ability of a porous medium to transmit fluid. 

plume. The spatial distribution of a release of airborne or waterborne material as it disperses in 

the environment. 

porosity. The ratio of the aggregate volume of interstices in rock, soil or other porous media to its 

total volume. 

post-closure period. The period of time following the closure of a repository and 

decommissioning of related surface facilities. Some type of surveillance or control will probably be 

maintained in this period, particularly for near surface repositories. See also closure; preclosure 

period. 

practice. Any human activity that introduces additional sources of exposure or exposure 

pathways or extends exposure to additional people or modifies the network of exposure pathways 

from existing sources, so as to increase the exposure or the likelihood of exposure of people or 

the number of people exposed. 

preclosure period. The period of time spanning the construction and operation of a repository up 

to and including the closure and decommissioning of related surface facilities. See also closure; 

post-closure period. 

predisposal. Any radioactive waste management steps carried out prior to disposal, such as 

pretreatment, treatment, conditioning, storage and transport activities. Decommissioning is 

considered to be a part of predisposal management of radioactive waste. 

pretreatment. Any or all of the operations prior to waste treatment, such as collection, 

segregation, chemical adjustment and decontamination. 

quality assurance (QA). Planned and systematic actions necessary to provide adequate 

confidence that an item, process or service will satisfy given requirements for quality, for example 

those specified in the licence. 

quality control (QC). The part of quality assurance intended to verify that systems and 

components correspond to predetermined requirements. 

radioactive contamination. See contamination. 

radioactive discharges. See discharges, radioactive. 

radioactive effluent. See effluent. 

radioactive half-life. See half-life, radioactive. 




July 2009 

131 


adioactive material. Material designated in national law or by a regulatory body as being 

subject to regulatory control because of its radioactivity. 

radioactive waste. See waste, radioactive. Low activity solid radioactive wastes are taken to 

include all wastes with an activity level lying below the defined Low Level Waste (LLW) category 

upper limit, but above either the levels specified for exclusion from the provisions of the 

Radioactive Substances Act 1993 (RSA93) or for exemption from specific regulatory action under 

the Act as a result of the Substances of Low Activity (SoLA) Exemption Order. This range 

includes, at the lower end, an officially recognised waste category termed Very Low Level Waste 

(VLLW). 

Low Level Waste (LLW) is a waste containing radioactive materials other than those suitable for 

disposal with ordinary refuse, but not exceeding 4GBq/te (gigabecquerels/tonne) of alpha or 12 

GBq/te of beta/gamma activity; i.e., wastes that can normally be accepted for authorised disposal 

at Drigg, Dounreay or other engineered landfill sites. 

Very Low Level Waste (VLLW) is a waste that can be disposed of with ordinary refuse, each 0.1 

cubic metre (m3) of material containing less than 400kBq (kilobecquerels) of beta/gamma activity 

or single items containing less than 40kBq. In the application of the VLLW upper threshold, there 

are separate, complementary, restrictions on the permissible content of carbon-14 and tritium; 

these are a factor of ten greater. VLLW disposal was originally intended for small volumes and is 

also known as “dustbin” disposal. 

In practice, there are other streams of low activity solid radioactive waste that are disposed of to 

routes other than Drigg and dustbin disposal. These waste streams are associated with landfill 

disposal, in-situ burial on licensed nuclear sites, and incineration. The waste streams deemed 

suitable for landfill or in-situ burial are generally characterised by radioactivity levels well below 

the defined LLW upper activity threshold, and by the fact that they may arise in large volumes. 

Incineration is essentially treatment of LLW and VLLW prior to landfill disposal of the secondary 

incineration products (hearth ash and gas cleaning residues) as VLLW dustbin disposal or 

exempt wastes. 

Landfill disposal processes for LLW were developed for those wastes arising principally in the 

non-nuclear sector which were above the limits for dustbin disposal and unsuitable for 

incineration. The activity limit is typically above VLLW, but well below the LLW upper bound. The 

development of this route depended on the availability of suitable landfill sites with good 

containment characteristics that had been subject to an environmental assessment satisfying the 

regulators that public safety was assured, and to an ongoing leachate monitoring programme 

carried out by the regulators. Disposal of LLW is subject to issue of an authorization under 

RSA93 by the regulators. 

radioactive waste management. See waste management, radioactive. 

radioactivity. The phenomenon whereby atoms undergo spontaneous random disintegration, 

usually accompanied by the emission of radiation. 

radiological survey. See survey, radiological. 

radionuclide. A nucleus (of an atom) that possesses properties of spontaneous disintegration 

(radioactivity). Nuclei are distinguished by their mass and atomic number. 

records. A set of documents, such as instrument charts, certificates, log books, computer 

printouts and magnetic tapes for each nuclear facility, organized in such a way that it provides 

past and present representations of facility operations and activities including all phases from 

design through closure and decommissioning (if the facility has been decommissioned). Records 

are an essential part of quality assurance. 




July 2009 

132 


egulatory body. An authority or a system of authorities designated by the government of a State 

as having legal authority for conducting the regulatory process, including issuing authorizations, 

and thereby for regulating the siting, design, construction, commissioning, operation, closure, 

decommissioning and, if required, subsequent institutional control of the nuclear facilities (e.g. 

near surface repositories) or specific aspects thereof. 

release. See discharge. 

remedial action. Action taken when a specified action level is exceeded, to reduce a radiation 

dose that might otherwise be received, in an intervention situation involving chronic exposure. 

Examples are: (a) actions which include decontamination, waste removal and environmental 

restoration of a site during decommissioning and/or closure efforts; (b) actions taken beyond 

stabilization of tailings impoundments to allow for other uses of the area or to restore the area to 

near pristine conditions. 

remediation. See cleanup. 

repository. A nuclear facility where waste is emplaced for disposal. 

repository, near surface. A facility for disposal of radioactive waste located at or within a few 

tens of metres from the earth’s surface. 

restoration. See cleanup. 

retardation. A reduction in the rate of radionuclide movement through the soil due to the 

interaction (e.g. by sorption) with an immobile matrix. 

retardation coefficient, Rd. A measure of capability of porous media to impede the movement of 

a particular radionuclide being carried by fluid. 

retrievability. The ability to remove waste from where it has been emplaced. 

risk. A multiattribute quantity expressing hazard, danger or chance of harmful or injurious 

consequences associated with actual or potential exposures. It relates to quantities such as the 

probability that specific deleterious consequences may arise and the magnitude and character of 

such consequences. (2) The combination of the frequency, or probability, of occurrence and the 

consequence of a specified hazardous event. The concept of risk always has two elements: the 

frequency or probability with which a hazardous event occurs and the consequences of the 

hazardous event. Risk = Probability x Consequence. 

risk assessment. See assessment, risk. 

rock. In geology, any mass of mineral matter, whether consolidated or not, which forms part of 

the earth’s crust. Rocks may consist of only one mineral species, in which case they are called 

monomineralic but they usually consist of several mineral species. 

RWMAC The Radioactive Waste Management Advisory Committee (RWMAC) was established 

in 1978 to offer independent advice to Ministers on radioactive waste management issues. 

Members of the Committee were drawn from a wide range of backgrounds and specialisms 

including radioactive waste management, radiological protection, earth sciences, environmental 

law & planning, medical physics and social sciences. Each year until 2004, RWMAC undertook 

a programme of work commissioned by Government Ministers. 

safety case. An integrated collection of arguments and evidence to demonstrate the safety of a 

facility. This will normally include a safety assessment, but could also typically include information 

(including supporting evidence and reasoning) on the robustness and reliability of the safety 

assessment and the assumptions made therein. 




July 2009 

133 


safety culture. The assembly of characteristics and attitudes in organizations and individuals 

which establishes that, as an overriding priority, protection and safety issues receive the attention 

warranted by their significance. 

safety report. A document required from the operating organization by the regulatory body 

containing information concerning a nuclear facility (e.g. a repository), the site characteristics, 

design, operational procedures, etc., together with a safety analysis and details of any provisions 

needed to restrict risk to personnel and the public. 

saturated zone. See zone, saturated. 

scenario. A postulated or assumed set of conditions and/or events. They are most commonly 

used in analysis or assessment to represent possible future conditions and/or events to be 

modelled, such as possible accidents at a nuclear facility, or the possible future evolution of a 

repository and its surroundings. 

screening. A type of analysis aimed at eliminating from further consideration factors that are less 

significant for the purpose of the analysis, in order to concentrate on the more significant factors. 

Screening is usually conducted at an early stage in order to narrow the range of factors needing 

detailed consideration in an analysis or assessment. 

secondary waste. See waste, secondary. 

segregation. An activity where waste or materials (radioactive and exempt) are separated or are 

kept separate according to radiological, chemical and/or physical properties which will facilitate 

waste handling and/or processing. For example, it may be possible to segregate radioactive from 

exempt material and thus reduce the waste volume. 

sensitivity analysis. See analysis, sensitivity. 

shielding. A material interposed between a source of radiation and persons, or equipment or 

other objects, in order to absorb radiation and thereby reduce radiation exposure. 

short lived waste. See waste, short lived. 

site. The area containing, or under investigation for its suitability for, a nuclear facility (e.g. a 

repository). It is defined by a boundary and is under effective control of the operating 

organization. 

site characterization. See characterization, site. 

solidification. Immobilization of gaseous, liquid or liquid-like materials by conversion into a solid 

waste form, usually with the intent of producing a physically stable material that is easier to 

handle and less dispersible. Calcination, drying, cementation, bituminization and vitrification are 

some of the typical ways of solidifying liquid waste. See also conditioning; immobilization. 

solidified waste. See waste, solidified. 

solubility. The amount of a substance that will dissolve in a given amount of another substance. 

The solubility of a waste form or a radionuclide is an important factor in determining the potential 

migration of radionuclides from a disposal area. 

sorption. The interaction of an atom, molecule or particle with the surface of a solid. A general 

term including absorption (sorption taking place largely within the pores of a solid) and adsorption 

(surface sorption with a non-porous solid). The processes involved may also be divided into 

chemisorption (chemical bonding with the substrate) and physisorption (physical attraction, for 

example by weak electrostatic forces). 

source. (1) Anything that may cause radiation exposure, such as by emitting ionizing radiation or 

by releasing radioactive substances or materials. (2) More specifically, radioactive material used 

as a source of radiation. 




July 2009 

134 


source, natural. A naturally occurring source of radiation, such as the sun and stars (sources of 

cosmic radiation) and rocks and soil (terrestrial sources of radiation). 

source term. A mathematical expression used to denote information about the actual or potential 

release of radiation or radioactive material from a given source, which may include further 

specifications, for example the composition, the initial amount, the rate and the mode of release 

of the material. 

specific activity. See activity, specific. 

storage. The holding of spent fuel or of radioactive waste in a facility that provides for its 

containment, with the intention of retrieval (3). Storage is by definition an interim measure, and 

the term interim storage would therefore be appropriate only to refer to short term temporary 

storage when contrasting this with the longer term fate of the waste. Storage as defined above 

should not be described as interim storage. 

surface water. Water which fails to penetrate into the soil and flows along the surface of the 

ground, eventually entering a lake, a river or the sea. 

survey, radiological. An evaluation of the radiological conditions and potential hazards 

associated with the production, use, transfer, release, disposal, or presence of radioactive 

material or other sources of radiation. 

sustainability The concept of meeting the needs of the present without compromising the ability 

of future generations to meet their needs. The term originally applied to natural resource 

situations, where the long term was the focus. Today, it applies to many disciplines, including 

economic development, environment, food production, energy, and social organization. Basically, 

sustainability/sustainable development refers to doing something with the long term in mind. 

transmissivity, hydraulic. The rate at which water is transmitted through a unit width of a water 

conducting feature (e.g. an aquifer) under a unit hydraulic gradient. 

transmutation. The conversion of one element into another. Transmutation is under study as a 

means of converting longer lived radionuclides into shorter lived or stable radionuclides. The term 

actinide burning is used in some countries. 

transport, radionuclide. The movement (migration) of radionuclides in the environment, for 

example radionuclide transport by groundwater. This could include processes such as advection, 

diffusion, sorption and uptake. This usage does not include intentional transport of radioactive 

materials by humans (transport of radioactive wastes in casks, etc). See also migration. 

treatment. Operations intended to benefit safety and/or economy by changing the characteristics 

of the waste. Three basic treatment objectives are: volume reduction, removal of radionuclides 

from the waste and change of composition. Treatment may result in an appropriate waste form. 

UKAEA The United Kingdom Atomic Energy Authority (UKAEA) was incorporated as a statutory 

corporation in 1954 and pioneered the development of nuclear energy in the UK. Today we are 

responsible for managing the decommissioning of the nuclear reactors and other radioactive 

facilities used for the UK's nuclear research and development programme in a safe and 

environmentally sensitive manner. UKAEA is a non-departmental public body, funded mainly by 

its lead department the Department of Trade and Industry under contract to the NDA. 

uptake. A general term for the processes by which radionuclides enter one part of a biological 

system from another. Used in a range of situations, particularly in describing the overall effect 

when there are a number of contributing processes, for example root uptake, the transfer of 

radionuclides from soil to plants through the plant roots. 

uranium, depleted. Uranium containing a lesser mass percentage of uranium-235 than in natural 

uranium. 




July 2009 

135 


uranium, enriched. Uranium containing a greater mass percentage of uranium-235 than 0.72%. 

uranium, natural. Chemically separated uranium containing the naturally occurring distribution of 

uranium isotopes (approximately 99.28% uranium-238 and 0.72% uranium-235 by mass). 

very low level waste (VLLW). See waste, very low level. 

volume reduction. A treatment method that decreases the physical volume of a waste. Volume 

reduction is employed because it is economical and facilitates subsequent handling, storage, 

transport and disposal of the waste. Typical volume reduction methods are mechanical 

compaction, incineration and evaporation. Volume reduction of a given waste results in a 

corresponding increase in radionuclide concentration. The total volume of waste may also be 

reduced through decontamination (with subsequent exemption) or through the avoidance of 

waste generation. See also minimization, waste. 

waste. Material in gaseous, liquid or solid form for which no further use is foreseen. 

waste, alpha bearing. Radioactive waste containing one or more alpha emitting radionuclides. 

Alpha bearing waste can be short lived or long lived. 

waste, exempt. Waste released from regulatory control in accordance with exemption principles. 

See also clearance levels; exemption. 

waste, long lived. Radioactive waste that contains significant levels of radionuclides with halflives 

greater than 30 years. Typical characteristics are long lived radionuclide concentrations 

exceeding limitations for short lived waste. 

waste, low level (LLW). See waste, low and intermediate level. 

waste, mixed. Radioactive waste that also contains non-radioactive toxic or hazardous 

substances. 

waste, radioactive. For legal and regulatory purposes, waste that contains or is contaminated 

with radionuclides at concentrations or activities greater than clearance levels as established by 

the regulatory body. It should be recognized that this definition is purely for regulatory purposes 

and that material with activity concentrations equal to or less than clearance levels is radioactive 

from a physical viewpoint — although the associated radiological hazards are considered 

negligible. 

waste, secondary. A form and quality of waste that results as a by-product from processing of 

waste. 

waste, short lived. Radioactive waste that does not contain significant levels of radionuclides 

with half-lives greater than 30 years. 

waste, very low level (VLLW). Radioactive waste considered suitable by the regulatory body for 

authorized disposal, subject to specified conditions, with ordinary waste in facilities not 

specifically designed for radioactive waste disposal. 

waste acceptance requirements. Quantitative or qualitative criteria specified by the regulatory 

body, or specified by an operator and approved by the regulatory body, for radioactive waste to 

be accepted by the operator of a repository for disposal, or by the operator of a storage facility for 

storage. Waste acceptance requirements might include, for example, restrictions on the activity 

concentration or the total activity of particular radionuclides (or types of radionuclide) in the waste 

or requirements concerning the waste form or waste package. 

waste characterization. See characterization, waste. 

waste form. Waste in its physical and chemical form after treatment and/or conditioning 

(resulting in a solid product) prior to packaging. The waste form is a component of the waste 

package. 




July 2009 

136 


waste generator. The operating organization of a facility or activity that generates waste. See 

also operator. 

waste inventory. Quantity, radionuclides, activity and waste form characteristics of wastes for 

which an operator is responsible. 

waste management, radioactive. All activities, administrative and operational, that are involved 

in the handling, pretreatment, treatment, conditioning, transport, storage and disposal of 

radioactive waste. 

water table. The upper surface of a zone of groundwater saturation. 

zone, saturated. A subsurface zone in which all the interstices are filled with water. This zone is 

separated from the unsaturated zone, i.e. the zone of aeration, by the water table. See also zone, 

unsaturated. 

zone, unsaturated. A subsurface zone in which at least some interstices contain air or water 

vapour, rather than liquid water. Also referred to as the ‘zone of aeration’. See also zone, 

saturated. 




July 2009 

137 


Annexes 

A Radiation, People and the Environment (IAEA, 2004) 

B Suitability Assessment – Galson Sciences 

C ENRMF, IRRs 1999, Radiation Risk Assessment for 

Low Level Waste Disposal, HPA 

D Dose Rate calculations in support of Low Level Waste 

disposal authorisation, TSG(09)0487 

E SNIFFER Methodology Information 

F Copy of Application Form 

G Example Capacity Calculation Layout 

H Calculation of dose rate at landfill, TSG(09)0488 

I Baseline Groundwater and Leachate Sample Results 

J Capability Statements 




July 2009 

138 


Annex A 

Radiation, People and the Environment 

IAEA 

This booklet is provided as a primer for readers who are new to the subject of 

radioactivity. The booklet is not specific to the application. 

The booklet has not been included as a hardcopy version and can be read on 

downloaded at: 

http://www.iaea.org/Publications/Booklets/RadPeopleEnv/radiation_booklet.html 






the 

 

 

 

 

 

 

 


















































































Annex B 

Suitability Assessment – Galson Sciences 


Radiological Assessment for 

Disposal of Solid Low-level 

Radioactive Waste 

at the Landfill at East Northants 

Resource Management Facility 

Galson 

S C I E N C E S L T D 

R D Wilmot & D Reedha 

July 2009 

5 Grosvenor House, Melton Road, Oakham, Rutland LE15 6AX, UK 

Tel: +44 (1572) 770649 Fax: +44 (1572) 770650 www.galson-sciences.co.uk 

0820-2 

Version 2 


Report History 



Radioactive Waste 

at the Landfill at East Northants 


0820-2 

Version 2 

This document has been prepared by Galson Sciences Limited for UKAEA Harwell under the 

terms of Contract No. CF12/07. 

Radiological Assessment for Disposal of Solid Low-level Radioactive Waste at the 

Landfill at East Northants Resource Management Facility 

Version: Date: 

Principal Author: 

R D Wilmot 

Reviewed by: 

D Reedha 

Approved by: 

D A Galson 

Sign Sign Sign 

0820-2 

Version 2 

14 July 2009 

Date 

14 July 2009 

Date 

14 July 2009 

Date 

14 July 2009 


Radiological Assessment 0820-2 

Version 2 

Executive Summary 

The nuclear decommissioning industry has significant future arisings of 

decommissioning wastes that fall into the category of low level waste. In the recently 

published “Policy for the Long Term Management of Solid Low Level Radioactive 

Waste” (Defra 2007), the government has confirmed the acceptability of the disposal 

of LLW to landfill including a new subset of LLW classified as high-volume very low 

level waste. 

Augean plc operates a hazardous waste disposal facility at the East Northants 

Resource Management Facility (ENRMF) in Northamptonshire and proposes that the 

site is used for the disposal of LLW with a specific activity up to 200 Bq / g. This 

report presents a radiological assessment for this site in order to assess the potential 

consequences of disposal and determine the radiological capacity of the site for 

different waste streams. In this report an example waste stream of LLW from 

UKAEA Harwell has been used for illustration. 

This report investigates the suitability of the landfill as a disposal route for LLW with 

a specific activity up to 200 Bq / g, based on the established approaches used 

previously for assessing “Special Precautions Burial” (SPB) which has been used for 

wastes of comparable activity. The methodology has been modified to take account 

of the likely waste volumes and also for certain site-specific aspects that differ from 

the generic assumptions used in the available model. 

The assessment model is a simplified and conservative model of the events and 

processes that will or might take place during and after operations. A number of 

simplifying assumptions are therefore required in order to represent the site and its 

surroundings. These assumptions are outlined in the report and the equations and 

parameter values used in the model are reported. Where there are significant 

uncertainties regarding aspects of the site, a range of assumptions have been used to 

test the sensitivity of the model. 

The report presents specific dose calculations. These are the doses that would be 

received from a disposal of 1MBq of each radionuclide of interest. The specific dose 

depends on the pathway by which radionuclides are released and the time of the 

release. Specific doses have been calculated for the groundwater release, intrusion, 

irradiation and gas release pathways and for pathways relating to leachate 

management. 

Specific doses are used to calculate the capacity of the site for individual 

radionuclides, based on a public dose criterion of 20 Sv / year for normal release 

scenarios and a 3 mSv / year dose criterion for pathways resulting from human 

intrusion. An illustrative overall site radiological capacity has also been calculated 

using preliminary data for the UKAEA Harwell Meashill Trenches waste stream. 

In addition to specific dose calculations for the potential exposure of humans to 

releases from the site, assessments of dose rates have been made for wildlife in the 

vicinity of the site. All of the dose rates calculated, for both terrestrial and the 

Galson Sciences Limited i 14 July 2009 



Version 2 

freshwater ecosystems, show that no organisms or wildlife groups are likely to receive 

dose rates in excess of the internationally recognised criterion of 10 Gy / hour. 

The actual use of the ENRMF for disposal of LLW waste remains subject to 

discussions with the Environment Agency and other stakeholders. 

Galson Sciences Limited ii 14 July 2009 



Version 2 

Contents 

Executive Summary..................................................................................................... i 

1 Introduction............................................................................................................1 

1.1 Project background..........................................................................................1 

1.2 Approach .........................................................................................................1 

1.3 Structure of report............................................................................................2 

2 Background ............................................................................................................3 

2.1 Site...................................................................................................................3 

2.1.1 Design and operations .........................................................................3 

2.1.2 Geology and hydrogeology .................................................................6 

2.1.3 Biosphere and receptors ....................................................................10 

2.2 Wastes............................................................................................................11 

3 Assessment Methodology.....................................................................................14 

3.1 Summary of SNIFFER methodology ............................................................14 

3.1.1 Assessment framework......................................................................14 

3.1.2 Scenarios............................................................................................16 

3.1.3 Dose calculations...............................................................................18 

3.2 Modifications to SNIFFER methodology .....................................................21 

3.2.1 Dose criteria and compliance points..................................................21 

3.2.2 Barrier design and performance ........................................................23 

3.2.3 Distribution of waste .........................................................................24 

3.2.4 Leachate concentration......................................................................25 

3.3 Supplementary calculations...........................................................................26 

4 Assessment Data and Assumptions ....................................................................28 

4.1 Site characteristics .........................................................................................28 

4.1.1 Size of site .........................................................................................28 

4.1.2 Construction ......................................................................................28 

4.1.3 Barrier................................................................................................28 

4.1.4 Cap.....................................................................................................30 

4.1.5 Operational period.............................................................................31 

4.1.6 Leachate collection and management procedures .............................32 

4.1.7 Leachate spillage ...............................................................................32 

4.1.8 Control over future site use ...............................................................35 

4.2 Hydrogeological setting.................................................................................35 

4.2.1 Underlying geology...........................................................................35 

4.2.2 Unsaturated zone characteristics .......................................................36 

4.2.3 Saturated zone characteristics............................................................36 

4.2.4 Groundwater discharges ....................................................................37 

4.2.5 Stream and river characteristics.........................................................37 

4.2.6 Groundwater flow and radionuclide transport...................................38 

4.3 Other scenarios and pathways .......................................................................40 

4.3.1 Gas.....................................................................................................40 

Galson Sciences Limited iii 14 July 2009 



Version 2 

4.3.2 Fire.....................................................................................................41 

4.3.3 Barrier failure ....................................................................................41 

4.3.4 Site remediation and re-engineering..................................................42 

4.3.5 Bathtubbing .......................................................................................42 

5 Dose Calculations.................................................................................................43 

5.1 Groundwater pathway ...................................................................................43 

5.2 Irradiation pathway........................................................................................45 

5.3 Intrusion.........................................................................................................46 

5.4 Leachate management and spillage ...............................................................48 

5.4.1 Leachate management .......................................................................48 

5.4.2 Leachate spillage ...............................................................................49 

5.4.3 Aerosol pathway................................................................................51 

5.5 Gas pathway ..................................................................................................52 

5.7 Dose rates to wildlife.....................................................................................53 

6 Radiological Capacity..........................................................................................59 

6.1 Introduction ...................................................................................................59 

6.2 Radionuclide-specific radiological capacities ...............................................60 

6.3 Overall radiological capacity.........................................................................63 

7 References.............................................................................................................66 

Appendix A Dose calculations ............................................................................67 

A.1 Doses during site operations..........................................................................67 

A.2 Doses to site residents after closure...............................................................68 

A.3 Doses during and after excavation of waste ..................................................69 

A.3.1 Dose to the Excavator........................................................................69 

A.3.2 Dose to Site Resident after Excavation .............................................71 

A.4 Doses arising from use of contaminated groundwater ..................................74 

Appendix B Radionuclide-specific data ............................................................76 

Appendix C Sensitivity Studies ..........................................................................79 

C.1 Groundwater Pathway ...................................................................................79 

C.2 Leachate Spillage...........................................................................................87 

Galson Sciences Limited iv 14 July 2009 



Version 2 

1 Introduction 

1.1 Project background 



Radioactive Waste to the 

Landfill at East Northants 


The nuclear decommissioning industry has significant future arisings of 

decommissioning wastes that fall into the category of low level waste. In the recently 

published “Policy for the Long Term Management of Solid Low Level Radioactive 

Waste” (Defra 2007), the government has confirmed the acceptability of the disposal 

of LLW to landfill including a new subset of LLW classified as high-volume very low 

level waste. 

Augean plc operates a hazardous waste disposal facility at the East Northants 

Resource Management Facility (ENRMF) in Northamptonshire and proposes that the 

site is used for the disposal of LLW with a specific activity up to 200 Bq / g. This 

report presents a radiological assessment for this site in order to assess the potential 

consequences of disposal and determine the radiological capacity of the site for 

different waste streams. In this report an example waste stream of LLW from 

UKAEA Harwell has been used for illustration. 

The radiological assessment presented in this report is based on the established 

approaches used previously for assessing “Special Precautions Burial” (SPB) which 

has been used for wastes of comparable activity. The methodology has been modified 

to take account of the likely waste volumes and also for certain site-specific aspects 

that differ from the generic assumptions used in the available model. 

The actual use of the ENRMF for disposal of LLW remains subject to discussions 

with the Environment Agency and other stakeholders. 

1.2 Approach 

The assessment presented in this report comprises two stages: 

Dose assessment 

Radiological capacity assessment 

Galson Sciences Limited 1 14 July 2009 



Version 2 

The dose assessment is based on an established methodology developed for SNIFFER 

(SNIFFER 2006a). Because the inventory, or amount of each radionuclide, is not 

known at this stage, the dose calculations are based on unit disposals (1 MBq) of each 

radionuclide, and the results are expressed as specific doses (μSv y -1 per MBq). 

Radiological capacity is the amount of radioactive material that can be consigned to a 

site without any of the potentially exposed groups considered receiving a dose above 

a specified criterion. 

For a single radionuclide, the radiological capacity (in Bq) is calculated by dividing 

the dose criterion (expressed in μSv y -1 ) by the maximum specific dose for that 

radionuclide (expressed in μSv y -1 per MBq). In the case of waste streams, however, 

in which the proportions of different radionuclides are fixed, the calculation of 

capacity must consider both the specific dose and the ratio of radionuclides in the 

waste stream. This means that there is not a single radiological capacity for the site 

and this assessment provides only an illustrative overall site radiological capacity 

based on preliminary data for the UKAEA Harwell Meashill Trenches waste stream. 

The radiological capacity determines how much radioactivity can be disposed of to 

the site without causing significant doses to workers or members of the public. The 

actual amount of waste that is disposed depends upon the specific activity of the waste 

(Bq / g or MBq / te). For very low activity wastes, the physical capacity of the site 

will impose a limit, and the radiological capacity may not be reached. For higher 

activity wastes, there are other constraints relating to waste handling and transport 

that impose an effective limit of 200 Bq / g. Because of the heterogeneous nature of 

the wastes envisaged for disposal at the ENRMF, the average activity would be 

significantly below 200 Bq / g but this value can be used to estimate the volume of the 

site that could be used for LLW disposals. 

1.3 Structure of report 

Following this Introduction, Section 2 of the report summarises available information 

concerning the proposed disposal site and also summarises information on the 

illustrative waste stream used in the radiological capacity calculations. Section 3 

summarises the SNIFFER methodology used for the radiological assessment, 

including a review of where the assumptions in the SNIFFER assessment model have 

been modified for this site-specific radiological assessment. Section 4 sets out the key 

assumptions, equations and parameter values for the scenarios adopted for the 

calculations of potential radionuclide releases from the site. Section 5 presents 

specific doses calculated for these release scenarios and also presents calculated dose 

rates for wildlife around the site. Section 6 presents radiological capacities based on 

the principal exposure pathways. Appendix A presents details of the exposure models 

used to calculate specific doses, including the equations and parameter values used. 

Appendix B presents the radionuclide-specific data used in the dose calculations. 

Appendix C presents results from sensitivity studies undertaken as part of the 

assessment. 




Version 2 

2 Background 

2.1 Site 

This section provides background information for the radiological assessment. 

Section 2.1 presents information on the operation and construction of the ENRMF, 

together with a summary of the geological and hydrogeological setting. A brief 

description of the environmental setting of the site and the populations that might be 

exposed to any release from the site is also presented. 

For the purpose of illustrating the overall radiological capacity of the site, and for 

estimating the volumes of waste that could be disposed, information on one waste 

stream currently being considered for disposal via the landfill route is presented in 

Section 2.2. 

This section provides a brief description of the ENRMF site and its surroundings. 

This information provides the basis for the assessment of potential doses from 

disposals of radioactive waste, and is derived principally from the Environmental 

Statement (Bullen Consultants Ltd, 2005) and Hydrogeological Risk Assessment 

(HRA) (ESI, 2004) made available by Augean. 

The ENRMF (formerly known as King’s Cliffe or Slipe Clay Pit) landfill site is about 

6 km from Stamford and began as a landfill site in 2002. Prior to that date it had been 

a clay pit used for the extraction of refractory clays. The landfill was initially used for 

the co-disposal of hazardous and non-hazardous wastes but, following a change in 

legislation, became a hazardous waste disposal site in 2004. 

2.1.1 Design and operations 

The site is divided into a series of cells, separated by clay bunds (Figure 2.1). These 

cells are progressively, constructed, infilled with waste and temporarily capped. 

The overall volume of the site is in the order of 1.8 x 10 6 m 3 (Table 2.1). At the time 

that the site became a hazardous waste site, Cells 1 and 2 were already full and had 

temporary caps. Hazardous waste only disposals started in Cell 3 which is now full. 

Cells 1, 2 and 3 have been permanently capped and partially restored. At the time the 

radiological assessment was initiated (October 2007), there was about 700,000 m 3 of 

remaining capacity. Planning permission for the site imposes a limit of 249,999 m 3 

on annual disposals and completion of inputs by 2013. 




Version 2 

WS010001/ENRMF/CONSAPPCRF 554 

Figure 2.1 Disposal cells at the ENRMF landfill site. Cells 1, 2 and 3 have been capped. 

Galson Sciences Limited 4 14 July 2009


Version 2 

Cell 

Basal area 

(m 2 ) 

Surface area 

(m 2 ) 

Void volume 

(m 3 ) 

1A 12,866 15,195 268,072 

Western 

extension 

2,000 2,396 13,080 

1B 11,934 11,934 193,865 

2A 12,449 12,449 187,900 

2B 10,464 11,822 143,010 

3A 8,412 13,922 160,877 

3B 9,090 11,624 174,685 

4A 11,144 14,097 174,564 

4B 12,552 12,552 206,102 

5A 6,929 11,669 144,211 

5B 8,294 9,887 147,221 

Table 2.1 Cell areas and volumes at the ENRMF landfill site. 

Cell construction comprises: 

Leachate drainage layer. This varies between cells, but in Cells 3, 4 and 5 it 

will be 500 mm of crushed granite. 

Artificial sealing liner. All cells include a 2 mm thick high density 

polyethylene (HDPE) geomembrane. 

Artificial mineral layer. All cells (except the western extension and Cell 2b) 

include at least 1.5m thickness of artificially emplaced geological barrier 

(Upper Lias clay sourced from the Slipe Clay Pit). This clay will be placed 

with a maximum design permeability of 3x10 -10 m/s and thickness of 1.5m or 

equivalent to meet the Environmental Permit requirement for permeability of 

less than or equal to 1.0 x l0 -9 m/s with a thickness of more than or equal to 

5m. 

The geological barrier also includes between 3 and 8 m of natural geological barrier 

(unsaturated zone) above the water table. 

Waste is examined on receipt at the site and after checking for compliance with the 

waste acceptance criteria is deposited within the current filling area. Temporary haul 

roads are formed within the fill area ensuring the de1ivery trucks do not traffic, 

unnecessarily, on the waste surface. 




Version 2 

Following deposit of the waste the material is levelled and compacted and 

intermediate inert cover materials placed over the top. Cover materials are currently 

sourced on site and are primarily made up of waste clays from the mineral excavation 

operations. Alternative cover materials are being investigated. Wastes are deposited 

in controlled layers to ensure adequate compaction and minimise settlement. 

Leachate forms in both open and capped cells as rainwater infiltrates through the cap, 

if present, and the waste. Leachate levels are monitored through a series of boreholes 

across the site, and excess leachate is removed by pumping. The leachate can be 

managed by recirculation in Cells 1 and 2 but is otherwise removed by tanker for 

treatment off site. The amount of leachate allowed to accumulate is regulated, with 

trigger levels of 2 m head in the sumps and 1 m in the monitoring wells. The Annual 

Monitoring Report for 2007 shows that these levels are maintained except during 

periods of unusually high rainfall. 

The “Kings Cliffe Landfill Site Annual Monitoring Report 2007” (ENRMF was 

formerly known as Kingscliffe or Slipe Clay pit) reports that approximately 

5000 tonnes of leachate were abstracted from the site during 2007 and transferred to a 

disposal facility. The “Pollution Inventory reporting form” submitted to the 

Environment Agency for the site reports 5402 tonnes of landfill leachate (EWC code 

19 07 03) being discharged in 2007. 

Leachate abstracted from the site is transferred by tanker to an off-site facility for 

treatment and discharge. The current arrangement is for transfer to a biological 

treatment plant at Avonmouth, discharge to trade effluent sewer and further treatment 

at a water treatment plant. There are feasibility studies underway for use of an 

alternative off-site facility and for construction of an on-site leachate treatment plant. 

The closed cells are capped with a composite cap consisting of a gas drainage layer, 

clay regulating layer, geotextile protector, geosynthetic clay liner, LDPE 

geomembrane liner and soil cover. 

There are lagoons on site that receive surface water that does not enter the disposal 

cells. Water from one lagoon is used for dust suppression. After capping, surface 

water will be directed to settling ponds and then discharged to a swallow hole 

(northern slopes) or surface water (southern areas). 

Following capping, the landfill site will enter a post closure managed stage. This stage 

will include the maintenance of leachate levels and gas abstraction, and will continue 

until it can be confirmed that the site no longer represents a significant risk of 

pollution of the environment or harm to human health. Leachate, surface water and 

groundwater quality will be monitored throughout the post closure managed stage. 

2.1.2 Geology and hydrogeology 

The regional geology comprises a sequence of Jurassic sedimentary rocks, including 

limestones, clays and mudstones (Table 2.2). On higher ground, the Jurassic rocks 

are overlain by Pleistocene glacial clays. 




Version 2 

ENRMF was originally a clay pit, which exploited the refractory clays at the base of 

the Upper Estuarine Series. The base of the pit is therefore effectively the top of the 

Lincolnshire Limestone, necessitating the need for an artificial geological barrier over 

most of the site, as described above. 

Group Formation Thickness Lithology Notes 

Great Oolite Group 

Inferior Oolite Group 

Lias 

Group 

Glacial Till 

(Pleistocene) 

Great Oolite Limestone 

(formerly Blisworth 

limestone) 

Upper Estuarine Series 

Upper and Lower 

Lincolnshire Limestone 

0 – 7m Yellow-brown clay with 

chalk and limestone 

fragments 

0 – 1.9 m Yellow micritic limestone 

9 – 12 m Grey-brown firm silty 

mudstone 

15 – 20 m Oolitic, pisolitic and 

massive limestones 

interbedded with sandy 

limestones 

Grantham Formation 0 – 2m Fine sands, sills, silty 

clays and mudstones. 

Northampton Sand ~2m Sands and sandstones 

with siderite nodules, 

some subordinate 

limestones and silts. 

Upper Lias >2m Grey mudstones and clays 

with subordinate thin 

limestone bands 

Patchy distribution; not 

present in the south east 

comer of the site 

Locally fissured. Has 

been excavated to win 

clay from the base of the 

unit. 

The only formation 

remaining beneath the 

excavation. Fractured, 

with some small voids 

and fissures. 

Noted in most boreholes 

drilled at King’s Cliffe. 

Sometimes present at the 

base of the Lower 

Lincolnshire Limestone. 

Table 2.2 Outline geological succession in the region of the ENRMF landfill site. 

Drilling near the site confirmed the presence of the Blisworth Limestone (Great Oolite 

Limestone) to the east of the site. The underlying Upper Estuarine Series 

(corresponding to the material exploited at the clay pit) ranges in thickness from 4.2 

m to 12.9 m, with a typical thickness of 11.5m. 

The Upper Estuarine Series is mainly argillaceous which has been divided into two 

parts. The lower part is the Lower Freshwater Sequence which appears to be devoid 

of marine fossils and is composed of dark or brown grey mudstones and seatearths 

with abundant rootlets and listric surfaces. Bioturbated laminae, load casts and sand 

filled cracks are common sedimentary features. This lower sequence is around 5 m 

thick. The upper division of the Series is composed of a cyclical sequence of marine 

and brackish/freshwater sediments. The marine beds are composed of shelly 




Version 2 

limestone and mudstones and the brackish sediments constitute mudstones and 

siltstones. This upper division ranges between 1 and 8 m in thickness. 

The Upper Estuarine Series is recorded in most borehole logs around the site. The 

interface between this unit and the overlying Glacial Boulder Clay and weathered 

clayey soils is difficult to discern. 

The upper part of the Lincolnshire Limestones underlying the Upper Estuarine Series 

and forming the base of the clay pit comprises a sequence of oolitic limestones. The 

lower part of the Lincolnshire Limestones is composed of fine grained sandy 

limestones. The Lincolnshire Limestone has been proven in the majority of the 

boreholes drilled within the site boundaries and ranges from 9 to 21 m in thickness. 

Discrete horizontal fissuring associated with bedding is present within the limestones. 

Fissures are generally clean and smooth with infilling material composed of 

fragments of limestone, crystalline calcite sand, silt and clay. Extensive fissuring and 

fragmented limestone has been seen in cores at certain elevations with decalcification 

producing cavities within the limestone. During preparation of the formation base of 

Cell 3A, two fissures were observed on the exposed surface of the limestone. These 

fissures were up to 1 m in length, 8 cm wide and estimated to be 30 cm deep. These 

fissures were infilled with a gravel concrete mix before emplacement of the basa1 

barrier. 

Beneath the Lincolnshire Limestones, the Grantham Formation is somewhat 

discontinuous around King’s Cliffe, and often the Lower Lincolnshire Limestone is in 

direct contact with the Northampton Sands. Below the Northampton Sand is the 

Upper Lias. 

Glacial Till deposits are relatively extensive, especially on higher ground to the 

southwest, east and southeast of the site. Glacial Till was encountered to the 

southwest of the quarry, where it has been described as firm to stiff, dark brown and 

grey, slightly sandy clay with limestone gravels, with a thickness of 8.6 m. 

The principal hydrogeological units in the Jurassic rocks of the area are listed in Table 

2.3. Although groundwater vulnerability maps show the ENRMF landfill site to be in 

a non-aquifer area (Upper Estuarine Series), the removal of clay during quarrying 

means that the base of the landfill situated on the Lincolnshire Limestone, classified 

as a Major Aquifer by the Environment Agency. 

The hydraulic properties of the Lincolnshire Limestone in the East Midlands area are 

summarised in Table 2.4. Because of the removal of the overlying Upper Estuarine 

Series, the Limestone in the region of the site is unconfined. Regional groundwater 

flow is down dip towards the confined area in the east (Allen et al., 1997). 

The Lincolnshire Limestone aquifer is characterised by fracture flow and there are 

also swallow holes in the vicinity which allow rapid flow of water from surface to the 

water table. The depth to the water table is estimated to be between 4m to 8m, and the 

thickness of the aquifer at the site is 15m to 22m. Permeability is lower than would be 




Version 2 

expected in other regions and the aquifer, although used for agricultural supply, is not 

used for public water supply in this area. 

Formation Hydrogeological 

classification 

Comment 

Great Oolite Limestone Aquifer Minor aquifer. Unsaturated 

Upper Estuarine Series 

(Silty Mudstone) 

Aquitard Potentially semi-confines the 

underlying Lincolnshire Limestone 

aquifer in the local area and to the 

south and southeast. Elsewhere may 

be absent. 

Lincolnshire Limestone Aquifer Major aquifer (EA classification). 

Locally has limited thickness. 

Dominated by fracture flow. 

Recharge through swallow holes 

extending through Upper Estuarine 

Series. The semi-confining clays of 

the Upper Estuarine Series were 

removed at quarry site resulting in 

water tab1e conditions. Elsewhere to 

the east conditions are 

confined/artesian. 

Grantham Formation Aquitard May be absent. 

Northampton Sand Aquifer Minor aquifer. It is semi-confined by 

the silts/clays of the Lower Estuarine 

Series. Where this is absent, it is in 

hydraulic continuity with the Lower 

Lincolnshire Limestone Aquifer. 

Upper Lias Aquiclude Basal Aquiclude to the Oolite Series 

aquifer/aquitard system. 

Table 2.3 Principal hydrogeological units in the region of the ENRMF landfill 

site. 




Version 2 

Hydraulic Property Sample method Samples Range (Mean) 

Transmissivity 

(m /day) 

Pumping tests 59 1 – 14 000* (665) a 

Porosity Core data 415 0.13 – 0.22 (0.18) b 

Fracture porosity Estimates 0.004 – 0.01 

Storage Pumping tests 37 2x10 -7 – 6x10 -1 

Hydraulic conductivity 

(m/day) 

Core data 415


Version 2 

2.2 Wastes 

The nuclear industry is generating significant quantities of radioactive waste from 

decommissioning, and these types of waste will continue to arise as decommissioning 

continues. Some of these wastes are potentially suitable for disposal at landfill sites. 

Wastes from other activities and industries, such as hospitals, universities and 

radiochemical manufacture, could also be considered for disposal at such sites. 

To ensure that the potential radiological consequences of the disposal of a 

representative range of LLW can be assessed, the radionuclides listed in Table 2.5 

have been considered in the radiological assessment. Radionuclides with half-lives 

less than one year have not been explicitly assessed. Where such radionuclides arise 

from ingrowth, they are included through the assumption that they will be in secular 

equilibrium with the parent radionuclide, and the dose coefficients used are adjusted 

accordingly. 

Radionuclide 

Half-life 

(years) 

3 H 12.3 

14 C 5,730 

36 Cl 3.01E+05 

55 Fe 2.73 

60 Co 5.27 

63 Ni 96.0 

90 Sr 28.8 

94 Nb 2.00E+04 

99 Tc 2.11E+05 

106 Ru 1.02 

108m Ag 418 

125 Sb 2.80 

126 Sn 2.07E+05 

129 I 1.57E+07 

133 Ba 10.7 

134 Cs 2.10 

137 Cs 30.0 

147 Pm 2.60 

152 Eu 13.3 

154 Eu 8.80 

155 Eu 4.96 

210 Pb 22.3 

Daughters assumed to be in secular 

equilibrium 

90 Y 

106 Rh 

126 Sb 

137m Ba 

210 Bi, 210 Po 




Version 2 

Radionuclide 

Half-life 

(years) 

226 Ra 1,600 

227 Ac 21.7 

229 Th 7,340 

230 Th 7.54E+04 

232 Th 1.40E+10 

231 Pa 3.27E+04 

232 U 68.9 

233 U 1.59E+05 

234 U 2.45E+05 

235 U 7.04E+08 

236 U 2.34E+07 

238 U 4.47E+09 

237 Np 2.14E+06 

238 Pu 87.7 

239 Pu 2.41E+04 

240 Pu 6,540 

241 Pu 14.4 

242 Pu 3.76E+05 

241 Am 432 

243 Cm 29.1 

244 Cm 18.1 

Daughters assumed to be in secular 

equilibrium 

222 Rn, 218 Po, 218 At, 214 Pb, 214 Bi, 214 Po, 210 Tl, 

210 Pb 

227 Th, 223 Fr, 223 Ra, 219 Rn, 215 Po, 211 Pb, 211 Bi, 

207 Tl 

225 Ra, 225 Ac, 221 Fr, 221 Ra, 217 Rn, 217 At, 213 Bi, 

213 Po, 209 Tl, 209 Pb 

228 Ra, 228 Ac, 228 Th, 224 Ra, 220 Rn, 216 Po, 212 Pb, 

212 Bi, 212 Po, 208 Tl 

231 Th 

234 Th, 234m Pa, 234 Pa 

233 Pa 

235m U 

Table 2.5 List of radionuclides for the radiological assessment. Radionuclides 

with half-lives of


Version 2 

industry decommissioning wastes that might be considered for disposal at the 

ENRMF. 

The waste stream used to illustrate the potential radiological capacity of the ENRMF 

has been compiled from information on material in the Meashill Trenches at Harwell. 

This waste stream comprises a mixture of activated synchrotron components, reactor 

components and decommissioning/land remediation wastes, but is dominated by Co- 

60 through the presence of activated steel. Other waste streams from Harwell are 

more typically dominated by Cs-137. To partly reduce this dominance by Co-60, the 

2000 inventory provided has been decayed to 2010 (Table 2.6). 

The inventory presented in Table 2.6 does not represent the complete inventory for 

the Meashill Trenches. There will be small quantities of long-lived daughter 

radionuclides from the plutonium, uranium and thorium decay series. These become 

of increasing importance as the inventory decays, and are considered in the long-term 

assessments, but are not significant after only 10 years of decay. There are likely to 

be other radionuclides present, but these would contribute much less to any potential 

dose than the radionuclides listed. 

Radionuclide Half-life 

(years) 

2000 inventory 2010 inventory 

MBq % MBq % 

H3 12.3 5.70E+00 0.02 3.25E+00 0.03 

Co60 5.27 3.00E+04 89.90 8.05E+03 72.25 

Cs137 30 1.20E+03 3.60 9.52E+02 8.55 

Ra226 1600 1.00E+02 0.30 9.96E+01 0.89 

Th232 1.4E+10 4.00E+01 0.12 4.00E+01 0.36 

U234 2.44E+05 5.00E+02 1.50 5.00E+02 4.49 

U235 7.04E+08 2.40E+01 0.07 2.40E+01 0.22 

U238 4.47E+09 5.00E+02 1.50 5.00E+02 4.49 

Pu238 87.7 4.00E+01 0.12 3.70E+01 0.33 

Pu239 2.41E+04 4.00E+02 1.20 4.00E+02 3.59 

Pu240 6540 4.00E+02 1.20 4.00E+02 3.59 

Pu241 14.4 6.18E+01 0.19 3.82E+01 0.34 

Am241 432 1.00E+02 0.30 9.92E+01 0.89 

Total 3.34E+04 1.11E+04 

Table 2.6 Illustrative inventory for wastes from the Meashill Trenches, Harwell. 

The 2000 inventory provided by UKAEA Harwell has been decayed to 

2010 to provide an input to radiological capacity calculations. 




Version 2 

3 Assessment Methodology 

This section describes the overall assessment methodology used to calculate potential 

doses from disposals of LLW at the ENRMF and to determine radiological capacities. 

Section 3.1 summarises the SNIFFER methodology, on which the assessment is 

based. Section 3.2 describes the changes made to the SNIFFER methodology to take 

account of specific features of the ENRMF and the proposed disposals. 

3.1 Summary of SNIFFER methodology 

3.1.1 Assessment framework 

The SNIFFER methodology was developed so as to provide the regulators, and other 

stakeholders, with a consistent approach to assessing the potential for landfill sites to 

accept the category of LLW known as Special Precautions Burial (SPB). The overall 

assessment approach is illustrated in Figure 3.1 (SNIFFER 2006a). 

It was originally envisaged that a screening stage would be useful if large numbers of 

sites were examined (SNIFFER 2006a). This might be done by site owners, seeking 

to put forward a few sites as potential disposal sites, or by planners, seeking to assess 

the overall availability of disposal capacity for LLW. The principal application of the 

methodology, however, would be for the assessment of particular sites and this 

screening stage would not be required. 

An important aim of the SNIFFER methodology was to provide regulators with a 

means of assessing radiological capacity for a landfill site and updating this capacity 

as more information becomes available and the available capacity is reduced though 

disposals. To ensure that the assessment is robust and fit for purpose, the approach 

developed by the IAEA and others of defining an assessment context forms an 

important part of the SNIFFER methodology. To provide as much consistency and 

flexibility as possible, the elements comprising the assessment context were 

incorporated as generic elements (providing consistency) or site-specific elements 

(providing flexibility) as considered most appropriate. 

Using the assessment terminology established by the IAEA, the generic aspects of the 

assessment context are the assessment purpose, endpoints, basis, and assumptions 

regarding future society. The site-specific aspects of the assessment context are the 

environmental system of interest, site context, nature of the wastes, and assessment 

timescales. 

The assessment endpoint and basis are established as generic factors so as to ensure as 

much consistency as possible between sites. The assessment end-point is dose, so that 

the results can be compared with an effective dose criterion. The SNIFFER 

methodology was based on a criterion of 20 Sv/year, representing the point at which 

doses arising from disposals can be regarded as being below regulatory concern 

(SNIFFER 2006a). 




Version 2 

Figure 3.1: The overall SNIFFER assessment approach (SNIFFER 2006a). 

General 

Site 

Information 

Site- 

Specific 

Data 

Existing 

Inventory 

Assess 

Mitigation 

Measures 

Screening 

Protocol 

Pass 

Develop 

Assessment 

Context 

Dose 

Calculations 

Radiological 

Capacity 

Calculations 

Authorisation 

Conditions 

Site 

Unacceptable 

Generic 

Data 

Dose 

Constraint 

The assessment basis, also established as a generic element, includes all of the 

scenarios (describing ways in which doses could be received) that should be 

considered in an assessment. Some scenarios may be excluded from particular 

assessments, but only if there is a documented reason for doing so. Scenarios are 

discussed in more detail below. 

The site context includes a range of features of the site and its surroundings that help 

to define the source-pathway-receptor system(s) used in the assessment calculations. 

These features include: 


Fail 



Version 2 

The identity and proximity of potentially affected populations or other 

environmental receptors. 

Potential exposure pathways associated with the potentially affected 

populations, such as stream and groundwater discharge points, drinking water 

wells and irrigation practices, and atmospheric pathways for gas and dust, 

including point source emissions from combustion of landfill gas. 

Site management practices, such as waste segregation, coverage of waste, liner 

type, permitted leachate head, and leachate management. 

Past disposals of radioactive wastes and other wastes that might interact with 

radioactive wastes (e.g., organic materials). 

3.1.2 Scenarios 

As noted above, the selection of applicable scenarios is a site-specific aspect of the 

assessment context. As an aid to uniformity of approach, and to make possible the 

development of a useable assessment model, a set of potential scenarios is defined 

within the SNIFFER methodology (SNIFFER 2006a). 

Scenarios are divided into operational and post-closure scenarios. Four exposed 

groups are considered. 

Site workers. At the type of facility considered using the SNIFFER 

methodology, site workers are not considered as radiation workers, and may 

have no specific information about the types of material being consigned. In 

terms of dose constraints, therefore, they are considered in the same way as 

members of the public. 

Members of the public living near the site. 

Members of the public exploiting potentially contaminated groundwater or 

surface water resources. Depending on the hydrological setting of the site, this 

group may be the same as the local resident group. 

Members of the public living on the site after closure and the withdrawal of 

controls. 

Potential operational scenarios are presented in Table 3.1 and post-closure scenarios 

are presented in Table 3.2. 




Version 2 

Scenario name Description Hazards 

Normal operations 

Barrier failure 

Leachate spillage 

Site remediation or 

re-engineering 

Fire 

Expected operation of the 

landfill up to capping and 

closure, as approved by the 

relevant Agency. Doses to site 

workers and to the public are 

considered. 

Failure of the artificial sealing 

liner and geological barrier 

during operations. Doses to the 

public are considered. 

Unintentional release of 

leachate to surface water. 

Doses to the public are 

considered. 

Workers expose waste during 

operations to remediate 

containment failure or to 

enlarge or otherwise reengineer 

site. 

Fire releases radioactivity. 

Doses to site workers and to 

the public are considered. 

Gas Release 

Liquid release (leachate) 

Aerosols (leachate) 

Direct irradiation 



Solid release (dust while 

uncovered) 


Solid release (dust), gases 

and vapour 

Table 3.1: Operational scenarios included in the SNIFFER methodology and the 

associated hazards (SNIFFER 2006a). 

The last two of the scenarios in Table 3.1 are considered to encompass the range of 

other events that may result in a site worker being exposed, such as short-term contact 

with leachate. 




Version 2 

Scenario name Description Hazards 

Normal post-closure 

evolution 

Bathtubbing 

Inadvertent 

excavation 

During this time, the landfill 

Gas Release 

engineering is assumed to 

gradually degrade. Doses to 


the public are considered. Direct irradiation (through 

cover) 

Blockage of the drainage 

system causes overflow of 

leachate laterally from the 

landfill onto the soil. Doses to 

the public are considered. 

Waste is inadvertently 

excavated and re-distributed, 

e.g., during building or 

farming. Doses to the intruder 

and the subsequent user of the 

site are considered. 



Solid release (dust) 

Solid release (waste) 

Table 3.2: Post-closure scenarios included in the SNIFFER methodology and the 

associated hazards (SNIFFER 2006a). 

3.1.3 Dose calculations 

This section describes the potential pathways identified within the SNIFFER 

methodology. Not all of these pathways will be necessarily be relevant to the 

assessment of a specific site, and the methodology requires both the identification and 

characterisation of the exposure pathways associated with a particular landfill. The 

pathways considered in the assessment of the ENRMF are discussed in Section 3.2 

and Section 4. 

External irradiation from standing near radioactively-contaminated waste. 

This pathway will be minimised when the waste is covered, and will then only 

apply to gamma-emitting wastes. 

Inhalation of contaminated dust. Because the waste will be emplaced in 

sacks/drums and be buried on emplacement, creation of contaminated dust is 

not considered as an exposure pathway during the normal operation of the 

landfill. However, deliberate intervention to maintain, remediate or reengineer 

the site (including the drilling of boreholes for landfill gas 

abstraction), or inadvertent excavation during unrelated development of the 

site after closure, could lead to the creation of contaminated dust. 

Inhalation of aerosols from leachate. Leachate treatment potentially generates 

aerosols that could be inhaled by workers or members of the public near the 

site or any off-site treatment facility. The spraying of leachate back onto the 




Version 2 

surface of the landfill is a practice that should be prevented through the 

Environmental Permitting process. Aerosols from leachate may, however, be 

generated during other types of leachate treatment either on or off-site, 

particularly if this involves aeration. Leachate treatment may continue after 

closure, but will end at the end of the control period. Use of leachate 

following the loss of control may also lead to aerosol formation but 

concentrations are likely to be lower than during leachate treatment. 

Inhalation of dust, particles and gases from fires. Accidental fires in the waste 

are a potential hazard at landfill sites with combustible wastes. A fire at the 

site could lead to the release of radioactive particles and dust that could be 

inhaled by workers and members of the public downwind of the site, and 

could also lead to some gaseous releases. Waste fires may be associated with 

the collection and utilisation of landfill gas at sites which accept biodegradable 

wastes. Gaseous releases of radioactive material from flaring or other use are 

included in the following pathway. 

Inhalation of radioactive gas, i.e., 14 CO2, 14 CH4, 3 H, and radon. The first three 

may be generated through microbial degradation or corrosion of the 

radioactive waste. Landfill sites which accept biodegradable wastes are 

required to collect and flare or utilise the gas, and this could disperse 

radioactive gases that could be inhaled by workers and members of the public 

downwind of the site. Radon is generated through the decay of Ra-226, which 

in turn is a decay product of Th-230. Radon could be inhaled by workers, 

members of the public downwind of the site, and occupants working or living 

on the site after loss of control. 

Ingestion of contaminated water. This pathway arises mainly through the 

leakage of leachate through the engineering and into groundwater (Figure 5). 

Once groundwater is contaminated, ingestion can occur through: 

- extraction of contaminated groundwater via a well for drinking; and 

- discharge of contaminated groundwater to surface water used for drinking. 

Surface water may also be contaminated by the unintentional release of 

contaminated leachate. Once surface water is contaminated, ingestion can 

occur through: 

- extraction of water for drinking. 

Spillage of leachate may also contaminate groundwater used for drinking 

water supply, but retardation and dilution are likely to mean that potential 

doses through this pathway are less than those from surface water. 

Ingestion of contaminated food. This pathway arises mainly through the 

leakage of leachate through the engineering and into groundwater. Once in 

the groundwater, radioactivity can contaminate food supplies through: 




Version 2 

- extraction of groundwater for irrigation, thereby contaminating soil used 

for farming, or for stock watering; 

- discharge of contaminated groundwater to surface water used for 

irrigation, thereby contaminating soil used for farming, or for stock 

watering; and 

- discharge of contaminated groundwater to surface water or marine water 

that is used for fishing. 

Surface water may also be contaminated by the unintentional release of 

contaminated leachate. Once surface water is contaminated, radioactivity can 

contaminate food supplies through: 

- use of surface water for irrigation, thereby contaminating soil used for 

farming, or for stock watering; 

- use of surface waters for fishing. 

Spillage of leachate may contaminate groundwater used for irrigation, but 

retardation and dilution are likely to mean that potential doses through this 

pathway are less than those from use of contaminated surface water. 

Soil may be contaminated by the lateral discharge of leachate directly from the 

site after blockage of the drainage system (bathtubbing). 

Inhalation of dust from contaminated soil. This pathway mainly arises 

indirectly through the leakage of leachate through the engineering and into 

groundwater. Once in the groundwater, radioactivity can contaminate soil 

through: 

- capillary rise of contaminated groundwater into the soil; 

- discharge of contaminated groundwater to surface water and subsequent 

flooding; 

- extraction of groundwater for irrigation, thereby contaminating soil; and 

- discharge of contaminated groundwater to surface water used for 

irrigation, thereby contaminating soil. 

Soil may also be contaminated indirectly through spillage or inadvertent discharge 

of leachate to surface water and subsequent irrigation. 

Soil may be contaminated directly by the lateral discharge of leachate from the 

site after blockage of the drainage system (bathtubbing). 

Details of the models and equations used to calculate doses via these pathways are 

given in the Technical Reference Manual for the SNIFFER methodology (SNIFFER 

2006b). For the radiological assessment of the disposal of LLW with an activity of 




Version 2 

up to 200Bq/g at the ENRMF, the models and assumptions underlying all of the 

scenarios described above have been re-examined and a number of changes made. 

These are described in Section 3.2 and Section 4. 

3.2 Modifications to SNIFFER methodology 

The assessment of a hazardous waste site for the disposal of significant volumes of 

LLW is sufficiently different to the original application of the SNIFFER methodology 

outlined above to require a re-examination of the key assumptions. 

This re-examination identified several aspects of the overall methodology and the 

assessment model where different assumptions are required: 

Dose criteria and compliance points 

Barrier design and performance 

Distribution of waste 

Leachate concentration 

Several aspects of the assessment model were also identified where the default 

parameter values included in the SNIFFER model required change to take account of 

site-specific features. These are highlighted in Section 4 and Appendices 1 and 2, 

where the assessment data are presented. 

It should also be noted that there were errors in the dose coefficients included in the 

original SNIFFER assessment model, which did not account for the contribution to 

dose from short-lived daughter radionuclides in secular equilibrium with the parent 

radionuclides. These dose coefficients have been corrected and updated for the 

radiological assessment of the ENRMF. 

3.2.1 Dose criteria and compliance points 

In the original application of the SNIFFER methodology, a single dose criterion of 

20 Sv/year was used as the basis for radiological capacity calculations and applied to 

all scenarios. More recently, the environment agencies have recognised that human 

intrusion into disposal facilities represents a different class of uncertainties about 

system behaviour than barrier degradation and natural processes. An alternative dose 

criterion for human intrusion has been specified in guidance for near-surface facilities 

intended solely for radioactive wastes (Environment Agency et al. 2009). This 

revised guidance states that the assessed effective dose to any person during and after 

an intrusion should not exceed a dose guidance level in the range of around 

3 mSv/year to around 20 mSv/year. Values towards the lower end of this range are 

applicable to assessed exposures continuing over a period of years (prolonged 

exposures), while values towards the upper end of the range are applicable to assessed 

exposures that are only short term (transitory exposures). 




Version 2 

The regulatory guidance notes that the following are events for which the dose 

guidance levels for human intrusion events apply: 

human intrusion directly into a disposal facility; 

other human actions that damage barriers or degrade their functions, such as 

removing material from a disposal facility cap. Barriers considered to be 

affected by these human actions may be engineered, natural or a combination 

of both. 

The dose criteria used do not affect the way in which the dose assessments are 

conducted, but they are important for assessing the radiological capacity of a disposal 

facility. For the assessment of the ENRMF, three potential events are assessed that 

are considered to fall within these definitions: 

Direct excavation of waste. 

Occupation and subsequent use of the site following removal of the cap or 

excavation and re-distribution of waste. 

Use of a borehole at the site boundary as a source of drinking water. 

The first two of these events are derived from the inadvertent intrusion scenario in the 

SNIFFER methodology. Although doses to those excavating the waste could be 

regarded as transitory according to the regulatory guidance, and therefore subject to a 

dose guidance level of up to 20 mSv/year, the lower dose criterion of 3 mSv/year has 

been used in the calculation of radiological capacities. 

The third event is included to provide a comparison with assessments of potential 

releases of non-radiological hazardous substances. Radiological assessments are 

based on calculating releases to the accessible environment and then determining 

doses to members of the critical group. For future releases, the same approach is used 

but a range of potentially exposed groups are considered at different release points 

where contaminated resources might be exploited in the future. In order to show 

compliance with the Groundwater Directive, assessments of potential releases of nonradiological 

hazardous substances must show that a site does not allow the discharge 

of List I substances into groundwater or the pollution of groundwater by List II 

substances. Such assessments therefore use compliance points at the water table, 

regardless of whether the groundwater is actually exploited at that point. 

To provide a comparison between the two types of assessment, the radiological 

assessment has been extended to include use of a borehole at the site boundary for 

drinking water. This provides a compliance point for groundwater, although 

boreholes for drinking water would not normally be permitted in such locations as 

they would degrade the function of the natural barriers that are a key part of providing 

long-term safety for radioactive waste disposal. In calculating the radiological 

capacity of the site, it is therefore appropriate to regard such a borehole as an intrusion 

event and to use the dose guidance level of 3 mSv/year. 




Version 2 

The radiological criteria in the GRA (Environment Agency et al. 2009) apply to a 

representative of the critical group or those at greatest risk. Although the criteria are 

expressed as annual doses, they are established on the basis that exposures may be 

prolonged (several years or lifetime exposures) rather than transitory (occurring in 

one specific year). Combined with the uncertainties inherent in these types of 

assessments, this means that the representative person is assumed to be an adult, and 

consumption rates and dose coefficients are set accordingly. 

In the case of accidental releases, particularly during the operational phase, exposures 

may be for shorter periods than from post-closure, normal evolution releases. In these 

cases, it may be appropriate to use alternative assumptions about consumption rates 

and the corresponding dose coefficients to determine whether infants or children 

receive significantly greater doses than adults. 

In the case of foetuses, the Health Protection Agency (HPA 2008) notes that for most 

radionuclides doses to the foetus are lower than to the mother, and that: 

… for solid waste disposals it will generally be unnecessary to consider the 

ernbryo/fetus/breastfed infant as any increases in doses over those to other age 

groups will be small compared to the overall uncertainty in the assessed doses. 

The radionuclides identified in this guidance as giving higher doses to the foetus than 

to the mother do not generally occur in decommissioning or similar wastes and are not 

included in the set of radionuclides considered in this assessment (Table 2.5). 

3.2.2 Barrier design and performance 

The principal differences between different types of landfill are the requirements 

relating to the barrier at the base of the landfill. Schedule 2 of the Landfill (England 

and Wales) Regulations 2002 states: 

(4) The landfill base and sides shall consist of a mineral layer which provides 

protection of soil, groundwater and surface water at least equivalent to 

that resulting from the following permeability and thickness 

requirements - 

(a) in a landfill for hazardous waste: k


Version 2 

The assessment model developed as part of the SNIFFER methodology was based on 

the disposal of LLW at non-hazardous waste sites (SNIFFER 2006a,b), and the 

default values for an effective geological barrier were set to a thickness of 1 m and a 

hydraulic conductivity of 1 x 10 -9 m s -1 . The assumption was that this barrier would 

be provided by an artificial mineral layer between a basal liner and the natural 

geological barrier. The presence of a natural geological barrier (unsaturated zone) is 

allowed for in the assessment model, but typical characteristics of this zone mean that 

it is likely to be less resistant to leachate transport than the artificial geological barrier 

and effectively redundant in terms of assessing potential doses via the groundwater 

pathway. 

In the case of the landfill cap, the default values used in the SNIFFER methodology 

assume that the cap is initially 95% efficient in terms of preventing infiltration into 

the landfill, and that the cap will gradually degrade over a period of 60 years from the 

time of emplacement until it is no more effective than a soil layer. Also, the 

SNIFFER methodology assumes that it is only the effectiveness of the cap that limits 

infiltration of leachate into groundwater after the end of the operational period – the 

liner is assumed to become ineffective at the time the cap is emplaced. 

At the ENRMF, the basal liner in the area considered for disposal of radioactive 

wastes is constructed with a 2 mm thick high density polyethylene (HDPE) 

geomembrane, and at least 1.5m thickness of artificially emplaced geological barrier 

(Upper Lias clay sourced locally). This clay is placed with a maximum design 

permeability of 3x10 -10 m/s. 

The final cap for the ENRMF comprises a gas drainage layer, clay regulating layer, 

geotextile protector, geosynthetic clay liner, LDPE geomembrane liner and soil cover. 

The cap design will aim for a minimum effectiveness of 99%. 

Following capping the assessment model assumes that there will be a minimum of 60 

years management of the site, which will include monitoring of leachate levels within 

the waste. In practice the management period will be considerably longer. This will 

enable the effectiveness of the cap and the bottom liner to be assessed and for 

mitigation measures to be taken if there is evidence of damage or deterioration. It is 

therefore reasonable to assume that the design performance of the cap and liner will 

be maintained during the management phase and that degradation will not take place 

until after the withdrawal of control. 

3.2.3 Distribution of waste 

The SNIFFER methodology does not require the actual volume of waste to be 

specified, because the dose calculations are based on the disposal of 1 MBq of each 

radionuclide. However, to determine the concentration of waste that might be 

excavated after site closure, the methodology assumes that all of the disposals at a 

particular site could be in part of a cell as small as 10 m 3 . 

The proposed disposals of LLW at the ENRMF could form a significant proportion of 

the material disposed of to the selected cell. The SNIFFER methodology has 




Version 2 

therefore been revised to account for waste deposited throughout the cell rather than 

in a particular volume. For the radiological assessment, it is again not necessary to 

assume the volume of waste, because the calculations are based on a unit disposal of 1 

MBq, homogeneously dispersed throughout the cell. Assumptions regarding waste 

activity are required to convert the calculated radiological capacity into waste 

volumes and to determine whether the assumption concerning homogeneity is 

reasonable. 

3.2.4 Leachate concentration 

There are significant uncertainties associated with modelling the release of 

radionuclides from radioactive waste and into leachate. The mechanisms by which 

this release would occur depend on the type of waste (e.g., waste composition, how it 

is contaminated and how it is packaged), on the conditions within the landfill (e.g., 

pH, Eh, degree of saturation) and on the radionuclides concerned (e.g., whether they 

are readily sorbed). Even with detailed mechanistic models of waste behaviour, 

significant variability (due to heterogeneities in the wastes and landfill conditions) 

and uncertainties (due to lack of information about the processes involved) would 

remain. 

In the SNIFFER methodology, these uncertainties are treated by means of 

conservative assumptions: 

For scenarios involving waste excavation, it is assumed that the entire 

radionuclide inventory remains in the solid waste and that there are no losses 

to leachate. 

For scenarios involving leakage of leachate, it is assumed that the entire 

radionuclide inventory is available for dissolution into leachate at site closure, 

with the concentration in leachate determined by the appropriate sorption 

coefficient (Kd). 

The assumption regarding the partitioning of radionuclides between waste and 

leachate would be conservative even if sorption coefficients could be determined for 

the actual wastes and conditions within the landfill, because not all of the radioactive 

contamination would be on the surface of the waste and available for immediate 

dissolution. Furthermore, because of the difficulties in determining sorption 

coefficients, the default values are set to zero, effectively meaning that in the 

SNIFFER methodology the entire radionuclide inventory enters the leachate at site 

closure. 

For the radiological assessment of the ENRMF, alternative assumptions have been 

made for the scenarios involving leakage of leachate: 

For pathways involving contamination of soil (including irrigation using 

contaminated groundwater), the assumption that the entire radionuclide 

inventory is available for dissolution into leachate at site closure is retained. 




Version 2 

For pathways involving drinking water or other transitory exposures such as 

aerosols, it is assumed that dissolution is gradual and an annual release to 

leachate is used to calculate releases and doses. 

These pathways are distinguished because in the former case radionuclides will 

accumulate in the soil and an assessment effectively based on a single year would not 

be demonstrably conservative. 

The closest analogue for landfill disposal is the trench disposals at the LLWR near 

Drigg. A comparison of the annual discharges through the marine pipeline (BNFL 

2002a) with estimates of the disposed inventory (BNFL 2002b) indicates that a factor 

of at least 1x10 -3 year -1 should be applied to determining what fraction of the 

inventory might be in leachate. Initial concentrations, and concentrations of more 

insoluble radioelements, would probably be lower than this, but this factor has been 

used in this assessment for all radionuclides as a conservative assumption. 

3.3 Supplementary calculations 

In addition to a radiological assessment based on the pathways and scenarios included 

within the SNIFFER methodology, two supplementary calculations have been 

undertaken. These relate to potential doses from the treatment and discharge of 

leachate at an off-site water treatment plant, and to possible radiological effects on 

wildlife. 

The SNIFFER methodology includes a leachate spillage scenario, which is modelled 

as a release of leachate to a water body (e.g., river, lake) that is then exploited as a 

water resource (e.g., drinking, fishing). This scenario and modelling treatment is 

intended to address accidental releases and not routine discharges. At the ENRMF, 

leachate is collected and sent by tanker to a water treatment plant at Avonmouth. 

Following treatment, water is then discharged to the Severn Estuary. Potential doses 

arising from the leachate treatment and discharge are not included within the 

SNIFFER methodology and have been separately assessed using the Environment 

Agency’s Initial Radiological Assessment - Sewer methodology (Environment 

Agency 2006a; 2006b). 

Discharges and migration of radionuclides from a disposal facility might have a 

detrimental effect on non-human species or more general environmental effects such 

as damaging habitat quality. The guidance from the Environment Agencies includes a 

requirement to ensure that all aspects of the accessible environment are protected: 

The developer/operator should carry out an assessment to investigate the 

radiological effects of a disposal facility on the accessible environment both 

during the period of authorisation and afterwards with a view to showing that 

all aspects of the accessible environment are adequately protected. 

Although there is no specific evidence that there might be a threat to populations of 

non-human species from the authorised release of radioactive substances if people are 

protected, environmental damage might occur to areas and habitats that are not 

extensively exploited by people. Furthermore, there is a specific need to be able to 




Version 2 

demonstrate that non-human species are protected under legislation related to 

conservation, for example that derived from the EC Habitats Directive (EC 1992). 

There are currently (June 2009) no internationally established criteria for determining 

radiological protection of the environment. However, a number of research studies 

and regulatory guidance documents have proposed that an incremental dose rate value 

of 10 Gyh -1 is appropriate as a screening criterion, although dose rates less than 

40 Gyh -1 are unlikely to exert any effect on the reproductive capacity of mammals 

and chronic effects for other organisms are unlikely at even greater dose rates 

(Copplestone et al. 2002). 

An assessment tool developed as part of the ERICA project (Environmental Risk from 

Ionising Contaminants: Assessment and Management) has been used to calculate 

potential dose rate values. The ERICA assessment tool allows three tiers of 

assessment. A Tier 1 assessment has been undertaken and the calculated incremental 

dose rate values are below the screening value indicated above. More detailed 

assessments (Tier 2 and Tier 3) are therefore not required. 




Version 2 

4 Assessment Data and Assumptions 

The assessment model is a simplified model of the events and processes that will or 

might take place during and after operations. A number of simplifying assumptions 

are therefore required in order to represent the site and its surroundings, as 

summarised in Section 2 of this report, in the model. These assumptions are outlined 

in this Section. Where there are significant uncertainties regarding aspects of the site, 

alternative sets of assumptions have been made and assessment calculations carried 

out. 

This section includes the equations used in modelling the release of radionuclides 

from the site according to the different scenarios and assumptions. Equations for 

calculating the potential doses from these releases are presented in Appendix A. 

Parameter values used in the calculations are included in this section and in 

Appendices A and B (radionuclide-specific data). Unless site-specific parameter 

values have been identified, the parameter values used in the modelling are those used 

in SNIFFER (2006b), which are derived in large part from IAEA (2003). 

4.1 Site characteristics 

4.1.1 Size of site 

The ENRMF landfill site has an overall disposal volume of 1,800,000 m 3 , with a 

surface area of about 125,000 m 2 . About 700,000 m 3 remains available for disposal, 

with a surface area of about 50,000 m 2 . 

For the purpose of the radiological assessment reported here, it is assumed that 

disposal of LLW will be restricted to Cells 4B, 5A and 5B with a volume of 

497,534 m 3 and a surface area of 34,108 m 2 . 

4.1.2 Construction 

It is assumed that Cells 4B, 5A and 5B are hydrologically isolated from the remainder 

of site. 

4.1.3 Barrier 

All cells include a 2 mm thick high density polyethylene (HDPE) geomembrane. 

Seepage through a geomembrane sealing layer is dominated by leaks through flaws 

(holes) in the liner. The number of holes will depend on the effectiveness of the 

quality control during emplacement, but some holes will occur in all cases. Large 

holes will generally be detected, and so smaller holes or pinholes will be most 

common. For a geomembrane liner underlain by a mineral layer or host geology, the 

flow, qliner (m 3 year -1 ), through holes in the liner is given by: 

q 

0. 

1 0. 

9 0. 

74 

liner c 

aholes 

h 

K 

barrier 

3. 

16E 

07 


 



Version 2 

where c is a constant depending on the contact between the liner and the 

material beneath (0.21 for good contact, 1.15 for poor contact) 

(dimensionless). 

aholes is the area of the holes (m 2 ). 

h is the head of leachate (m). 

Kbarrier is the hydraulic conductivity of the barrier (material beneath the 

liner) (m s -1 ). 

3.16E+07 is the number of seconds in a year (s year –1 ). 

The parameter values assumed for the assessment are: 

c 0.5 

aholes 4.2 x 10 -4 m 2 

h 1 m 

Kbarrier 3 x 10 -10 m s -1 

Other assumed characteristics of the engineered clay barrier are presented in Table 

4.1. 

Parameter Value Rationale Reference 

Thickness 1.5 m Representative value based 

on design parameter 

Clay liner thickness is in the range 

0.5 to 2.5 m ( 1.5 m for future 

cells) – HRA, pp 62-70 

Hydraulic conductivity 3×10 -10 m/s Mean value Minimum (1×10 -11 m/s), mean 

(3×10 -10 m/s) and maximum 

(6.6×10 -10 m/s) permeability 

values for 115 clay samples from 

Cells 1A, 1B and 3A – HRA, p 107 

Porosity 0.05 Mean value Minimum (0.01), mean (0.05) and 

maximum (0.1) estimates based on 

specific yield values for clay – 

HRA, p 107 

Density 1560 kg/m 3 

Table 4.1 Properties of artificial clay barrier. 

Calculated mean value Minimum (1260 kg/m 3 ) and 

maximum (1860 kg/m 3 ) values for 

445 clay samples from Cells 1A, 

1B and 3A – HRA, p 107 

The geological barrier and unsaturated zone (Section 4.2 below) are treated as a single 

unit of thickness, D (m), and the advective transfer of radionuclide Rn through the 

unit, barrier (year -1 ), is given by: 

where qbarrier 

 

Rn 

barrier 

 

D 

a 

landfill 

 

barrier 

Rn 

K barrier 

d , barrier 


q 

barrier 

is the volume of the water flowing through the barrier (m 3 year -1 ). 



Version 2 

barrier is the porosity of the barrier (dimensionless). 

is the degree of saturation of the barrier (dimensionless). 

Kd,barrier is the distribution coefficient for radionuclide Rn in the barrier 

(m 3 kg -1 ). 

barrier is the bulk density of the barrier (kg m -3 ). 

alandfill is the area of the landfill (m 2 ). 

D is the depth of the barrier (m). 

The maximum value of qbarrier is determined by the product of the area of the landfill, 

alandfill (m 2 ) and the hydraulic conductivity of the unit, Kbarrier (m year -1 ). The actual 

value of qbarrier varies over the assessment period: 

For the period of operation of the landfill, qbarrier is set to qliner. 

After emplacement of the cap and while the site is still managed and 

monitored, qbarrier is set to the minimum of qliner and infiltration through the 

intact cap (see below). 

After the management phase, and before complete degradation of the cap, 

qbarrier is set to the minimum of the infiltration through the degraded cap (see 

below) and the value determined by the barrier properties. 

After degradation of the cap, qbarrier is determined by the barrier properties. 

The release of radioactivity over time into the geological barrier and radioactive decay 

result in a change to the inventory remaining in the landfill: 

where 

4.1.4 Cap 

 

Rn 

waste 

 

V 

A 

Rn 

landfill 

( t) 

 

barrier 

Rn 

K waste 

d , waste 


q 

ARn, 

initial 

e 

waste 

Rn 

t 

Rn 

waste is the rate constant for radionuclide Rn from loss of leachate (year -1 ). 

qbarrier is the volume of the water flowing through the geological barrier 

(m 3 year -1 ). 

t is the time (years). 

Rn is the radioactive decay constant of radionuclide Rn (year -1 ). 

ARn,initial is the initial inventory of each radionuclide (Bq). 

The assessment model does not require explicit details of cap construction. The 

volume of water available to infiltrate the landfill is assumed to be a function of the 

annual precipitation and the efficiency of the cap in diverting this precipitation. 

Peff total 

q P 

inf 

Rn 

eff alandfill 

PAE runoff 

1 

E ) 

( 0 

waste 

for tc < t te 



Version 2 

 

 

t te 

P 

eff Ptotal 

AE runoff 

E 1 

 

 

 

t f te 

 

Peff total 

PAE runoff 

 

1 0 for te < t tf 

for t > tf 

where qinf is the volume of water entering the landfill through the cap 

(m 3 year -1 ). 

alandfill is the area of the landfill (m 2 ). 

Peff is the potential rate of water infiltration through the cap of the 

landfill (m year -1 ). 

Ptotal is the total precipitation (m year -1 ). 

AE is the amount of precipitation that is lost by evapotranspiration 

(m year -1 ). 

runoff is the amount of precipitation lost by runoff (m year -1 ). 

E0 is the initial cap efficiency (a dimensionless fraction of the 

infiltration water initially deflected by the cap). 

t is the time after closure (years). 

tc is the time cap is emplaced (years). 

te is the time cap starts to degrade (years). 

tf is the time of cap failure (years). 

The HRA (p 31) provides a table of average monthly effective rainfall (rainfall minus 

potential evapotranspiration) for the period 1961 to 1990, indicating a mean annual 

value of ~ 0.072 m. In comparison with the annual rainfall for the area (~ 0.6 

m/year), this value is low for effective rainfall but is considered appropriate for net 

infiltration (effective rainfall minus runoff). 

Based on the assumptions for infiltration made in the HRA, the initial cap efficiency 

is set at 99%. The period of cap effectiveness is assumed to be 60 years and the 

period of cap degradation is assumed to be 100 years. 

In addition to providing protection to the landfill against infiltration, the cap also 

reduces potential doses from external radiation to members of the public living and 

working on the cap after closure. For these calculations, the cap is assumed to have a 

minimum thickness of 1.5 m. 

4.1.5 Operational period 

For the purpose of the assessment, it is assumed that the facility will continue in 

operation for a further 5 years, and that LLW disposals will take place throughout this 

period. 

It is proposed that waste will be transported to the site in suitable transport packages 

and disposed directly, and that loose waste will not be handled at the site. It is also 

proposed that waste will be covered by 0.3 m of soil material. These practices will 

ensure that members of the public off-site will not be exposed to groundshine. On- 




Version 2 

site workers, using mechanical handling to emplace the waste packages and soil cover 

or working near-by, may be exposed. Occupational dose assessments will be made 

separately from this report. 

4.1.6 Leachate collection and management procedures 

Leachate levels at the ENRMF are maintained by pumping excess leachate to tankers 

and transporting this leachate to a water treatment plant at Avonmouth. There may be 

periods when this route is unavailable – leachate could be re-circulated to the upper 

parts of the site but leachate would not be used for dust suppression or other processes 

that could lead to aerosol formation at the site. 

The radiological assessment includes two scenarios that could result in doses to offsite 

exposed groups: 

Tanker accident resulting in spillage of leachate and contamination of a water 

resource. 

Routine treatment of leachate and discharge of treated water to an estuary. 

The first of these scenarios has been considered using the SNIFFER methodology as 

discussed below. The second of the leachate scenarios has been considered by means 

of supplementary calculations described in Section 4.4. 

4.1.7 Leachate spillage 

Notwithstanding any radioactive components, landfill leachate poses a hazard to the 

environment if spilt and any road accident involving loss of an entire load would be 

subject to mitigation measures. Leachate that did enter water resources would also 

become diluted. For this assessment, it is conservatively assumed that an entire 

tanker load of leachate (30 m 3 of leachate) reaches a small reservoir (2 x 10 6 m 3 ) that 

is used for drinking water, irrigation and fishing. 

The dissolved radionuclide concentration, CRn,leachate (Bq m -3 ) in the leachate 

associated with an inventory ARn (Bq), is given by: 

C Rn, 

leachate 

 

V 

ARn 

Df 

 

 


landfill 

where Vlandfill is the volume of the waste (m 3 ). 

Df is the dissolution factor (-). 

waste is the porosity of the waste (dimensionless). 

is the degree of saturation of the waste (dimensionless). 

Parameter values used in the calculation of radionuclide concentrations in leachate are 

listed in Table 4.2. 

waste 



Version 2 


Vlandfill volume of the waste 672,098 m 3 

Df dissolution factor 0.001 - 

waste porosity of the waste 0.5 - 

degree of saturation of the waste 0.5 - 

Table 4.2 Parameter values used in calculating leachate concentrations. 

The contamination is assumed to relate to a one-off event, but the resulting 

radioactive contamination, CRn,water,spill (Bq m -3 ), is assumed to remain constant for one 

year (i.e., no dilution by throughflow): 

C 

Rn, 

water, 

spill 

C 

 

Rn, 

leachate 


V 

( t) 

V 

where CRn,leachate(t) is the concentration of radionuclide in the leachate at the time of the 

spill, t (Bq m -3 ). 

Vspill is the volume of leachate in the spill (m 3 ). 

Vwater is the volume of the surface water body (m 3 ). 

Dose calculations for drinking contaminated water, or ingesting fish taken from 

contaminated water are described in Appendix A. 

If the contaminated water body is used for irrigation, then a one-off soil 

concentration, CRn,soil,spill (Bq kg -1 ), is calculated from: 

where Irrigrate 

C 

Rn, 

soil, 

spill 

C 

Rn, 

water, 

spill 

water 

spill 

Irrig 

 

 

 

soil d 

is the amount of irrigation in one year (m). 

dsoil is the depth of the soil layer being irrigated (m). 

soil is the density of the soil (kg m -3 ). 

Dose calculations for ingestion of crops grown on irrigated soil and ingestion of 

contaminated soil are described in Appendix A. Decay constants and other 

radionuclide-specific parameter values are presented in Appendix B. Other parameter 

values used in the calculations of specific doses for the ENRMF are listed in Table 

4.3. 


Vspill volume of leachate in the spill 30 m 3 

Vwater volume of the surface water body 2 x 10 6 m 3 

Irrigrate amount of irrigation in one year 0.3 m 

dsoil depth of the soil layer being irrigated 1 m 

soil density of the soil 1300 kg m -3 

Table 4.3 Parameter values used in the calculation of the effects of leachate 

spillage. 

rate 

soil 

 

 

 

 



Version 2 

There is potential, during leachate management or spillage, for the production of 

aerosols which could lead to doses via the inhalation pathway. Other pathways, such 

as external irradiation from deposited aerosols or ingestion of foodstuffs contaminated 

by aerosols would give specific doses that are comparable to or less than the 

inhalation pathway. 

The concentration of aerosols at time t, CRn,air,aero(t) (Bq m -3 ), created during leachate 

management or spillage is assumed to be equivalent to the concentration of the 

leachate diluted by the aerosol load: 

aerosol 

t) 

C 

1000 

CRn, air, 

aero ( Rn, 

leachate 

where aerosol is the aerosol concentration (kg m -3 of air). 

CRn,leachate is the activity of radionuclide, Rn, in the leachate at time t (Bq m -3 ). 

1000 is the density of water (kg m -3 ). 

The above equation cautiously assumes that the aerosols are non-depleting during 

passage towards the exposed individual. An aerosol concentration of 0.001 kg m -3 of 

air is assumed. 

Exposure to aerosols at the ENRMF or during a tanker accident will be abnormal and 

short-lived. An initial assessment of the potential impacts from routine, off-site 

leachate management has been made using the Environment Agency’s methodology 

and the assumption that doses from water treatment would be similar to doses from 

sewage treatment. 

The Environment Agency’s methodology allows for a range of exposure groups 

affected by releases to a public sewer, depending on the discharge route for treated 

effluent. For this assessment, only the groups associated directly with operation of 

the treatment plant, farming of land conditioned by sludge or using the estuary are 

considered. These groups and the relevant exposure pathways are: 

Sewage treatment workers (adults only) 


( t) 

External irradiation from radionuclides in raw sewage and sludge 

Inadvertent inhalation and ingestion of raw sewage and sludge containing 

radionuclides 

Farming family living on land conditioned with sewage sludge 

Consumption of food produced on land conditioned with sludge and 

incorporating radionuclides 

External irradiation from radionuclides in sludge conditioned soil 

Inadvertent inhalation and ingestion of sludge conditioned soil 

Fisherman family (estuary/coastal water receives treated effluent from sewage 

works, typically via a river) 



Version 2 

External irradiation from radionuclides deposited in sediments 

Consumption of fish incorporating radionuclides 

A key assumption in assessing potential doses from off-site leachate management is 

the extent of dilution with other inputs to the water treatment plant. For this 

assessment, it is assumed that the Avonmouth facility treats some 1,080 m 3 per day, 

based on the use of six anaerobic digesters, with a daily feed of about 180 m 3 per day 

each. 

4.1.8 Control over future site use 

It is intended that the future use of the ENRMF site, after closure, would be for 

agriculture, and that normal agricultural practices, combined with knowledge of the 

previous site use, would prevent intrusion into the waste or the excavation of 

radioactive material. However, for the purpose of the assessment, it is assumed that 

knowledge of the site will be lost and there will be no control over use of the site at 

some time after closure. 

Loss of control means that there is a potential for the site to be disturbed and for 

radioactive material to be incorporated into soil used to grow crops and graze animals. 

The calculations of effective doses to workers engaged in excavation activities and to 

members of the public residing on the disturbed facility are described in Appendix A. 

The principal assumption is that control over the site will be lost 60 years after 

closure. To illustrate the sensitivity to this assumption, alternative cases in which it is 

assumed that control over the site will be lost 20 years and 100 years after closure 

have also been considered. 

4.2 Hydrogeological setting 

Hydrogeology data were derived from the HRA (ESI, 2004) and the Environmental 

Statement (Bullen Consultants Ltd, 2005) prepared for assessments of hazardous 

waste disposal at the ENRMF. 

4.2.1 Underlying geology 

ENRMF landfill site was originally a clay pit, which has been largely quarried out, 

exposing the top of the underlying Lincolnshire Limestone, the thickness of which is 

between 15 m to 20 m. The upper part of the formation comprises a sequence of 

oolitic limestones, whereas the lower part is composed of fine grained sandy 

limestones. The Lincolnshire Limestone has been classified as a Major Aquifer by 

the Environment Agency. It is characterised by fracture flow and there are also 

swallow holes in the vicinity which allow rapid flow of water from the surface to the 

water table. 




Version 2 

4.2.2 Unsaturated zone characteristics 

Water levels records from monitoring boreholes at the site indicate fluctuation in 

groundwater levels in the range of approximately 3 m to 7 m below the base of Cells 

1 and 2 (HRA, p. 40). Although, given the uncertainties in the dataset, it is possible 

for the unsaturated zone to be no more than 3 m thick across the entire site, it is 

considered that assuming this minimum thickness would be overly conservative. For 

the purpose of the assessment the mean value, 5.5 m, of the range of values reported 

has been used. 

The unsaturated zone is modelled together with the geological barrier as a single unit 

(see Section 4.1). The approach adopted ignores any dispersion effects in the 

unsaturated zone. 

4.2.3 Saturated zone characteristics 

The thickness of the saturated zone across the base of the landfill site can be derived 

from the estimates for the thickness of the Lincolnshire Limestone and that of its 

unsaturated zone. As mentioned above, the latter is in the range of 3-7 m and the 

former in the range of 15-20 m, suggesting a thickness for the saturated zone in the 

range of 7-17 m, with a mean value of 12 m. (Note that the Environmental Statement 

suggests a thickness for this layer of approximately 7-18 m.) 

There is uncertainty on the range of hydraulic conductivity values for the Lower 

Lincolnshire Limestone formation (saturated zone) underneath the landfill site. 

Values reported in 1998 suggest a hydraulic conductivity of 0.01 to 0.1 m/day 

(Environmental Statement, p. 44), whereas the results of the most recent slug tests 

indicate hydraulic conductivity in the range of 1 to 7 m/day (HRA, p. 38). The 

assessment uses a mean value derived from the more recent findings. 

The permeability and transmissivity of the Lincolnshire Limestone are mainly due to 

fractures and, as such, the fracture porosity of this formation is used as the effective 

porosity in the assessment. 

Parameter values assumed for the saturated zone are presented in Table 4.4. 




Version 2 


Thickness 12 m Calculated mean value The Lincolnshire Limestone 

formation is 15-20 thick, with an 

unsaturated zone thickness of 3-7 m 

– HRA, pp 34 and 40 

Hydraulic conductivity 4.63×10 -5 m/s Calculated mean value Minimum (1 m/day) and maximum 

(7 m/day) values derived from slug 

tests – HRA, p 38 

Hydraulic gradient 0.0025 Calculated mean value 0.002 - 0.003 – HRA, p 41 

Porosity 0.007 Calculated mean value 0.004 – 0.01 – HRA, p 107 

Density 2000 kg/m 3 Estimated density for 

sedimentary rock 

Table 4.4 Properties of saturated zone. 

4.2.4 Groundwater discharges 

Estimated density for sedimentary 

rock – HRA, p 108 

Local groundwater level contours at the landfill site indicate that groundwater flows 

approximately southwards to south-eastwards (HRA, p. 41). Hence, for assessment 

purposes, only groundwater abstractions south to south-easterly of the landfill site are 

considered. The nearest such active licensed groundwater abstraction point is at 

Law’s Lawn, 1487 m south-east of the site – the water is abstracted from the confined 

aquifer under artesian conditions and is used solely for agricultural purposes 

(Environmental Statement, p 48) - Table 4.5. 


Distance to nearest 

groundwater abstraction 

point 

Abstracted water usage 

Table 4.5 Groundwater abstraction. 

4.2.5 Stream and river characteristics 

1487 m Exact value Groundwater is abstracted at Law's 

Lawn (1,487 m south-east of the 

site) for agricultural usage 

Irrigation Assumed sub-activity under 

agricultural usage 

Livestock Assumed sub-activity under 

agricultural usage 

As above 

As above 

There are no natural surface water features on, and no known springs in the vicinity 

of, the site. The nearest spring is about 1 km south-east of the site, but no further data 

are available. The nearest natural surface water course is the River Welland, 

approximately 2.5 km to the west of the site – only data on water quality are available 




Version 2 

for this stream. Hence, for the assessment, it is assumed that groundwater does not 

discharge to a stream or river, but is instead abstracted from a borehole (see above). 

4.2.6 Groundwater flow and radionuclide transport 

The groundwater flow and radionuclide transport model is based on a series of units 

or compartments, with the transfer of radionuclide Rn from each unit to the next 

downstream, gw (year -1 ), being given by: 

 

Rn 

gw 

 

L 

K 

Rn 

K gw 

d , rock 


gw 

H 

where Kgw 

is the hydraulic conductivity of the rock in which the groundwater 

flow is occurring (m year -1 ). 

H is the hydraulic gradient (dimensionless). 

gw is the porosity of the groundwater pathway (dimensionless). 

Kd,rock is the distribution coefficient for radionuclide Rn in the rock 

(m 3 kg -1 ). 

gw is the bulk density of the groundwater pathway (kg m -3 ). 

L is the length of each groundwater compartment (m). 

Longitudinal dispersion is approximated implicitly by dividing the path length into 

ten units. Transverse dispersion is approximated by successively increasing the width 

(and, thereby, the volume) of each downstream unit to account for spreading of the 

plume of contaminated groundwater. The width, W (m), at a distance, x (m), 

downstream is given by: 

gw 

W W 

T x 24 

2 2 

0 

where Wo 

is the initial width of the unit in which the groundwater flow is 

occurring (m). 

x is the distance downstream (m). 

T is the transverse dispersion length (m), assumed to be one tenth of the 

initial width. 

For the calculation of radionuclide concentrations in a borehole at the site boundary, 

the overall groundwater path length is assumed to be 100 m, representing flow from a 

point below the centre of the site to the site boundary. 

The concentration of a radionuclide in water abstracted from groundwater at time t, 

CRn,water(t) (Bq m -3 ), is given by: 

C 

Rn, 

water 

( t) 

 

V 

gw 

 

A 

Rn 

K gw 

Rn, 

gw 

( t) 

gw 

d , rock 

where ARn,gw(t) is the activity in the groundwater compartment at time t (Bq). 

Vgw is the volume of the groundwater compartment (m 3 ). 

gw is the porosity of the groundwater pathway (dimensionless). 



Version 2 

Kd is the distribution coefficient for radionuclide Rn in the rock 

(m 3 kg -1 ). 

gw is the bulk density of the rock for the groundwater pathway (kg m -3 ). 

The change in concentration of radionuclides in soil, CRn,soil (Bq kg -1 ), that is irrigated 

with contaminated water is given by: 

dC 

Rn, 

soil 

dt 

Irrig rate 

CRn, 

water ( t) 

 

eff C 

Rn, 

soil d 

 

soil 

where CRn,water(t) is the concentration of radionuclide in the water used for irrigation at 

time t (Bq m -3 ). 

Irrigrate is the rate of irrigation (m year -1 ). 



eff is an effective decay coefficient that considers radioactive decay, 

leaching from the soil, uptake by plants, and erosion (year -1 ), given 

by: 

eff Rn 

Ptotal 

AE runoff TFplant 

Yield 

plant 

 

 

Rn 

 

 

 

Rn 

erosion 

d soil soil soil K d soil soil d 

 

, 

soil 

where Ptotal is the total precipitation (m year -1 ). 

AE is the amount of precipitation that is lost by evapotranspiration 

(m year -1 ). 

runoff is the amount of precipitation lost by runoff (m year -1 ). 

Yieldplant is the plant yield (kg m -2 year -1 ). 

TFplant is the soil to plant transfer factor for radionuclide, Rn (Bq kg -1 fresh 

weight of crop per Bq kg -1 of soil). 


soil is the bulk density of the soil (kg m -3 ). 

soil is the porosity of the soil (dimensionless). 

is the degree of saturation (dimensionless). 

Kd,soil is the distribution coefficient for radionuclide Rn in the soil (m 3 kg -1 ). 

Rn is the decay constant of radionuclide Rn (year -1 ). 

erosion is the loss of radioactivity owing to erosion of the soil (year -1 ). 

The removal of activity from the groundwater through the irrigation process is not 

tracked. 

Dose calculations for the groundwater pathway are described in Appendix A. Decay 

constants and other radionuclide-specific parameter values are presented in Appendix 

B. Other parameter values used in the calculations of specific doses for the ENRMF 

are listed in Table 4.6. 



soil 

( t)


Version 2 

Parameter Description Value Unit 

Irrigrate rate of irrigation 0.3 m year -1 

dsoil depth of the soil layer being irrigated 1.0 m 

soil density of the soil 1300 kg m -3 

Yieldplant plant yield 

Pasture 1.7 

kg m 

Grain 0.4 

Green veg 3.0 

Root veg 3.5 

-2 year -1 

soil porosity of the soil 0.3 dimensionless 

degree of saturation 0.5 dimensionless 

erosion 

loss of radioactivity owing to erosion 

of the soil 

2.0x10 -4 

year -1 

Table 4.6 Parameter values used in the calculation of radionuclide concentrations 

in groundwater used for irrigation and in irrigated soil. 

4.3 Other scenarios and pathways 

4.3.1 Gas 

The gas pathway has been considered only for tritium and radon. It is not envisaged 

that there would be sufficient organic waste material in the LLW to generate 

radiogenic CO2 or CH4. Potentially exposed groups for the gas pathway are site 

workers, members of the public spending time immediately downwind of the site 

during the operational period, and members of the public living in a house built on the 

site after closure. 

Radioactive Gas Release 

For H-3 (in hydrogen, water, or methane) and C-14 (in carbon dioxide or methane), 

the release rate of radioactive gas, RRn,gas (Bq year -1 ), at time t (years) is given by: 

R 

Rn, 

gas 

( t) 

 

A 

Rn, 

waste 

where: ARn,waste is the initial activity of radionuclide Rn in the waste (Bq). 

Rn is the decay constant of radionuclide Rn (year -1 ). 

fgas is the fraction of the activity associated with each gas 


gas is the average timescale of generation of each gas (years). 


e 

 

Rnt 

For radon (Rn-222), the release rate at time t is given by: 

Ra 

226t 

R ( t) 

a C e 

H 

radon 

Rn222 

Ra-226, 

waste 

where: is the decay constant of the indicated radionuclide(year –1 ). 

a is the surface area of the disposal unit (m 2 ). 

CRa-226,waste is the initial Ra-226 concentration in the waste (Bq kg –1 ). 

gas 

 

f 

gas 

waste 


1 

e 

-h 

H 

2 

2

4.3.2 Fire 


Version 2 

waste is the bulk density of the waste (kg m -3 ). 

is the emanation factor, defined as the fraction of the radon atoms 

produced which escape from the solid phase of the waste into the 

pore spaces (dimensionless). 

H1 is the effective diffusion relaxation length for the waste (m). 

h2 is the thickness of the cover (m). 

H2 is the effective relaxation length of the cover (m). 

Dose calculations for the gas pathway are described in Appendix A. Decay constants 

and other radionuclide-specific parameter values are presented in Appendix B. Other 

parameter values used in the calculations of specific doses for the ENRMF landfill are 

listed in Table 4.7. 


fgas fraction of the activity associated with 

tritium 

3.9x10 -2 dimensionless 

gas average timescale of generation of 

tritium 

50 years 

a surface area of the disposal unit 34,108 m 2 

waste bulk density of the waste 700 kg m -3 

emanation factor, defined as the 

fraction of the radon atoms produced 

which escape from the solid phase of 

the waste into the pore spaces 

H1 effective diffusion relaxation length 

for the waste 

0.1 dimensionless 

0.2 m 

h2 thickness of the cover 1.5 m 

H2 effective relaxation length of the cover 0.2 m 

Table 4.7 Parameter values used in the calculation of gas release rates during 

operations and after closure. 

Fire is a potential issue at landfill sites where LLW is disposed of alongside municipal 

and other wastes with large amounts of combustible material. It is not envisaged that 

there would be significant amounts of combustible material amongst the LLW or the 

hazardous waste, and fires within existing disposal cells would not affect the cells 

containing LLW. The consequences of a fire starting within or affecting the LLW 

have therefore not been assessed. There is a potential for accidents such as aircraft 

impact to release material in a similar manner to a fire, but the scale and nonradiological 

consequences of this type of accident means that they are more 

appropriately discussed in qualitative terms in the overall safety case rather than 

modelled within the radiological assessment. 

4.3.3 Barrier failure 

This scenario was included in the SNIFFER methodology to account for the 

possibility of damage or defects in the lining and a damaged or inadequate geological 

barrier could lead to leachate release during operations. This is a conservative 




Version 2 

scenario even for a non-hazardous waste site with LLW disposals. It is considered 

unreasonable for a hazardous waste site receiving LLW where the construction, 

operation and monitoring will all reduce the possibility of the barrier failing in a 

manner that allows the release of large amounts of leachate. Even if damage did 

occur, the potential for environmental damage from leachate from such a site would 

ensure that remediation would occur before members of the public were exposed. 

The barrier failure scenario has therefore not been assessed. 

4.3.4 Site remediation and re-engineering 

This scenario was included in the SNIFFER methodology because it was possible that 

a site operator would have no records of radioactive waste disposals or their location. 

In the case of comparatively large volumes of LLW disposed of to a hazardous waste 

landfill, records would be maintained. Any remediation work would be done with the 

knowledge that there was radioactive material on the site and it can be assumed that 

appropriate precautions against exposure would be adopted. 

4.3.5 Bathtubbing 

This scenario was included in the SNIFFER methodology to account for the 

possibility of excessive infiltration through the cap at a time when the barrier still 

prevents leakage to the underlying formation. For a hazardous waste site, it is 

envisaged that controls on cap construction and leachate monitoring would prevent or 

identify releases through this pathway. For the purpose of the assessment, it has been 

assumed that remediation, cognisant of radioactive material, would occur before 

members of the public were exposed via this pathway. 




Version 2 

5 Dose Calculations 

This section presents results from the calculations of specific dose arising from the 

different scenarios and pathways described in Section 4 and the dose calculations 

described in Appendix A. 

Radiological capacity calculations for the ENRMF, based on these dose calculations, 

are presented in Section 6. 

5.1 Groundwater pathway 

Specific doses calculated for members of the public via the groundwater pathway are 

presented in Table 5.1. These calculations are based on assumptions described in 

Sections 4.1 and 4.2 and in Appendix A (Section A.4). Sensitivity studies showing 

the effect of variations in leachate head, cap lifetime, cap efficiency, the length of the 

assessment period and the exposed individual on the results for the groundwater 

pathway are presented in Appendix C. 

Radionuclide 

Specific dose 

(μSv y -1 per MBq) 

Borehole 

1500m 

Irrigation 

Site 

boundary 

Drinking 

Radionuclide 

Specific dose 


Borehole 

1500m 

Irrigation 

Site 

boundary 

Drinking 

H-3 3.66E-30 6.89E-23 Ra-226 1.56E-08 1.36E-06 

C-14 2.06E-09 1.39E-07 Ac-227 5.66E-26 6.70E-19 

Cl-36 6.52E-08 1.69E-06 Th-229 1.44E-08 1.29E-06 

Fe-55 1.04E-43 4.45E-36 Th-230 8.24E-09 7.35E-07 

Co-60 1.21E-39 3.79E-32 Th-232 4.04E-08 3.59E-06 

Ni-63 7.94E-21 3.47E-15 Pa-231 3.60E-08 3.08E-06 

Sr-90 3.04E-24 2.19E-17 U-232 5.07E-20 6.19E-14 

Nb-94 3.76E-10 9.07E-09 U-233 4.62E-09 4.13E-07 

Tc-99 2.12E-09 1.52E-07 U-234 4.35E-09 3.89E-07 

Ru-106 3.44E-46 1.67E-38 U-235 4.35E-09 3.87E-07 

Ag-108m 1.68E-17 1.51E-12 U-236 4.22E-09 3.78E-07 

Sb-125 4.81E-42 2.03E-34 U-238 4.35E-09 3.89E-07 

Sn-126 1.20E-08 4.87E-07 Np-237 1.22E-06 9.52E-05 

I-129 1.38E-05 3.02E-04 Pu-238 1.86E-12 1.67E-10 

Ba-133 1.44E-31 3.25E-24 Pu-239 1.54E-08 1.37E-06 

Cs-134 1.82E-43 8.12E-36 Pu-240 1.04E-08 9.33E-07 

Cs-137 1.28E-25 9.32E-19 Pu-241 1.59E-10 3.62E-08 

Pm-147 3.41E-44 1.47E-36 Pu-242 1.69E-08 1.51E-06 




Version 2 

Radionuclide 

Specific dose 


Borehole 

1500m 

Irrigation 

Site 

boundary 

Drinking 

Radionuclide 

Specific dose 


Borehole 

1500m 

Irrigation 

Site 

boundary 

Drinking 

Eu-152 2.77E-32 4.84E-25 Am-241 5.37E-11 1.08E-08 

Eu-154 3.28E-35 7.98E-28 Cm-243 1.82E-10 1.63E-08 

Eu-155 3.63E-41 1.29E-33 Cm-244 1.29E-09 1.15E-07 

Pb-210 1.25E-25 1.43E-18 

Table 5.1 Specific doses to members of the public via the groundwater pathway. 

Results include doses arising from ingrowth of daughter radionuclides 

for 100 years. 

The results from the sensitivity studies show that the calculated doses for most 

radionuclides are not sensitive to leachate head. This is because the most significant 

releases take place after the engineered barriers (cap and liner) have degraded and 

radionuclide transport into groundwater is governed by the infiltration rate through 

soil and the properties of the geological barrier. Radionuclides with short half-lives 

do show some sensitivity to leachate head, because they have largely decayed by the 

time the barriers degrade. Calculated doses for these radionuclides are determined by 

the relatively small releases while the barriers are effective, and the magnitude of 

these releases is governed by the leachate head. The small changes in calculated 

doses for longer-lived radionuclides arise because some of the inventory is lost from 

the site while the barriers are effective, and this reduces the inventory available for 

later release. 

Results from the sensitivity studies for cap lifetime again show a dependency on halflife. 

Calculated doses for long-lived radionuclides show little variation with cap 

lifetime because they are not released in significant amounts during the period the cap 

is effective. Radionuclides with shorter half-lives show a sensitivity to cap lifetime; 

this affects whether a significant inventory is still available for release once the 

engineered barriers have degraded. 

Sensitivity studies for cap efficiency show that this has relatively little effect on 

calculated doses. This is because it is the barriers at the base of the landfill that have 

most effect on the release of radionuclides during the period when the engineered 

barriers are effective. In practice, a less effective cap would allow more infiltration 

which would lead to an increase in leachate head and potentially to bath-tubbing if the 

site was not monitored and managed. The assessment methodology used does not 

explicitly model these links and so the secondary effects of changes in cap efficiency 

are not apparent in the calculated doses. 

In the case of the sensitivity studies on the effects of varying the assessment period, 

the reverse of the effects discussed above is apparent – calculated doses for shortlived 

radionuclides show little or no sensitivity and those for long-lived radionuclides 

are very sensitive. This is because radionuclides are released only slowly to the 




Version 2 

groundwater system, even after the engineered barriers become degraded. The 

geological barrier in particular has an important role in retarding transport of 

radionuclides. For short-lived radionuclides, the barriers are effective enough that 

radioactive decay reduces the inventory to less significant levels during the 

assessment period. For long-lived radionuclides, even a 5,000 year assessment period 

does not lead to significant radioactive decay and there is no peak in the calculated 

dose. For these radionuclides, the assessment model is effectively assuming that the 

inventory is transferred to a part of the accessible environment where it leads to 

increasing calculated doses with time. In practice, there will be more dispersion of 

the radionuclides over long periods, reducing calculated doses. 

A comparison of calculated specific doses for exposed individuals in different age 

groups shows that for the majority of the radionuclides assessed, specific doses to 

adults are higher than those to infants or children. This arises because the adult rates 

of consumption for foodstuffs grown on contaminated soil are sufficiently higher then 

those for infants and children to off-set the higher dose coefficients for these age 

groups. In the case of Cl-36, specific doses to children and infants are higher than 

those to adults, but the difference is less than a factor of 10. 

5.2 Irradiation pathway 

Specific doses through external irradiation to members of the public living on the site 

after closure, are presented in Table 5.2. These calculations are based on the 

assumptions described in Section 4.1 and Appendix A (Section A.2). 

Radionuclide 

Specific dose 


Radionuclide 

Specific dose 


H-3 0.00E+00 Ra-226 3.14E-18 

C-14 1.29E-45 Ac-227 7.80E-23 

Cl-36 3.09E-20 Th-229 2.24E-21 

Fe-55 0.00E+00 Th-230 8.06E-20 

Co-60 1.17E-14 Th-232 3.41E-22 

Ni-63 0.00E+00 Pa-231 1.70E-17 

Sr-90 2.22E-25 U-232 9.70E-47 

Nb-94 2.49E-12 U-233 1.27E-23 

Tc-99 2.61E-33 U-234 1.39E-22 

Ru-106 1.05E-32 U-235 2.15E-20 

Ag-108m 3.23E-13 U-236 4.15E-23 

Sb-125 1.07E-20 U-238 2.96E-24 

Sn-126 1.76E-26 Np-237 1.65E-27 

I-129 1.96E-91 Pu-238 1.42E-22 

Ba-133 5.59E-15 Pu-239 6.36E-28 




Version 2 

Radionuclide 

Specific dose 


Radionuclide 

Specific dose 


Cs-134 8.72E-22 Pu-240 2.04E-27 

Cs-137 1.97E-21 Pu-241 2.55E-30 

Pm-147 2.92E-34 Pu-242 8.14E-33 

Eu-152 1.90E-13 Am-241 4.84E-30 

Eu-154 4.30E-14 Cm-243 4.80E-16 

Eu-155 5.99E-31 Cm-244 4.43E-28 

Pb-210 6.65E-62 

Table 5.2 Specific doses to workers and members of the public via the external 

irradiation pathway. Results for doses to the public include doses 

arising from ingrowth of daughter radionuclides for 60 years. 

5.3 Intrusion 

Specific doses to workers intruding into the waste and to members of the public living 

on excavated waste after intrusion are presented in Table 5.3. These calculations 

assume that the LLW is disposed of to all of the remaining cells at the site. Other 

assumptions are described in Appendix A (Section A.3). 

Radionuclide 

Intruder 

20 years 

Specific dose 


Intruder 

60 years 

Resident 

60 years 

Resident 

100 years 

H-3 4.31E-11 4.52E-12 1.28E-10 1.35E-11 

C-14 7.09E-09 7.05E-09 8.68E-09 8.63E-09 

Cl-36 3.02E-08 3.02E-08 4.28E-07 4.28E-07 

Fe-55 1.40E-12 4.85E-17 8.13E-18 2.82E-22 

Co-60 2.44E-06 1.27E-08 4.19E-11 2.17E-13 

Ni-63 1.16E-09 8.68E-10 2.60E-10 1.95E-10 

Sr-90 2.49E-07 9.59E-08 3.31E-07 1.28E-07 

Nb-94 7.53E-05 7.52E-05 2.44E-07 2.44E-07 

Tc-99 1.09E-08 1.09E-08 3.68E-07 3.68E-07 

Ru-106 1.16E-14 1.39E-26 7.44E-29 8.90E-41 

Ag-108m 6.38E-05 5.13E-05 1.68E-07 1.35E-07 

Sb-125 1.13E-08 5.68E-13 1.89E-15 9.47E-20 

Sn-126 1.34E-04 1.34E-04 4.90E-07 4.90E-07 

I-129 1.06E-06 1.06E-06 1.36E-06 1.36E-06 

Ba-133 2.20E-06 1.65E-07 5.63E-10 4.22E-11 




Version 2 

Radionuclide 

Intruder 

20 years 

Specific dose 


Intruder 

60 years 

Resident 

60 years 

Resident 

100 years 

Cs-134 3.70E-09 6.83E-15 2.76E-17 5.09E-23 

Cs-137 1.41E-05 5.60E-06 2.61E-08 1.04E-08 

Pm-147 1.52E-12 3.54E-17 4.55E-19 1.06E-23 

Eu-152 1.14E-05 1.42E-06 4.63E-09 5.76E-10 

Eu-154 5.63E-06 2.41E-07 7.87E-10 3.37E-11 

Eu-155 2.15E-08 8.04E-11 2.74E-13 1.03E-15 

Pb-210 7.61E-06 2.19E-06 1.28E-07 3.70E-08 

Ra-226 1.10E-04 1.08E-04 5.40E-06 5.31E-06 

Ac-227 7.65E-05 2.14E-05 1.81E-08 5.07E-09 

Th-229 9.52E-05 9.48E-05 8.84E-08 8.80E-08 

Th-230 3.31E-05 3.49E-05 1.56E-07 2.43E-07 

Th-232 1.87E-04 1.87E-04 4.75E-07 4.75E-07 

Pa-231 8.60E-05 6.82E-05 1.66E-06 1.65E-06 

U-232 1.34E-05 8.99E-06 1.57E-08 1.05E-08 

U-233 3.54E-06 3.89E-06 4.32E-09 4.65E-09 

U-234 3.28E-06 3.29E-06 3.71E-09 3.78E-09 

U-235 8.92E-06 8.96E-06 2.47E-08 2.61E-08 

U-236 1.38E-06 1.38E-06 3.14E-09 3.14E-09 

U-238 3.88E-06 3.88E-06 6.87E-09 6.88E-09 

Np-237 2.47E-05 2.47E-05 4.65E-08 4.65E-08 

Pu-238 2.79E-05 2.03E-05 9.86E-09 7.19E-09 

Pu-239 3.85E-05 3.84E-05 1.86E-08 1.86E-08 

Pu-240 3.84E-05 3.82E-05 1.85E-08 1.84E-08 

Pu-241 1.74E-06 1.47E-05 8.29E-09 5.27E-08 

Pu-242 3.54E-05 3.54E-05 1.76E-08 1.76E-08 

Am-241 2.97E-05 2.79E-05 1.57E-08 1.47E-08 

Cm-243 7.50E-06 3.03E-06 4.49E-09 1.92E-09 

Cm-244 6.43E-06 1.57E-06 9.45E-10 2.38E-09 

Table 5.3 Specific doses to intruders into the landfill and to members of the 

public living on excavated waste. Results include doses arising from 

ingrowth of daughter radionuclides over 20, 60 and 100 years as 

appropriate. 




Version 2 

5.4 Leachate management and spillage 

5.4.1 Leachate management 

Specific doses to members of three exposed groups at a sewage treatment works 

receiving leachate from the ENRMF are presented in Table 5.4. These results are 

based on the Environment Agency’s Initial Assessment methodology (Environment 

Agency 2006a; 2006b), which does not include the same range of radionuclides as 

used in the remainder of the radiological assessment. 

These results are based on the assumption that the sewage treatment works has an 

overall throughput of 1,080 m 3 of effluent per day and that the average exchange rate 

in the estuary is 30 m 3 / s. The specific doses for the STW worker and farming family 

are sensitive to the throughput (decreasing as overall throughput increases), and the 

specific doses for the fisherman are sensitive to the exchange rate (decreasing as the 

exchange rate increases). 

Radionuclide STW worker 

Specific dose 

(microSv / y per MBq / y) 

Farming 

family Fisherman 

H-3 2.11E-09 2.83E-06 2.52E-09 

C-14 7.78E-08 4.72E-03 1.30E-03 

Cl-36 1.33E-06 7.78E-02 4.80E-09 

Fe-55 2.00E-07 1.33E-03 1.00E-07 

Co-60 4.94E-02 7.78E-01 1.87E-03 

Sr-90 2.28E-05 2.17E-02 1.83E-05 

Tc-99 1.17E-07 2.83E-01 2.10E-05 

Ru-106 6.11E-04 3.06E-03 1.44E-04 

I-129 2.44E-05 6.11E-02 6.67E-05 

Cs-134 1.11E-02 1.17E-01 2.80E-04 

Cs-137 4.11E-03 1.00E-01 3.50E-04 

Pm-147 2.11E-07 1.67E-05 6.50E-07 

Eu-152 1.39E-02 2.67E-01 3.67E-03 

Eu-154 1.50E-02 2.72E-01 3.33E-03 

Eu-155 3.33E-04 5.06E-03 6.17E-05 

Pb-210 4.44E-04 5.33E-01 6.33E-02 

Ra-226 2.22E-02 5.56E-01 1.83E-03 

Th-230 3.22E-04 1.28E-02 3.67E-05 

Th-232 4.89E-04 1.39E+00 2.23E-03 




Version 2 

U-234 1.11E-05 1.17E-03 3.90E-05 

U-235 3.61E-04 7.78E-03 6.60E-05 

U-238 8.33E-05 2.06E-03 4.20E-05 

Np-237 3.67E-04 7.22E-02 6.00E-04 

Pu-238 4.44E-04 2.00E-02 2.67E-03 

Pu-239 4.83E-04 2.28E-02 2.83E-03 

Pu-240 4.83E-04 2.28E-02 2.83E-03 

Pu-241 8.89E-06 3.39E-04 5.33E-05 

Pu-242 4.67E-04 2.22E-02 2.67E-03 

Am-241 8.33E-04 3.94E-02 2.37E-05 

Cm-243 2.44E-03 6.67E-02 1.00E-04 

Cm-244 4.50E-04 1.78E-02 9.00E-06 

Table 5.4 Specific doses to workers at a sewage treatment works (STW) 

receiving leachate from the ENRMF and to members of the public 

using resources contaminated by effluent from the STW. 

5.4.2 Leachate spillage 

Specific doses calculated for members of the public via the leachate spillage pathway 

are presented in Table 5.5. These calculations are based on the assumptions described 

in Section 4.1.8 and in Appendix A. 

Radionuclide 

Pathways associated 

with water contaminated 

by leachate 

Drinking 

water 

Specific dose 


Fish Crops 

Pathways associated with soil 

contaminated by irrigation with 

contaminated water 

Livestock and 

associated 

products 

H-3 1.58E-12 4.34E-15 3.53E-11 5.17E-13 1.50E-17 

C-14 5.11E-11 1.26E-09 1.12E-09 2.08E-11 4.84E-16 

Cl-36 8.19E-11 1.12E-11 1.82E-09 4.26E-11 7.76E-16 

Fe-55 2.91E-11 7.96E-12 6.39E-10 1.54E-12 2.76E-16 

Co-60 2.99E-10 2.46E-10 6.59E-09 8.64E-12 2.84E-15 

Ni-63 1.32E-11 3.62E-12 2.91E-10 1.92E-12 1.25E-16 

Sr-90 2.70E-09 4.44E-10 5.96E-08 2.02E-10 2.56E-14 

Nb-94 1.50E-10 1.23E-10 3.29E-09 6.25E-16 1.42E-15 

Tc-99 5.63E-11 3.09E-12 1.28E-09 7.94E-14 5.34E-16 

Ru-106 6.16E-10 1.69E-11 1.36E-08 8.19E-11 5.84E-15 


Soil 



Version 2 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 




Livestock and 

associated 

products 

Ag-108m 2.02E-10 2.77E-12 4.46E-09 1.03E-13 1.92E-15 

Sb-125 2.74E-10 7.50E-11 6.03E-09 8.54E-14 2.60E-15 

Sn-126 6.25E-10 1.71E-09 1.38E-08 8.63E-12 5.93E-15 

I-129 9.68E-09 7.96E-10 2.13E-07 1.86E-09 9.18E-14 

Ba-133 1.32E-10 1.45E-12 2.91E-09 7.24E-13 1.25E-15 

Cs-134 1.67E-09 9.17E-09 3.68E-08 3.31E-10 1.59E-14 

Cs-137 1.14E-09 6.27E-09 2.52E-08 2.27E-10 1.09E-14 

Pm-147 2.29E-11 1.88E-12 5.04E-10 3.06E-13 2.17E-16 

Eu-152 1.23E-10 1.01E-11 2.71E-09 2.04E-13 1.17E-15 

Eu-154 1.76E-10 1.45E-11 3.88E-09 2.91E-13 1.67E-15 

Eu-155 2.82E-11 2.32E-12 6.20E-10 4.65E-14 2.67E-16 

Pb-210 1.66E-07 1.37E-07 3.66E-06 5.88E-10 1.58E-12 

Ra-226 1.91E-07 2.62E-08 4.21E-06 2.52E-09 1.81E-12 

Ac-227 1.06E-07 2.33E-07 2.34E-06 4.52E-11 1.01E-12 

Th-229 5.40E-08 4.44E-09 1.19E-06 3.87E-10 5.12E-13 

Th-230 1.85E-08 1.52E-09 4.07E-07 1.32E-10 1.75E-13 

Th-232 9.36E-08 7.70E-09 2.06E-06 6.70E-10 8.88E-13 

Pa-231 6.25E-08 1.71E-09 1.38E-06 1.09E-11 5.93E-13 

U-232 4.05E-08 1.11E-09 8.91E-07 1.66E-10 3.84E-13 

U-233 4.49E-09 1.23E-10 9.88E-08 1.83E-11 4.26E-14 

U-234 4.31E-09 1.18E-10 9.49E-08 1.76E-11 4.09E-14 

U-235 4.17E-09 1.14E-10 9.17E-08 1.70E-11 3.95E-14 

U-236 4.14E-09 1.13E-10 9.11E-08 1.69E-11 3.92E-14 

U-238 4.26E-09 1.17E-10 9.38E-08 1.74E-11 4.04E-14 

Np-237 9.76E-09 2.67E-10 2.15E-07 2.62E-11 9.26E-14 

Pu-238 2.02E-08 2.22E-10 4.46E-07 7.17E-13 1.92E-13 

Pu-239 2.20E-08 2.41E-10 4.84E-07 7.80E-13 2.09E-13 

Pu-240 2.20E-08 2.41E-10 4.84E-07 7.80E-13 2.09E-13 

Pu-241 4.23E-10 4.63E-12 9.30E-09 1.50E-14 4.01E-15 

Pu-242 2.11E-08 2.32E-10 4.65E-07 7.49E-13 2.00E-13 


Soil 



Version 2 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 




Livestock and 

associated 

products 

Am-241 1.76E-08 1.45E-09 3.88E-07 2.08E-12 1.67E-13 

Cm-243 1.32E-08 1.09E-09 2.91E-07 3.59E-12 1.25E-13 

Cm-244 1.07E-08 8.76E-10 2.34E-07 2.90E-12 1.01E-13 

Table 5.5 Specific doses via exposure pathways associated with spillage of 

leachate into a surface water resource. Results do not include the 

effects of ingrowth of long-lived daughter radionuclides. 

Sensitivity studies to show the effect of different assumptions about the exposed 

individual are presented in Appendix C, where specific doses for an adult, an infant (1 

year old) and child (10 year old) are given for the different potential exposure 

pathways following a leachate spill to a surface water body. A comparison of the 

calculated specific doses for different age groups shows that for pathways associated 

with the consumption of fish and crops the greater consumption rates for adults 

outweigh the higher dose coefficients for infants and children. For the pathways 

associated with livestock and drinking water, the age group with the highest specific 

dose depends on the radionuclide, but in all cases the ratio of specific doses between 

age groups is less than a factor of 10. Only in the case of exposure through the 

consumption of contaminated soil are specific doses to infants and children higher 

than those to adults by a factor of more than 10, because of the greater consumption 

rates assumed for these age groups. 

5.4.3 Aerosol pathway 

Specific doses calculated for workers and members of the public via the aerosol 

pathway are presented in Table 5.6. These calculations are based on the assumptions 

described in Section 4.1.8 and in Appendix A, and are presented as specific doses per 

hour of exposure to aerosols. 

Radionuclide 

Specific dose 

(μSv y -1 per MBq per 

hour) 

Workers Public 

Radionuclide 


Soil 

Specific dose 


hour) 


H-3 2.51E-12 2.09E-12 Ra-226 1.88E-07 1.57E-07 

C-14 5.6E-11 4.66E-11 Ac-227 5.49E-06 4.57E-06 

Cl-36 7.04E-11 5.87E-11 Th-229 2.47E-06 2.06E-06 

Fe-55 7.43E-12 6.19E-12 Th-230 9.65E-07 8.04E-07 

Co-60 2.99E-10 2.49E-10 Th-232 1.64E-06 1.36E-06 



Version 2 

Radionuclide 

Specific dose 


hour) 


Radionuclide 

Specific dose 


hour) 


Ni-63 4.63E-12 3.86E-12 Pa-231 1.35E-06 1.13E-06 

Sr-90 1.56E-09 1.30E-09 U-232 4.52E-07 3.77E-07 

Nb-94 1.06E-10 8.84E-11 U-233 9.26E-08 7.72E-08 

Tc-99 1.25E-10 1.05E-10 U-234 9.07E-08 7.56E-08 

Ru-106 6.37E-10 5.31E-10 U-235 8.20E-08 6.83E-08 

Ag-108m 3.57E-10 2.97E-10 U-236 3.09E-08 2.57E-08 

Sb-125 5.27E-11 4.39E-11 U-238 7.73E-08 6.44E-08 

Sn-126 3.01E-10 2.51E-10 Np-237 4.82E-07 4.02E-07 

I-129 3.47E-10 2.89E-10 Pu-238 1.06E-06 8.84E-07 

Ba-133 2.99E-11 2.49E-11 Pu-239 1.16E-06 9.65E-07 

Cs-134 6.56E-11 5.47E-11 Pu-240 1.16E-06 9.65E-07 

Cs-137 3.76E-10 3.14E-10 Pu-241 2.22E-08 1.85E-08 

Pm-147 4.82E-11 4.02E-11 Pu-242 1.06E-06 8.84E-07 

Eu-152 4.05E-10 3.38E-10 Am-241 9.26E-07 7.72E-07 

Eu-154 5.11E-10 4.26E-10 Cm-243 3.00E-07 2.50E-07 

Eu-155 6.66E-11 5.55E-11 Cm-244 2.61E-07 2.18E-07 

Pb-210 9.64E-08 8.03E-08 

Table 5.6 Specific doses to workers and members of the public through exposure 

to aerosols during the operational phase. Results do not include the 

effects of ingrowth of long-lived daughter radionuclides. 

5.5 Gas pathway 

Specific doses via the gas pathway to workers, members of the public living near the 

site and members of the public living on the site after closure are presented in Table 

5.7. These calculations are based on the assumptions described in Appendix A 

(Sections A.1 and A.2). 

Although carbon-based gases (e.g., CO, CO2, CH4) are likely to be present within the 

landfill, and may be collected and flared, it is unlikely that the processes generating 

such gases would take place within cells dominated by LLW. C-14 would therefore 

not be released as a gas and is excluded from this assessment. 




Version 2 

Radionuclide 

Specific dose 

(microSv y -1 per MBq) 

Worker Public near site Resident 

after closure 

H-3 5.44E-08 1.13E-07 3.43E-08 

Ra-226 1.19E-07 2.48E-07 2.85E-06 

Th-230 - - 7.3E-08 

U-234 - - 2.02E-11 

U-238 - - 2.57E-12 

Pu-238 - - 1.64E-15 

Pu-242 - - 6.46E-21 

Table 5.7 Specific doses to workers and to members of the public via the gas 

pathway. Doses to residents after closure include the effects of 

ingrowth of daughter radionuclides for 60 years. 

5.7 Dose rates to wildlife 

As noted in Section 3.3, a radiological assessment of the potential effects of LLW 

disposal at the ENRMF on wildlife has been undertaken. This assessment has been 

undertaken using the Tier 1 approach within the assessment tool developed as part of 

the ERICA project (Environmental Risk from Ionising Contaminants: Assessment and 

Management). 

The ERICA toolkit allows for consideration of three ecosystems: terrestrial, 

freshwater and marine. Only the first two of these have been considered for the 

ENRMF. Within these ecosystems, the ERICA tool considers a range of organisms 

and wildlife groups (Table 5.9). 

Terrestrial Freshwater 

Bird Amphibian 

Bird egg Benthic fish 

Detritivorous invertebrate Bird 

Flying insects Bivalve mollusc 

Gastropod Crustacean 

Grasses & Herbs Gastropod 

Lichen & Bryophytes Insect larvae 

Mammal (Deer) Mammal 

Mammal (Rat) Pelagic fish 

Reptile Phytoplankton 

Shrub Vascular plant 




Version 2 

Soil Invertebrate (worm) Zooplankton 

Tree 

Table 5.9 Organisms and wildlife groups included in the two ecosystems 

considered in the assessment of impacts on wildlife. 

Within the Tier 1 assessment, the ERICA tool compares environmental concentrations 

for individual radionuclides with limiting concentrations calculated using generic 

assumptions about the ecosystems. The limiting concentrations are based on a 

screening dose rate of 10 Gy h -1 . Table 5.10 presents the limiting concentrations for 

the radionuclides considered in the wildlife assessment 1 . Note that the calculated 

dose rate for the same environmental concentration differs between organisms and 

therefore the limiting concentration does not necessarily apply to the same organism 

or wildlife group for each radionuclide. Within the Tier 1 assessment, it is assumed 

that all of the organisms and groups listed in Table 5.9 are present, so that the limiting 

concentration is the lowest calculated for any organism. 

Radionuclide 

Bq l -1 

Freshwater ecosystem Terrestrial ecosystem 

Limiting organism Bq kg -1 

Limiting organism 

Am-241 2.63E-03 Phytoplankton 6.25E+02 Flying insects 

C-14 1.56E+01 Bird 8.33E+01 Mammal (Deer) 

Cl-36 1.06E+02 Vascular plant 1.47E+03 Grasses & Herbs 

Cm-243 5.13E-03 Zooplankton 7.19E+02 Flying insects, Gastropod 

Cm-244 5.18E-03 Zooplankton 7.30E+02 Flying insects, Gastropod 

Co-60 1.87E-02 Insect larvae 7.35E+03 Mammal (Rat) 

Cs-134 2.06E-02 Insect larvae 1.67E+03 Mammal (Deer) 

Cs-137 5.10E-02 Insect larvae 3.13E+03 Mammal (Deer) 

Eu-152 7.19E+00 Vascular plant 1.72E+04 Soil Invertebrate (worm), 

Detritivorous invertebrate 

Eu-154 7.14E+00 Vascular plant 1.56E+04 Detritivorous invertebrate 

H-3 3.45E+05 Phytoplankton 2.60E+03 Detritivorous invertebrate 

I-129 2.75E+01 Phytoplankton 4.26E+02 Bird egg 

Ni-63 4.17E+01 Gastropod 1.08E+06 Grasses & Herbs 

Np-237 3.05E-03 Phytoplankton 3.77E+02 Shrub 

Pb-210 7.87E-02 Phytoplankton 3.88E+03 Lichen & bryophytes 

Pu-238 2.14E-02 Phytoplankton 1.02E+03 Lichen & bryophytes 


1 The ERICA tool does not include data for the complete range of radionuclides considered in the other 

parts of the radiological assessment. 




Version 2 


Pu-241 8.47E+01 Phytoplankton 4.05E+06 Lichen & bryophytes 

Ra-226 1.40E-02 Vascular plant 2.27E+02 Lichen & bryophytes 

Ru-106 1.28E-01 Insect larvae 8.20E+02 Lichen & bryophytes 

Sb-125 8.40E-01 Insect larvae 3.73E+04 Detritivorous invertebrate 

Sr-90 3.51E+00 Insect larvae 3.76E+02 Reptile 

Tc-99 5.05E+01 Vascular plant 2.11E+03 Bird egg 

Th-230 3.10E-02 Phytoplankton 1.63E+03 Lichen & bryophytes 

Th-232 3.64E-02 Phytoplankton 1.90E+03 Lichen & bryophytes 

U-234 4.22E-02 Vascular plant 1.67E+03 Lichen & bryophytes 



Table 5.10 Limiting concentrations in freshwater and terrestrial ecosystems, based 

on a dose rate of 10 Gy h -1 to the limiting organism or wildlife group. 

For the purposes of the wildlife assessment, the modified SNIFFER model described 

in Section 4.2 has been used to calculate radionuclide concentrations in a hypothetical 

stream close to the site boundary. This stream is assumed to receive baseflow from 

groundwater using the same assumptions as used for the drinking water pathway 

(Section 5.1). For the terrestrial ecosystem, it is assumed that this stream periodically 

floods an adjacent area of land. The SNIFFER model does not explicitly model 

flooding, and the irrigation model is therefore used to determine radionuclide 

concentrations in soil. 

Radionuclide concentrations in the freshwater and terrestrial ecosystems calculated 

using the SNIFFER model do not reflect actual concentrations as they are based on 

unit disposal (1 MBq) of each radionuclide. To allow comparison with the limiting 

concentrations presented in Table 5.10, it is conservatively assumed that the site 

receives the maximum amount of each radionuclide that keeps the site below the 

radiological capacity (see Section 6.2). Environmental concentrations based on 

disposal of radionuclides at capacity are presented in Tables 5.11 and 5.12, together 

with calculated risk quotients. The risk quotient for a radionuclide is the highest 

value of the ratio between calculated dose rate and the 10 Gy h -1 criterion (i.e., a risk 

quotient of 1 or greater would indicate that the screening criterion was exceeded). 

Radionulide Soil 

concentration 

(Bq kg -1 ) 

Risk 

quotient 

Am-241 4.92E-03 7.87E-06 Flying insects 

Limiting reference organism 

C-14 2.22E+00 2.66E-02 Mammal (Deer) 

Cl-36 5.28E-02 3.59E-05 Grasses & Herbs 

Cm-243 1.60E-13 2.22E-16 Flying insects, Gastropod 




Version 2 

Radionulide Soil 

concentration 

(Bq kg -1 ) 

Risk 

quotient 


Cm-244 3.41E-17 4.67E-20 Flying insects, Gastropod 

Co-60 1.12E-23 1.52E-27 Mammal (Rat) 

Cs-134 1.41E-22 8.45E-26 Mammal (Deer) 

Cs-137 5.11E-13 1.64E-16 Mammal (Deer) 

Eu-152 6.76E-18 3.92E-22 Soil Invertebrate (worm), Detritivorous 

invertebrate 

Eu-154 3.25E-20 2.08E-24 Detritivorous invertebrate 

H-3 8.14E-13 3.13E-16 Detritivorous invertebrate 

I-129 1.27E-04 2.98E-07 Bird egg 

Ni-63 4.79E-05 4.42E-11 Grasses & Herbs 

Np-237 1.52E-03 4.03E-06 Shrub 

Pb-210 8.45E-16 2.18E-19 Lichen & bryophytes 

Pu-238 4.42E-08 4.34E-11 Lichen & bryophytes 




Ra-226 5.04E-04 2.22E-06 Lichen & bryophytes 

Ru-106 1.62E-13 1.98E-16 Lichen & bryophytes 

Sb-125 4.03E-22 1.08E-26 Detritivorous invertebrate 

Sr-90 2.83E-13 7.53E-16 Reptile 

Tc-99 2.06E-02 9.74E-06 Bird egg 

Th-230 1.29E-02 7.92E-06 Lichen & bryophytes 

Th-232 2.76E-03 1.45E-06 Lichen & bryophytes 

U-234 2.25E-02 1.35E-05 Lichen & bryophytes 



Table 5.11 Calculated soil concentrations based on disposal of radionuclides at the 

individual radiological capacity. Risk quotients based on the 

10 Gy h -1 dose rate criterion. 




Version 2 

Radionulide Water 

concentration 

(Bq m -3 ) 

Risk 

quotient 

Am-241 4.50E-02 1.71E-02 Phytoplankton 

C-14 1.41E+01 9.07E-04 Bird 

Cl-36 3.43E+00 3.23E-05 Vascular plant 

Cm-243 1.66E-11 3.24E-12 Zooplankton 

Cm-244 5.45E-15 1.05E-15 Zooplankton 

Co-60 5.29E-21 2.83E-22 Insect larvae 

Cs-134 1.74E-19 8.44E-21 Insect larvae 

Cs-137 5.20E-11 1.02E-12 Insect larvae 

Eu-152 1.42E-15 1.97E-19 Vascular plant 

Eu-154 9.62E-18 1.35E-21 Vascular plant 

H-3 5.67E-10 1.64E-18 Phytoplankton 

I-129 2.43E-02 8.82E-07 Phytoplankton 

Ni-63 1.67E-03 4.01E-08 Gastropod 

Np-237 8.70E-02 2.85E-02 Phytoplankton 

Pb-210 1.12E-13 1.42E-15 Phytoplankton 

Pu-238 1.66E-06 7.75E-08 Phytoplankton 




Ra-226 2.23E-03 1.59E-04 Vascular plant 

Ru-106 3.42E-10 2.67E-12 Insect larvae 

Sb-125 3.91E-19 4.65E-22 Insect larvae 

Sr-90 4.07E-11 1.16E-14 Insect larvae 

Tc-99 1.25E+01 2.48E-04 Vascular plant 

Th-230 4.81E-02 1.55E-03 Phytoplankton 

Th-232 1.03E-02 2.83E-04 Phytoplankton 

U-234 2.24E-01 5.31E-03 Vascular plant 




Table 5.12 Calculated water concentrations based on disposal of radionuclides at 

the individual radiological capacity. Risk quotients based on the 

10 Gy h -1 dose rate criterion. 




Version 2 

The ERICA assessment tool allows three tiers of assessment. A Tier 1 assessment has 

been undertaken and the calculated incremental dose rate values are all below the 

recommended screening value. More detailed assessments (Tier 2 and Tier 3) are 

therefore not required. 




Version 2 

6 Radiological Capacity 


Section 5 presents the calculated doses for a range of scenarios and exposure 

pathways. These are presented as specific doses, the annual dose calculated to arise 

from a disposal of 1 MBq of the radionuclide concerned. The actual doses will 

depend on how much radioactive waste is actually disposed of to the site. Because of 

the conservative assumptions involved in the assessment model any doses received in 

the future are likely to be significantly less that these calculated doses. These specific 

doses do, however, provide a basis for decisions about the amount of waste that can 

be disposed. 

The radiological capacity is the amount that can be disposed without the calculated 

doses exceeding the appropriate dose criterion. Two types of radiological capacity 

can be calculated. The first is the capacity for individual radionuclides, which 

represents how much of any one radionuclide could be disposed of. If radionuclides 

are completely independent, then these capacities can be apportioned directly to the 

radionuclides – e.g., 50% of the capacity to radionuclide A and 50% to radionuclide 

B. 

In practice, radionuclides are not independent and are present in waste streams in 

certain ratios. In this case, the capacity cannot be directly apportioned to the 

radionuclides and must take account of both the specific dose for each radionuclide 

and the proportion of the radionuclide in the waste stream. 

The radiological capacity for radionuclide Rni in a waste stream (RCi) is given by: 

RC 

i 

 

where: 

fi 

SDi 

 

f 

DC 

i 

SD 

f 

i 

i 

is the fraction of the overall activity arising from Rni (such that fi=1) 

is the specific dose from Rni 


Radiological capacities for mixtures of waste streams can be calculated by 

apportioning part of the overall capacity to each waste stream. 

The following sections present radiological capacities at the ENRMF for individual 

radionuclides (Section 6.2) and overall capacities based on an illustrative waste 

stream from the UKAEA decommissioning programme at Harwell (see Section 2.2). 

Section 5 presents specific doses for a range of scenarios and exposure pathways, but 

radiological capacities have only been calculated for the principal exposure routes: 




Version 2 

groundwater and intrusion. Specific doses through at least one of these routes are 

higher than the corresponding doses through irradiation, and irradiation will not 

therefore determine the overall capacity. The aerosol and leachate spillage pathways 

are all highly uncertain, both in terms of the possibility of occurring and duration. 

The specific doses presented in Section 5 are illustrative, and might be considered in 

establishing mitigation measures, but should not be used to determine overall 

capacities. 

The specific doses calculated as a result of off-site leachate management are based on 

a generic model for a sewage treatment plant, with significant uncertainties regarding 

the extent of dilution by other waste streams and the type and fat of effluents. This 

model again calculates specific doses based on unit radioactivity inputs. As discussed 

elsewhere, there are large uncertainties about the rate at which radionuclides would 

enter the leachate at the ENRMF. Conservative assumptions have been made so that 

future doses arising from releases to groundwater or accidental spillage of leachate 

can be calculated. It would, however, be unreasonable to apply these same 

assumptions to the routine management of leachate during the operational period. 

Even if the same generic model for the sewage treatment plant is used to estimate 

specific doses, it would be more appropriate to determine a permissible level of 

radioactivity in leachate and then to develop authorisation conditions from these. 

Monitoring of leachate would ensure compliance with these conditions and give more 

control than applying conservative assumptions. 

As discussed in Section 3.2.1, two dose criteria have been used to determine 

radiological capacity. For exposures arising from releases to groundwater and 

subsequent use of an existing borehole for irrigation purposes, a dose criterion of 

20 Sv / year has been used. For exposures resulting from intrusion into the waste, 

from excavation and subsequent use of waste as soil, and from consumption of 

groundwater extracted from close to the site boundary, a dose criterion of 3 mSv / 

year has been used. 

6.2 Radionuclide-specific radiological capacities 

Tables 6.1 to 6.3 present specific doses and the corresponding radionuclide-specific 

capacities for the two groundwater pathways and the principal intrusion pathway 

described in Sections 4 and 5. 

Radionuclide 

Specific 

dose 

(μSv y -1 per 

MBq) 

Radiological 

capacity 

(MBq) 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

H-3 3.66E-30 5.47E+30 Ra-226 1.56E-08 1.28E+09 

C-14 2.06E-09 9.69E+09 Ac-227 5.66E-26 3.53E+26 

Cl-36 6.52E-08 3.07E+08 Th-229 1.44E-08 1.39E+09 

Fe-55 1.04E-43 1.92E+44 Th-230 8.24E-09 2.43E+09 

Co-60 1.21E-39 1.65E+40 Th-232 4.04E-08 4.95E+08 




Version 2 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Ni-63 7.94E-21 2.52E+21 Pa-231 3.60E-08 5.56E+08 

Sr-90 3.04E-24 6.59E+24 U-232 5.07E-20 3.94E+20 

Nb-94 3.76E-10 5.33E+10 U-233 4.62E-09 4.33E+09 

Tc-99 2.12E-09 9.45E+09 U-234 4.35E-09 4.60E+09 

Ru-106 3.44E-46 5.81E+46 U-235 4.35E-09 4.60E+09 

Ag-108m 1.68E-17 1.19E+18 U-236 4.22E-09 4.74E+09 

Sb-125 4.81E-42 4.16E+42 U-238 4.35E-09 4.59E+09 

Sn-126 1.20E-08 1.67E+09 Np-237 1.22E-06 1.64E+07 

I-129 1.38E-05 1.45E+06 Pu-238 1.86E-12 1.07E+13 

Ba-133 1.44E-31 1.39E+32 Pu-239 1.54E-08 1.30E+09 

Cs-134 1.82E-43 1.10E+44 Pu-240 1.04E-08 1.92E+09 

Cs-137 1.28E-25 1.57E+26 Pu-241 1.59E-10 1.26E+11 

Pm-147 3.41E-44 5.87E+44 Pu-242 1.69E-08 1.19E+09 

Eu-152 2.77E-32 7.21E+32 Am-241 5.37E-11 3.73E+11 

Eu-154 3.28E-35 6.09E+35 Cm-243 1.82E-10 1.10E+11 

Eu-155 3.63E-41 5.50E+41 Cm-244 1.29E-09 1.55E+10 

Pb-210 1.25E-25 1.60E+26 

Table 6.1 Specific doses to members of the public via use of water from a 

borehole 1500 m from the site boundary for irrigation, and 

corresponding radiological capacities. Radiological capacities are 

based on a 20 Sv / year dose criterion. Results include doses arising 

from ingrowth of daughter radionuclides for 100 years. 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

H-3 6.89E-23 4.36E+25 Ra-226 1.36E-06 2.20E+09 

C-14 1.39E-07 2.16E+10 Ac-227 6.70E-19 4.48E+21 

Cl-36 1.69E-06 1.78E+09 Th-229 1.29E-06 2.33E+09 

Fe-55 4.45E-36 6.74E+38 Th-230 7.35E-07 4.08E+09 

Co-60 3.79E-32 7.91E+34 Th-232 3.59E-06 8.36E+08 

Ni-63 3.47E-15 8.65E+17 Pa-231 3.08E-06 9.74E+08 

Sr-90 2.19E-17 1.37E+20 U-232 6.19E-14 4.85E+16 




Version 2 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Nb-94 9.07E-09 3.31E+11 U-233 4.13E-07 7.26E+09 

Tc-99 1.52E-07 1.97E+10 U-234 3.89E-07 7.72E+09 

Ru-106 1.67E-38 1.80E+41 U-235 3.87E-07 7.75E+09 

Ag-108m 1.51E-12 1.98E+15 U-236 3.78E-07 7.94E+09 

Sb-125 2.03E-34 1.48E+37 U-238 3.89E-07 7.71E+09 

Sn-126 4.87E-07 6.16E+09 Np-237 9.52E-05 3.15E+07 

I-129 3.02E-04 9.94E+06 Pu-238 1.67E-10 1.80E+13 

Ba-133 3.25E-24 9.22E+26 Pu-239 1.37E-06 2.18E+09 

Cs-134 8.12E-36 3.69E+38 Pu-240 9.33E-07 3.21E+09 

Cs-137 9.32E-19 3.22E+21 Pu-241 3.62E-08 8.29E+10 

Pm-147 1.47E-36 2.04E+39 Pu-242 1.51E-06 1.99E+09 

Eu-152 4.84E-25 6.20E+27 Am-241 1.08E-08 2.77E+11 

Eu-154 7.98E-28 3.76E+30 Cm-243 1.63E-08 1.84E+11 

Eu-155 1.29E-33 2.32E+36 Cm-244 1.15E-07 2.61E+10 

Pb-210 1.43E-18 2.09E+21 

Table 6.2 Specific doses to members of the public via use of water from a 

borehole at the site boundary for drinking, and corresponding 

radiological capacities. Radiological capacities are based on a 3 mSv / 

year dose criterion. Results include doses arising from ingrowth of 

daughter radionuclides for 100 years. 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

H-3 1.28E-10 2.34E+13 Ra-226 5.40E-06 5.56E+08 

C-14 8.68E-09 3.46E+11 Ac-227 1.81E-08 1.66E+11 

Cl-36 4.28E-07 7.01E+09 Th-229 8.84E-08 3.39E+10 

Fe-55 8.13E-18 3.69E+20 Th-230 1.56E-07 1.92E+10 

Co-60 4.19E-11 7.16E+13 Th-232 4.75E-07 6.32E+09 

Ni-63 2.60E-10 1.15E+13 Pa-231 1.66E-06 1.80E+09 

Sr-90 3.31E-07 9.06E+09 U-232 1.57E-08 1.91E+11 

Nb-94 2.44E-07 1.23E+10 U-233 4.32E-09 6.94E+11 

Tc-99 3.68E-07 8.15E+09 U-234 3.71E-09 8.09E+11 




Version 2 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Radionuclide 

Specific 

dose 


MBq) 

Radiological 

capacity 

(MBq) 

Ru-106 7.44E-29 4.03E+31 U-235 2.47E-08 1.21E+11 

Ag-108m 1.68E-07 1.79E+10 U-236 3.14E-09 9.55E+11 

Sb-125 1.89E-15 1.59E+18 U-238 6.87E-09 4.37E+11 

Sn-126 4.90E-07 6.12E+09 Np-237 4.65E-08 6.45E+10 

I-129 1.36E-06 2.21E+09 Pu-238 9.86E-09 3.04E+11 

Ba-133 5.63E-10 5.33E+12 Pu-239 1.86E-08 1.61E+11 

Cs-134 2.76E-17 1.09E+20 Pu-240 1.85E-08 1.62E+11 

Cs-137 2.61E-08 1.15E+11 Pu-241 8.29E-09 3.62E+11 

Pm-147 4.55E-19 6.60E+21 Pu-242 1.76E-08 1.71E+11 

Eu-152 4.63E-09 6.48E+11 Am-241 1.57E-08 1.91E+11 

Eu-154 7.87E-10 3.81E+12 Cm-243 4.49E-09 6.68E+11 

Eu-155 2.74E-13 1.09E+16 Cm-244 9.45E-10 3.17E+12 

Pb-210 1.28E-07 2.34E+10 

Table 6.3 Specific doses to members of the public living on excavated waste after 

60 years, and corresponding radiological capacities. Radiological 

capacities are based on a 3 mSv / year dose criterion. Results include 

doses arising from ingrowth of daughter radionuclides for 60 years. 

6.3 Overall radiological capacity 

To illustrate the potential overall capacity of the ENRMF, waste stream data for the 

UKAEA Harwell Meashill Trenches (Table 2.6) has been used to calculate overall 

radiological capacities based on the groundwater and human intrusion pathways using 

the assumptions and dose criteria described above. The results presented in Tables 

6.4 to 6.6 show that the calculated capacities for the groundwater and human intrusion 

pathways are very similar, in large part because of the different dose criteria applied. 

Table 6.6 also shows the contributions to the overall dose of the individual 

radionuclides. For the groundwater pathway, Pu-239 and Pu-240 are the key 

contributors to dose. In the case of human intrusion, Ra-226 is the principal 

contributor to dose. 

Radionuclide 

Specific dose 


2010 

Inventory 

(MBq) 

Radiological 

capacity 

(MBq) 

Contribution 

to dose 

(μSv) 

H-3 5.47E+30 3.25 3.62E+06 0.00 

Co-60 1.65E+40 8090 9.01E+09 0.00 

Cs-137 1.57E+26 952 1.06E+09 0.00 

Ra-226 1.28E+09 99.6 1.11E+08 1.73 




Version 2 

Radionuclide 

Specific dose 


2010 

Inventory 

(MBq) 

Radiological 

capacity 

(MBq) 

Contribution 

to dose 

(μSv) 

Th-232 4.95E+08 40 4.46E+07 1.80 

U-234 4.60E+09 500 5.57E+08 2.42 

U-235 4.60E+09 24 2.67E+07 0.12 

U-238 4.59E+09 500 5.57E+08 2.42 

Pu-238 1.07E+13 37 4.12E+07 0.00 

Pu-239 1.30E+09 400 4.46E+08 6.84 

Pu-240 1.92E+09 400 4.46E+08 4.65 

Pu-241 1.26E+11 38.2 4.25E+07 0.01 

Am-241 3.73E+11 98.4 1.10E+08 0.01 

Total 1.25E+10 20 

Table 6.4 Radiological capacity of the site based on specific doses to members of 

the public via use of water from a borehole 1500 m from the site 

boundary for irrigation. Radiological capacity of the site is based on an 

illustrative waste inventory for the Meashill Trenches. Results include 

doses arising from ingrowth of daughter radionuclides for 100 years. 

Radionuclide 

Specific dose 


2010 

Inventory 

(MBq) 

Radiological 

capacity 

(MBq) 

Contribution 

to dose 

(μSv) 

H-3 4.36E+25 3.25 6.08E+06 0.00 

Co-60 7.91E+34 8090 1.51E+10 0.00 

Cs-137 3.22E+21 952 1.78E+09 0.00 

Ra-226 2.20E+09 99.6 1.86E+08 254.30 

Th-232 8.36E+08 40 7.49E+07 268.72 

U-234 7.72E+09 500 9.36E+08 363.85 

U-235 7.75E+09 24 4.49E+07 17.38 

U-238 7.71E+09 500 9.36E+08 364.15 

Pu-238 1.80E+13 37 6.92E+07 0.01 

Pu-239 2.18E+09 400 7.49E+08 1028.38 

Pu-240 3.21E+09 400 7.49E+08 698.63 

Pu-241 8.29E+10 38.2 7.15E+07 2.59 

Am-241 2.77E+11 98.4 1.84E+08 2.00 

Total 2.09E+10 3000 


the public via use of water from a borehole at the site boundary for 




Version 2 

Radionuclide 

drinking. Radiological capacity of the site is based on an illustrative 

waste inventory for the Meashill Trenches. Results include doses 

arising from ingrowth of daughter radionuclides for 100 years. 

Specific dose 


2010 

Inventory 

(MBq) 

Radiological 

capacity 

(MBq) 

Contribution 

to dose 

(μSv) 

H-3 2.34E+13 3.25 1.61E+07 0.00 

Co-60 7.16E+13 8090 4.01E+10 1.68 

Cs-137 1.15E+11 952 4.72E+09 123.31 

Ra-226 5.56E+08 99.6 4.94E+08 2666.92 

Th-232 6.32E+09 40 1.98E+08 94.20 

U-234 8.09E+11 500 2.48E+09 9.19 

U-235 1.21E+11 24 1.19E+08 2.94 

U-238 4.37E+11 500 2.48E+09 17.03 

Pu-238 3.04E+11 37 1.83E+08 1.81 

Pu-239 1.61E+11 400 1.98E+09 36.95 

Pu-240 1.62E+11 400 1.98E+09 36.74 

Pu-241 3.62E+11 38.2 1.89E+08 1.57 

Am-241 1.91E+11 98.4 4.88E+08 7.67 

Total 5.54E+10 3000 


the public living on excavated waste after 60 years. Radiological 

capacity of the site is based on an illustrative waste inventory for the 

Meashill Trenches. Results include doses arising from ingrowth of 

daughter radionuclides for 60 years. 

As noted above, the radiological capacities based on the Meashill Trenches data are 

illustrative. They do demonstrate that the ENRMF could use the whole of Cells 4B, 

5A and 5B for the disposal of LLW at up to 200 Bq / g and remain, subject to an 

appropriate mixture of radionuclides, within an acceptable radiological capacity. 




Version 2 

7 References 

Allen, D.J., Brewerton, L.J., Coleby, L.M., Gibbs, B.R., Lewis, M.A., MacDonald, 

A.M., Wagstaff, S.J. and Williams, A.T., (1997). The physical properties of major 

aquifers in England and Wales. British Geological Survey Technical Report 

WD/97/34. Environment Agency R&D Publication 8. 

Bullen Consultants Ltd (2005). Environmental Statement: Slipe Clay Pit Landfill Site. 

Report no. 104b028/Re01-Rev A/ arb. 

Copplestone, D., Bielby, S., Jones, S.R., Patton, D., Daniel, P., and Gize, I. 2001. 

Impact assessment of ionising radiation on wildlife. Environment Agency R&D 

Publication 128. Environment Agency, Bristol. 

Defra (2007) Policy for the Long Term Management of Solid Low Level Radioactive 

Waste in the United Kingdom. Department for Environment, Food and Rural Affairs, 

London. 

Environment Agency (2006a) Initial Radiological Assessment Methodology – Part 1 

User Report. Environment Agency Science Report, SC030162/SR Part 1. 

Environment Agency (2006b) Initial Radiological Assessment Methodology – Part 2 

Methods and Input Data. Environment Agency Science Report, SC030162/SR Part 2. 

Environment Agency, Scottish Environment Protection Agency and Northern Ireland 

Environment Agency (2009). Near-Surface Disposal Facilities on Land for Solid 

Radioactive Wastes: Guidance on Requirements for Authorisation. Environment 

Agency, Bristol. 

Environmental Simulations International Ltd. (ESI) (2004). Hydrogeological Risk 

Assessment and risk based monitoring scheme: King’s Cliffe Landfill. Report 

reference: 6490R3rev1. 

HPA (Health Protection Agency) (2008). Guidance on the Application of Dose 

Coefficients for the Embryo, Fetus and Breastfed Infant in Dose Assessments for 

members of the Public. Health Protection Agency, Didcot. 

IAEA (International Atomic Energy Agency). (2003). Derivation of Activity Limits 

for the Disposal of Radioactive Waste in Near Surface Disposal Facilities. IAEA- 

TECDOC-1380. ISBN 92-0-113003-1. 

SNIFFER (2006a) Development of a Framework for Assessing the Suitability of 

Controlled Landfills to Accept Disposals of Solid Low-Level Radioactive Waste: 

Principles Document. SNIFFER, Edinburgh. 

SNIFFER (2006b) Development of a Framework for Assessing the Suitability of 

Controlled Landfills to Accept Disposals of Solid Low-Level Radioactive Waste: 

Technical Reference Manual. SNIFFER, Edinburgh. 




Version 2 

Appendix A Dose calculations 

A.1 Doses during site operations 

The air concentration of a radionuclide, CRn,gas,outdoors (Bq m -3 ), can be approximated 

by dividing by the air volume into which the activity released per year is diluted: 

C 

Rn , gas, 

outdoors 

 

Rn, 

gas 

Wuh3.16E 07 

where: RRn,gas is the release rate of radionuclide Rn in gas (Bq year –1 ) at the time of 

interest. 

W is the width of the source perpendicular to the wind direction (m). 

u is the mean wind speed (m s –1 ). 

h is the height for vertical mixing (m). 

3.16E+07 is the number of seconds in a year (s year –1 ). 

The dose from gases other than radon is given by: 

Dose B O 

D 

gas, 

outdoors 


R 

CRn, 

gas, 

outdoors 

where: Oout 

is the time spent in the gas plume (years year –1 ). 

B is the breathing rate (m 3 year -1 ). 

Dinh is the dose coefficient for inhalation of radionuclide Rn (Sv Bq -1 ). 

The dose calculation for radon must account for the effect of the daughters of Rn-222 

in the body, and has several additional terms: 

where: K1 

Doseradon, outdoors Cradon, 

outdoors K1 

B Oout 

out 

Rn 

inh 

 

K 

is the effective dose equivalent corresponding to an absorbed energy 

of 1 joule (Sv J -1 ). 

is the equilibrium factor (dimensionless). 

K2 is the potential -energy of Rn-222 in equilibrium with its daughters 

(J Bq -1 ). 

Radionuclide-specific data are presented in Appendix 2. Other parameter values used 

in the calculation of specific doses for the ENRMF site are presented in Table A.1. 


2


Version 2 


W width of the source perpendicular to 

the wind direction 

10 m 

u mean wind speed 6.2 m s –1 

h height for vertical mixing 2.0 m 

Oout time spent in the worker 880 

hours year 

gas plume public 2192 

–1 

B breathing rate worker 1.2 

m 

public 1.0 

3 hour -1 

K1 effective dose equivalent 

corresponding to an absorbed energy 

of 1 joule 

2.0 Sv J -1 

equilibrium factor 0.8 dimensionless 

K2 potential -energy of Rn-222 in 

equilibrium with its daughters 

5.5x10 -9 

J Bq -1 

Table A.1 Parameter values used in calculations of doses through the gas pathway 

during site operations. 

A.2 Doses to site residents after closure 

For calculation of peak dose, it is assumed that a house is constructed on top of the 

landfill cap immediately after closure. Irradiation doses are calculated for a resident 

spending 75% of the time indoors and 25% outdoors. Doses from gas inhalation are 

calculated for indoor exposure of the house resident to gas accumulating in the 

dwelling. 

where: Oout 

Dose 

irr 

 

D 

Rn 

irr , slab 

 

Rn 

Rn, 

waste 

 

x( 

t ) 

OOsf 

e 

out 

in 

A 

 

V 


waste 

( t ) 

 

 

waste 

is the time spent outside exposed to the waste (years year –1 ). 

Oin is the time spent inside (years year –1 ). 

sf is the shielding factor from the ground when indoors (dimensionless). 

ARn, waste(t) is the activity of the radionuclide, Rn (Bq), in the waste at time t 

Dirr,slab The dose conversion factor for irradiation from radionuclide Rn (Sv 

year -1 Bq -1 kg), based on the receptor being 1 m from the ground and 

the contamination being spread out so as to approximate a semiinfinite 

slab. 

Vwaste is the volume of material in which radioactivity is present (m 3 ). 

waste is the bulk density of the waste (kg m -3 ). 

Rn is the attenuation coefficient for radionuclide Rn (m -1 ). 

x is the thickness of the cap (m). 

Doses from inhalation of radioactive gases (excluding radon) are calculated from: 

Dose 

gas, 

indoors 

 

D 

Rn 

inh 

B 

O 

in 

 

R 

 

where: B is the breathing rate (m 3 year -1 ). 

Rn, 

gas 

a 

H 1 

( t) 

 

 

 

 

a kV 

 



Version 2 

Oin is the occupancy of the house (years year -1 ). 

RRn,gas(t) is the release rate of gas at time t (Bq year -1 ). 

aH/a is the horizontal area of a dwelling divided by the area over which 

the radioactive gas is being released (i.e., the facility footprint) 


k is a turnover rate to account for release of the gas by ventilation 

(year -1 ). 

V is the volume of the house (m 3 ). 


As for the outdoor calculation, the dose calculation for radon must account for the 

effect of the daughters of Rn-222 in the body: 

Dose 

radon 

, indoors K1 

 

K 

2 

BO 

 

R 

 

a 

H 1 

( t) 

 

 

 

 

a kV 

 


in 

radon 

where the terms are the same as those in the equations above. 


in the calculation of specific doses for the ENRMF site are presented in Table A.3. 


Vwaste volume of material in which 

radioactivity is present 

497,534 m 3 

waste bulk density of the waste 700 kg m -3 

x thickness of the cap 1.5 m 

B breathing rate 1.0 m 3 hour -1 

Oin occupancy of the house 6575 hours year -1 

Oout time spent outside exposed to the waste 2192 hours year -1 

sf shielding factor from the ground when 

indoors 


house 

50 m 2 

aH/a horizontal area of a 

dwelling divided by the 

area over which the 

radioactive gas is being facility 

released (i.e., the 

facility footprint) 

3.4x10 4 m 2 

1.04x10 -3 

dimensionless 

k is a turnover rate to account for release 

of the gas by ventilation 

8.8x10 3 

year -1 

V volume of the house 130 m 3 

Table A.3 Parameter values used in calculations of doses to site residents after 

site closure. 

A.3 Doses during and after excavation of waste 

A.3.1 Dose to the Excavator 

The excavator may receive a dose from irradiation, inhalation, and ingestion: 



Version 2 

Rn 

Rn 

Rn 

Doseexcavator Dirr, 

slabTC 

Rn, 

waste ( t) 

DinhTBM 

inhC 

Rn, 

waste ( t) 

DingTM 

ingC 

Rn, 

waste ( t) 

where Minh is the dust load of contaminated waste inhaled by the excavator 

(kg m -3 ). 

Ming is the rate of ingestion of dust from the material (kg hour -1 ). 

T is the time the excavator is exposed to the material (hours year -1 ). 

B is the breathing rate (m 3 hour -1 ). 

Dirr,slab, Dinh, and Ding are the dose coefficients for radionuclide Rn (Sv hour -1 Bq -1 kg; 

Sv Bq -1 ; and Sv Bq -1 , respectively). 

CRn,waste(t) is the concentration of radionuclide Rn (Bq kg -1 ) in the excavated 

material at the time of excavation, t: 

ARn( 

t) 

CRn, 

waste( 

t) 

 

V 

where ARn(t) is the activity of radionuclide Rn in the landfill at the time of 

excavation, t (Bq). 

Vlandfill is the volume of landfill in which the activity ARn(t) is 

homogeneously distributed (m 3 ). 

waste is the density of the waste (kg m -3 ). 

The exposure from external irradiation is assumed to come from proximity to 

contaminated material, approximated by a semi-infinite slab. 

The excavator might also receive a dose through direct contact with contaminated 

waste dust on hands and face: 

Dose 

 

CRn, 

waste ( t) 

d 

hands 

 

 

 

4 

 

10 


 

 

 

 

landfill 

waste 

waste 

skin, hands gamma7 

beta40 

 

 

 

Rn 

Rn 

Areahands 

DDWT skin 

Area 

where CRn,waste(t) is the concentration of radionuclide Rn (Bq kg -1 ) in the waste at the 

time of excavation, t. 

Dgamma7 is the skin equivalent dose rate for radionuclide Rn to the basal layer 

of the skin epidermis for gamma irradiation (Sv h -1 per Bq cm -2 ). 

Dbeta40 is the skin equivalent dose rate for radionuclide Rn to the basal layer 

of the skin epidermis for hands for beta irradiation, skin thickness 

400 m (40mg cm -2 ), (Sv h -1 per Bq cm -2 ). 

10 4 converts Bq m -2 to Bq cm -2 . 

dhands is the thickness of the contaminated layer on the hands (m). 


Wskin is the tissue weighting factor for skin (dimensionless). 

Areahands is the area of skin in contact with the contaminated dust (cm 2 ). 

Areabody is the total exposed skin area of the adult body (cm 2 ). 

T is the time the worker is exposed to the material (hours year -1 ). 

Dose 

 

C Rn, 

waste ( t) 

d 

face 

 

 

 

4 

 

10 

waste 

skin, face gamma7 

beta4 

 

 

 

 

 

 

 

Area 

Rn 

Rn 

face 

DDWT skin 

Area 

where CRn,waste(t) is the concentration of radionuclide Rn (Bq kg -1 ) in the waste at the 

time of excavation, t. 


body 

body


Version 2 

Dgamma7 is the skin equivalent dose rate for radionuclide Rn to the basal layer 

of the skin epidermis for gamma irradiation (Sv h -1 per Bq cm -2 ). 

Dbeta4 is the skin equivalent dose rate for radionuclide Rn to the basal layer 

of the skin epidermis for face for beta irradiation, skin thickness 40 

m (4 mg cm -2 ), (Sv h -1 per Bq cm -2 ). 

10 4 converts Bq m -2 to Bq cm -2 . 

dface is the thickness of the contaminated layer on the face (m). 


Wskin is the tissue weighting factor for skin (dimensionless). 

Areaface is the area of skin in contact with the contaminated dust (cm 2 ). 

Areabody is the total exposed skin area of the adult body (cm 2 ). 

T is the time the worker is exposed to the material (hours year -1 ). 

A.3.2 Dose to Site Resident after Excavation 

It is assumed that following, or as part of the reason for, the excavation, the waste and 

the cover are mixed together and re-laid, creating a soil layer partly contaminated with 

the radioactivity that was in the waste. The initial concentration of radionuclide Rn in 

the material, CRn,soil,excavate (Bq kg -1 ), immediately after the excavation event is 

calculated by: 

C 

Rn, 

soil, 

excavate 

A 

 

V 


Rn 

( t) 

Dil 

 

where ARn(t) is the activity of radionuclide Rn in the landfill at the time of 

excavation, t (Bq). 

Dil is the dilution factor given by the ratio of the volume of contaminated 

landfill waste to the volume of other material that is mixed in to form 

the soil (dimensionless). 

Vlandfill is the volume of the landfill (m 3 ). 


landfill 

Dose from ingesting contaminated soil that may be attached to crops is given by: 

where Qsoil 

soil 

Dose D 

ing, 

soil Qsoil 

CRn, 

soil, 

excavate 

is the soil consumption rate (kg year –1 ). 

Ding is the dose coefficient for ingestion of radionuclide, Rn (Sv Bq -1 ). 

Dose from crops grown on contaminated soil is given by: 

Rn 

Q 

crop C 

Rn, 

soil excavate TFcrop 

 

Dose ing , crops 

, 

crop 

D 

where Qcrop is the crop consumption rate (kg year –1 ). 

TFcrop is the soil to crop transfer factor for radionuclide, Rn (Bq kg -1 fresh 



Rn 

ing 

Rn 

ing 



Version 2 

Dose from livestock and associated products (e.g., milk) raised on contaminated 

ground is given by: 

 

 

Doseing , animal 

 

animal 

 

 

 

crop 

 

 

Rn Rn Rn 

Qanimal 

 

qsoil 

CRn 

soil excavate qcrop CRn 

soil excavate TFcrop 

, , , , TFanimal 

Ding 

where Qanimal is the animal foodstuff consumption rate (kg year –1 ). 

qsoil is the soil consumption rate by the animal (kg day –1 ). 

qcrop is the crop consumption rate by the animal (kg day –1 ). 



TFanimal is the animal product transfer factor for radionuclide, Rn (days kg -1 ). 


Dose from external irradiation while living or working on contaminated soil is given 

by: 

Rn 

OOsf CD Dose irr, 

soil out in 

Rn, 

soil, 

excavate irr, 

slab 

where: Oout 

is the time spent outside exposed to the soil (years year –1 ). 



Dirr,slab is the dose conversion factor for irradiation from radionuclide Rn 

(Sv year -1 Bq -1 kg), based on the receptor being 1 m from the ground 

and assuming a semi-infinite slab of contamination. 

Dose from inhaling dust derived from contaminated soil is given by: 

Dose dustload 

D 

inh, 

soil BOdust 

C 

Rn, 

soil, 

excavate 

where: Odust is the time spent exposed to dust from the soil (hours year –1 ). 

B is the breathing rate (m 3 hour -1 ). 

dustload is the dust concentration (kg m -3 of air). 



in the calculation of specific doses arising from excavation and intrusion into the 

ENRMF site are presented in Table A.4. 



Rn 

inh


Version 2 


Minh dust load of contaminated waste 

inhaled by the excavator 

1.0x10 -6 kg m -3 

Ming rate of ingestion of dust from 

excavated material 

3.45x10 -5 kg hour -1 

T time the excavator is exposed to 

excavated material 

88 hours year -1 

B breathing rate (worker) 1.2 m 3 hour -1 

Vlandfill volume of landfill in which the 

activity is homogeneously distributed 

497,534 m 3 

waste density of the waste 1700 kg m -3 

dhands 

thickness of the contaminated layer 

on the hands 

1.0x10 -4 m 

Wskin tissue weighting factor for skin 1x10 -2 dimensionless 

Areahands area of skin in contact with the 2x10 2 

cm 2 

contaminated dust 

Areabody total exposed skin area of the adult 

body 

dface thickness of the contaminated layer 

on the face 

Areaface area of skin in contact with the 

contaminated dust 

Dil dilution factor given by the ratio of 

the volume of contaminated landfill 

waste to the volume of other material 

that is mixed in to form the soil 

3x10 3 


cm 2 

5x10 -5 m 

1x10 2 

cm 2 


Qsoil soil consumption rate 3.0x10 -2 kg year –1 

Qcrop 

Qanimal 

crop consumption rate 

animal foodstuff 

consumption rate 

Grain 50 

Green veg 30 

Root veg 120 

Meat 32 

Milk 100 

kg year –1 

kg year –1 

qsoil soil consumption rate by the animal 0.6 kg day –1 

qcrop crop consumption rate by the animal 55 kg day –1 

Oout time spent outside exposed to the soil 0.25 years year –1 

Oin time spent inside 0.75 years year –1 

sf shielding factor from the ground 

when indoors 

Odust time spent exposed to dust from soil 2.2x10 3 

B Breathing rate (public) 1.0 

dustload dust concentration 1x10 -7 


hours year –1 

m 3 hour -1 

kg m -3 of air 

Table A.4 Parameter values used in calculations of doses during and after 

excavation of waste. Consumption rates for animal foodstuffs and 

grain are about one-third of the average rates cited in IAEA (2003). 

Consumption rates for vegetables are about one-half of the average 

rates cited in IAEA (2003). 



Version 2 

A.4 Doses arising from use of contaminated groundwater 

If a well or river is used for irrigation, then doses can result from ingestion of 

foodstuffs raised on contaminated soil, inhalation of dust from the soil, and external 

exposure to the soil. Drinking of contaminated water from a well or river is also a 

potential exposure pathway. If contaminated groundwater discharges to surface water 

(spring, river, sea), then ingestion of foodstuffs from the surface water is a potential 

exposure pathway. 

Dose from ingesting contaminated soil that may be attached to crops is given by: 

Dose Q C 

( t) 

D 

ing, 

soil 

soil 

Rn, 

soil 

where Qsoil is the soil consumption rate (kg year –1 ). 

CRn,soil(t) is the concentration of radionuclide in the soil at time t (Bq kg -1 ). 


Dose from crops grown on contaminated soil is given by: 

 

Irrig rate Intcrop 

Fcrop 

 

Dose 

 

ing, 

crops Qcrop 

CRn, 

water ( t) 

 

C 

 

 

Rn, 

soil ( t) 

TF 

crop Yieldcrop 

 


Rn 

ing 

Rn 

crop 

 

 

 

D 

 

 

where Qcrop is the crop consumption rate (kg year –1 ). 

Irrigrate is the rate of irrigation (m year -1 ). 

Intcrop is the effective interception factor (dimensionless). 

Fcrop is the fraction remaining after processing (dimensionless). 

Yieldcrop is the crop yield (kg m -2 ). 



CRn,water(t) is the concentration of radionuclide in the water used for irrigation at 

time t (Bq m -3 ). 

CRn,soil(t) is the concentration of radionuclide in the crop soil at time t 

(Bq kg -1 ). 


Dose from livestock and associated products (e.g., milk) raised on contaminated 

ground and fed with contaminated crops is given by: 

 

 

Doseing, animal animal soil Rn soil 

crop Rn soil crop animal 

animal 

 

, ( , 

 

 

crop 

 

 

Rn Rn Rn 

Q 

q C 

t) 

qC( t) 

TF 

TF D 

where Qanimal is the animal foodstuff consumption rate (kg year –1 ). 

qwater is the water consumption rate by the animal (m 3 day –1 ). 

qsoil is the soil consumption rate by the animal (kg day –1 ). 

qcrop is the crop consumption rate by the animal (kg day –1 ). 

TFcrop is the soil to crop transfer factor for radionuclide Rn (Bq kg -1 fresh 


TFanimal is the animal product transfer factor for radionuclide Rn (days kg -1 ). 


ing 

Rn 

ing


Version 2 

CRn,soil(t) is the concentration of radionuclide Rn in the pasture and crop soil at 

time t (Bq kg -1 ). 


Dose from external irradiation while living or working on contaminated soil is given 

by: 

Rn 

OOsf C( t) 

D 

Dose irr, 

soil out in 

Rn, 

soil irr, 

slab 

where: Oout 

is the time spent outside exposed to the soil (years year –1 ). 




Dirr,slab is the dose conversion factor for irradiation from radionuclide Rn 

(Sv year -1 Bq -1 kg), based on the receptor being 1 m from the ground 

and assuming a semi-infinite slab of contamination. 

Dose from inhaling dust derived from contaminated soil is given by: 

Dose BO 

C ( t) 

dustload 

D 

inh, 

soil 

dust 

Rn, 

soil 

where: Odust 

is the time spent exposed to dust from the soil (years year –1 ). 

B is the breathing rate (m 3 year -1 ). 


dustload is the dust concentration (kg m -3 of air). 



Irrigrate rate of irrigation 0.3 m year -1 

Intcrop effective interception factor 0.33 dimensionless 

Fcrop fraction remaining after Grain 1.0 

dimensionless 

processing 

Green veg 0.3 

Yieldcrop 

crop yield 

Root veg 1.0 

Pasture 1.7 

Grain 0.4 

Green veg 3.0 

Root veg 3.5 


Rn 

inh 

kg m -2 

Table A.5 Parameter values used in calculations of doses arising from use of 

contaminated groundwater for irrigation. 



Version 2 

Appendix B Radionuclide-specific data 

Radionuclide Half-life 

DC 

inhalation 

DC 

ingestion 

Irradiation 

slab 

Dgamma7 Dbeta4 Dbeta40 

name y (Sv Bq-1) (Sv Bq-1) (Sv y-1 Bq-1 kg) (Sv h-1 Bq-1 cm2) (Sv h-1 Bq-1 cm2) (Sv h-1 Bq-1 cm2) H-3 1.23E+01 2.60E-10 1.80E-11 0.00E+00 0.00E+00 0.00E+00 0.00E+00 

C-14 5.73E+03 5.80E-09 5.80E-10 3.64E-12 0.00E+00 9.02E-07 0.00E+00 

Cl-36 3.01E+05 7.30E-09 9.30E-10 6.46E-10 1.10E-11 2.51E-06 5.37E-07 

Fe-55 2.70E+00 7.70E-10 3.30E-10 0.00E+00 1.60E-08 0.00E+00 0.00E+00 

Co-60 5.27E+00 3.10E-08 3.40E-09 4.38E-06 1.30E-07 1.83E-06 2.85E-08 

Ni-63 9.60E+01 4.80E-10 1.50E-10 0.00E+00 0.00E+00 1.83E-08 0.00E+00 

Sr-90 2.91E+01 1.62E-07 3.07E-08 6.65E-09 2.40E-12 5.14E-06 1.76E-06 

Nb-94 2.00E+04 1.10E-08 1.70E-09 2.62E-06 9.47E-08 2.17E-06 1.83E-07 

Tc-99 2.13E+05 1.30E-08 6.40E-10 3.39E-11 3.49E-14 1.60E-06 1.37E-08 

Ru-106 1.01E+00 6.60E-08 7.00E-09 3.49E-07 1.20E-08 2.85E-06 1.60E-06 

Ag-108m 1.27E+02 3.70E-08 2.30E-09 2.61E-06 1.28E-07 2.76E-07 1.15E-07 

Sb-125 2.80E+00 5.46E-09 3.11E-09 6.61E-07 3.54E-08 2.05E-06 8.45E-08 

Sn-126 1.00E+05 3.12E-08 7.10E-09 4.66E-06 1.33E-07 4.54E-06 1.43E-06 

I-129 1.57E+07 3.60E-08 1.10E-07 3.50E-09 9.70E-09 6.51E-07 0.00E+00 

Ba-133 1.07E+01 3.10E-09 1.50E-09 5.32E-07 0.00E+00 0.00E+00 0.00E+00 

Cs-134 2.10E+00 6.80E-09 1.90E-08 2.56E-06 8.79E-08 1.83E-06 3.08E-07 

Cs-137 3.00E+01 3.90E-08 1.30E-08 9.75E-07 3.30E-08 2.54E-06 3.90E-07 

Pm-147 2.60E+00 5.00E-09 2.60E-10 1.35E-11 4.91E-13 1.26E-06 4.11E-10 

Eu-152 1.33E+01 4.20E-08 1.40E-09 1.89E-06 1.18E-07 1.60E-06 1.71E-07 

Eu-154 8.80E+00 5.30E-08 2.00E-09 2.08E-06 9.02E-08 3.42E-06 3.77E-07 

Eu-155 4.96E+00 6.90E-09 3.20E-10 4.93E-08 1.77E-08 8.68E-07 3.20E-10 

Pb-210 2.23E+01 9.99E-06 1.89E-06 1.65E-09 8.30E-09 2.63E-06 8.45E-07 

Ra-226 1.60E+03 1.95E-05 2.17E-06 3.02E-06 1.64E-07 5.89E-06 1.64E-06 

Ac-227 2.18E+01 5.69E-04 1.21E-06 5.43E-07 3.81E-08 6.59E-06 2.00E-06 

Th-229 7.34E+03 2.56E-04 6.13E-07 4.33E-07 7.31E-08 8.56E-06 1.36E-06 

Th-230 7.70E+04 1.00E-04 2.10E-07 3.27E-10 3.83E-09 1.04E-07 0.00E+00 

Th-232 1.40E+10 1.70E-04 1.06E-06 4.37E-06 2.20E-09 3.08E-08 0.00E+00 

Pa-231 3.27E+04 1.40E-04 7.10E-07 5.15E-08 6.27E-08 1.48E-07 5.14E-09 

U-232 6.89E+01 4.69E-05 4.60E-07 2.44E-10 9.36E-08 3.20E-08 0.00E+00 

U-233 1.58E+05 9.60E-06 5.10E-08 3.78E-10 1.70E-09 5.25E-07 0.00E+00 

U-234 2.44E+05 9.40E-06 4.90E-08 1.09E-10 2.70E-09 7.42E-09 0.00E+00 

U-235 7.04E+08 8.50E-06 4.73E-08 2.05E-07 5.31E-08 2.52E-06 1.09E-08 

U-236 2.34E+07 3.20E-06 4.70E-08 5.78E-11 3.55E-09 4.57E-09 0.00E+00 

U-238 4.47E+09 8.01E-06 4.84E-08 3.58E-08 9.23E-09 3.82E-06 1.26E-06 

Np-237 2.14E+06 5.00E-05 1.11E-07 2.97E-07 3.20E-08 3.46E-06 9.93E-08 

Pu-238 8.77E+01 1.10E-04 2.30E-07 4.09E-11 2.70E-09 1.06E-07 0.00E+00 

Pu-239 2.41E+04 1.20E-04 2.50E-07 7.98E-11 1.00E-09 4.34E-10 0.00E+00 

Pu-240 6.54E+03 1.20E-04 2.50E-07 3.96E-11 2.60E-09 0.00E+00 0.00E+00 

Pu-241 1.44E+01 2.30E-06 4.80E-09 1.60E-12 3.30E-12 0.00E+00 0.00E+00 

Pu-242 3.76E+05 1.10E-04 2.40E-07 3.46E-11 3.07E-09 0.00E+00 0.00E+00 

Am-241 4.32E+02 9.60E-05 2.00E-07 1.18E-08 1.70E-08 5.48E-08 0.00E+00 

Cm-243 2.91E+01 3.11E-05 1.50E-07 1.58E-07 7.99E-09 1.94E-06 3.42E-08 

Cm-244 1.81E+01 2.71E-05 1.21E-07 3.41E-11 2.17E-09 0.00E+00 0.00E+00 




Version 2 

Radionuclide 

UF freshwater 

fish 

name (m 3 kg -1) 

TF cow meat TF cow milk UF green veg WR green veg UF root veg UF grain UF grass 

(d kg -1 fresh 

weight) 

(d kg -1) (Bq kg -1Bq -1 kg) (y -1) 

(Bq kg-1Bq-1 kg) 




H-3 1.00E-03 2.90E-02 1.00E-02 5.00E+00 1.83E+01 5.00E+00 1.00E-02 5.00E+00 

C-14 9.00E+00 1.20E-01 1.00E-02 1.00E-01 1.83E+01 1.00E-01 1.60E-01 1.00E-01 

Cl-36 5.00E-02 4.30E-02 1.70E-02 5.00E+00 1.83E+01 5.00E+00 8.80E-02 5.00E+00 

Fe-55 1.00E-01 2.00E-02 3.00E-05 2.00E-04 1.83E+01 3.00E-04 1.00E-01 4.00E-04 

Co-60 3.00E-01 1.00E-02 3.00E-04 3.00E-02 1.83E+01 3.00E-02 8.00E-02 6.00E-03 

Ni-63 1.00E-01 5.00E-03 1.60E-02 3.00E-02 1.83E+01 3.00E-02 5.00E-02 2.00E-02 

Sr-90 6.00E-02 8.00E-03 3.00E-03 3.00E+00 1.83E+01 9.00E-02 1.20E-01 3.00E+00 

Nb-94 3.00E-01 3.00E-07 4.10E-07 1.00E-02 1.83E+01 1.00E-02 1.00E-02 1.00E-02 

Tc-99 2.00E-02 1.00E-04 2.30E-05 1.00E+01 1.83E+01 1.00E+01 1.00E+01 1.00E+01 

Ru-106 1.00E-02 5.00E-02 3.30E-06 4.00E-03 1.83E+01 1.00E-02 1.00E-01 4.00E-02 

Ag-108m 5.00E-03 3.00E-05 5.00E-05 2.70E-04 1.83E+01 1.30E-03 8.80E-02 1.50E-01 

Sb-125 1.00E-01 4.00E-05 2.50E-05 1.00E-02 1.83E+01 1.00E-02 1.00E-02 1.00E-02 

Sn-126 1.00E+00 1.90E-03 1.00E-03 1.00E-01 1.83E+01 1.00E-01 2.00E-01 2.00E-01 

I-129 3.00E-02 4.00E-02 1.00E-02 1.00E-01 1.83E+01 1.00E-01 2.80E-01 1.00E-01 

Ba-133 4.00E-03 5.00E-04 5.00E-04 4.00E-03 1.83E+01 1.00E-02 1.00E-01 4.00E-02 

Cs-134 2.00E+00 5.00E-02 7.90E-03 3.00E-02 1.83E+01 3.00E-02 2.00E-02 3.00E-02 

Cs-137 2.00E+00 5.00E-02 7.90E-03 3.00E-02 1.83E+01 3.00E-02 2.00E-02 3.00E-02 

Pm-147 3.00E-02 5.00E-03 2.00E-05 3.00E-03 1.83E+01 3.00E-03 3.00E-03 3.00E-03 

Eu-152 3.00E-02 4.70E-04 5.00E-05 3.00E-03 1.83E+01 3.00E-03 4.80E-02 3.00E-03 

Eu-154 3.00E-02 4.70E-04 5.00E-05 3.00E-03 1.83E+01 3.00E-03 4.80E-02 3.00E-03 

Eu-155 3.00E-02 4.70E-04 5.00E-05 3.00E-03 1.83E+01 3.00E-03 4.80E-02 3.00E-03 

Pb-210 3.00E-01 4.00E-04 3.00E-04 1.00E-02 1.83E+01 1.00E-02 1.00E-02 1.00E-02 

Ra-226 5.00E-02 9.00E-04 1.30E-03 4.00E-02 1.83E+01 4.00E-02 4.00E-02 4.00E-02 

Ac-227 8.00E-01 1.60E-04 4.00E-07 1.00E-03 1.83E+01 1.00E-03 1.00E-03 1.00E-03 

Th-229 3.00E-02 2.70E-03 5.00E-06 5.00E-04 1.83E+01 5.00E-04 5.00E-04 5.00E-04 

Th-230 3.00E-02 2.70E-03 5.00E-06 5.00E-04 1.83E+01 5.00E-04 5.00E-04 5.00E-04 

Th-232 3.00E-02 2.70E-03 5.00E-06 5.00E-04 1.83E+01 5.00E-04 5.00E-04 5.00E-04 

Pa-231 1.00E-02 5.00E-05 5.00E-06 4.00E-02 1.83E+01 4.00E-02 4.00E-02 4.00E-02 

U-232 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03 

U-233 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03 

U-234 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03 

U-235 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03 

U-236 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03 

U-238 1.00E-02 3.00E-04 4.00E-04 1.00E-03 5.11E+01 1.00E-03 1.00E-04 1.00E-03 

Np-237 1.00E-02 1.00E-03 5.00E-06 1.00E-02 5.11E+01 1.00E-03 3.00E-04 5.00E-03 

Pu-238 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03 

Pu-239 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03 

Pu-240 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03 

Pu-241 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03 

Pu-242 4.00E-03 1.00E-05 1.10E-06 1.00E-04 5.11E+01 1.00E-03 3.00E-05 1.00E-03 

Am-241 3.00E-02 4.00E-05 1.50E-06 1.00E-03 1.83E+01 1.00E-03 1.00E-05 5.00E-03 

Cm-243 3.00E-02 1.00E-04 1.00E-06 1.00E-04 1.83E+01 1.00E-03 3.00E-05 1.00E-03 

Cm-244 3.00E-02 1.00E-04 1.00E-06 1.00E-04 1.83E+01 1.00E-03 3.00E-05 1.00E-03 



Version 2 

Radionuclide Kd soil Kd barrier 

Irradiation 

cloudshine 

Irradiation groundshine Attenuation 

coefficient 

name (m 3 kg -1) (m 3 kg -1) (Sv hr -1 Bq -1 m 3) (Sv y -1 Bq -1 m 2) (m -1) 

H-3 1.00E-04 1.00E-04 1.19E-15 0.00E+00 0.00E+00 

C-14 1.00E-01 1.00E-01 8.08E-16 5.08E-13 -5.59E+01 

Cl-36 1.50E-02 1.50E-02 8.03E-14 2.12E-11 -2.04E+01 

Fe-55 2.20E-01 8.00E-01 0.00E+00 0.00E+00 0.00E+00 

Co-60 6.00E-02 1.00E+01 4.55E-10 7.43E-08 -1.20E+01 

Ni-63 4.00E-01 6.00E-01 0.00E+00 0.00E+00 0.00E+00 

Sr-90 1.30E-02 1.40E-01 7.12E-13 1.77E-10 -2.88E+01 

Nb-94 1.60E-01 7.60E+00 2.77E-10 4.84E-08 -1.38E+01 

Tc-99 1.40E-04 1.90E-01 5.83E-15 2.46E-12 -3.85E+01 

Ru-106 5.50E-02 4.00E-01 3.73E-11 6.69E-09 -1.40E+01 

Ag-108m 9.00E-02 1.80E-01 2.81E-10 5.03E-08 -1.49E+01 

Sb-125 4.50E-02 2.40E-01 7.27E-11 1.34E-08 -1.50E+01 

Sn-126 1.30E-01 6.70E-01 5.03E-10 8.94E-08 -3.59E+01 

I-129 1.00E-03 1.00E-03 1.37E-12 8.15E-10 -1.31E+02 

Ba-133 4.10E-03 4.00E-02 6.42E-11 1.25E-08 -1.40E+01 

Cs-134 2.70E-01 2.00E+00 2.72E-10 4.81E-08 -1.40E+01 

Cs-137 2.70E-01 2.00E+00 1.04E-10 1.85E-08 -2.61E+01 

Pm-147 2.40E-01 1.30E+00 2.49E-15 1.08E-12 -2.78E+01 

Eu-152 2.40E-01 7.80E+00 2.03E-10 3.48E-08 -1.30E+01 

Eu-154 2.40E-01 7.80E+00 2.21E-10 3.75E-08 -1.29E+01 

Eu-155 2.40E-01 7.80E+00 8.96E-12 1.86E-09 -3.37E+01 

Pb-210 2.70E-01 4.90E+00 3.23E-13 1.12E-10 -8.36E+01 

Ra-226 4.90E-01 9.00E+00 3.20E-10 5.26E-08 -2.29E+01 

Ac-227 4.50E-01 5.00E+00 7.50E-11 1.40E-08 -2.75E+01 

Th-229 3.00E+00 1.43E+01 4.15E-10 6.87E-08 -2.65E+01 

Th-230 3.00E+00 1.43E+01 6.28E-14 2.37E-11 -2.82E+01 

Th-232 3.00E+00 1.43E+01 8.70E-10 1.34E-07 -2.93E+01 

Pa-231 5.40E-01 1.00E+01 8.78E-12 1.29E-09 -1.91E+01 

U-232 3.30E-02 6.00E+00 5.10E-14 3.19E-11 -6.00E+01 

U-233 3.30E-02 6.00E+00 5.87E-14 2.26E-11 -3.74E+01 

U-234 3.30E-02 6.00E+00 2.74E-14 2.36E-11 -2.29E+01 

U-235 3.30E-02 6.00E+00 2.78E-11 5.25E-09 -3.58E+01 

U-236 3.30E-02 6.00E+00 1.80E-14 2.05E-11 -2.32E+01 

U-238 3.30E-02 6.00E+00 3.40E-10 5.89E-08 -3.16E+01 

Np-237 4.10E-03 4.60E-02 3.74E-11 7.07E-09 -8.74E+01 

Pu-238 5.40E-01 7.60E+00 1.75E-14 2.64E-11 -2.18E+01 

Pu-239 5.40E-01 7.60E+00 1.53E-14 1.16E-11 -5.37E+01 

Pu-240 5.40E-01 7.60E+00 1.71E-14 2.54E-11 -2.96E+01 

Pu-241 5.40E-01 7.60E+00 2.61E-16 6.09E-14 -5.58E+01 

Pu-242 5.40E-01 7.60E+00 1.44E-14 2.10E-11 -4.44E+01 

Am-241 2.00E+00 3.20E+00 2.95E-12 8.67E-10 -3.73E+01 

Cm-243 4.00E-01 4.00E+00 2.12E-11 3.95E-09 -1.66E+01 

Cm-244 4.00E-01 4.00E+00 1.77E-14 2.77E-11 -2.89E+01 




Version 2 

Appendix C Sensitivity Studies 

C.1 Groundwater Pathway 

This section presents the results from a set of sensitivity studies to assess how 

calculated doses via the groundwater pathway are affected by alternative assumptions 

about the disposal system. Results are presented that show the effect of variations in 

leachate head within the landfill, of changes to the lifetime and efficiency of the cap, 

in different assessment periods, and in the assumptions about the exposed group. 

The results of these sensitivity studies are discussed in Section 5.1. 

Radionuclide 

Specific dose 


1m head 2m head 5m head 10m head 

Increase 

1m - 10m 

H-3 5.99E-23 6.01E-23 6.07E-23 6.17E-23 1.03 

C-14 1.47E-07 1.47E-07 1.47E-07 1.47E-07 1.00 

Cl-36 1.60E-06 1.60E-06 1.60E-06 1.60E-06 1.00 

Fe-55 1.90E-36 6.63E-36 3.45E-35 1.20E-34 63.08 

Co-60 2.92E-32 3.00E-32 3.63E-32 1.26E-31 4.33 

Ni-63 2.92E-15 2.92E-15 2.92E-15 2.92E-15 1.00 

Sr-90 1.70E-17 1.70E-17 1.71E-17 1.71E-17 1.01 

Nb-94 6.84E-09 6.84E-09 6.84E-09 6.84E-09 1.00 

Tc-99 1.15E-07 1.15E-07 1.15E-07 1.15E-07 1.00 

Ru-106 7.67E-39 2.67E-38 1.39E-37 4.84E-37 63.09 

Ag-108m 1.11E-12 1.11E-12 1.11E-12 1.11E-12 1.00 

Sb-125 8.25E-35 2.87E-34 1.49E-33 5.20E-33 63.08 

Sn-126 3.72E-07 3.72E-07 3.72E-07 3.72E-07 1.00 

I-129 2.72E-04 2.72E-04 2.72E-04 2.72E-04 1.00 

Ba-133 2.45E-24 2.46E-24 2.49E-24 2.53E-24 1.03 

Cs-134 3.95E-36 1.37E-35 7.15E-35 2.49E-34 63.08 

Cs-137 8.29E-19 8.29E-19 8.31E-19 8.34E-19 1.01 

Pm-147 6.04E-37 2.10E-36 1.09E-35 3.81E-35 63.08 

Eu-152 3.63E-25 3.64E-25 3.66E-25 3.71E-25 1.02 

Eu-154 6.00E-28 6.03E-28 6.13E-28 6.29E-28 1.05 

Eu-155 9.68E-34 1.00E-33 2.18E-33 7.58E-33 7.83 

Pb-210 1.07E-18 1.07E-18 1.08E-18 1.08E-18 1.01 

Ra-226 1.04E-06 1.04E-06 1.04E-06 1.04E-06 1.00 




Version 2 

Radionuclide 

Specific dose 


1m head 2m head 5m head 10m head 

Increase 

1m - 10m 

Ac-227 5.00E-19 5.01E-19 5.02E-19 5.05E-19 1.01 

Th-229 9.80E-07 9.80E-07 9.80E-07 9.80E-07 1.00 

Th-230 5.15E-07 5.15E-07 5.15E-07 5.15E-07 1.00 

Th-232 2.73E-06 2.73E-06 2.73E-06 2.73E-06 1.00 

Pa-231 2.32E-06 2.32E-06 2.32E-06 2.32E-06 1.00 

U-232 4.58E-14 4.58E-14 4.59E-14 4.59E-14 1.00 

U-233 3.04E-07 3.04E-07 3.04E-07 3.04E-07 1.00 

U-234 2.94E-07 2.94E-07 2.94E-07 2.94E-07 1.00 

U-235 2.88E-07 2.88E-07 2.88E-07 2.88E-07 1.00 

U-236 2.86E-07 2.86E-07 2.86E-07 2.86E-07 1.00 

U-238 2.94E-07 2.94E-07 2.94E-07 2.95E-07 1.00 

Np-237 7.22E-05 7.22E-05 7.22E-05 7.22E-05 1.00 

Pu-238 1.47E-13 1.47E-13 1.47E-13 1.47E-13 1.00 

Pu-239 1.04E-06 1.04E-06 1.04E-06 1.04E-06 1.00 

Pu-240 7.04E-07 7.04E-07 7.04E-07 7.04E-07 1.00 

Pu-241 4.19E-24 4.20E-24 4.23E-24 4.27E-24 1.02 

Pu-242 1.14E-06 1.14E-06 1.14E-06 1.14E-06 1.00 

Am-241 5.52E-09 5.52E-09 5.52E-09 5.52E-09 1.00 

Cm-243 2.79E-18 2.79E-18 2.80E-18 2.81E-18 1.01 

Cm-244 5.23E-21 5.24E-21 5.26E-21 5.30E-21 1.01 

Table C.1 Sensitivity studies on the effect of increased leachate head in the 

landfill. Specific doses are to members of the public via use of water 

from a borehole at the site boundary for drinking. Results do not 

include the effects of ingrowth of daughter radionuclides. 

Radionuclide 

Specific dose 


20 y cap 60 y cap 100 y cap 

Difference 

20 - 100 y 

H-3 5.56E-22 5.99E-23 5.63E-24 98.76 

C-14 1.49E-07 1.47E-07 1.45E-07 1.03 

Cl-36 1.61E-06 1.60E-06 1.59E-06 1.01 

Fe-55 1.06E-34 1.90E-36 4.80E-36 22.04 

Co-60 5.26E-30 2.92E-32 2.99E-33 1758.56 




Version 2 

Radionuclide 

Specific dose 



Difference 

20 - 100 y 

Ni-63 3.88E-15 2.92E-15 2.16E-15 1.80 

Sr-90 4.37E-17 1.70E-17 6.26E-18 6.97 

Nb-94 6.93E-09 6.84E-09 6.74E-09 1.03 

Tc-99 1.16E-07 1.15E-07 1.13E-07 1.03 

Ru-106 1.70E-39 7.67E-39 3.97E-38 0.04 

Ag-108m 1.38E-12 1.11E-12 8.85E-13 1.56 

Sb-125 5.99E-33 8.25E-35 2.03E-34 29.57 

Sn-126 3.77E-07 3.72E-07 3.67E-07 1.03 

I-129 2.72E-04 2.72E-04 2.72E-04 1.00 

Ba-133 3.18E-23 2.45E-24 1.62E-25 196.36 

Cs-134 2.80E-35 3.95E-36 1.20E-35 2.33 

Cs-137 2.07E-18 8.29E-19 3.14E-19 6.58 

Pm-147 2.52E-35 6.04E-37 1.56E-36 16.13 

Eu-152 2.85E-24 3.63E-25 4.07E-26 70.00 

Eu-154 1.35E-26 6.00E-28 2.21E-29 613.41 

Eu-155 2.40E-31 9.68E-34 1.89E-34 1275.31 

Pb-210 3.67E-18 1.07E-18 2.91E-19 12.61 

Ra-226 1.05E-06 1.04E-06 1.02E-06 1.04 

Ac-227 1.76E-18 5.00E-19 1.32E-19 13.36 

Th-229 9.93E-07 9.80E-07 9.66E-07 1.03 

Th-230 5.22E-07 5.15E-07 5.07E-07 1.03 

Th-232 2.76E-06 2.73E-06 2.69E-06 1.03 

Pa-231 2.36E-06 2.32E-06 2.29E-06 1.03 

U-232 6.82E-14 4.58E-14 3.00E-14 2.27 

U-233 3.08E-07 3.04E-07 2.99E-07 1.03 

U-234 2.98E-07 2.94E-07 2.90E-07 1.03 

U-235 2.92E-07 2.88E-07 2.84E-07 1.03 

U-236 2.90E-07 2.86E-07 2.82E-07 1.03 

U-238 2.99E-07 2.94E-07 2.90E-07 1.03 

Np-237 7.30E-05 7.22E-05 7.13E-05 1.02 

Pu-238 2.01E-13 1.47E-13 1.06E-13 1.91 

Pu-239 1.05E-06 1.04E-06 1.02E-06 1.03 

Pu-240 7.13E-07 7.04E-07 6.94E-07 1.03 

Pu-241 2.82E-23 4.19E-24 5.57E-25 50.61 




Version 2 

Radionuclide 

Specific dose 



Difference 

20 - 100 y 

Pu-242 1.15E-06 1.14E-06 1.12E-06 1.03 

Am-241 5.88E-09 5.52E-09 5.16E-09 1.14 

Cm-243 7.16E-18 2.79E-18 1.03E-18 6.97 

Cm-244 2.38E-20 5.23E-21 1.05E-21 22.70 

Table C.2 Sensitivity studies on the effect of changes in the cap lifetime. Specific 

doses are to members of the public via use of water from a borehole 

100 m from the site boundary for drinking. Results do not include the 

effects of ingrowth of daughter radionuclides. 

Radionuclide 

Specific dose 


99% 95% 90% 

Difference 

99 - 90 % 

H-3 5.99E-23 6.94E-23 8.31E-23 1.39 

C-14 1.47E-07 1.47E-07 1.47E-07 1.00 

Cl-36 1.60E-06 1.6E-06 1.6E-06 1.00 

Fe-55 1.90E-36 1.9E-36 1.9E-36 1.00 

Co-60 2.92E-32 4.5E-32 7.16E-32 2.45 

Ni-63 2.92E-15 2.96E-15 3.02E-15 1.03 

Sr-90 1.70E-17 1.79E-17 1.91E-17 1.12 

Nb-94 6.84E-09 6.84E-09 6.85E-09 1.00 

Tc-99 1.15E-07 1.15E-07 1.15E-07 1.00 

Ru-106 7.67E-39 7.67E-39 7.67E-39 1.00 

Ag-108m 1.11E-12 1.12E-12 1.14E-12 1.03 

Sb-125 8.25E-35 8.25E-35 8.25E-35 1.00 

Sn-126 3.72E-07 3.73E-07 3.73E-07 1.00 

I-129 2.72E-04 0.000272 0.000272 1.00 

Ba-133 2.45E-24 2.89E-24 3.53E-24 1.44 

Cs-134 3.95E-36 3.95E-36 3.95E-36 1.00 

Cs-137 8.29E-19 8.72E-19 9.29E-19 1.12 

Pm-147 6.04E-37 6.04E-37 6.04E-37 1.00 

Eu-152 3.63E-25 4.12E-25 4.82E-25 1.33 




Version 2 

Radionuclide 

Specific dose 


99% 95% 90% 

Difference 

99 - 90 % 

Eu-154 6.00E-28 7.39E-28 9.43E-28 1.57 

Eu-155 9.68E-34 1.53E-33 2.5E-33 2.59 

Pb-210 1.07E-18 1.15E-18 1.26E-18 1.17 

Ra-226 1.04E-06 1.04E-06 1.04E-06 1.00 

Ac-227 5.00E-19 5.37E-19 5.88E-19 1.18 

Th-229 9.80E-07 9.8E-07 9.81E-07 1.00 

Th-230 5.15E-07 5.15E-07 5.15E-07 1.00 

Th-232 2.73E-06 2.73E-06 2.73E-06 1.00 

Pa-231 2.32E-06 2.32E-06 2.33E-06 1.00 

U-232 4.58E-14 4.68E-14 4.8E-14 1.05 

U-233 3.04E-07 3.04E-07 3.04E-07 1.00 

U-234 2.94E-07 2.94E-07 2.94E-07 1.00 

U-235 2.88E-07 2.88E-07 2.88E-07 1.00 

U-236 2.86E-07 2.86E-07 2.86E-07 1.00 

U-238 2.94E-07 2.95E-07 2.95E-07 1.00 

Np-237 7.22E-05 7.22E-05 7.22E-05 1.00 

Pu-238 1.47E-13 1.49E-13 1.53E-13 1.04 

Pu-239 1.04E-06 1.04E-06 1.04E-06 1.00 

Pu-240 7.04E-07 7.04E-07 7.05E-07 1.00 

Pu-241 4.19E-24 4.71E-24 5.44E-24 1.30 

Pu-242 1.14E-06 1.14E-06 1.14E-06 1.00 

Am-241 5.52E-09 5.53E-09 5.56E-09 1.01 

Cm-243 2.79E-18 2.94E-18 3.14E-18 1.12 

Cm-244 5.23E-21 5.72E-21 6.39E-21 1.22 

Table C.3 Sensitivity studies on the effect of differences in cap efficiency. Cap 

lifetime is assumed to be 60 years. Specific doses are to members of 

the public via use of water from a borehole 100 m from the site 

boundary for drinking. Results do not include the effects of ingrowth of 

daughter radionuclides. 




Version 2 

Radionuclide 

Specific dose 


5,000 y 1,000 y 500 y 

Difference 

5,000 - 500 y 

H-3 5.99E-23 5.99E-23 5.99E-23 1.0 

C-14 1.47E-07 9.76E-11 9.90E-14 1.5E+06 

Cl-36 1.60E-06 1.01E-09 9.92E-13 1.6E+06 

Fe-55 1.90E-36 1.90E-36 1.90E-36 1.0 

Co-60 2.92E-32 2.92E-32 2.92E-32 1.0 

Ni-63 2.92E-15 2.72E-15 9.58E-17 3.0E+01 

Sr-90 1.70E-17 1.70E-17 1.70E-17 1.00 

Nb-94 6.84E-09 2.96E-12 2.86E-15 2.4E+06 

Tc-99 1.15E-07 4.58E-11 4.37E-14 2.6E+06 

Ru-106 7.67E-39 7.67E-39 7.67E-39 1.0 

Ag-108m 1.11E-12 7.16E-13 1.04E-14 1.1E+02 

Sb-125 8.25E-35 8.25E-35 8.25E-35 1.0 

Sn-126 3.72E-07 1.46E-10 1.39E-13 2.7E+06 

I-129 2.72E-04 1.03E-06 1.33E-09 2.0E+05 

Ba-133 2.45E-24 2.45E-24 2.45E-24 1.0 

Cs-134 3.95E-36 3.95E-36 3.95E-36 1.0 

Cs-137 8.29E-19 8.29E-19 8.29E-19 1.0 

Pm-147 6.04E-37 6.04E-37 6.04E-37 1.0 

Eu-152 3.63E-25 3.63E-25 3.63E-25 1.0 

Eu-154 6.00E-28 6.00E-28 6.00E-28 1.0 

Eu-155 9.68E-34 9.68E-34 9.68E-34 1.0 

Pb-210 1.07E-18 1.07E-18 1.07E-18 1.0 

Ra-226 1.04E-06 2.16E-09 2.55E-12 4.1E+05 

Ac-227 5.00E-19 5.00E-19 5.00E-19 1.0 

Th-229 9.80E-07 5.38E-10 5.36E-13 1.8E+06 

Th-230 5.15E-07 2.01E-10 1.92E-13 2.7E+06 

Th-232 2.73E-06 1.03E-09 9.75E-13 2.8E+06 

Pa-231 2.32E-06 9.52E-10 9.15E-13 2.5E+06 

U-232 4.58E-14 4.58E-14 6.11E-15 7.50 

U-233 3.04E-07 1.16E-10 1.11E-13 2.7E+06 

U-234 2.94E-07 1.12E-10 1.07E-13 2.8E+06 

U-235 2.88E-07 1.08E-10 1.03E-13 2.8E+06 




Version 2 

Radionuclide 

Specific dose 


5,000 y 1,000 y 500 y 

Difference 

5,000 - 500 y 

U-236 2.86E-07 1.08E-10 1.02E-13 2.8E+06 

U-238 2.94E-07 1.11E-10 1.05E-13 2.8E+06 

Np-237 7.22E-05 3.26E-08 3.13E-11 2.3E+06 

Pu-238 1.47E-13 1.45E-13 7.17E-15 2.1E+01 

Pu-239 1.04E-06 4.38E-10 4.22E-13 2.5E+06 

Pu-240 7.04E-07 4.05E-10 4.06E-13 1.7E+06 

Pu-241 4.19E-24 4.19E-24 4.19E-24 1.0 

Pu-242 1.14E-06 4.32E-10 4.11E-13 2.8E+06 

Am-241 5.52E-09 1.70E-10 3.61E-13 1.5E+04 

Cm-243 2.79E-18 2.79E-18 2.79E-18 1.0 

Cm-244 5.23E-21 5.23E-21 5.23E-21 1.0 

Table C.4 Sensitivity studies on the effect of changes in the assessment period. 

Specific doses are to members of the public via use of water from a 

borehole 100 m from the site boundary for drinking. Results do not 

include the effects of ingrowth of daughter radionuclides. 

Radionuclide 

Specific dose 


Adult Infant Child 

Infant - 

Adult 

H-3 3.66E-30 2.90E-30 2.62E-30 0.79 

C-14 2.06E-09 7.29E-10 7.23E-10 0.35 

Cl-36 6.52E-08 1.64E-07 6.88E-08 2.51 

Fe-55 1.04E-43 5.62E-44 5.78E-44 0.54 

Co-60 1.21E-39 7.44E-40 6.86E-40 0.61 

Ni-63 7.94E-21 3.66E-21 2.58E-21 0.46 

Sr-90 3.04E-24 8.49E-25 1.32E-24 0.28 

Nb-94 3.76E-10 3.18E-10 3.08E-10 0.85 

Tc-99 2.12E-09 1.60E-09 9.71E-10 0.76 

Ru-106 3.44E-46 1.55E-46 1.18E-46 0.45 

Ag-108m 1.68E-17 1.08E-17 1.04E-17 0.65 

Sb-125 4.81E-42 2.29E-42 1.68E-42 0.48 

Sn-126 1.2E-08 8.99E-09 8.11E-09 0.75 

I-129 1.38E-05 2.14E-06 4.07E-06 0.16 




Version 2 

Radionuclide 

Specific dose 


Adult Infant Child 

Infant - 

Adult 

Ba-133 1.44E-31 1.23E-32 2.07E-32 0.09 

Cs-134 1.82E-43 1.29E-44 2.37E-44 0.07 

Cs-137 1.28E-25 4.17E-26 6.74E-26 0.33 

Pm-147 3.41E-44 1.85E-44 1.24E-44 0.54 

Eu-152 2.77E-32 1.28E-32 1.08E-32 0.46 

Eu-154 3.28E-35 1.58E-35 1.26E-35 0.48 

Eu-155 3.63E-41 1.86E-41 1.30E-41 0.51 

Pb-210 1.25E-25 4.85E-26 5.74E-26 0.39 

Ra-226 1.56E-08 4.58E-09 8.22E-09 0.29 

Ac-227 5.66E-26 1.19E-26 1.29E-26 0.21 

Th-229 1.44E-08 2.26E-09 3.08E-09 0.16 

Th-230 8.24E-09 1.32E-09 1.80E-09 0.16 

Th-232 4.04E-08 6.30E-09 8.79E-09 0.16 

Pa-231 3.6E-08 5.45E-09 8.50E-09 0.15 

U-232 5.07E-20 9.40E-21 1.46E-20 0.19 

U-233 4.62E-09 9.50E-10 1.17E-09 0.21 

U-234 4.35E-09 8.70E-10 1.10E-09 0.20 

U-235 4.35E-09 9.21E-10 1.11E-09 0.21 

U-236 4.22E-09 8.81E-10 1.05E-09 0.21 

U-238 4.35E-09 8.80E-10 1.10E-09 0.20 

Np-237 1.22E-06 1.74E-07 2.04E-07 0.14 

Pu-238 1.86E-12 3.73E-13 4.70E-13 0.20 

Pu-239 1.54E-08 1.97E-09 2.78E-09 0.13 

Pu-240 1.04E-08 1.34E-09 1.89E-09 0.13 

Pu-241 1.59E-10 2.27E-11 2.73E-11 0.14 

Pu-242 1.69E-08 2.14E-09 3.06E-09 0.13 

Am-241 5.37E-11 7.65E-12 9.17E-12 0.14 

Cm-243 1.82E-10 2.33E-11 3.29E-11 0.13 

Cm-244 1.29E-09 1.65E-10 2.33E-10 0.13 

Table C.5 Sensitivity studies on the effect of changes in the exposed individual. 

Specific doses are to members of the public via use of water from a 

borehole 1500 m from the site for irrigation. Results include the effects 

of ingrowth of daughter radionuclides over 100 years. 




Version 2 

C.2 Leachate Spillage 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 




Livestock and 

associated 

products 

H-3 2.96E-12 3.44E-15 1.92E-11 1.37E-12 9.52E-17 

C-14 4.15E-11 4.34E-10 2.58E-10 2.09E-11 1.34E-15 

Cl-36 9.85E-11 5.73E-12 6.40E-10 7.37E-11 3.17E-15 

Fe-55 5.70E-11 6.63E-12 3.55E-10 3.37E-12 1.84E-15 

Co-60 5.70E-10 1.99E-10 3.55E-09 1.92E-11 1.84E-14 

Ni-63 1.45E-11 1.69E-12 9.03E-11 3.48E-12 4.68E-16 

Sr-90 3.41E-09 2.38E-10 2.13E-08 3.65E-10 1.10E-13 

Nb-94 1.76E-10 6.15E-11 1.10E-09 1.17E-15 5.68E-15 

Tc-99 6.74E-11 1.57E-12 4.57E-10 1.29E-13 2.17E-15 

Ru-106 7.49E-10 8.71E-12 4.66E-09 1.11E-10 2.41E-14 

Ag-108m 2.23E-10 1.30E-12 1.39E-09 1.81E-13 7.18E-15 

Sb-125 3.36E-10 3.91E-11 2.09E-09 1.57E-13 1.08E-14 

Sn-126 7.03E-10 8.17E-10 4.38E-09 1.43E-11 2.26E-14 

I-129 9.85E-09 3.44E-10 6.13E-08 2.59E-09 3.17E-13 

Ba-133 5.73E-11 2.67E-13 3.56E-10 4.89E-13 1.85E-15 

Cs-134 7.58E-10 1.76E-09 4.72E-09 1.96E-10 2.44E-14 

Cs-137 2.07E-09 4.81E-09 1.29E-08 5.35E-10 6.66E-14 

Pm-147 2.96E-11 1.03E-12 1.84E-10 4.43E-13 9.52E-16 

Eu-152 1.35E-10 4.70E-12 8.39E-10 2.81E-13 4.34E-15 

Eu-154 2.13E-10 7.42E-12 1.32E-09 4.42E-13 6.85E-15 

Eu-155 3.53E-11 1.23E-12 2.19E-10 7.34E-14 1.14E-15 

Pb-210 2.70E-07 9.42E-08 1.68E-06 1.45E-09 8.70E-12 

Ra-226 3.22E-07 1.87E-08 2.00E-06 6.76E-09 1.04E-11 

Ac-227 8.55E-08 7.95E-08 5.32E-07 4.07E-11 2.75E-12 

Th-229 4.02E-08 1.40E-09 2.50E-07 3.22E-10 1.30E-12 

Th-230 1.24E-08 4.34E-10 7.74E-08 9.96E-11 4.01E-13 

Th-232 6.95E-08 2.43E-09 4.33E-07 5.56E-10 2.24E-12 

Pa-231 4.77E-08 5.55E-10 2.97E-07 1.04E-11 1.54E-12 

U-232 4.12E-08 4.79E-10 2.56E-07 2.67E-10 1.33E-12 


Soil 



Version 2 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 




Livestock and 

associated 

products 

U-233 4.04E-09 4.70E-11 2.52E-08 2.62E-11 1.30E-13 

U-234 3.84E-09 4.46E-11 2.39E-08 2.49E-11 1.24E-13 

U-235 3.71E-09 4.31E-11 2.31E-08 2.40E-11 1.19E-13 

U-236 3.63E-09 4.22E-11 2.26E-08 2.35E-11 1.17E-13 

U-238 3.79E-09 4.41E-11 2.36E-08 2.46E-11 1.22E-13 

Np-237 5.75E-09 6.69E-11 3.58E-08 1.73E-11 1.85E-13 

Pu-238 1.24E-08 5.79E-11 7.74E-08 5.57E-13 4.01E-13 

Pu-239 1.40E-08 6.51E-11 8.71E-08 6.27E-13 4.51E-13 

Pu-240 1.40E-08 6.51E-11 8.71E-08 6.27E-13 4.51E-13 

Pu-241 2.64E-10 1.23E-12 1.64E-09 1.18E-14 8.52E-15 

Pu-242 1.35E-08 6.27E-11 8.39E-08 6.04E-13 4.34E-13 

Am-241 1.14E-08 3.98E-10 7.10E-08 1.58E-12 3.67E-13 

Cm-243 8.30E-09 2.89E-10 5.16E-08 2.55E-12 2.67E-13 

Cm-244 7.32E-09 2.55E-10 4.55E-08 2.25E-12 2.36E-13 

Table C.6 Specific doses to a child (10 years) via exposure pathways associated 

with spillage of leachate into a surface water resource. Results do not 

include the effects of ingrowth of long-lived daughter radionuclides. 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 


Soil 




Livestock and 

associated 

products 

H-3 4.63E-12 2.89E-15 1.80E-11 2.74E-12 1.00E-15 

C-14 6.17E-11 3.47E-10 2.30E-10 2.57E-11 1.34E-14 

Cl-36 2.43E-10 7.60E-12 9.45E-10 2.41E-10 5.26E-14 

Fe-55 9.26E-11 5.79E-12 3.45E-10 2.51E-12 2.00E-14 

Co-60 1.04E-09 1.95E-10 3.88E-09 2.15E-11 2.25E-13 

Soil 



Version 2 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 




Livestock and 

associated 

products 

Ni-63 3.24E-11 2.03E-12 1.21E-10 1.32E-11 7.01E-15 

Sr-90 3.09E-09 1.16E-10 1.15E-08 4.31E-10 6.68E-13 

Nb-94 3.74E-10 7.02E-11 1.39E-09 3.99E-15 8.10E-14 

Tc-99 1.85E-10 2.32E-12 7.51E-10 4.07E-13 4.01E-14 

Ru-106 1.64E-09 1.02E-11 6.09E-09 1.09E-10 3.54E-13 

Ag-108m 4.24E-10 1.33E-12 1.58E-09 5.65E-13 9.18E-14 

Sb-125 7.66E-10 4.79E-11 2.85E-09 5.20E-13 1.66E-13 

Sn-126 1.52E-09 9.50E-10 5.66E-09 4.35E-11 3.29E-13 

I-129 8.49E-09 1.59E-10 3.16E-08 2.62E-09 1.84E-12 

Ba-133 4.87E-11 1.22E-13 1.81E-10 6.47E-13 1.05E-14 

Cs-134 6.77E-10 8.46E-10 2.52E-09 1.79E-10 1.46E-13 

Cs-137 2.07E-09 2.59E-09 7.72E-09 5.49E-10 4.49E-13 

Pm-147 7.33E-11 1.37E-12 2.73E-10 5.19E-13 1.59E-14 

Eu-152 2.86E-10 5.35E-12 1.06E-09 5.34E-13 6.18E-14 

Eu-154 4.63E-10 8.68E-12 1.72E-09 8.66E-13 1.00E-13 

Eu-155 8.49E-11 1.59E-12 3.16E-10 1.59E-13 1.84E-14 

Pb-210 3.81E-07 7.14E-08 1.42E-06 3.06E-09 8.24E-11 

Ra-226 2.87E-07 8.98E-09 1.07E-06 9.76E-09 6.22E-11 

Ac-227 1.31E-07 6.57E-08 4.89E-07 2.90E-11 2.84E-11 

Th-229 4.83E-08 9.06E-10 1.80E-07 1.78E-10 1.04E-11 

Th-230 1.58E-08 2.97E-10 5.89E-08 5.82E-11 3.42E-12 

Th-232 8.03E-08 1.51E-09 2.99E-07 2.95E-10 1.74E-11 

Pa-231 5.02E-08 3.14E-10 1.87E-07 9.65E-12 1.09E-11 

U-232 4.41E-08 2.76E-10 1.64E-07 4.59E-10 9.54E-12 

U-233 5.40E-09 3.38E-11 2.01E-08 5.62E-11 1.17E-12 

U-234 5.02E-09 3.14E-11 1.87E-08 5.21E-11 1.09E-12 

U-235 5.05E-09 3.16E-11 1.88E-08 5.25E-11 1.09E-12 

U-236 5.02E-09 3.14E-11 1.87E-08 5.21E-11 1.09E-12 

U-238 4.98E-09 3.11E-11 1.85E-08 5.18E-11 1.08E-12 

Np-237 8.17E-09 5.11E-11 3.04E-08 1.18E-11 1.77E-12 


Soil 



Version 2 

Radionuclide 



by leachate 

Drinking 

water 

Specific dose 


Fish Crops 




Livestock and 

associated 

products 

Pu-238 1.54E-08 3.86E-11 5.74E-08 6.28E-13 3.34E-12 

Pu-239 1.62E-08 4.05E-11 6.03E-08 6.59E-13 3.51E-12 

Pu-240 1.62E-08 4.05E-11 6.03E-08 6.59E-13 3.51E-12 

Pu-241 2.20E-10 5.50E-13 8.19E-10 8.94E-15 4.76E-14 

Pu-242 1.54E-08 3.86E-11 5.74E-08 6.28E-13 3.34E-12 

Am-241 1.43E-08 2.68E-10 5.31E-08 1.29E-12 3.09E-12 

Cm-243 1.27E-08 2.39E-10 4.74E-08 1.99E-12 2.76E-12 

Cm-244 1.13E-08 2.12E-10 4.20E-08 1.77E-12 2.44E-12 

Table C.7 Specific doses to a infant (1 year) via exposure pathways associated 

with spillage of leachate into a surface water resource. Results do not 

include the effects of ingrowth of long-lived daughter radionuclides. 


Soil 


Annex C 

ENRMF, IRRs 1999, Radiation Risk 

Assessment for Low Level Waste Disposal, 

HPA 


EAST NORTHANTS RESOURCE MANAGEMENT FACILITY 

IONISING RADIATIONS REGULATIONS 1999 

1 

VERSION 3, 14 July 2009 

RADIATION RISK ASSESSMENT FOR LLW WITH A SPECIFIC ACTIVITY UP TO 

200Bq/g 

1 SCOPE AND DEFINITIONS 

1.1 INTRODUCTION 

The East Northants Resource Management Facility (ENRMF) operated by Augean plc is 

intending to dispose of low level radioactive wastes (LLW) with a specific activity of up to 

200Bq/g. An application under the Radioactive Substances Act 1993 has been prepared, and 

this includes an assessment of the potential radiation exposure of workers and members of 

the public. In addition to this, the Ionising Radiations Regulations 1999 (IRR99) require that a 

radiological risk assessment is undertaken for any work involving ionising radiation. 

Specifically, Regulation 7 requires the radiation employer (Augean plc) to carry out a prior risk 

assessment before commencing work with radioactive materials at the ENRMF site. This 

document is intended to meet the requirements of this Regulation in relation to the operational 

phase of the controlled burial operation. 

1.2 RADIOACTIVE MATERIALS AND RADIATION HAZARDS 

The type and quantities of radioactive materials that may be accepted at ENRMF are 

described in the RSA93 application and supporting documents. In brief, the application 

includes a range of potential radionuclides from nuclear and non-nuclear practices (including 

radionuclides of natural origin) with a maximum total activity concentration of 200 Bq/g. This 

assessment pessimistically assumes that the waste received contains radionuclides at the 

maximum activity concentrations, which is unlikely to be the case in practice. 

The radionuclides considered emit a combination of alpha and beta particles and gamma 

rays. The handling of these materials can potentially give rise to a radiation hazard from: 

- external gamma exposure from proximity to the waste (either during handling waste 

containers or occupancy of the disposal areas); 

- internal radiation exposure from the inhalation of contaminated dust (air contamination) 

arising during the work; 

- internal radiation from the transfer and inadvertent ingestion of material (surface 

contamination) during the work; and 

- internal radiation from any contaminated wounds incurred during the work. 

This risk assessment focuses on the exposure of workers and other persons visiting the 

ENRMF site. The potential radiation exposure of persons off-site (i.e. members of the public) 

from a range of exposure pathways has been considered in detail in the RSA93 application. 

This demonstrated that the maximum dose to a member of the public is expected to be 

below 0.02 mSv per year 1 . This is well below the relevant IRR99 dose limit of 1 mSv per 

1 A higher dose of up to 1 mSv was associated with accidental (public) intrusion into the 

landfill. This is a post-closure scenario and is beyond the scope of this risk assessment. 


2 


year, and consequently doses to persons off-site are not considered in detail in this risk 

assessment. 

1.2 RISK ASSESSMENT REQUIREMENTS 

The purpose of this risk assessment is to identify the measures needed to restrict the 

exposure of employees and other persons to ionising radiation from the controlled burial 

waste (LLW) operations at ENRMF. 

IRR99 Regulation 7 also requires that potential radiation accidents are identified and 

quantified, and that steps are taken to prevent accidents, limit the consequences of any 

accidents that do occur, and to provide any necessary information, instruction, training and 

equipment to deal with such accidents. 

Paragraph 44 of the Approved Code of Practice to IRR99 recommends that the following 

matters should be considered when carrying out this risk assessment. The parts of this 

document that correspond to these matters are listed in the table below. 

Nature of the radiation source 1.2 

Estimated radiation dose rates to which anyone can be exposed 2.1 

Likelihood of contamination arising and being spread 2.2 

Results of previous personal dosimetry or area monitoring 2.1.1 

Advice from manufacturers or suppliers 4.5 

Engineering control measures and design features 4.1 

Any planned systems of work 4.1, 4.4.1 

Estimated levels of airborne and surface contamination 2.2.1, 3 

Effectiveness and suitability of personal protective equipment 4.5 

Extent of unrestricted access to working areas where dose rates or 

contamination levels are likely to be significant 

Possible accident situations, their likelihood and potential severity 3 

The failure of control measures or systems of work 3 

Steps to prevent identified accident situations or limit their consequences 3, 4.1.3 


4.2

3 


Paragraph 45 of the ACoP states that the risk assessment should enable the radiation 

employer to determine the following outcomes. Again, the relevant parts of this risk 

assessment are indicated in the table below. 

What action is needed to ensure that radiation exposures are as low as 

reasonably practicable (ALARP) 

What engineering controls, design features, safety and warning devices, and 

systems of work are needed 

Whether it is appropriate to provide personal protective equipment 4.1, 4.5 

Whether dose constraints for planning purposes are needed 4.1.4 

The need to alter the working conditions of any female employee who 

declares she is pregnant or breastfeeding 

4.1 

4.1 

4.1.6 

A dose investigation level to check that exposures are ALARP 4.1.5 

What maintenance and testing schedules are required 4.5 

What contingency plans are necessary 4.1 

The training of classified and non-classified employees 4.6 

The need to designate specific areas as controlled or supervised and the 

need for local rules 

The actions needed to ensure restriction of access for controlled or 

supervised areas 

4.2 

4.2, 4.4.1 

The need to designate certain employees as classified persons 4.3 

The need for individual dose assessment 4.3 

The responsibilities of managers 4.4.2 

An appropriate programme of monitoring or auditing of arrangements. 4.7 


2 RADIATION RISKS FROM NORMAL OPERATIONS 

2.1 RADIATION DOSE RATES AND EXTERNAL RADIATION RISKS 

2.1.1 Augean employees engaged in the LLW operation 

4 


The radiation dose rates from a range of radionuclides have been calculated as part of the 

supporting documents to the RSA93 application. Extracts from these documents, relevant to 

the estimation of external dose, are reproduced in Appendix 1. 

For all external dose scenarios, cobalt-60 is the limiting radionuclide (i.e. it gives rise to the 

highest dose rates). For the purpose of this risk assessment, the following representative 

dose rates, working patterns and estimated doses are used. 

Work activity 

Receipt of waste 

consignments, including QA 

and monitoring, etc. 

Dose rate 

(Sv/h) 

10 

2 

Occupancy 

(hours/year) 

50 

100 

Estimated 

annual dose 

(mSv) 

Transfer and placement of 

waste in landfill. 2 100 0.2 

Occupancy of covered waste 

area 

Summary 

0.5 

0.2 

2 100 0.2 

TOTAL ESTIMATED ANNUAL DOSE 1.1 mSv 

The above estimates are likely to be conservative, and it is unlikely that the same 

person(s) will be exposed during all the work activities listed above. Nevertheless, it 

is reasonable to assume, for planning purposes, that annual external doses of the 

order of 1 mSv per year might be associated with the LLW operation. 

2.1.2 External radiation risks to other persons 

Such persons might include other employees (i.e. not involved in the LLW operation), visitors 

to site, etc. Such persons would be unlikely to be exposed to dose rates above 1 μSv/h, and 

exposure times would be expected to be short. Consequently, it is expected that external 

doses to such persons should be negligible. 

External doses to members of the public (during and after the LLW operation) were 

estimated in the RSA93 application, and are a small fraction of the 20 μSv/y dose constraint. 

2.2 CONTAMINATION LEVELS AND INTERNAL RADIATION RISKS 

The waste will be in closed containers (either steel drums or bulk bags) throughout the LLW 

operation. Furthermore, waste consigners will be required to demonstrate that the external 

surfaces of these containers are effectively free of loose contamination. Consequently, 

contamination levels, and hence internal radiation doses, during normal operations are 

expected to be negligible. 

There is the potential for contamination and internal exposures arising from accidents, in 

particular a damaged container. This is considered in the Section 3. 


3 RADIATION RISKS FROM ACCIDENTS 

The following reasonably foreseeable incidents/accidents have been identified: 

5 


3.1 The delivery of waste containing unexpectedly high levels of radioactivity 

The likelihood of receiving waste that is more radioactive than expected is limited by the strict 

pre-acceptance criteria and associated procedures that are to be put in place. In addition, it is 

expected that incoming consignments will be monitored, and a dose rate acceptance test 

applied. Thus any radiation exposures from this scenario, should be limited to a brief external 

exposure to increased dose rates at the receiving stage. Even if the dose rate is 10x the 

acceptance criteria, the resulting doses to workers from the monitoring and subsequent 

quarantine of the consignment would be expected to be negligible. 

3.2 Dropping or otherwise damaging a container of waste and spilling the contents 

The “dropped bag” scenario is specifically considered in the RSA93 application using a 

pessimistic dispersion model to estimate the radiation doses (from dust inhalation) to workers 

and persons off-site. This assessment is principally concerned with the exposure of workers, 

in particular those that may be involved in cleaning up any spills. Consequently, for this risk 

assessment the following general “spillage” scenario is assumed: 

either type of waste container (drum or bag) could be damaged; 

contaminated dust is released producing a localised dust loading of 10 mg/m 3 , which 

is considered a pessimistic assumption for an accident outdoors; 

workers remain in the above dust loading for a total of 4 hours (to allow for any cleanup). 

the worker breathing rate is 1.2 m 3 /h and no respiratory protective equipment (RPE) 

is worn; and 

dust is inadvertently ingested (e.g. during the clean-up) at a rate of 3.45 x 10 -5 kg/h 

(the same rate as assumed in the RSA application for excavation scenarios) 

The above assumptions produce an inhaled dust mass of 48 mg, and an ingested dust mass 

of 138 mg. ICRP dose coefficients for inhalation and ingestion (the same as those used in 

the RSA93 application) are given in the Appendix to this risk assessment. Combining these 

with the mass of dust inhaled and ingested, and an activity concentration of 200 Bq/g (i.e. a 

worst case assumption) gives the following (rounded) internal doses: 

Radionuclide 

Estimated internal dose from a single spillage (mSv) 

Inhalation Ingestion Total 

Ac-227 5 5 

Th-229 2 2 

Th-230,232 

Pa-231 

Pu-238, 239, 240, 242 

Am-241 

Ra-228, Th-228 

U-232, Np-237 

Cm-243, 244 

All other radionuclides

6 


Thus, dust inhalation is the dominant exposure pathway. The highest estimated doses are for 

actinium-227 and thorium-229. However, it is considered highly unlikely that waste would 

contain these radionuclides at 200 Bq/g. 

There would also be an external dose associated with the clean up. Assuming a 4 hour 

exposure to an average dose rate of 10 μSv/h gives an external dose of 0.04 mSv. Taking all 

these factors into account, it is concluded that the radiation exposure (internal plus 

external) from a spillage of waste containing up to 200 Bq/g is unlikely to exceed 1 

mSv. This includes any exposures from the subsequent clean-up of the spill. 

3.3 Internal exposure from contaminated wounds 

Under normal circumstances this is not a reasonably foreseeable exposure scenario. 

However, if contamination does arise, for example because of the spill scenario in 3.2 above, 

then this additional accident exposure pathway becomes a possibility. It is considered that 

doses from this pathway would be likely to be the same order of magnitude as from 

inadvertent ingestion, i.e., less than 0.1 mSv. 

The UKAEA Safety Assessment Handbook (UKAEA/SAH/D9, Issue 1, March 2006) gives 

dose factors for contaminated wounds. Assuming that 0.1 g of material (at 200 Bq/g) 

becomes incorporated into a wound, the highest estimated dose is approximately 3 mSv, from 

actinium-227. As mentioned above, this radionuclide is most unlikely to predominate, and it is 

concluded that internal doses from a contaminated wound would be very unlikely to exceed 1 

mSv in practice. 


4. RECOMMENDED ACTIONS - REQUIREMENTS OF IRR99 

4.1 RESTRICTION OF EXPOSURE (IRR99 REGULATION 8) 

4.1.1 Summary of estimated doses 

7 


Regulation 8 requires that every radiation employer shall take all necessary steps to ensure 

that the radiation exposure of employees and other persons is as low as reasonably 

practicable (ALARP). The preceding dose assessment produced the following estimated 

effective doses: 

Augean LLW workers 

Normal operations: 1 mSv/y from external exposure 

Accidents: Negligible (

8 


protection in the event of a spillage of waste. This is existing practice at the ENRMF site 

for al operatives. 

Radiation monitoring (individual and environmental) is required – see 4.3 and 4.7 below. 

Local rules and training should be provided - see 4.4 and 4.6 below. 

Other persons 

Other persons should be excluded from the immediate area during the LLW operation. 

The dust suppression measures for spills, as described below, should also ensure that 

the spread of airborne dust is minimised. No other specific protection measures are 

required. 

4.1.3 Accidents – prevention and mitigation 

Dose rate checks on incoming consignments of waste should be undertaken, as 

recommended above. 

Contingency plans should be prepared for dealing with spillages of waste. These should 

include the following precautions: 

o Simple dust suppression measures (e.g. damping down, and avoiding dust 

resuspension during clean-up) should be applied, where practicable. 

o As an additional precaution, workers should wear respiratory protective 

equipment when cleaning up spills – see 4.5 below. 

o Spilled material must be placed into suitable containers for disposal, and the 

affected area should be monitored to ensure that all contaminated material has 

been removed. 

The above precautions should ensure that the radiation doses from accidents are 

negligible (

4.1.6 Pregnant and breast-feeding employees 

9 


Regulation 8(5) contains additional dose restriction provisions for such employees. For 

pregnant women, it is recommended the dose to the foetus should be kept below 1 mSv. 

Whilst exposures of over 1 mSv are unlikely to occur, as a precaution it is recommended that 

pregnant employees are not allowed in the LLW work areas. 

For breastfeeding women, it is recommended that they avoid situations where significant 

bodily contamination might occur. As a general precaution, it is recommended that such 

women are not allowed in the LLW work areas. 

The risks associated with radiological hazards should be incorporated in the company risk 

assessment for pregnant and breastfeeding employees. 

4.2 DESIGNATED AREAS 

4.2.1 Controlled areas 

Regulation 16 requires the designation of a controlled area where either: 

a) radiation doses are likely to exceed three-tenths of the annual dose limits for workers 

(e.g. 6mSv/y effective dose); or 

b) special working procedures are required to restrict radiation exposures. 

Worker doses are not expected to exceed 6 mSv/y. However, it is considered that special 

working procedures (as defined in Regulation 16(1)) are appropriate in respect of certain 

operations. Consequently the following recommendations are made: 

Incoming waste consignments should be rapidly processed, and should not remain in any 

one area for an extended period of time. On this basis, a controlled area (for example, 

around arriving vehicles) is not recommended. 

A quarantine area should be provided for waste consignments that do not meet the dose 

rate limits described previously, and this should be designated as a controlled area 

whenever such consignments are quarantined. It should be ensured that the dose rate 

outside this area is below 2 μSv/h. 

During the deposition of waste containers, elevated dose rates are present, and there is 

the potential for accidents (dropped containers, etc.). It is recommended that this area is 

designated as a controlled area during the disposal operation, and should remain 

designated until a satisfactory covering layer has been applied (see 4.1.2). 

Controlled areas should, where practicable be physically demarcated and warning signs 

posted at the points of entry. For the above areas, the following is recommended: 

Quarantine area: the perimeter should be fully demarcated, ideally with fencing, but if 

not, with rope barriers or similar. A controlled area warning sign should be posted at all 

points of potential access. 

Disposal area: during the operational period the area will be occupied or under 

surveillance and it is considered sufficient to temporarily post controlled area warning 

signs at the access points to the area. 


10 


Access to the controlled areas should be restricted to authorised personnel. Local rules and 

Radiation Protection Supervisors (Regulation 17) should be provided for controlled areas – 

see 4.4 below. 

4.2.1 Supervised areas 

The Regulations also require that a supervised area should be designated where it is 

considered necessary to keep the radiological conditions under periodic review. Although 

some confirmatory monitoring is recommended outside the controlled areas (see 4.7 below), 

the designation of a supervised area is not considered necessary provided that the 

aforementioned dose rate limits are met. 

4.3 CLASSIFIED PERSONS AND INDIVIDUAL MONITORING 

4.3.1 Designation of classified persons 

Regulation 20 requires workers to be designated as classified persons if they are likely to 

receive an effective dose in excess of 6 mSv per year. This risk assessment indicates that 

doses are expected to be well below this value and, therefore, it is not recommended that 

Augean employees are designated as classified persons. 

4.3.2 Monitoring of individual dose 

As a means of confirming the restriction of exposures, and for checking against the Dose 

Investigation Level, it is recommended that a programme of individual dose monitoring is 

implemented for all Augean employees engaged in the LLW operation. For monitoring 

external exposure, it is recommended that passive whole body dosemeters (e.g. TLDs) are 

worn, and Augean should make the necessary arrangements with an appropriate dosimetry 

service. 

Internal exposures during normal operations are expected to be negligible, and the 

precautions listed in Section 4.1.3 should ensure that this is also the case for internal 

exposures from accidents. The systematic assessment of individual internal dose is not, 

therefore, warranted (see ACoP paragraph 386). 

4.4 WORKING PROCEDURES AND SUPERVISION 

4.4.1 Local rules 

Regulation 17 requires that Local Rules are written for work in controlled areas. Augean 

should draft Local Rules, consulting the RPA as required, to ensure that the format and 

content of the rules (as specified in IRR99) are appropriate. The Local Rules must include: 

- the dose investigation level; 

- a description of each controlled area, and the means by which access is restricted; 

- names of the Radiation Protection Supervisors (see below); 

- for each controlled area, appropriate working instructions (PPE, good working 

practice, monitoring arrangements, etc) including written arrangements for the entry of 

non-classified persons into the controlled areas; 

- details of any contingency arrangements, for example for dealing with spillages. 


4.4.2 Radiation Protection Supervisors (RPSs) 

11 


Regulation 17 requires that Augean appoint one or more employees as RPS. The main role 

of the RPS is to ensure that the Local Rules are being observed, and whoever is appointed 

should be suitable for the role. In practice, this means that they are appropriately trained and 

are able to properly supervise the work being undertaken. There should be an RPS present 

on the ENRMF site whenever LLW is being processed. 

It should be noted that the RPSs are not a substitute for line-management responsibilities. 

Augean must ensure that line managers involved in the LLW operation project are familiar 

with the contents of this risk assessment and the local rules, and their responsibilities for 

health and safety. 

4.5 PERSONAL PROTECTIVE EQUIPMENT 

The internal dose to Augean employees from inhalation of dust during normal operations is 

expected to be negligible. As indicated in Section 3.2, the inhalation dose from dealing with a 

waste spill is likely to be below 1 mSv. Although this is well below the 20 mSv/y dose limit, it 

is recommended that respiratory protective equipment be worn in the interests of keeping 

exposures ALARP, and to ensure compliance with the dose constraint and dose investigation 

level. 

The RPE should be readily available in the event of a spill occurring, and must be put on 

before attempting to clean up any spilt LLW material. 

RPE with a minimum protection factor of 5 is recommended: this, combined with the dust 

suppression measures described in Section 4.1.3, should ensure that inhalation doses are 

below 0.1 mSv. In addition: 

- RPE must be CE marked; 

- the comfort of the wearer should be taken into account when choosing a particular type 

of respirator; 

- RPE should be fit-tested to ensure a good seal to individual faces; 

- If the RPE is reusable, it should be thoroughly examined at suitable intervals and 

properly maintained in accordance with the manufacturer’s instructions, and as required 

by Regulation 10(2). Suitable records of examinations and maintenance should be 

made and kept for at least 2 years; and (very importantly) 

- training in the proper use and maintenance of RPE must be provided. 

In addition to RPE, protective clothing should also be worn by Augean employees when 

working in the area, as follows: 

- coveralls must be worn, the type being selected according to the nature of the work. 

- protective gloves must be worn, the type being selected according to the nature of the 

work. Gloves should be impermeable and be sufficiently strong to withstand wear and 

tear and provide protection against cuts/wounds; 

- footwear – normal safety footwear is considered sufficient; and 

- suitable washing and changing facilities should be provided for use by workers before 

lunch breaks, ends of shift, etc.. This should include facilities for separate storage of 


12 


clean and dirty clothing, and hand/face washing facilities with elbow-operated taps. It is 

suggested that a contamination monitor should also be considered for reassurance 

purposes, i.e. so that workers can check themselves if they wish. 

After dealing with a spill, coveralls and gloves may need to be disposed of. It is suggested 

that disposable outer coveralls and gloves should be provided for use when cleaning up spills. 

Gloves should be taped to coveralls where there is a risk of up-sleeve contamination during a 

clean-up. Footwear should be washed down before leaving the area. 

4.6 INFORMATION, INSTRUCTION AND TRAINING 

To meet the requirements of Regulation 14, the following arrangements are recommended: 

All Augean employees engaged in LLW work should receive training in radiation 

protection prior to the work. This should cover: 

o the nature of the radiation hazards associated with LLW; 

o the risks to health associated with exposure to radiation; 

o the precautions that need to be taken to restrict exposures, including the contents 

of this risk assessment and the local rules; 

o the correct use of RPE; and 

o the regulatory requirements associated with the work, and the importance of 

complying with these requirements. 

In addition, specifically appointed Augean employees should receive additional 

training to act as a Radiation Protection Supervisor(s) and (if applicable) to examine and 

maintain RPE. 

Other persons working on the ENRMF site should be provided with information to 

indicate that the certain areas are designated as controlled areas, that access to these 

areas is restricted, and that warning signs should be observed. 

4.7 WORKPLACE MONITORING 

The following programme of workplace monitoring is recommended. 

Dose rates 

All incoming LLW containers should be subject to dose rate monitoring, and the results 

recorded. The dose rate a 1 metre from a container must not exceed 10 μSv/h. 

Any containers that do not meet the above criteria should be placed in quarantine. The 

dose rate around the perimeter of the quarantine area must be measured (and recorded) 

whenever containers are placed inside. The dose rate at the perimeter must not exceed 

2 μSv/h. 

The dose rate on top of any newly deposited material must be measured after the 

minimum 300mm cover is applied. The dose rate at a height of 1 metre must not exceed 

2 μSv/h. If necessary, additional cover should be applied. The measured dose rate and 

the thickness of cover applied should be recorded. 


13 


Annual environmental-level dose rate monitoring will be undertaken by the RPA at 

representative locations around the site boundary. 

Surface contamination monitoring 

Surface contamination is not expected to arise during routine operations. However, 

confirmatory monitoring should be undertaken once every month in the following areas: 

o At the exit point from the disposal area 

o After the vehicle wheel wash 

o Change rooms including PPE 

o At the main exit from the site. 

In addition, contamination monitoring should be undertaken after cleaning up any waste 

spillages. This should include: 

P V Shaw 

14 July 2009 

o Monitoring the affected area, i.e. to confirm that all contamination has been 

removed. 

o Monitoring all persons and items leaving the area to ensure that the spread of 

contamination is avoided. 

Document History 

Version 1: 27 March 2009. First complete draft produced by RPA 

Version 2: 7 July 2009. Incorporating comments by Augean 

Version 3: 14 July 2009. Revised by RPA to incorporate comments. 


APPENDIX TO ENRMF RISK ASSESSMENT 

14 


SUPPORTING RADIOLOGICAL DATA TAKEN FROM RSA93 APPLICATION 

A.1 EXTERNAL DOSE DATA 

WASTE IN CONTAINERS 

Specific calculations were undertaken for cobalt-60 (the most restrictive radionuclide) and 

caesium-137 (for comparison) at 200 Bq/g – for both high and low density waste in drums and 

bulk bags. A summary of the results is given below. 

Drums 

Estimated dose rate (μSv/h) 2 

Exposure 

scenario Cobalt-60 Caesium-137 

- contact (1 cm) 

- 1 metre 

- 2 metres 

Bulk waste bags 

- contact (1 cm) 

- 1 metre 

- 2 metres 

100 

6 

2 

125 

14 

5 

DEPOSITED WASTE 

Specific calculations were also undertaken to estimate the dose rate above deposited waste 

covered with 30 cm of compacted topsoil. The results are summarised in the following table. 

25 

1.5 

0.5 

Radionuclides Calculated dose rate (μSv/h) 

Cobalt-60 at 200 Bq/g 

Other radionuclides at 200 Bq/g 

5 to 10 

In this assessment a maximum dose rate of 2 μSv/h above the covered waste has been 

recommended (see 4.1.2). This value has, therefore, been used (in part 2.1.1) to estimate 

doses to workers. 

2 The values have been rounded and represent the average dose rate calculated for high 

density (2g/cm 3 ) and low density (1 g/m 3 ) waste. In the case of cylindrical drums, the average 

values calculated for the (curved) sides and (flat) ends are given. 

A.2 INTERNAL DOSE DATA - ICRP INTERNAL DOSE COEFFICIENTS 

15 


For consistency purposes, the data below are the same as those used in the RSA93 

application, and are the relevant ICRP dose coefficients for members of the public. The ICRP 

dose coefficients for workers are slightly different, but this does not materially affect the 

outcome of this risk assessment. 

Radionuclide 

Dose coefficient (Sv/Bq) 

Inhalation Ingestion 

H-3 2.6E-10 1.8E-11 

C-14 5.8E-09 5.8E-10 

Cl-36 7.3E-09 9.3E-10 

Fe-55 7.7E-10 3.3E-10 

Co-60 3.1E-08 3.4E-09 

Ni-63 4.8E-10 1.5E-10 

Sr-90 1.6E-07 2.8E-08 

Nb-94 1.1E-08 1.1E-08 

Tc-99 1.3E-08 6.4E-10 

Ru-106 6.6E-08 7.0E-09 

Ag-108m 3.7E-08 2.3E-09 

Sb-125 5.5E-08 3.1E-09 

Sn-126 3.1E-08 7.1E-09 

I-129 3.6E-08 1.1E-07 

Ba-133 3.1E-09 1.5E-09 

Cs-134 6.8E-09 1.9E-08 

Cs-137 3.9E-08 1.3E-08 

Pm-147 5.0E-09 2.6E-10 

Eu-152 4.2E-08 1.4E-09 

Eu-154 5.3E-08 2.0E-09 

Eu-155 6.9E-09 3.2E-10 

Pb-210 5.6E-06 6.9E-07 

Ra-226 9.5E-06 2.8E-07 

Ac-227 5.5E-04 1.1E-06 

Th-229 2.6E-04 6.1E-07 

Th-230 1.0E-04 2.1E-07 

Pa-231 1.4E-04 7.1E-07 

Th-232 1.1E-04 2.3E-07 

U-232 4.7E-05 4.6E-07 

U-233 9.6E-06 5.1E-08 

U-234 9.4E-06 4.9E-08 

U-235 8.5E-06 4.7E-08 

U-236 3.2E-06 4.7E-08 

U-238 8.0E-06 4.5E-08 

Np-237 5.0E-05 1.1E-07 


Radionuclide 

Dose coefficient (Sv/Bq) 

Inhalation Ingestion 

Pu-238 1.1E-04 2.3E-07 

Pu-239 1.2E-04 2.5E-07 

Pu-240 1.2E-04 2.5E-07 

Pu-241 2.3E-06 4.8E-09 

Pu-242 1.1E-04 2.4E-07 

Am-241 9.6E-05 2.0E-07 

Cm-243 3.1E-05 1.5E-07 

Cu-244 2.7E-05 1.2E-07 

16 



Annex D 

Dose Rate calculations in support of 

Low Level Waste disposal authorisation, 

TSG(09)0487 


Technical Services Group 

NOT PROTECTIVELY MARKED 

Reference: TSG(09)0487 

Issue: Issue 2 

Date: 15 th July 2009 

DOSE RATE CALCULATIONS IN SUPPORT OF A LOW LEVEL 

WASTE DISPOSAL AUTHORISATION 

UK-10497 

SUMMARY 

Dose rate calculations were performed in MicroShield to support a low level waste disposal 

authorisation. The dose rate was calculated on contact, 1m and 2m from a 200-litre drum 

and a bulk waste bag of soil and rubble waste. Dose was found to be highest when dealing 

with a 60 Co source. 

Prepared By: 

Checked By: 

Approved By: 

Name and Organisation Signature Date 

Tony Lansdell 

TSG 

Barry Cook 

TSG 

Gráinne Carpenter 

TSG 

ELECTRONIC 

COPY 



Table of Contents 

1 Introduction.......................................................................................................................3 

2 Methodology.....................................................................................................................3 

2.1 Background ..............................................................................................................3 

2.2 Materials ...................................................................................................................3 

2.3 200-litre drum case...................................................................................................3 

2.4 Bulk waste bag case.................................................................................................4 

2.5 MicroShield calculation details and uncertainties .....................................................5 

3 Results..............................................................................................................................6 

3.1 Low density case ......................................................................................................6 

3.2 High density case .....................................................................................................6 

4 References .......................................................................................................................7 


2

1 INTRODUCTION 


Dose rate calculations were required to support a low level waste disposal 

authorisation. Cases were run using MicroShield v7.02 [1] to determine the maximum 

dose rate at a series of distances from the wasteform, for two different wasteform 

geometries. 

2 METHODOLOGY 


MicroShield was used to determine the maximum dose rates at various distances from 

packaged contaminated soil and rubble waste. Two cases were defined, one for waste 

packaged in a 200-litre drum, and one for waste in a flexible bulk waste bag. In each 

case, the dose rate was required on contact, at 1m and at 2m from the wasteform. 

2.2 Materials 

Two sub-cases were defined; one for soil/rubble waste containing 200 Bq/g of 60 Co, 

and one for soil/rubble containing 200 Bq/g of 137 Cs. As soil is not a material type 

available to MicroShield, concrete was chosen to represent the waste material 

composition. 

The bulk density of the soil and rubble wastes will vary depending on the composition 

of the waste, the level of compaction used, and the packing efficiency. Cases were 

assessed for two wasteform density values to provide bounding results. 

A search of literature revealed that the bulk density of soil was typically 1.0 g/cm 3 for 

loose soil, 1.3 g/cm 3 for undisturbed soil, and 1.6 g/cm 3 for compacted soil [2]. 

Concrete rubble was assumed to be the same as normal density concrete, 2.35 g/cm 3 . 

The minimum density case was taken to be packaged loose soil, with a density of 1.0 

g/cm 3 . 

The maximum density wasteform was taken to contain the maximum amount of 

concrete rubble, with the remaining space taken up by compacted soil. It was assumed 

solid pieces of rubble would have a packing efficiency no better than 50%, hence 50% 

of the volume was assumed to be rubble (2.35 g/cm 3 ), with the remaining 50% 

consisting of compacted soil (1.6 g/cm 3 ). The maximum density of the wasteform was 

therefore predicted to be 1.98 g/cm 3 , and 2.0 g/cm 3 was used for simplicity. The 

maximum range of wasteform density used was therefore between 1 and 2 g/cm 3 . 

2.3 200-litre drum case 

200-litre drums are steel-walled cylindrical drums of diameter of 67 cm and height 

87 cm. The shielding effect of the drum was ignored to be conservative, hence the 

drum wall was not modelled in MicroShield, and the wasteform was taken to be a 

cylindrical volume of the above drum dimensions. Dose points were positioned on 

contact, 1m, and 2m from the wasteform, both in a radial direction ( Figure 1) 

and an 

axial direction ( Figure 2). 

Radial dose points were located at half the height of the 

cylinder, where the dose rate is maximised. Axial dose points were on axis with the 

centre of the cylinder, where the dose rate is maximised. 


3


Figure 1: Radial dose points for 200-litre drum (images from MicroShield) 

2.4 Bulk waste bag case 

Figure 2: Axial dose points for 200-litre drum 

The bulk waste bag is a cube of side length 1m, and the wasteform was modelled in 

MicroShield as a rectangular volume with all sides 1m in length ( Figure 3). 

Again, the 

wall of the bag was not explicitly modelled to be conservative. The dose points were 

positioned in line with the centre of a flat face, where the dose rate is maximised. 


4


Figure 3: Dose points for bulk waste bag 

2.5 MicroShield calculation details and uncertainties 

Energy deposition to dose rate conversion was performed automatically in MicroShield 

using built-in tables of effective dose rate, taken from ICRP-51 [3]. This presents a 

series of possible dose rates depending on the assumed irradiation geometry. The 

highest biological dose rate is produced assuming anterior-posterior geometry (with the 

gamma rays entering a person from the front and exiting through the back), and to be 

conservative it was this maximum dose rate that was reported. Dose rates can vary by 

approximately 30%, depending on which geometry is assumed. 

MicroShield approximates the contribution of scattered radiation to the resulting dose 

rate by the use of build-up tables. The dose rate is dependant on which material is 

chosen as the dominant scattering medium. In accordance with the MicroShield 

manual, the material containing the highest number of gamma ray mean free paths 

should be used as the build-up material – hence in these cases, the source was 

chosen as build-up material. If the air gap is chosen as the scattering medium, it was 

found that the resulting dose rates increased by 6% for 60 Co cases, and increased by 

12% for 137 Cs cases, but these results would be over-pessimistic. 

MicroShield uses a point-kernel integration technique to determine the dose rate. This 

involves splitting the geometry into pieces (kernels). The quadrature order of the 

calculation determines the number of kernels used and hence the accuracy of the 

approximation; the default quadrature order was used for the reported results. The 

order of the calculation was increased by a factor of two in each dimension, and the 

contact results only increased by 0.3%, which is well within the range of other sources 

of uncertainty in the calculation. Further increases in accuracy produced no change to 

the results. 

In all cases assessed, the ‘contact’ dose rate point was actually positioned at 1 cm 

from the surface, as the method of calculation used by MicroShield is known to become 

unstable at distances closer than 1 cm, though this will strongly depend on the 

integration order used.. 

137 Cs is a beta emitter. Its daughter, 137m Ba is the source of the gamma radiation. 

Where a source containing 137 Cs was specified, its daughter product 137m Ba was also 

included in equilibrium concentration with 137 Cs. Since the half-life of 137m Ba is short 

(2.5 minutes), it will almost always be found in equilibrium with its parent radionuclide. 


5

3 RESULTS 

3.1 Low density case 

Case 


Density = 1.0 g/cm 3 , specific activity = 200 Bq/g = 200 Bq/cm 3 . 

Dose Rate (Sv/hr) 

Curved cylinder face Flat cylinder face 

Contact* 100 cm 200 cm Contact* 100 cm 200 cm 

60 Co 91.65 6.019 1.95 98.35 4.766 1.465 

137 Cs 21.6 1.421 0.458 23.57 1.109 0.335 

Case 


Flat cube face 

Contact* 100 cm 200 cm 

60 Co 120.7 12.45 4.065 

137 Cs 27.6 2.875 0.925 

3.2 High density case 

Case 

Density = 2 g/cm 3 , specific activity = 200 Bq/g = 400 Bq/cm 3 . 


Curved cylinder face Flat cylinder face 

Contact* 100 cm 200 cm Contact* 100 cm 200 cm 

60 Co 109.2 7.293 2.316 123.9 5.682 1.654 

137 Cs 24.35 1.639 0.515 28.01 1.283 0.368 

Case 


Flat cube face 

Contact* 100 cm 200 cm 

60 Co 131.3 14.5 4.526 

137 Cs 28.72 3.261 1.01 

* Contact doses were located at 1 cm from the wasteform. 


6

4 REFERENCES 


1 MicroShield v7.02, Grove Software Inc, 2007 

2 “Soils and Soil Fertility”, page 54, Sixth Edition, F.R. Troeh and L.M. Thompson, 

Blackwell Publishing, 1979. 

3 ICRP-51 (1987) Data for use in protection against external radiation 


7

Annex E 

Development of a Framework for 

Assessing the Suitability of Controlled 

Landfills to Accept Disposals of Solid 

Low-Level Radioactive Waste: 

Principles and Technical Manual 

SNIFFER, 2005 

These manuals describe in detail the models used for the assessment in the 

application. Hardcopies have not been included. The information can be 

read and downloaded at: 

http://www.sniffer.org.uk/Resources/UKRSR03/Layout_EnvironmentalRegulati 

on/11.aspx?backurl=http%3a%2f%2fwww.sniffer.org.uk%3a80%2fthemes%2f 

environmental-regulation.aspx&selectedtab=completed 


Annex F 

Application Form 


Environment Agency 

Radioactive Substances Form RSA3 (interim) 

Application for authorisation to accumulate and dispose of radioactive 

waste from non-nuclear premises 

Radioactive Substances Act 1993 Sections 13 & 14 

Note 

This application form should be read and completed in conjunction with the current Environment 

Agency guidance, available on the Environment Agency web site 

http://www.environment-agency.gov.uk/business/444304/945840/1064273/?version=1&lang=_e 

or on request from Environment Agency offices (including the Environment Agency’s Interim 

Guidance to users of sealed sources on the High-activity Sealed Radioactive Sources and Orphan 

Sources Regulations 2005). Words used in this form have the same meaning as in the above 

guidance. 

To get an authorisation to accumulate and dispose of radioactive waste you generally also need to 

hold a registration for the premises. If you do not already hold a relevant and suitable registration for 

radioactive substances, you should fill in an application form RSA1 to cover the open or sealed 

sources you use or intend to use, and send it in with this form. You should note that the Environment 

Agency may inspect the premises and/or ask the Police to review security during consideration of this 

application for authorisation. 

The issue of the certificate of authorisation under Sections 13 and 14 of the Radioactive Substances 

Act 1993 does not allow you to contravene any other statutory legislation that might also apply to the 

premises. 

This form should only be used for the accumulation and disposal of radioactive waste from a single 

defined premise by a single organisation. You do not need to hold an authorisation to cover 

accumulation and disposal of radioactive waste which is within the scope of an exemption order, 

provided you can comply with all of the conditions in such an order. 

If you need more space than this form allows, please continue on separate sheets. Please write the 

number of the question you are answering on the top of each continuation sheet. 


Contents 

A For Office Use 

B Company or Organisation Details 

C Type of Application 

D Premises Details 

E Contact Details 

F Producing Radioactive Waste 

G Incineration on the Premises 

H Disposal of Gaseous Waste 

I Aqueous Waste 

J Organic Liquid Waste 

K Very Low Level Solid Waste 

L Solid Waste (excluding HASS and sources of similar potential hazard) 

M NAIR Arrangements 

N Checklist 

O Data Handling 

P Payment 

Q Declaration 

R Signature 

Annex 

S Sealed Sources 

A For Environment Agency use only 

New application number 

Date received – Agency date stamp 

Existing authorisation for premises? Yes 

Existing number 

No 

New operator account? Yes 

Invoice code 

Date 

No 

Commercially confidential? Yes 

No 

Sign 

National security? Yes 

No 

Sign 

Fee £ 

Date received 

Amount received £ 

Sign 

Declaration signed? Yes 

No 

Nuclear site tenant? Yes 

Yes No 


B Company or organisation details 

B1 Please give the name of the company or organisation carrying on the undertaking creating 

radioactive waste on the premises. 

Name Augean South Limited 

Registered office or business address 

If no registered office please give principal place of business 

4 Rudgate Court, Walton, Wetherby 

LS23 7BF 

Postcode 

Companies House registration number 

if you have one 

4636789 

B2 On behalf of what type of organisation are you applying? 

Tick the option which is most appropriate 

Sole trader 

Partnership 

Limited liability company 

Public limited company 

District or county council or unitary authority 

Educational establishment 

NHS trust 

Private hospital 

Other medical establishment please give details 

Non-governmental public body 

Ministry of Defence 

Other Government department 

Other please give details 

C Premises Details 

C1 Where are the premises you want to accumulate and dispose of radioactive waste? 

Address 

East Northants Resource Management Facility, Stamford Road, 

Kings Cliffe, Northamptonshire 

Postcode PE8 6XX 

Ordnance Survey national grid reference 

For example SJ 123 456 

TF 010 000 

C2 Are the premises located on a nuclear licensed site? 

ie as a tenant 

Yes 

No 

C3 Which council or unitary authority are the premises in? 

If premises are on a boundary please give names of all relevant authorities. 

Borough or district council or unitary authority 


East Northamptonshire Council 

County council unless there is a unitary authority 

Northamptonshire County Council 

C4 Who is the sewerage undertaker for the premises? 

This is often the local water supply company 

This is currently tankered to an offsite treatment works 

C5 Which Police Force area are the premises in? 

Avon & Somerset 

Bedfordshire 

Cambridgeshire 

Central Scotland 

Cheshire 

City of London 

Civil Nuclear Constabulary 

Cleveland 

Cumbria 

Derbyshire 

Devon & Cornwall 

Dorset 

Dumfries & Galloway 

Durham 

Dyfed-Powys 

Essex 

Fife 

Gloucestershire 

Grampian 

Greater Manchester 

Gwent 

Hampshire 

Hertfordshire 

Humberside 

Kent 

Lancashire 

Leicestershire 

Lincolnshire 

Lothian & Borders 

Merseyside 

Metropolitan 

Ministry of Defence Police 

Norfolk 

Northamptonshire 

Northern 

Northumbria 

North Wales 

North Yorkshire 

Nottinghamshire 

Northern Ireland 

South Wales 

South Yorkshire 

Staffordshire 

Strathclyde 

Suffolk 

Surrey 

Sussex 

Tayside 

Thames Valley 

Warwickshire 

West Mercia 

West Midlands 

West Yorkshire 

Wiltshire 


D Contact Details 

We need the names and details of members of your organisation to help us deal with your application 

and authorisation quickly and efficiently. 

Application contact 

D1 Who can we contact with questions on your application? 

Name Dr Gene Wilson 

Position Group Technical Director 

Address 

Postcode 

Phone 

Fax 

E-mail 

Operational contact 

D2 Who will be responsible for day to day supervision of the accumulation and disposal of 

radioactive waste? If different people are responsible for some wastes, please give details of each 

such person 

Name Simon Moyle 

Position 

Address 

Postcode 

Phone 

Fax 

E-mail 

Payments and invoices 

D3 Who can we contact about payment of fees and charges? 

Name Adam Emmott 

Position 

Address 

East Northants Resource Management Facility, 

Stamford Road, Kings Cliffe, Northamptonshire 

PE8 6XX 

01780 444905 

01780 444901 

genewilson@augeanplc.com 

Site Manager 

East Northants Resource Management Facility, Stamford 

Road, Kings Cliffe, Northamptonshire 

PE8 6XX 

01780 444900 

01780 444901 

simonmoyle@augeanplc.com 

Head of Group Finance 

4 Rudgate Court, Walton, Wetherby 

Postcode LS23 7BF 

Phone 01937 844980 

Fax 01937 844241 

E-mail adamemmott@augeanplc.com 


E Type of Application 

E1 When would you like the authorisation to start? 

We will try to meet your needs but it can take up to 4 months from date of receiving a valid application 

with all of the information we need (and fee), before you receive your authorisation. After you receive 

your authorisation there is usually another 28 days before you can start accumulating and disposing 

of radioactive waste. 

Date 17 November 2009 

E2 When would you like any current authorisation cancelled? 

This will be the same date on which your new authorisation starts unless you tell us otherwise. 

Date to cancel any existing authorisation No current authorisation 

E3 Have you made any other application to the Environment Agency (or previously HMIP) for 

any permission under the Radioactive Substances Acts, 1960 or 1993? 

Yes 

No go to ** 

E4 Where relevant, please give details for a current or previous authorisation for these 

premises. 

User 

N/A 

Authorisation number 

Date of authorisation 

E5 Are you applying for 

a new authorisation for premises you do not hold a current authorisation for? 

a variation to an authorisation for your existing premises? 

a new authorisation for a new legal entity? 

a variation to an authorisation because you have changed your name but not your legal 

status? 

other please give details 


F Producing Radioactive Waste 

F1 Practice 

Please indicate the practice or work activity which creates radioactive waste. 

Please tick each relevant box. 

Note this list is adapted from the definitive list of existing practices on the Defra web site (link to 

Defra) 

1.1 Enrichment of uranium - Use of the centrifuge process 

2.1 Production of nuclear fuel - Manufacture of uranium metal and oxide fuel for power reactors 

2.2 Production of nuclear fuel - Manufacture of mixed oxide fuel for power reactors 

2.3 Production of nuclear fuel - Manufacture of uranium fuel for research or materials testing 

reactors 

2.4 Production of nuclear fuel - Manufacture of experimental nuclear fuel 

3.1 Generation of electricity by nuclear reactors - Operation of Magnox power stations 

3.2 Generation of electricity by nuclear reactors - Operation of advanced gas-cooled power 

stations 

3.3 Generation of electricity by nuclear reactors - Operation of pressurised water power stations 

4.1 Recovery of usable products from spent nuclear fuel - Reprocessing of uranium metal from 

power reactors 

4.2 Recovery of usable products from spent nuclear fuel - Reprocessing of uranium oxide fuel 

from power reactors 

4.3 Recovery of usable products from spent nuclear fuel - Reprocessing of fuel from 

research/materials testing/prototype reactors 

5.1 Production of radioisotopes - Manufacture of radioisotopes using nuclear reactors and 

accelerators 

6.1 Production of radioactive products - Manufacture of radioactive sources, substances and 

radiopharmaceuticals 

7.1 Non-destructive testing - Use of radioactive sources and substances for radiography 

8.1 Radiation processing of food - Use of gamma radiation sources to reduce bacterial levels, 

sterilise, disinfect or modify foods 

9.1 Radiation processing of products - Use of gamma radiation sources to reduce bacterial levels, 

sterilise, disinfect or modify materials 

10.1 Substance measurement and process control - Use of sealed sources for thickness gauging, 

density gauging, mass gauging, level gauging, flow measurement, borehole and well logging, control 

of pipeline crawlers 

10.2 Substance measurement and process control - Use of neutron sources for moisture gauging 

11.1 Detection and analysis - Use of sealed sources for analysis 

11.2 Detection and analysis - Use of beta sources for gas chromatography detectors 

11.3 Detection and analysis - Use of radioactive sources for leak detection, chemical and 

explosives detection 

11.4 Detection and analysis - Use of neutron sources for activation analysis 

12.1 Elimination of static electricity - Use of radioactive sources to eliminate static electricity 

13.1 Illumination - Use and repair of gaseous tritium light sources for illumination, in safety signs 

and equipment, sighting and location markers, watches and instruments 


13.2 Illumination - Use of radioluminous paint (tritium and promethium-147) for the luminising of 

timepieces and the repair of radioluminised timepieces 

14.1 Electronic apparatus - Use of electronic apparatus containing radioactive substances e.g. 

tritium in spark gap devices 

15.1 Safety devices - Use of ionising radiation in smoke and fire detectors and other safety 

instruments 

16.1 Security screening - Use of gamma rays or neutron sources to examine packages, baggage, 

containers or vehicles 

16.2 Security screening - Use of gamma rays to detect people seeking illegal entry to the UK in 

vehicles or freight 

16.3 Security screening - Use of back-scatter imaging for the detection of concealed items on the 

person 

16.4 Security screening - Use of gamma rays or neutron sources to detect concealed items in 

buildings 

[17 The Environment Agency does not register practices in category 17 since they do not involve 

radioactive materials.] 

18.1 Radioactive tracers - Use of radioactive tracers in industrial process controls 

18.2 Radioactive tracers - Use of radioactive tracers for medical or biological techniques 

18.3 Radioactive tracers - Use of radioactive tracers for environmental tests 

19.1 Diagnosis – medical - Use of ionising radiation in radiography, fluoroscopy, computed 

tomography, in-vivo nuclear medicine and in-vitro nuclear medicine 

20.1 Treatment – medical - Use of ionising radiation in interventional radiology; in-vivo nuclear 

medicine; teletheraphy; brachytherapy; radiography (for planning purposes); fluoroscopy (for planning 

purposes); computed tomography (for planning purposes) 

21.1 Occupational health screening - Use of ionising radiation in radiography and in-vitro nuclear 

medicine. 

22.1 Health screening - Use of ionising radiation in radiography and in-vitro nuclear medicine 

23.1 Medical and biomedical research - Use of ionising radiation in radiography; fluoroscopy; 

interventional radiography; computed tomography; in-vivo nuclear medicine; in-vitro nuclear medicine; 

teletherapy; brachytherapy and neutron activation analysis. 

24.1 Medico-legal procedures - Use of ionising radiation in radiography; fluoroscopy; interventional 

radiography; computed tomography and in-vivo nuclear medicine 

25.1 Diagnosis and therapy – veterinary - Use of ionising radiation in radiography, fluoroscopy, 

computed tomography, in-vivo nuclear medicine, in-vitro nuclear medicine, teletherapy and 

brachytherapy 

26.1 Teaching, including further and higher education and training - Use of radioactive sources and 

substances 

27.1 Research and development - Operation of nuclear fission or fusion reactors for R & D 

purposes 

28.1 Ionising radiation metrology - Use of all types of radiation sources to support National 

Measurement System and use of calibration sources in the testing of equipment 

29.1 Storage in transit of radioactive materials 


The numbers between 30 and 100 have been left for future use 

101.1 Use of NORM - as a chemical reagent 

101.2 Use of NORM - as a balance weight 

101.3 Use of NORM - as radiation shielding 

101.4 Use of NORM - Adventitious arising from gas and oil production 

101.5 Other uses of NORM 

NORM means naturally occurring radioactive material 

102.1 Use by MOD or the armed services 

102.2 Use for military purposes by a contractor to the MOD 

F2 Please indicate which Associated Activity(ies) are carried out and produce radioactive 

waste 

Please tick each relevant box. 

This list is intended to give the Environment Agency more detailed information about the production of 

radioactive waste. 

A Research and development 

B Manufacture 

C Repair 

D Maintenance 

E Supply 

F Assembly 

G Handling 

H Holding 

I Testing (operation and quality assurance) 

J Storage 

K Use 

L 

Decommissioning and waste disposal 

M Other Please specify 

F3 Please describe how the radioactive waste is produced. 

Radioactive waste is not produced at the premises. The premises are intended to be a final disposal facility for 

radioactive waste of low specific activity produced from various sources and primarily from the UK civil nuclear 

decommissioning programme. The premises are an existing permitted PPC hazardous waste disposal landfill. See 

Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East 

Northants Resource Management Facility, Supporting Information attached to this application for further details. 

Secondary waste/emissions could arise from the disposal facility under normal conditions through the management of 

leachate and the emission of landfill gas. Secondary waste/emissions could arise from the disposal under non-normal/ 

post closure conditions which are assessed in detail in the application supporting information. 

F4 Please attach a brief statement covering the following issues (which will be included as 

conditions if we decide to issue an authorisation): 

A statement of the existence and scope of your management system for compliance with 

authorisation requirements. 

A diagram or description of your organisational structure with respect to compliance with 

authorisation requirements. 

An indication of the resources available to achieve compliance with authorisation requirements. 

Assurance that consultation with an RPA or other qualified expert can take place when 

necessary. 

Assurance that written operating procedures are in place to cover the accumulation and disposal 

of radioactive waste. 

The arrangements for adequate supervision of disposal of radioactive waste. 

The information is provided in the application document Disposal of LLW including HV-VLLW Under the 

Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information. 


Please note that Environment Agency Officers may seek more detailed information on compliance 

with relevant authorisation conditions during determination of the application or subsequently. 

F5 Please enclose your assessment of how you plan to use best practicable means to 

minimise the disposal of radioactive waste. 

See Environment Agency guidance on BPM (in Chapter 4 of RASAG (http://www.environmentagency.gov.uk/business/444304/945840/1064273/?version=1&lang=_e) 

F6 Will the radioactive waste be produced for a limited time? 

Yes, how long? 

Indirect radioactive waste arising from leachate management and land fill gas management will arise over the operational 

period of the landfill and over the aftercare period. The current closure date for the landfill (subject to revision) is 2013 and 

the current No aftercare period extends to completion as defined in the Landfill Regulations. 

F7 Do you intend to receive and dispose of radioactive waste from another person or 

premises? 

Yes 

No 

F8 Please give details of each such person 

See Company attached information, 1 Application for Disposal of LLW including Company HV-VLLW 2 Under the Radioactive Substances 

Act 1993, for the East Northants Resource Management Facility, Supporting Information. There are no specific named 

consignors Addressof 

radioactive waste at the time of application. Consignors Addresswill 

be established after the disposal route has 

been authorised and will comprise the nuclear decommissioning industry operating under the NDA programme and 

other consignors from the UK. The waste route is intended to be open to all potential users in the same manner as the 

LLWR facility or typical hazardouswaste facilities, with quality assurance for waste receipt established through 

“conditions Postcodefor 

acceptance” derived from the authorisation requirements Postcode and established through commercial contracts. 

This Phone would include all NDA nuclear decommissioning sites, UK Phone nuclear power producing sites, commercial users of 

radioactivity, Fax hospitals, MOD, the oil/gas industry, legacy wastes, Fax other “small users” and other producers. 

E-mail E-mail 

Please use a continuation sheet if necessary 

F9 What is the chemical and physical nature of the waste you intend to receive? 

See attached information, Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances 

Act 1993, for the East Northants Resource Management Facility, Supporting Information. The waste will be solid, low 

specific activity low level radioactive waste that is (in so far as is reasonably practicable) non putrescible and that 

complies with the non-radiological acceptance criteria for the landfill based upon existing non-radiological risk 

assessments. The waste may have non-radiological properties that would be classified as inert, non-hazardous or 

hazardous were the waste not a radioactive waste. The existing landfill is a permitted hazardous waste facility. 

G Incineration on the Premises 

G1 Do you intend to use an incinerator on your own premises? 

Yes 

No please go to next section 

G2 Is there any environmental licence covering the use of your incinerator? 

Yes please give details 

No 

G3 What type of incinerator do you have? 

Please give the manufacturer and model or type number 

G4 When was the incinerator installed? 


G5 Briefly describe any gas clean-up system or filtration on your incinerator. 

H Disposal of gaseous waste 

H1 Do you intend to dispose of radioactive waste in the form of gas, mist or dust? 

Yes Please continue with the rest of this section 

No Please go to next section 

H2 How many discharge points do you intend to use to dispose of gaseous waste? 

Number of discharge points 

Please supply the information in questions H3 to H5 for each discharge point if you have more than 

one. 

H3 Identify or describe the discharge point. 

H4 List the radionuclides you intend to discharge. 

You should only include intentional and unavoidable discharges of radioactive waste that you expect 

to need to make after the application of Best Practicable Means to your processes. The Environment 

Agency does not authorise the accidental release of radioactive material. The quantities specified 

should be the maximum realistically likely within the normal range of operations. 

Radionuclide Maximum discharge 

in a single day in 

becquerels 

Maximum discharge in 

a year in becquerels 

H5 Maximum number of days in a year on which you intend to discharge 

Is the proposed annual 

discharge greater than 

one tenth of the 

relevant Pollution 

Inventory Threshold? 

(available from the 

Environment Agency’s 

web site) 


H6 How do you intend to measure or estimate the activity of the discharge? Please explain the 

methods to be used and state whether the methods are capable of demonstrating compliance with 

any proposed discharges greater than one tenth of the Pollution Inventory Threshold. 

Assessment 

H7 Please attach your radiological assessment of the proposed discharges to this form. 

You should assess the dose to the most likely exposed individual(s) who are not involved in your work 

with radioactive material or waste. For each discharge point you should give details of 

• the height of the discharge point 

• the height of the discharge point above the highest part of the nearest building 

• the discharge rate 

• details of any filtration on the discharge system 

Please give details of the calculations you use. 

I Aqueous Waste 

Accumulation of aqueous waste 

I1 Do you intend to accumulate radioactive aqueous waste? 

This includes accumulation of waste to enable short-lived radionuclides to decay. 



I2 Why do you intend to accumulate aqueous waste? 

It is not common to accumulate aqueous waste before you dispose of it. Please explain why you want 

to do it. Any proposed accumulation should be part of the BPM assessment supplied under question 

F4. 

I3 How do you intend to accumulate aqueous waste? 

Please explain what facilities and controls you will use to store the accumulated aqueous waste 

safely. 

I4 How long do you intend to accumulate aqueous waste for? 

Please give the maximum time that radioactive aqueous waste will be stored from creation or receipt 

until final disposal. 


I5 How much radioactive waste do you intend to accumulate? 

• Where one or a few radionuclides dominate the waste you should detail each of them. 

• You must detail all alpha-emitting radionuclides. 

• If you use just a few megabecquerels of similar radionuclides, you can list them as a group. 

Examples could be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will 

give you flexibility. 

Radionuclide Maximum activity in becquerels 

I6 What is the maximum volume you intend to accumulate at any one time? 

Disposal of aqueous waste 

Cubic metres 

I7 Do you intend to dispose of radioactive aqueous waste? 



I8 What is the chemical and physical nature of the waste you intend to dispose of? 

The aqueous waste will be leachate collected from the operating landfill and will have the 

chemical properties typically associated with landfill leachate from a hazardous waste site. The 

leachate could conceivably contain leached radioactivity. 

I9 How do you intend to measure or estimate the activity of the discharge? Please explain 

The maximum activity of the discharge has been estimated for risk assessment purposes in Application for “ 

Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants 

Resource Management Facility, Supporting Information. The estimate is based upon conservative assumptions 

for overall risk assessment purposes. The actual amount of the discharge is uncertain and will depend on the 

actual amount of waste disposed in the void at any one time and the fraction of the inventory which transfers to 

the leachate. The activity in the leachate will be monitored. 

I10 Where will you dispose of the radioactive aqueous waste? 

Please tick all that apply and answer the questions in the relevant section. 

to a public sewer 

direct to a watercourse or water body 

to your premises’ own sewage treatment works 

other method(s) Leachate is currently tankered to an offsite treatment works 

Disposal to a public sewer 

I11 What is the name of your sewerage undertaker? 

I12 What is the OS national grid reference of the sewage treatment works discharge point? 



I13 What is the total monthly volume of water which you intend to discharge from the premises 

into the sewer? 

Zero 

Cubic metres 

I14 What is the maximum monthly total of each radionuclide which you intend to discharge? 




Examples 

could be ‘Tritium/Carbon 14’ or ‘total other radionuclides excluding alpha emitters’. This will give you 

flexibility. 

Radionuclide Maximum total activity in any single month 

In becquerels 

Discharge limits for leachate are indicated within Application for Disposal of LLW including 

HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource 

Management Facility, Supporting Information and are subject to agreement. 

I15 Please attach your radiological assessment of the proposed discharge to this form. 


with radioactive material or waste. Please give details of the calculations you use. 

Disposal direct to a watercourse or water body 

I16 What is the name of the watercourse or body of water that you intend to discharge into? 

I17 Is the body of water a pond or lake? 

Yes 

No 

I18 What is the OS national grid reference of the discharge point? 


I19 What is the maximum volume of water you intend to discharge from the premises in a 

month? 

Cubic metres 



In becquerels 




Disposal to a sewage treatment works on the premises 


I22 What is the name of the watercourse or body of water that your sewage treatment works 

discharges into? 

I23 What is the OS national grid reference of your sewage treatment works discharge point? 


I24 What is the total monthly volume of water which you intend to discharge from your sewage 

treatment works? 

Cubic metres 



In becquerels 

I26 What do you intend to do with any sludge or solids which are left after treatment? 

I27 How do you plan to assess the activity of any sludge or solids which are left after treatment 

before final disposal? 




Disposal of aqueous waste by other methods 

I29 Please give details of the method on a separate sheet and attach it to this form, including 

• a description of the type and quantity of radioactive waste 

• a description of the disposal route, including water and residual solids 

• a description of the measurement methods for the radioactivity 

• a brief summary of any agreement with a contractor and attach it to this form 

• your radiological assessment of the proposed discharge to this form. 




J Organic Liquid Waste 

Accumulation of organic liquid waste 

J1 Do you intend to accumulate radioactive organic liquid waste? 



J2 How do you intend to accumulate organic liquid waste? 

Please include details of measures and controls used to help keep the waste safe, for example 

security, fire precautions and alarms etc. 

J3 How long do you intend to accumulate organic liquid waste for? 

Please give the maximum time that radioactive organic liquid waste will be stored from creation or 

receipt until final disposal. 

J4 How will you measure the activity of the organic liquid waste? 

J5 How much radioactive waste do you intend to accumulate? 




Examples 


flexibility. 


J6 What is the maximum volume you intend to accumulate at any one time? 

Cubic metres 

Disposal of organic liquid waste 

J7 Do you intend to dispose of organic liquid waste? 



J8 What is the chemical and physical nature of the waste that you intend to dispose of? 

J9 How do you intend to dispose of organic liquid waste? 


Please tick all that apply and answer the relevant questions. 

incineration on the premises 

transfer to a contractor 

by other means 

Incineration of organic liquid on the premises 

J10 What is the maximum daily and monthly activity of each radionuclide which you intend to 

incinerate? 




Examples 


flexibility. 


J11 What is the maximum volume you intend to dispose of in the following periods? 

Day Month 

metres 

Cubic 

metres 

Cubic 

J12 How do you intend to assess the activity content of the ash from the incinerator or solids 

from any filtration system? 

J13 How do you intend to dispose of ash from the incinerator or solids from any filtration 

system? 

J14 What will you do if your incinerator fails or breaks down? 

J15 Please attach your radiological assessment of the proposed disposal to this form. 

You should assess the dose to the most likely exposed individual(s) who are not involved in the work 

with the radioactive material or waste. 

You should give details of 

• the height of the incinerator discharge point 

• the height of the discharge point above the highest point of the nearest building 

• the discharge rate 

• details of any filtration on the incinerator 



Transfer to a contractor 

Please provide relevant details for each contractor if you want more than one on the authorisation. 

J16 Please attach a brief summary of your agreement with the contractor to this form. 

J17 How much radioactive waste do you intend to transfer to your contractor? 




Examples 


flexibility. 

Radionuclide Maximum annual activity 

In becquerels 

J18 What is the maximum volume you intend to dispose of in any one year? 

Cubic metres 

J19 What is the company name of the contractor? 

J20 What is the address of the contractor’s site which will receive the waste? 

Address 

Postcode 

Phone 

Fax 

E-mail 

J21 In which County, borough, district or unitary authority areas is the contractor’s premises? 

J22 Please describe contingency arrangements if your planned transfer routes become 

unavailable. 

For example failure of contractor’s incinerator 

Disposal of organic liquid waste by other means 

J23 Please describe any other method you intend to use to dispose of liquid organic waste on 

a separate sheet. Attach your description to this form. 


K Very Low Level Solid Waste 

Please contact us if you wish to dispose of alpha-emitting radionuclides via this route. 

K1 Do you intend to accumulate or dispose of very low level solid waste? 



The application is for the premises to be a disposal facility for controlled burial wastes which would include HV-VLLW wastes. See Application for Disposal of 

LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information. 

K2 What is the chemical and physical nature of the waste? 

K3 What categories of very low level waste do you intend to accumulate or dispose of? 

VLLW Category 1 waste in which 

• there are no alpha-emitting radionuclides 

• the sum of all radionuclides in any 0.1 cubic metre of refuse is less than 400kBq and 

less than 40kBq in any one article 

VLLW Category 2 (higher limits for Tritium and Carbon 14) waste in which 

• the sum of all Tritium and Carbon 14 in any 0.1 cubic metre of refuse is less than 4 

MBq and less than 400 kBq in any one article 

• there are no other radionuclides 

K4 If you are seeking category 2 please tell us why you need these higher limits 

K5 How will you measure or assess the activity of the waste? 

Accumulation of very low level solid waste 

K6 Do you intend to accumulate very low level solid waste? 



K7 How much very low level waste do you intend to accumulate at any one time? 

metres 

Cubic 

K8 How long do you intend to accumulate the waste before you dispose of it? 

The usual time is two weeks 

Weeks 

Where the accumulation time is longer than two weeks please tell us why you need the extra time 

K9 How will you store the accumulated very low level waste until it is disposed of? 

Please give details of measures and controls used to help keep the waste safe, for example security, 

fire precautions and alarms, etc. 


Disposal of very low level solid waste 

K10 What is the maximum amount of very low level solid waste you intend to dispose of with 

normal refuse in any one month? 

Cubic metres 

K11 How do you intend to dispose of very low level solid waste? 

Landfill at a site under your control 

Collection by a Local Authority or its contractor 

Transfer to another contractor for landfill 

L Solid Waste (excluding Sealed Sources) 

L1 Do you intend to accumulate or dispose of solid waste? 

Please note that solid waste in the form of sealed sources is covered in Annex S to this form - please 

complete that Annex if you need an authorisation to accumulate or dispose of sealed sources. Do not 

include waste which can be disposed of without an authorisation under the terms of an exemption 

order. Waste accumulated to enable short-lived radionuclides to decay should be included. 


No Please go to the next section 

Accumulation of solid waste 

L2 Do you intend to accumulate solid waste? 

This includes accumulation of waste to enable short-lived radionuclides to decay. 



L3 What is the chemical and physical nature of the waste? 

L4 How much radioactive waste do you intend to store? 






Radionuclide Maximum activity in becquerels Maximum Time of 

Accumulation 

L5 How much waste do you intend to accumulate at any one time? 

Cubic metres 

L6 Why are you suggesting this time period(s) for accumulating the waste? 


L7 How will you record and label this solid waste? 

Disposal of solid waste 

L8 Do you intend to dispose of solid waste? 


Application No for Please Disposal go of to LLW next including section HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants 

Resource Management Facility, Supporting Information. 

L9 How do you intend to dispose of solid waste? 

Please tick all that apply and answer the relevant sections below. 

incineration on the premises 

transfer to a person authorised to receive them 

transfer to a manufacturer or supplier of similar sources 

transfer to Drigg or Sellafield sites 

controlled disposal on premises 

Application by for other Disposal means of LLW Please including describe HV-VLLW any other Under method the Radioactive you intend Substances to use to Act dispose 1993, of for solid the East waste Northants 

Resource on a separate Management sheet Facility, and attach Supporting it to this Information. form. You must give relevant details. 

Incineration on the premises 

L10 What is the maximum daily and monthly activity of each radionuclide which you intend to 

incinerate? 

Where one or a few radionuclides dominate the waste you should detail each of them. 

You must detail all alpha-emitting radionuclides. 

If you use just a few megabecquerels of similar radionuclides, you can list them as a group. Examples 


flexibility. 

You must indicate which are sealed sources and if any of those are high-activity sealed sources. 

Radionuclide Maximum discharge 

in a single day 

in becquerels 

Maximum discharge in a month 

In becquerels 

L11 How much radioactive solid waste do you intend to incinerate each month? 

Cubic metres 


L13 How do you intend to assess the activity in the ash from the incinerator and solids from 

any filtration system? 


1 

0 Solid waste continued 

L14 How do you intend to dispose of ash from the incinerator and solids from any filtration 

system? 

L15 What will you do if your incinerator fails or breaks down? 

L16 Please attach your radiological assessment of the proposed disposal to this form. 


with radioactive material or waste. You should give details of: 

the height of the incinerator discharge point 

the height of the discharge point above the highest point of the nearest building 

the discharge rate 

details of any filtration on the incinerator 


Transfer to a person authorised under RSA 93 to receive them or a manufacturer or supplier of 

similar sealed sources 

For the purposes of this form the person who will be receiving the waste is referred to as “the 

contractor”. 

Give full separate details for each contractor 

L17 How much radioactive waste do you intend to transfer to the contractor? 







In becquerels 

L18 What is the name of the company or organisation which will receive your solid waste? 

L19 What is the address of the company or organisation which will receive your solid waste? 

Postcode 

Contact numbers and e-mail 


Phone 

Fax 

E-mail 

L20 What is the address of the site where solid waste will be sent (if different)? 

Postcode 


Phone 

Fax 

E-mail 

L21 What is the National Grid Reference Number of the site where solid waste will be sent? 

L22 Is the site where solid waste will be sent on a nuclear licensed site (except LLWR Drigg or 

Sellafield)? 

Yes 

No 

L23 What is the recipient’s Environment Agency authorisation number for the site where solid 

waste will be sent, if known? Not needed for nuclear sites 

L24 In which borough, district or unitary authority area is the site where solid waste will be 

sent? 



Please give the county council unless there is a unitary authority 

L25 Please attach a brief summary of your agreement with any relevant contractor 

L26 Please describe contingency arrangements if your planned contractor is unavailable. 

Transfer to Low Level Waste Repository (LLWR) Drigg or Sellafield sites 

Please attach a brief summary of your agreement with the site operator to this form. 

L27 Will any consignment of waste contain alpha emitting radionuclides in excess of 4 

gigabecquerels per tonne or all other radionuclides in excess of 12 gigabecquerels per tonne? 

Yes 

No 



L29 What is the maximum annual disposal activity (at the time of transfer) for each of the 

following? in becquerels 

Uranium 

Radium 226 plus Thorium 232 

Other alpha emitters 

Carbon 14 

Iodine 129 

Tritium 

Cobalt 60 

Other beta-emitting radionuclides (half-life greater than 3 

months) 

Other beta-emitting radionuclides (half-life less than 3 months) 

L30 What is the maximum amount of waste you plan to send to the site operator at LLWR 

Drigg or Sellafield in any one year? 

Cubic metre 

L31 How many consignments are intended for BNFL at Drigg or Sellafield in a year? 

Controlled Burial 

Please attach a brief summary of your agreement with the site operator to this form. 

L32 How much radioactive waste do you intend to bury at the operator’s disposal site? 


• You Please must detail refer to all the alpha-emitting supporting radionuclides. 

application document for details. 


Examples 


flexibility. 

Radionuclide Maximum activity in Concentration 

any one month Bq per cubic metre 

In becquerels 


The waste will be solid, low specific activity low level radioactive waste that is (in so far as is reasonably practicable) non putrescible 

and that complies with the non-radiological acceptance criteria for the landfill based upon existing non-radiological risk assessments. 

The waste may have non-radiological properties that would be classified as inert, non-hazardous or hazardous were the waste not a 

radioactive waste. The existing landfill is a permitted hazardous waste facility. 

L34 What is the name of the company or organisation which will receive your solid waste? 

Within our own organisation 

L35 What is the address of the company or organisation which will receive your solid waste? 

Within our own organisation 


Postcode 


Phone 

Fax 

E-mail 

L36 What is the address of the site where solid waste will be sent (if different)? 

Postcode 


Phone 

Fax 

E-mail 

L37 What is the National Grid Reference Number of the site where solid waste will be sent? 

TF 010 000 

L38 What is the recipient’s Environment Agency authorisation number for the site where solid 

waste will be sent, if known? 

This authorisation application applies 

L39 In which borough, district or unitary authority area is the site where solid waste will be 

sent? 



East Northamptonshire Council 


Northamptonshire County Council 

L40 Please attach a brief summary of your agreement with any relevant contractor 

Application for Disposal of LLW including HV-VLLW Under the Radioactive Substances Act 1993, for the East Northants Resource Management Facility, Supporting Information. 

 

L41 Please describe contingency arrangements if your planned contractor is unavailable. 

n/a 

L42 Please attach your radiological assessment of the proposed disposal to this form. 

Application You should for assess Disposal the of dose LLW to including the most HV-VLLW likely exposed Under the individual(s) Radioactive who Substances are not Act involved 1993, for in the your East work 

Northants with radioactive Resource material Management or waste. Facility, You Supporting should give Information. details of 

• the disposal arrangements at the disposal site 

• the type and approximate depth of the overlying material 

• any measurable radiation dose rates from the closed containers holding the waste. 

M NAIR Arrangements 

M1 Do you have an emergency role as a participant under the National Arrangements for 

Incidents involving Radioactivity (NAIR)? 




M2 Do you wish to have the standard conditions for NAIR participants to be entered into this 

authorisation? 

This allows you to accumulate waste arising from your participation in the NAIR scheme. 

Yes 

No 

M3 Do you have a separate current Variation Notice for the accumulation and disposal of NAIR 

waste? 

Yes What is the reference number of the Notice? 

No 

N Checklist 

This section is to help you check that you have 

• completed the correct parts of the form ( ) 

• attached the right documents to help us process your application quickly ( ). 

Company or organisation details 

Type of application 

About the application 

About the premises 

Contact details 

Producing radioactive waste 

Gaseous waste 

Disposal of gaseous waste 

Discharge point description(s) 

Radiological assessment of discharge 

Aqueous waste 

Accumulation of aqueous waste 

Disposal of aqueous waste 

Disposal to a public sewer 


Disposal direct to a watercourse or water body 


Disposal to a sewage treatment works on the premises 


Disposal of aqueous waste by other methods 


Organic liquid waste 

Accumulation of organic liquid waste 

Disposal of organic liquid waste 




Disposal of organic liquid waste by other means 

Description of method 

Radiological assessment of disposal 

Very low level solid waste 

Accumulation of very low level solid waste 

Disposal of very low level solid waste 

Solid waste 

Accumulation of solid waste 

Disposal of solid waste 





Transfer to LLWR Drigg or Sellafield sites 

Controlled burial 


Other methods of solid waste disposal 

Description of method 

NAIR arrangements 

Data handling 

Claim of confidentiality 

National security direction 

Payment for your application 

Cheque 

Declaration 

O Data Handling 

O1 Commercial in confidence 

Is there any information in the application which you believe should be restricted on the 

grounds that the information relates to a “relevant process” or trade secret? 

“Relevant process” means any process applied for the purposes of, or in connection with, the 

production or use of radioactive material. 

Yes Please describe the information and explain why you believe it should be restricted 

 

No 

O2 National security 

Is there any information in the application which you believe should be restricted on the 

grounds of national security? 

Yes 

Please enclose a copy of any request for a Direction which you have made to the Secretary of State 

or National Assembly for Wales. The Environment Agency already holds a Direction requiring it to 

ensure that no information relating to sealed source applications/registrations is to be included in 

public registers. Pursuant to Section 25(3)(b) of RSA 93 no such information will be sent to Local 

Authorities. Nor will we send similar information relating to accumulation or disposal of radioactive 

waste to Local Authorities. 

No 

O3 Data Protection Notice 

The Environment Agency is responsible for regulating environmental protection, flood defence, water 

resources and fisheries. It has a duty to discharge its functions to protect and enhance the 

environment and to promote conservation and recreation. 

The information provided will be processed by the Environment Agency to deal with your application, 

to monitor compliance with the licence/permit/registration conditions and to process renewals. 

We may also process and/or disclose it in connection with the following: 

• offering/providing you with our literature/services relating to environmental matters. 

• consulting with the public, public bodies and other organisations (eg Health and Safety Executive, 

local authorities, emergency services, DEFRA on environmental issues) 

• carrying out statistical analysis, research and development on environmental issues 

• providing public register information to enquirers 

• investigating possible breaches of environmental law and taking any resulting action 

• preventing breaches of environmental law 

• assessing customer service satisfaction and improving our service. 

• reporting to the European Commission on the experience gained in implementing Council Directive 

2003/122/Euratom. 

• exchanging information and co-operation with European Union Member States, third countries or 

relevant international organisations. 



Annex S - Sealed Sources 

Accumulation and Disposal of Waste Sealed Sources 

S1 Do you intend to accumulate or dispose of waste sealed sources? 

This does not include waste sealed sources which can be accumulated or disposed of under an 

exemption order without an authorisation. You will be responsible for complying with the conditions of 

any such exemption order. It does include accumulation of sealed sources to enable short-lived 

radionuclides to decay. 

Yes Please continue with this Annex 

No Please do not complete this Annex 

Accumulation of Waste Sealed Sources 

If your premises is located on a nuclear licensed site please do not complete Questions S2 to S** 

S2 Do you intend to accumulate waste sealed sources on the authorised premises? 

This includes accumulation of sources to enable short-lived radionuclides to decay. 

Yes Please continue with this Annex 

No Please describe how you dispose of sources without accumulating them. Then go to 

question ** 

S3 What waste sealed sources do you intend to store at any one time? 

Including those covered by any existing authorisatons, but excluding sources exempt from 


In order, starting with the highest activity material and finishing with the lowest activity material. 

If you intend to accumulate several sources of the same radionuclide with approximately the same 

activity you can describe them together in a single line in the table below. Refer to the maximum 

activity of an individual source. (For example, Caesium-137, three sources, maximum activity for each 

100 Megabecquerels would cover sources of 75, 85, and 95 megabecquerels activity). 

You do not need to include radionuclides which are present as a result of radioactive decay of the 

listed radionuclides. 

You may apply for the maximum number of sources that you reasonably expect to accumulate in the 

foreseeable future (ie the next 1-2 years). 

If you want to accumulate large numbers of relatively small sources, you can opt to authorise them as 

a group. (For example, beta/gamma emitting radionuclides, alpha emitting radionuclides.) However, it 

will help us process your application if you provide as much information as possible about the 

proposed individual radionuclides you intend to accumulate. If you do this the maximum activity of any 

single source must not exceed the HASS threshold (see Environment Agency HASS guidance annex) 

for that radionuclide. 

Using becquerels 

You should list activity in SI units (becquerels). Write the prefix kilo, mega, giga, tera or peta clearly 

(in full) to minimise the risk of error. 

Rounding up substances of nominal activity 

If you accumulate radioactive substances of nominal activity (particularly with radionuclides of short 

half life), you may round up the figure to ensure you do not risk exceeding your authorised limit. If you 

do round up a figure, please make sure you say how and where you have done this. 

Depleted uranium 

You should be aware that some sources may be supplied in depleted uranium containers. Where 

necessary you should give the masses for depleted uranium (for example, in source containers, 

counterbalance weights) in kilogrammes. 

Radionuclide Maximum 

activity in 

Becquerels 

Maximum time of 

accumulation 

Is the source a 

“new HASS”, 

existing HASS or 

other type? * 

Number of 

Waste 

Sources of 

each type 


*Note – These terms are explained in the Environment Agency’s interim guidance on HASS. 

(http://www.environment-agency.gov.uk/commondata/acrobat/hass_guidance_1155126.pdf). 

Please put in new, existing or other as appropriate 

S4 Why are you suggesting this time period(s) for accumulating waste sealed sources? 

S5 How will you record and label the waste sealed sources? 

S6 How will you store the accumulated waste sealed sources until they are disposed of? 

Security of Sources 

The Environment Agency now has regulatory powers over protective security of certain waste sealed 

sources. Consideration of security is required for waste high-activity sources and for other sources 

which, in the opinion of the Environment Agency, constitute a similar level of potential hazard. See the 

Environment Agency's Interim Guidance on HASS. 

(http://www.environment-agency.gov.uk/commondata/acrobat/hass_guidance_1155126.pdf) . It is our 

opinion that any source, or aggregation of sources in a single premises, which falls in any of source 

categories 1 to 4 in the scheme set out in the NSAC Document constitutes a similar level of potential 

hazard to a HASS. All users, applicants and other interested parties who need to see the NSAC 

Document should ask their Police Force Counter Terrorism Security Adviser for a copy. 

Where sources are not considered to constitute a similar level of potential hazard to that from highactivity 

sources, the Environment Agency will be requiring users to take simple precautions to protect 

them. 

S7 Please provide the following details of the maximum holding of waste sealed sources at 

any time 

Building or Facility 

name or number 

Radionuclide(s) 

and Practice(s) from 

Table 1 of NSAC 

Document * 

Maximum total 

activity of each 

radionuclide 

GBq 

Source 

Category 

(1- 5) * 

Security 

Group 

(A – D) * 


* Note – Security Requirements for Radioactive Sources (October 2005), NSAC 

The NSAC Document (see Note above) describes how to calculate the category and relevant security 

group. You should do this on the basis of aggregating all the sealed sources that may be held in a 

single building on the premises at any one time (include both registered and waste sealed sources). 

If you hold HASS or sources of similar level of potential hazard, then we will need to consider whether 

or not all of your sources are vulnerable to the same threat, and our assessment of security group 

may differ from your initial one. If this means that you need additional security measures, we will give 

you the opportunity to amend your application. If your assessment of the category of your sources 

indicates that you need significant expenditure to meet the requirements of the NSAC Document, or if 

your sources are distributed around more than one building on the premises, you may consider 

discussing your situation with your local Police Counter Terrorism Security Adviser before completing 

this form. 

S8 Please confirm that you hold a copy of NSAC Document “Security Requirements for Sites 

and Sectors working with Radioactive Substances”, October 2005 and that you understand its 

requirements 

This is available from your local Police Counter Terrorism Security Adviser 

Yes 

No 

S9 If you consider your premises to be in Security Groups A, B or C, have you met all of the 

requirements of the NSAC Document for the security group you consider your premises to be? 

Yes 

No 

S10 Please indicate if you have the following security measures in place for the sources both 

when they are in use and when they are being stored while not being used and waste sources. 

General measures to prevent loss of the sources, ie care of sources 

Physical security: 

Fence 

Building 

Room 

Store 

Security provided by source container 

Access control 

Storage of information and databases 

Security of essential utilities (eg. electricity) 

Site procedures and security plan to: 

Prevent unauthorised access to or loss or theft of the sources 

Detect unauthorised access to or loss or theft of the sources 

Include options to upgrade the site security plan in response to increased threat 

Information security plan covering: 

Unauthorised access to information on the sources 

Unauthorised access to information on the security measures taken 

Personnel checks 

Detection: 

Patrols 

Alarms 

CCTV 

Response to detection: 

Local 

Police 


Documentary evidence of measures taken 

Other measures 

The NSAC Document specifies how to determine which of these security features are required at 

premises in Security Groups A, B or C. 

YOU SHOULD NOT INCLUDE DETAILS OF YOUR SECURITY MEASURES WITH YOUR 

APPLICATION. 

High-Activity Sealed Sources 

See the Environment Agency's Interim Guidance on HASS 

(http://www.environment-agency.gov.uk/commondata/acrobat/hass_guidance_1155126.pdf). 

You need to complete these questions if you intend to accumulate or dispose of a “new HASS”, or 

both new and existing HASS, under the terms of an authorisation under RSA 93. You do not need to 

complete them if you can dispose of your waste HASS without authorisation under the terms of an 

exemption order. 

S11 Please confirm you have read the requirements of the Defra guidance on financial and 

other provision (High-activity Sealed Radioactive Sources and Orphan Sources Directive (Council 

Directive 2003/122/Euratom) Guidance to the Environment Agency) for each waste high-activity 

sealed source you intend to accumulate. 

Yes 

No 

S12 Which mechanism are you proposing to use for this purpose? 

You will need to include with the application, sufficient documentation to enable the Environment 

Agency to assess whether your proposed provision is adequate. 

Disposal of waste sealed sources 

S13 Do you intend to dispose of waste sealed sources? 


No Please describe what happens to the sources after accumulation. Then leave the 

remaining questions blank 

Method of Disposal of Waste Sealed Sources 

S14 How do you intend to dispose of waste sealed sources? 

Please tick all that apply and answer relevant questions below in addition to the following general 

questions 

1 transfer to a person authorised under section 13 of RSA 93 to receive them 

2 transfer to a manufacturer or supplier of similar sources 

3 transfer to a nuclear licensed site except LLWR at Drigg 

4 controlled burial 

5 transfer to LLWR at Drigg 

6 disposal with local authority refuse (in the form of VLLW) 

7 by other means Please specify and attach full details 


Please give the relevant details for each route to be used. 

S15 What waste sealed sources do you intend to dispose of? 

Include those covered by any existing authorisatons, but excluding sources exempt from 

authorisation. Please list them in order, starting with the highest activity material and finishing with the 

lowest activity material. 

If you intend to dispose of several sources of the same radionuclide with approximately the same 

activity you can describe them together in a single line in the table below. Refer to the maximum 

activity of an individual source. (For example, Caesium-137, three sources, maximum activity for each 

100 Megabecquerels would cover sources of 75, 85, and 95 megabecquerels activity.) 

You do not need to include radionuclides which are present as a result of radioactive decay of the 

listed radionuclides. 

You may apply for the maximum number of sources that you reasonably expect to dispose of in the 

foreseeable future (ie the next 1-2 years). 

If you want to dispose of large numbers of relatively small sources, you can opt to authorise them as a 

group. (For example, beta/gamma emitting radionuclides, alpha emitting radionuclides.) However, it 

will help us process your application if you provide as much information as possible about the 

proposed individual radionuclides you intend to dispose of. If you do this the maximum activity of any 

single source must not exceed the HASS threshold (see Environment Agency HASS guidance annex) 

for that radionuclide. You must detail all alpha-emitting radionuclides. 

Using becquerels 

You should list activity in SI units (becquerels). Write the prefix kilo, mega, giga, tera or peta clearly 

(in full) to minimise the risk of error. 

Rounding up substances of nominal activity 

If you dispose of radioactive substances of nominal activity (particularly with radionuclides of short 

half life), you may round up the figure to ensure you do not risk exceeding your authorised limit. If you 

do round up a figure, please make sure you say how and where you have done this. 

Depleted uranium 

You should be aware that some sources may be supplied in depleted uranium containers. Where 

necessary you should give the masses for depleted uranium (for example, in source containers, 

counterbalance weights) in kilogrammes. 


in becquerels 

Disposal by Transfer (Methods 1, 2 or 3 above) 

S16 What is the name of the company or organisation whch will receive your waste sealed 

sources? 

S17 What is the address of the company or organisation which will receive your waste sealed 

sources? 


Phone 

Fax 

Postcode 


E-mail 

S18 What is the address of the site where waste sealed sources will be sent (if different)? 


Phone 

Fax 

E-mail 

Postcode 

S19 What is the National Grid Reference Number of the site where sources will be sent? 

S20 What is the recipient’s Environment Agency authorisation number for the site where the 

sources will be sent? If known 

S21 In which County, borough, district or unitary authority area is the site where waste sealed 

sources will be sent? 




S22 Please attach a brief summary of your agreement with any relevant contractor 

S23 Please describe contingency arrangements if your planned contractor is unavailable. 

Disposal of Waste Sealed Sources to LLWR (Low Level Waste Repository) at Drigg 

You should answer these questions if you intend to dispose of sealed sources to LLWR at Drigg. 

S24 Will any consignment of waste sealed sources transferred to the Site Operator of the 

LLWR at Drigg contain alpha emitting radionuclides in excess of 4 gigabecquerels per tonne 

or all other radionuclides in excess of 12 gigabecquerels per tonne? 

Yes 

No 

S25 What is the maximum annual disposal activity (at the time of transfer) for each of the 

following? in becquerels 

Uranium 

Radium 226 plus Thorium 232 

Other alpha emitters 

Carbon 14 

Iodine 129 

Tritium 

Cobalt 60 


Other beta-emitting radionuclides (half-life greater than 3 

months) 

Other beta-emitting radionuclides (half-life less than 3 months) 

S26 What is the maximum amount of waste you plan to send to LLWR at Drigg in any one 

year? 

metres 

Cubic 

S27 How many consignments are intended for LLWR at Drigg in a year? 

Disposal of Sealed Sources by Controlled Burial (Method 4) 

You should answer these questions if you intend to dispose of sealed sources by controlled burial. 

S28 What is the name of the company or organisation whch will receive your waste sealed 

sources? 

S29 What is the address of the company or organisation which will receive your waste sealed 

sources? 


Phone 

Fax 

E-mail 

Postcode 

S30 What is the address of the site where waste sealed sources will be sent (if different)? 


Phone 

Fax 

E-mail 

Postcode 

S31 What is the National Grid Reference Number of the site where sources will be sent? 

S32 What is the recipient’s Environment Agency authorisation number for the site where the 

sources will be sent? If known 

S33 In which County, borough, district or unitary authority area is the site where waste sealed 

sources will be sent? 





S34 Please attach a brief summary of your agreement with any relevant contractor 

S35 Please describe contingency arrangements if your planned contractor is unavailable. 

S36 What is the maximum volume in any one year to be sent for burial? 

metres 

Cubic 

S37 Please attach your radiological assessment of the proposed disposal to this form. 

You should assess the dose to the most likely exposed individual(s) who are not involved in the work 

with the radioactive material. You should give details of 

• the disposal arrangements at the disposal site 

• the type and approximate depth of the overlying material 

• any measurable radiation dose rates from the closed containers holding the waste. 

Disposal of Sealed Sources in VLLW (Method 6) 

S38 What categories of very low level waste do you intend to accumulate or dispose of? 

VLLW Category 1 waste in which 

• there are no alpha-emitting radionuclides 

• the sum of all radionuclides in any 0.1 cubic metre of refuse is less than 400kBq and 

less than 40kBq in any one article 

VLLW Category 2 (higher limits for Tritium and Carbon 14) waste in which 

• the sum of all Tritium and Carbon 14 in any 0.1 cubic metre of refuse is less than 4 

MBq and less than 400 kBq in any one article 

• there are no other radionuclides 

S39 If you are seeking category 2 please tell us why you need these higher limits 

S40 How do you intend to dispose of very low level solid waste? 

Landfill at a site under your control 

Collection by a Local Authority or its contractor 

Transfer to another contractor for landfill 


Annex G 

Example Capacity Calculation Layout 


Example Capacity Calculation Layout 

The following table is a possible layout for a spreadsheet to administer the 

radiological capacity of the landfill. A table is required for each of the 

scenarios that could be restrictive to landfill capacity and the conditions noted 

in the table must be satisfied for all the scenarios in order for the landfill to 

have remaining capacity. 

The table forecasts remaining capacity on the assumption that the “fingerprint” 

(radionuclide distribution) of the wastestream to be received is the same as 

that received to date. By changing the “current inventory” to include proposed 

shipments or hypothetical forecast future waste arisings the table will forecast 

the remaining capacity (if any) based on the resulting “overall” fingerprint of 

the waste received to date and to be received in the future. In a finalised 

version additional columns could be added to archive the current inventory at 

any particular date and consider proposed shipments separately. Additional 

features to codify particular shipments and their final location in the landfill 

could also be added. 

If the “fingerprint” of waste to be shipped over the life of the facility were 

known in advance the table would forecast the capacity for each nuclide and 

the overall capacity. However, since the radionuclide distribution is not known 

that far in advance (it is known prior to shipment) the table enables ongoing 

optimisation. 


Example Radiological Capacity Table for Scenario “X” 

Qi / Qi,l 

Qi,l (MBq) 

RCi - Qi 

fi / ( SDi fi) x DC 

SDi x fi 

fi 

Current 

Radiological 

Inventory of 

the Landfill 

Specific Dose 

SDi 

Type of 

Radionuclide 

Qi (MBq) is the 

actual activity of 

radionuclide Rni 

disposed 

The activity 

limit for 

radionuclide 

Rni if it were 

the only 

radionuclide to 

be disposed 

of. 

(MBq) 

The fraction 

of the 

overall 

activity 

arising from 

Rni (such 

that fi=1) 

Rni 

RCi (MBq) 

The remaining 

radiological 

capacity for each 

nuclide Rni based 

on the waste 

stream received to 

date represented 

by fi. 

Qi 

(microSv yr -1 

per MBq) 

For example Am- 

241 

The radiological 

capacity for 


(MBq) 

Scenario specific 

value obtained 

from Annex B 

Qi,l (MBq) is the 

activity limit for 


if it were the 

only 

radionuclide to 

be disposed of. 

Where DC is the dose 

constraint (microSv/yr) which 

is specific to the scenario: 

= DC / SDi 

Numbers must remain 

>0. 

DC = 20 microSv/yr or 3000 

microSv/yr for intrusion 

scenarios 

This is all waste 

received to date 

and could 

include a future 

amount 

proposed to be 

received to test 

the remaining 

capacity is 

adequate for 

that shipment 

Rni SDi Qi fi = Qi / Qi SDi x fi RCi = fi / ( SDi fi) x DC RCi - Qi Qi,l = DC / SDi = Qi / Qi,l 







= (Qi / Qi,) 

= (RCi - Qi) 

Totals = Qi = fi= 1 = (SDi fi ) = RCi 

Must be 0.

Annex H 

Calculation of dose rate at landfill, 

TSG(09)0488 


Technical Services Group 


Reference: TSG(09)0488 

Issue: Issue 2 

Date: 15 th July 2009 

CALCULATION OF DOSE RATE AT LANDFILL IN SUPPORT OF A 

LOW LEVEL WASTE DISPOSAL AUTHORISATION 

UK-10497 

SUMMARY 

Dose rate calculations were performed in MicroShield to support a low level waste disposal 

authorisation. An estimate of the dose rate at the landfill site was calculated based on 

lightly-contaminated rubble being covered by a 30 cm layer of soil material. Dose was found 

to dependant on the soil material density and largely independent on the distance from the 

source. 

Prepared By: 

Checked By: 

Approved By: 

Name and Organisation Signature Date 

Tony Lansdell 

TSG 

Barry Cook 

TSG 

Gráinne Carpenter 

TSG 

ELECTRONIC 

COPY 



Table of Contents 

1 Introduction.......................................................................................................................3 

2 Methodology.....................................................................................................................3 

2.1 Background ..............................................................................................................3 

2.2 Case details..............................................................................................................3 

2.3 MicroShield calculation details, uncertainties and assumptions...............................4 

3 Results..............................................................................................................................5 

3.1 Cobalt-60 case .........................................................................................................5 

3.2 Caesium-137 case....................................................................................................5 

3.3 Covering soil material thickness ...............................................................................6 

4 References .......................................................................................................................7 

Appendix A...............................................................................................................................8 


2

1 INTRODUCTION 


Dose rate calculations were required to support a low level waste disposal 

authorisation. Cases were run using MicroShield v7.02 [1], to determine the dose rate 

above the layer of lightly contaminated soil at a landfill waste disposal site. 

2 METHODOLOGY 


MicroShield was used to determine the maximum resulting dose rate from disposal of 

soil and rubble to a landfill site. This can be assumed to be uniformly contaminated to 

200 Bq/g of either 60 Co or 137 Cs. An infinite slab of contaminated soil and rubble was 

assumed to be covered with 30 cm of uncontaminated soil material, and the dose rate 

assessed. 

2.2 Case details 

MicroShield was used to model a slab of waste, infinite in horizontal extent, and 100 

cm thick. Preliminary study found that if the slab thickness was increased above 50 cm 

thick, the resulting dose rate was effectively unchanged, and hence a thickness of 100 

cm was used to be conservative. 

Preliminary studies also indicated that when dose rate was determined on contact, 1m 

and 2m above the shielding soil layer, dose rate was independent of dose point height 

and so only the contact dose was reported. This work can be found in Appendix A. 

It has been outlined that MicroShield did not correctly include the effects of build-up 

(scattered flux) when using the infinite slab geometry, and that the calculation was 

instead performed using a (finite) rectangular slab that was chosen to have a very large 

extent such that it was effectively infinite. The extent was chosen such that the results 

were unchanged with further increases in size, and it was found that beyond 200 cm in 

width the dose rate on contact was effectively constant, and 1000 cm was chosen to be 

conservative. 

After these initial tests, on the assumption that dose rates were in the worst case for 

each nuclide greater than 2.5 μSv/hr, it was to be found what thickness of soil material 

would result in a dose rate of 2.5 μSv/hr. 


3


2.3 MicroShield calculation details, uncertainties and assumptions 

Energy deposition to dose rate conversion was performed automatically in MicroShield 

using built-in tables of effective dose rate, taken from ICRP-51 [2]. This presents a 

series of possible dose rates depending on the assumed irradiation geometry. The 

highest biological dose rate is produced assuming anterior-posterior geometry (with the 

gamma rays entering a person from the front and exiting through the back), and to be 

conservative it was this maximum dose rate that was reported. Dose rates can vary by 

approximately 30%, depending on which geometry is assumed. 

MicroShield approximates the contribution of scattered radiation to the resulting dose 

rate by the use of build-up tables. The dose rate is dependant on which material is 

chosen as the dominant scattering medium. In accordance with the MicroShield 

manual, the material containing the highest number of gamma ray mean free paths 

should be used as the build-up material – hence in these cases, the source was 

chosen as build-up material. It was found that choosing the shielding soil as the buildup 

material produced identical results; hence the results are insensitive to this 

assumption. 

MicroShield uses a point-kernel integration technique to determine the dose rate. This 

involves splitting the geometry into pieces (kernels). The quadrature order of the 

calculation determines the number of kernels used and hence the accuracy of the 

approximation, at the expense of a longer calculation time. Due to the extent of the 

source relative to the dose rate distance, the quadrature order was increased in the y 

and z axes until the result was unchanged. Beyond a quadrature of 30, the results were 

unchanged, and 50 was used to be conservative. 

In all cases assessed, the ‘contact’ dose rate point was actually positioned at 1 cm 

from the surface, as the method of calculation used by MicroShield is known to become 

unstable at distances closer than 1 cm. The dose rate was found to be nearly 

independent of distance, with only a 1-2% drop in dose rate from contact to 2m, hence 

the results are insensitive to this assumption as well. 


4

3 RESULTS 

3.1 Cobalt-60 case 


The contact dose rate from high density (2 g/cm 3 ) soil containing 200 Bq/g 60 Co 

covered with 30 cm of uncontaminated soil was determined for a series of soil material 

densities from 1.0 to 1.6 g/cm 3 . The results are given in Table 1. 

Soil material 

density (g/cm 3 Contact dose rate 

) (Sv/hr) 

1.0 18.35 

1.2 13.02 

1.4 9.28 

1.6 6.63 

Table 1: Contact dose rates for various soil material densities 

The resulting dose rate above the shielding layer of soil will be between 18.35 Sv/hr 

for loose soil and 6.63 Sv/hr if the shielding surface soil has been compacted. 

3.2 Caesium-137 case 

The contact dose rate from high density (2 g/cm 3 ) soil containing 200 Bq/g 137 Cs 

covered with 30 cm of uncontaminated soil was determined for a series of soil material 

densities from 1.0 to 1.6 g/cm 3 . 137 Cs is a beta emitter. Its daughter, 137m Ba is the 

source of the gamma radiation. Where a source containing 137 Cs was specified, its 

daughter product 137m Ba was also included in equilibrium concentration with 137 Cs. 

Since the half-life of 137m Ba is short (2.5 minutes), it will almost always be found in 

equilibrium with its parent radionuclide. The results are given in Table 2. 

Soil material 

density (g/cm 3 Contact dose rate 

) (Sv/hr) 

1.0 2.58 

1.2 1.67 

1.4 1.08 

1.6 0.70 

Table 2: Contact dose rates for various soil material densities 

The resulting dose rate above the shielding layer of soil will be between 2.58 Sv/hr for 

loose soil and 0.70 Sv/hr if the shielding surface soil has been compacted. 


5

3.3 Covering soil material thickness 


Following on from the scenario outlined in Section 3.1, the contact dose rate from high 

density (2 g/cm 3 ) contaminated soil and rubble containing 200 Bq/g 60 Co at a density of 

2 g/cm 3 covered by uncontaminated soil of various thickness was determined to find a 

relationship between the shielding material thickness and dose rate. The density of the 

covering soil material was taken as 1 g/cm 3 , which presented the worst case in section 

3.1. Figure 1 and Table 3 show the relationship between dose and soil material 

thickness. 

Dose Rate (uSv/hr) 

20 

18 

16 

14 

12 

10 

8 

6 

4 

2 

0 

25 35 45 55 65 75 85 

Soil thickness (cm) 

Figure 1: The relationship between soil thickness and contact dose rate for soil and 

rubble contaminated by 60 Co isotopes resulting in a uniform activity of 200 Bq/g 

Soil Thickness (cm) Dose Rate (Sv/hr) 

30 18.35 

35 13.78 

40 10.39 

45 7.84 

50 5.93 

55 4.49 

60 3.40 

65 2.58 

70 1.96 

75 1.48 

Table 3: Tabulated data for Figure 1 

These data show that to get a dose rate of 10 Sv/hr, the soil material must be at least 

40cm thick, and to get a dose rate of 2.5 Sv/hr, the soil material must be at least 

65cm thick [3]. 


6

4 REFERENCES 


1 MicroShield v7.02, Grove Software Inc, 2007 

2 ICRP-51 (1987) Data for use in protection against external radiation 

3 Personal communication, Paul Atyeo, 23rd March, 2009 


7


APPENDIX A – Calculations to show that dose is geometry and air distance independent 

Depth: 200cm 

Length: 2000cm 

Breadth: 2000cm 

Based on the soil material having a thickness of 30cm and a density of 1 g/cm3 

Dose point Dose (μSv/h) 

‘Contact’ 18.35 

1m 18.21 

2m 17.82 


8

Annex I 

Baseline Groundwater and Leachate 

Sample Results 


Report Determination of 238 U, 235 U, 234 U, 

232 Th, 230 Th, 228 Th, 226 Ra, 3 H and gross 

alpha and gross beta in 8 water 

samples. 

(Samples: KO2A etc…) 

UKAEA Harwell 

Customer Jon Blackmore 

UKAEA B175 

Harwell International Business Centre 

Didcot 

Oxfordshire 

OX11 0RA 

Customer reference number Quote620 

GAU job number GAU1278 (Final) 

Date samples received 18 th August 2008 

Report date 1 st October 2008 

Report produced by Dr P. Gaca 

(Radiochemist, GAU-Radioanalytical) 

Signed 

Report checked by 

Signed 


Methodology 

Job reference number 

GAU1278 (Final) 

Samples were received at the National Oceanography Centre, Southampton on 18 th 

August 2008 in good condition. 

Gamma spectrometry (Method GAU/RC/2032: Accredited to ISO/IEC 17025:2005) 

100ml of the sample was evaporated down to less than 20ml and transferred to a 

scintillation vial. The sample was then counted on a well-type HPGe detector 

previously calibrated with a mixed nuclide standard of identical geometry. The 

resulting spectrum was analysed using Fitzpeaks spectral analysis software. All 

anthropogenic radionuclides were identified and quantified. In addition 60 Co, and 

137 Cs were specifically searched for and limits of detection reported where no activity 

was detected. 

Gross alpha / beta in waters (Method GAU/RC/2034) 

200 ml of the sample was acidified with H2SO4 and evaporated to dryness and the 

residue ignited at 350 C. The ignited residue was ground and mounted onto a 47 mm 

filter paper. The source was then counted on a gas flow proportional counter 

previously calibrated against 241 Am (alpha) and 137 Cs (beta). 

3 H in aqueous samples (Method GAU/RC/2004) 

50ml of the sample was removed for 3 H analysis. The sub-sample was purified by 

distillation. The 3 H content of the distillate was then measured using a Quantulus 

ultra-low level liquid scintillation counter. 

226 Ra in aqueous samples (Method GAU/RC/2038) 

An aliquot of the aqueous sample is mixed with a water-immiscible scintillation 

cocktail in a glass vial. The vial is sealed and immediately counted on a Perkin Elmer 

Quantulus liquid scintillation counter with alpha-beta discrimination activated to 

determine the total 222 Rn activity. The sample is then stored for two weeks and 

recounted to determine the activity of supported 222 Rn/ 226 Ra. 

Th isotopes by alpha spectrometry (Method GAU/RC/2027) 

An aliquot of the sample is spiked with 229 Th and acidified. An iron hydroxide 

precipitation followed by anion exchange chromatography is used to isolate Th from 

the solution. The activities of 230 Th and 232 Th are then determined by alpha 

spectrometry. 

U by alpha spec & ICPMS (Method GAU/RC/2026) 

An aliquot of the sample is spiked with 232 U and acidified. A combination of anion 

exchange and extraction chromatography is used to isolate U from the solution. 238 U 

and 234 U are determined by alpha spectrometry, and the 235 U content is determined 

relative to 238 U by ICP-MS. 

Geosciences Advisory Unit, National Oceanography Centre, Southampton, European Way, SO14 3ZH 

Page 2 of 8 1/10/08 

C:\Documents and Settings\carolearp\Desktop\Permit Application - LLW\Annex I.doc 




Limits of detection / quantification 

For gamma data, limits of quantification, LQ, is calculated as defined by Currie (1968) 

and Gilmore & Hemingway (2000) 

 

n C 

2 

1 100 100 1 

LQ 

gamma 0. 5 

( ) 

1 

1 

4 1 

 

2 

2m 

 

. 

 

 

t E Y M g 

where is set at 2.00, C is the background counts, n is the number of channels 

covering the peak, m is the number of background channels taken either side of the 

photopeak, t is the count time in seconds, E is the counting efficiency, Y is the gamma 

emission probability and Mg is the mass of sample analysed in grams 

Limits of detection for H-3 analyses are quoted as LD as defined by Currie, 1968. 

2. 

71 

4. 

65 C 100 100 1 

LD 

( Bq / g) 

 

 

t E R M 

where C is the background count, t is the count time in seconds, E is the measurement 

efficiency, R is the chemical recovery and m is the sample mass in grams. 

References 

Currie L.A. (1968). Limits of qualitative detection and quantitative determination. Anaytical Chemistry, 

40 (3), 586-593. 

Gilmore G. and Hemingway J. (2000). Practical gamma-ray spectrometry. John Wiley, Chichester, UK 

Geosciences Advisory Unit, National Oceanography Centre, Southampton, European Way, SO14 3ZH 

Page 3 of 8 1/10/08 

C:\Documents and Settings\carolearp\Desktop\Permit Application - LLW\Annex I.doc 

g 


Summary of samples and results 



All uncertainties quoted are propagated method uncertainties unless otherwise stated. 

* Indicates results obtained using an accredited method. 

Results 

GAU ID Customer ID Sample type 

GAU1278/1 KO2a Water 

GAU1278/2 KO3 Water 





GAU1278/7 KCLW2A2 Water 

GAU1278/8 KCLW3A1 Water 

Gross alpha/beta 

GAU ID 

Gross alpha 

[Bq/L] 

+/- 

Gross beta 

[Bq/L] 

+/- 

GAU1278/1

3 H 



GAU ID 

3 

H [Bq/L] +/- 

GAU1278/1

238 U, 235 U, 234 U 

238 U 

235 U 



GAU ID 

[Bq/L] 

+/- 

[Bq/L] 

+/- 

[Bq/L] 

+/- 

GAU1278/1 0.039 0.012



Gamma Spectrometry* 

Artificial Radionuclides 

241 60 137 154 54 65 

GAU ID Am +/- Co +/- Cs +/- Eu +/- Mn +/- Zn +/- 

GAU1278/1



Gamma spectrometry* 

Natural Radionuclides 

228 40 210 212 214 226 208 234 235 

GAU ID Ac +/- K +/- Pb +/- Pb +/- Pb +/- Ra +/- Tl +/- Th +/- U +/- 

GAU1278/1

Annex J 

Capability Statements 


Research Sites Restoration Limited (RSRL) 

(also referred to as UKAEA Harwell) 

RSRL provided Augean with technical support in relation to the Low Level Wastes 

from the perspective of a waste producer and consignor. RSRL attended the public 

exhibition to provide information to the public on the wastes. 

Research Sites Restoration Limited (RSRL) is the site licence company responsible 

for the closure programme at Harwell and Winfrith. Winfrith was a major centre for 

groundbreaking reactor development from the late 1950s to the 1990s whilst 

Harwell’s origins go back to the dawn of the UK’s nuclear industry in the 1940s. 

RSRL is a wholly-owned subsidiary of UKAEA (the United Kingdom Atomic Energy 

Authority) and operates under contract to the Nuclear Decommissioning Authority 

(NDA). 

UKAEA 

The UKAEA group is a world leader both in decommissioning and regenerating 

nuclear sites and in developing fusion as a sustainable, secure and carbon-free 

energy source. With decades of experience as pioneers in these fields, UKAEA is 

making a key contribution to meeting the twin challenges of sustainable development 

and climate change. 

UKAEA has over 50 years’ experience in nuclear site management, operations and 

decommissioning. Through projects spanning the nuclear lifecycle, UKAEA provides 

industry-leading technical, design, engineering, safety, and programme and project 

management consultancy services to organisations around the world. 

The UKAEA Ltd. subsidiary of the UKAEA group provided technical assessments in 

support of the authorisation application for the proposal. 

The involvement of RSRL was fronted by Paul Atyeo. 

Paul Atyeo, RSRL 

Paul is a Chartered Mechanical Engineer and Chartered Environmentalist with a first 

degree in Mechanical Engineering and a masters degree in Business Administration. 

Paul has worked in the nuclear industry for 21 years, specialising in nuclear reactor 

experimental systems, land remediation, nuclear waste management, nuclear 

decommissioning and site delicensing. Paul currently manages decommissioning of 

the Harwell nuclear site. 


Peter Shaw 

Job Title – Group Leader, Consultancy Development Group 

Qualifications 

1979 HNC Chemistry 

1986 Post Graduate Course in Radiological Protection (PGRP) 

1993 Advanced Course in Radiological Protection (ARP) 

1999 Diploma in Pollution Control 

Professional Qualifications 

2000 Member, Chartered Society for Radiological Protection (CRadP) 

March 2001 and February 2006 - RPA2000 Certificate of Competence to be a Radiation 

Protection Adviser 

Key skills 

Expert in radiological 

protection. International 

expertise in ALARA. 

Special expertise in 

NORM. 

Input to national and 

international standards 

Communications, 

delivering presentations 

at national and 

international events, 

chairing meetings, etc. 

Emergency response 

adviser 

Project and team 

management 

Customer liaison 

Career history 

Peter Shaw 

Profile 

Peter Shaw began his career with the National Radiological 

Protection Board in 1979, where he spent his formative years 

developing a sound grounding in the fields of health physics 

and radiological protection, including metrology, dosimetry, 

radiochemistry and radiological assessments. He has 

developed his expertise over the last 30 years, to become a 

well respected expert in radiological and environmental 

protection and hazard assessment. He develops and delivers 

professional level training modules for customers operating in 

the non-nuclear industrial sectors, and sits on a number of 

internal and external committees dedicated to developing 

expertise in this very specialised area. He sits on national and 

international committees and provides input to radiation 

protection standards. He is a highly qualified and experienced 

Radiological Protection Adviser and participates in national 

emergency exercises at off-site control centres. In addition to 

his well respected technical expertise, he is an accomplished 

manager, managing a multi-disciplinary team of technical 

employees as well as a portfolio of projects for internal and 

external customers. 

Group Leader HPA, Leeds 1984–present 

Consultancy Development Group 

Manages the Consultancy Development Group within HPA, with specific responsibility for the 


development of radiation protection services, including the Radiation Protection Advisor (Qualified 

Expert) service as well as various support activities. 

A certificated RPA with extensive experience of advising on the use of industrial radioactive 

sources and x-ray equipment including gauging, non-destructive testing, security and analytical 

equipment. Also specialises in advising users of unsealed radioactive materials and NORM. 

Extensive experience in the provision of radiation protection training at all levels. Provides 

professional level training to internal and external customers. Has managed the development and 

running of the HPA Radiological Protection Training Scheme Module on “Principle for Protection 

against Internal Radiation Sources”. 

Participates in nuclear emergency exercises at off-site control centres and also participates in the 

multi-agency CBRN preparedness group for local government. Contributes to the development of 

national security standards for radioactive sources. Assists in the investigation of international 

radiological accidents. 

Provides input to radiation protection standards and guidance at a national and international level. 

Currently Secretary of the European ALARA Network. 

Scientific Officer 

Peter Shaw 

National Radiological 

Protection Board, Leeds 1979–1984 

Operation and (later) management of radiation protection services, including metrology, external 

and internal dosimetry, radiochemistry and consumer product testing. 

Experienced in developing environmental transfer.models and undertaking assessments of the 

radiological impact (public and worker) from the release of radionuclides into the environment. 

Undertook environmental measurements and sampling following radiological accidents. 

. 


Galson Sciences Limited July 2009 

Dr. Roger D. Wilmot, BA, PhD Principal Consultant 

Dr. Roger D. Wilmot has degrees in Earth Sciences from Cambridge University (BA) and Imperial 

College, London (PhD). He is a geologist with over 20 years experience in providing a broad range of 

research, consultancy and management services to a range of clients, starting with site 

characterisation work for the four proposed UK shallow sites in the 1980s. 

Dr. Wilmot is chair of SSM’s OVERSITE international panel responsible for regulatory review of SKB’s 

risk assessments for radioactive waste disposal in Sweden. He has also worked fro the Swedish 

regulators on Quality Assurance, development of a strategy for consideration of future human actions 

in assessments and the conduct of risk assessments. 

Dr Wilmot provided technical support to the Environment Agency and SEPA in the recent revision of 

the agencies’ Guidance on Requirements for Authorisation, and was part of the management team for 

the review on behalf of the Environment Agency of BNFL’s safety case for the LLWR. 

Dr. Wilmot developed a computer code for UKAEA to undertake radiological performance assessments 

of waste disposal and storage facilities, and assessments of the impact of radioactivity in the 

environment. He has led a variety of waste management options appraisals for UKAEA Dounreay, 

including an evaluation of the transport of radioactive waste, as part of BPEO studies. He authored a 

draft ESC for the Pits facility at Dounreay. 

Dr. Wilmot was responsible for development, implementation and trial application of an methodology 

for assessing the doses associated with landfill sites for Special Precautions Burial of LLW, and has 

extended and used this methodology for PA of on-site disposal of radioactive wastes and for assessing 

the dose implications of dustbin disposal of VLLW. 

Dr. Dev Reedha, BEng(Hons), PhD Senior Consultant 

Dr. Dev Reedha has degrees in Mechanical Engineering and Energy Systems (BEng First Class 

Honours) and Engineering (PhD) from the University of Manchester. He has six years experience in 

radioactive waste management and nuclear consultancy and research, and ten years experience in 

computational fluid dynamics, mathematical modelling and simulation of fluid flow in sedimentary units, 

and in computer code development. He also has experience in software consultancy. 

Dr. Reedha has carried out key technical work in radioactive waste management. On behalf of the UK 

environmental regulators (through SNIFFER), he developed a computational model to evaluate 

allowable disposals of radioactive waste to landfill, and provided technical support for the evaluation of 

doses associated with VLLW disposals. On behalf of Defra, he provided technical support on a project 

to advise Government on the feasibility of gathering data on the geographical generation of nonnuclear 

LLW, including VLLW, within the UK. For ONDRAF-NIRAS, he has reviewed the treatment of 

concrete degradation in safety assessments, and is currently working (lead author) on the Belgian 

Category A inventory report. 

Dr. Reedha is an experienced groundwater flow and contaminant transport modeller using state-of-theart 

software packages such as FEFLOW and GoldSim-RT. He contributed to an assessment of postclosure 

safety of the LLW disposal facility near Drigg, on behalf of the Environment Agency. For 

DSRL, he worked on the development of a risk assessment model in GoldSim-RT for solid LLW 

disposal at Dounreay, and is currently undertaking hydrogeological modelling analysis of the proposed 

LLW facilities using FEFLOW. On behalf of the Nuclear Decommissioning Authority’s Radioactive 

Waste Management Directorate (NDA RWMD), he recently undertook transient three-dimensional 

groundwater flow calculations using FEFLOW to evaluate potential hydrological interactions in a 

geological disposal facility during the operational phase and after facility closure. 


NAME: GENE BARRY WILSON 

BORN: 1957 

NATIONALITY: British 

QUALIFICATIONS & PROFESSIONAL AFFILIATIONS: 

Doctor of Philosophy - Imperial College London 

Diploma of Imperial College - Imperial College London 

B.Sc. Honours in Botany/Genetics - University College Cardiff 

Chartered Town Planner; Member of the Royal Town Planning Institute 

Chartered Waste Manager; Member of the Chartered Institution of Wastes Management 

Chartered Biologist; Member of the Institute of Biology 

Chartered Environmentalist 

Member of the Institute of Quarrying 

Member of the Institute of Ecology and Environmental Management 

Registered Principal Environmental Auditor 

CAREER SUMMARY: 

2005 – Present: Group Technical Director, Augean plc 

Dr Wilson is an experienced environmental manager with particular expertise in quarrying 

and waste management together with skills in industrial and applied ecology. 

Dr Wilson is responsible for the planning and permitting strategy and delivery for the two 

divisions of the Group, Landfill and Treatment. This involves regular interface with regulator 

bodies and the public. 

A key part of his role is monitoring and advising on compliance and regulatory matters for 

the Group. Dr Wilson manages a team of auditors and monitoring technicians who 

continuously assess the Group’s environmental and health and safety performance. The 

results of these assessments are reported annually in the Group Corporate Responsibility 

report. 

Dr Wilson is responsible for the management and monitoring of the Group’s Integrated 

Management System which satisfies the requirements of ISO 14001 (Environmental 

Management System Standard), ISO 9001 (Quality Management System Standard) and 

OHSAS 18001 (Health, Safety and Welfare Management System Standard). 

Dr Wilson manages a team of highly trained chemists within our laboratory services who 

provide our clients with accurately assessed data to identify and understand the nature of 

the waste they produce which then allows us to offer the best management solution. 

Dr Wilson actively engages with the industry, regulators and government departments at a 

national level promoting high standards and new technologies for the sector. He is a 

member of the Regulatory and Planning Committees of the Environmental Services 

Association and regularly comments on planning policy and technical guidance notes. He is 

also a member of the DEFRA Hazardous Waste Steering Group which is developing a 

strategy for the modernisation of the sector. 

July 2009 Page 1 of 2 


1987 - 2005: Director of Environmental Planning and Principal 

Ecological Consultant with MJCA 

Dr Wilson directed the Company services in planning, environmental assessment, 

environmental audit, environmental management systems and waste management 

licensing bringing his strong organisational skills to manage successfully these complex 

environmental projects. Dr Wilson is an experienced expert witness and has given 

evidence on the subjects of minerals and waste management operations, planning policy, 

need and ecology. 

Dr Wilson was responsible for the Environmental Planning services of the Company which 

include town and country planning, minerals planning and in particular waste management 

planning. An essential part of his role at MJCA was the assessment of local policy and 

strategy in matters of land use and waste management and he has considerable 

experience of the structure, analysis and use of local plans. A significant proportion of Dr 

Wilson's work involved the preparation and negotiation of environmental assessments and 

planning applications where his knowledge of environmental science and understanding of 

the technical and practical demands of industrial development are critical. 

Dr Wilson is an experienced environmental manager who has assisted a range of 

companies from single site to multinational in the development and implementation of 

environmental management systems. Working closely with the client company he 

undertook environmental reviews, developed environmental policies, prepared manuals 

and codes of practice and provided advice on ISO 14001 and EMAS. Dr Wilson has 

managed several hundred environmental audits of companies and their facilities to 

demonstrate compliance with legislation and environmental policy, for the purpose of due 

diligence prior to acquisition and for insurance purposes. 

Dr Wilson is an experienced applied and industrial ecologist and provided advice on 

matters such as reclamation, conservation, bioengineering and landscape planting with a 

strong emphasis on the integration of land development with vegetation and wildlife. Dr 

Wilson has particular expertise in the restoration of quarries and landfill sites and in habitat 

creation. He lectured regularly to the waste management industry on reclamation and 

chaired the landfill reclamation course run by the Environmental Services Association. 

1983 - 1987: Research Scientist at the School of Agriculture, 

Nottingham University 

Dr Wilson was responsible for the design and management of experiments and 

reclamation procedures, including the propagation and establishment of native plant 

species in a fully active dolerite quarry in Wales, the organisation of surveys to describe 

native plant communities in the areas adjacent to the quarry and evaluation of their 

conservation status and the monitoring of meteorological and edaphic factors on field sites. 

Dr Wilson liaised closely with conservation bodies, local councils, landowners and in 

particular the sponsoring quarry company for whom periodic reports were produced and 

lectures given. This work culminated in the production of a manual for the continuing 

reclamation and conservation of the site. 

July 2009 Page 2 of 2

Appendix CRF - Part 3 - Northamptonshire County Council

Create successful ePaper yourself

Delete template?

Save as template?