atw - International Journal for Nuclear Power | 06.2021

nucmag.com 

2021 

6 

ISSN · 1431-5254 

32.50 € 

First Repatriation of 

Vitrified Reprocessing 

Waste from Sellafield 

Interview with 

Massimo Garribba 

Practical Response 

to a Dirty Bomb

atw Vol. 66 (2021) | Issue 6 ı November 

Nuclear Energy: A Path to Reason 

3 

Dear Reader, in the year 10 after Fukushima and the political reaction in Germany, unique in the world at that time, 

with the immediate closure of 8 of the 17 nuclear power plants in operation at that time and less than a year before the 

last three plants are to be shut down by 31 December 2022 at the latest, the facts seem to be catching up with this policy 

and the mistakes of the national nuclear phase-out are clearly revealing themselves. The facts are not based on political 

decisions, political wishful thinking or visions, but simply on our natural living conditions. We heat or cool because 

nature only offers pleasant weather throughout the year in a few places in the world. We need “mechanical energy” to 

move ourselves, sometimes over long distances, because there are no alternatives to sensible transport, neither for 

ourselves nor for the things we need every day or for what we want. We use a variety of energy sources and technologies 

because man as an energy source with an overall efficiency in the low single-digit range is the worst of all alternatives 

for energy supply – just ask yourself why battery-powered bicycles are currently booming when they originally relied on 

muscle power? 

EDITORIAL 

Germany is currently confronted with two facts more than 

ever: 

1. Politics in Germany has set ambitious targets for the 

reduction of climate-impacting emissions, and is 

currently even aiming for so-called climate neutrality 

by 2050 at the latest or earlier. At the same time, greenhouse 

gas emissions in Germany have been significantly 

reduced since 1990. Total emissions converted into 

carbon dioxide equivalents (excluding carbon dioxide 

emissions from land use, land use change and forestry) 

fell by around 439 million tonnes (mt) or 35.1 % by 

2019. These reductions were largely achieved by the 

energy industry, which is often berated in public on this 

issue, mainly through the expansion of low-emission 

renewables and the switch from coal to natural gas. But 

this balance could be much better, because in total, the 

nuclear power plants in operation in Germany before 

Fukushima had avoided around 140 million tonnes of 

CO 2 emissions year after year – emissions equal, for 

example, to those of all German car traffic. This 

“ shortfall”, moreover provided by highly flexible and 

highly available nuclear power plants, i.e. important for 

balancing volatile generation from renewables, is 

missing and will be missing. 

2. A development in the prices of fossil fuels and biomass 

that is almost unique in terms of speed and scope is 

burdening all consumers. Expressed in figures and 

using the example of heating oil, this means an increase 

from € 38 per 100 l at the beginning of November 2020 

to € 91 per 100 l now, in mid-October 2021, i.e. an 

increase of 139 %! The last time crude oil and crude oil 

product prices reached this level was before the banking 

and financial crisis in 2008. 

However, while politicians in Germany persist in their 

rigid course, almost in rigidity, looking at the development 

of energy prices, other countries and regions have long 

since taken other paths. Within two decades, China has 

built up an efficient nuclear industry. 50 high-performance 

nuclear power plants are an important part of the 

electricity supply without fear of contact with renewables, 

and in the coming years four to eight units can be added – 

every year. Moreover, China is open to new reactor 

concepts. High-temperature reactor technology with 

pebble fuel elements, which originated in Germany, as 

well as sodium-cooled reactors and now even “molten salt 

reactors” are being designed, built and commissioned. 

In addition, China is also slowly pushing into the worldwide 

export of its plants. The same is true for nuclear technology 

from Russia. In recent years, the Russian nuclear industry 

has impressively shown that it can build plants without 

major delays even in “green field” countries, i.e. in nuclear 

newcomer countries without their own power reactors 

so far. 

And now Great Britain: Nuclear energy is to play a 

central role here on the way to climate neutrality with 

simultaneously high supply security at affordable prices. 

To this end, investments in nuclear energy are to be 

promoted and secured by the state, and reactor projects of 

small and medium capacity (Small Modular Reactors, 

SMR) are to be supported. 

Last but not least, there is also France, where President 

Emmanuel Macron announced a 30 billion innovation 

programme at the beginning of October, with SMR reactors 

as an important infrastructure component. 

In the end, the German “Energiewende” will not be able 

to escape the laws of nature and the facts. It will then be 

interesting to see with what verbal acrobatics possible 

future problem developments will be conveyed by the 

actors who in Germany are pursuing the end of nuclear 

energy, but noticeably also an end to all other proven 

power plants and a secure energy supply. 

Worldwide, on the other hand, it is still true and 

reinforced that the energy supply will ultimately be based 

on nuclear energy in combination with all other energy 

sources – a path of reason. 

Christopher Weßelmann 

– Editor in Chief – 

Editorial 

Nuclear Energy: A Path to Reason


4 

CONTENTS 

Issue 6 

2021 

November 

Contents 

Editorial 

Nuclear Energy: A Path to Reason 3 

Inside Nuclear with NucNet 

How UAE’s Barakah Nuclear Project has Set Standard 

for Newcomer Countries 6 

Did you know? 7 

Calendar 8 

Feature | Decommissioning and Waste Management 

First Repatriation of Vitrified Reprocessing Waste from Sellafield 9 

Marco Wilmsmeier and Michael Köbl 

Interview with Massimo Garribba 

“EU Member States Can Choose Their Energy Sources and 

Can Include Nuclear in Their Energy Mix as Part of Their Effort 

to Achieve Decarbonisation and Carbon Neutrality by 2050.” 14 

Decommissioning and Waste Management 

Investigations of the Tailskin Seal During the Retrieval 

Concept ‘Shield Tunnelling with Partial-face Excavation’ 

in the Asse II Mine 18 

Birte Froebus 

Environment and Safety 

Practical Response to a Dirty Bomb 22 

James Conca 

Equipment Selection Methodology of Seismic Probability 

Safety Assessment for Nuclear Power Plant 27 

Junghyun Ryu and Moosung Jae 

Research and Innovation 

Experimental Study of Convective Heat Transfer 

Through Fuel Pins of a Nuclear Power Plant 37 

Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus 

Preliminary CFD Analysis of Innovative Decay Heat Removal 

System for the European Sodium Fast Reactor Concept 41 

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, 

Janos Bodi and Enrico Girardi 

Cover: 

Return transport of vitrified high-level radioactive 

waste from reprocessing of German fuel elements 

in Sellafield in November 2020 (Courtesy of GNS 

Gesellschaft für Nuklear Service mbH). 

Study on Verification of SPACE Code Based on an 

MSGTR Experiment at the ATLAS-PAFS Facility 49 

Kyungho Nam 

News 57 

Nuclear Today 

For the Sake of Nuclear Safety and Security, 

it’s Time for Rogue Actors to Leave the Stage 62 

Imprint 36 

Contents


Feature 

Decommissioning and 

Waste Management 

9 First Repatriation of Vitrified Reprocessing Waste 

from Sellafield 

5 

CONTENTS 



14 “EU Member States Can Choose Their Energy Sources and 

Can Include Nuclear in Their Energy Mix as Part of Their Effort 

to Achieve Decarbonisation and Carbon Neutrality by 2050.” 


18 Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield 

Tunnelling with Partial-face Excavation’ in the Asse II Mine 

Birte Froebus 


22 Practical Response to a Dirty Bomb 

James Conca 

27 Equipment Selection Methodology of 

Seismic Probability Safety Assessment for Nuclear Power Plant 



49 Study on Verification of SPACE Code Based on an 

MSGTR Experiment at the ATLAS-PAFS Facility 

Kyungho Nam 

Contents


6 

How UAE’s Barakah Nuclear Project has Set Standard 

for Newcomer Countries 

‘They did it in a way that is measurable and can be looked at by the rest of the international 

community’ 

INSIDE NUCLEAR WITH NUCNET 

When Unit 2 of the four-unit Barakah nuclear power station 

in the United Arab Emirates was connected to the grid earlier 

this year, delivering its first megawatts of electricity, project 

owner Emirates Nuclear Energy Corporation (ENEC) said 

the milestone took it another step closer to its goal of using 

nuclear to supply up to a quarter of the country’s electricity 

needs 24/7, while driving reductions in carbon emissions – 

the leading cause of climate change. 

Unit 2’s grid connection came five months after Unit 1 

became the first commercial nuclear power reactor in the 

Arab World to begin commercial operation. 

For ENEC, these recent operational milestones show the 

world that the UAE is serious about efforts to use nuclear 

energy to further climate change mitigation efforts through 

the decarbonization of the electricity sector. But the benefits 

of the Barakah project – one of the largest nuclear energy 

stations in the world – go deeper than that, with the facility 

providing energy security and powering social and economic 

growth by providing high-value jobs and supporting an 

entirely new industry. 

More than 2,000 UAE-based companies are part of the 

supply chain for Barakah and local contracts have been 

awarded to a value of $ 4.8 bn. “The local supply chain is, 

and will continue to be essential for the operations and 

maintenance of Barakah over the coming 60 years,” ENEC 

told NucNet. “It will help ensure we have the localized 

services and components we need.” 

Unit 1 at Barakah is already the largest single generator 

connected to the UAE grid and adding Unit 2 in the coming 

months will drive significant energy security, enabling the 

electrification of key industries. 

“The dual role of decarbonization and electrification 

offers a proven solution to cutting carbon emissions,” ENEC 

said. “It is the culmination of more than a decade of policy 

development, planning and program management.” 

Once fully operational, the four units at Barakah will 

prevent about 21 million tonnes of carbon emissions 

annually and potentially drive other clean energy 

technologies such as green hydrogen. In June, Enec signed 

an initial agreement with French state- controlled power 

group EDF to eventually cooperate on research and 

development in the nuclear energy sector – research that 

could include exploring the production of green hydrogen 

powered by nuclear energy. 

ENEC chief executive officer Mohamed al-Hammadi 

told the Atlantic Council Global Energy Forum earlier this 

year that the right environment exists for commercial 

nuclear power plants to produce green hydrogen with 

capital costs making it viable after falling 50 % in the past 

five years. 

So the benefits go far beyond clean electricity. The 

creation of a new local nuclear energy sector, along with 

thousands of skilled jobs and career paths for UAE nationals, 

and the national knowledge and expertise that has been 

developed, mean that Barakah is powering social and 

economic growth. ENEC believes nuclear energy can also 

support innovation in related sectors, such as deep space 

exploration and agriculture. 

Construction at Barakah began in 2012. Units 3 and 4 are 

in the final stages of commissioning at 94 % and 90 % 

complete respectively and the development of the station as 

a whole is now more than 96 % complete. 

For the UAE, as a newcomer country to nuclear, capacity 

development was vital. From the start, ENEC was training 

UAE nationals to ensure qualified professionals, working 

alongside international experts, could operate and maintain 

the facility. The UAE now has multiple academic institutions 

offering qualifications in nuclear sciences, and vocational 

training in related applications. The aim was always to 

ensure that the UAE has the educational facilities to support 

generations of staff entering the nuclear sector. 

Almost 60 % of the workforce are UAE nationals and more 

than 20 % are women, one of the highest percentages of 

female nuclear energy professionals globally. By collaborating 

with universities and schools across the UAE, Enec has 

trained and qualified more than 500 young engineers. 

The other key lesson for newcomer countries is that clear 

policies and full government support are needed from the 

beginning – not just in terms of a commitment to new-build 

and its financing, but also in areas such as safety, quality, 

transparency and security. 

The UAE worked directly with the International Atomic 

Energy Agency (IAEA), using the agency’s milestones 

approach to the construction of new nuclear power plants to 

establish a comprehensive and systematic guide that can be 

followed by other new comer countries planning commercial 

reactors. 

Hamad al Kaabi, the UAE’s permanent represen tative to 

the IAEA, told a webinar to discuss the UAE’s commercial 

nuclear energy programme that the IAEA was instrumental 

in helping the UEA prepare infrastructure, establish nuclear 

laws and oversee the development, construction and startup 

phases at Barakah. 

He said the task now is for the UAE to share its experience 

with other IAEA member states. “We now have a case study, 

not just theoretical guidelines,” Mr al Kaabi said. “Other 

countries can use it to develop nuclear infrastructure that s 

in line with best practices.” 

IAEA director-general Mariano Grossi said the UAE’s 

journey to nuclear power “is already being looked at as 

an exemplary one”. He said: “They did it in a way that is 

measur able and can be looked at by the rest of the international 

community.” 

Mr Grossi said that because of what the UAE had achieved 

“we now have a complete, successful example, a logical 

sequence of steps that can be followed by those who would 

like to build nuclear plants. 

Lessons learned during construction meant experience 

from the development of each unit could be applied to the 

next one. This has brought significant benefits in terms of 

manpower reduction, efficiency and sustainability. 

According to ENEC, Barakah provides a successful case 

study for the development of new nuclear projects, 

particularly for newcomer countries. It demonstrates that it 

is possible for a country to establish a nuclear energy 

industry as part of efforts to ensure energy security, to 

diversify its energy mix and decarbonize electricity 

generation. The Barakah model, says ENEC, stands as an 

example for others to follow. 

Authors 

By Kamen Kraev and David Dalton 

Inside Nuclear with NucNet 

How UAE’s Barakah Nuclear Project has Set Standard for Newcomer Countries


Net-Zero Germany 2045 – Contradictory McKinsey-Study 

Recently McKinsey & Company published the study “Net-Zero Deutschland – 

Chancen und Herausforderungen auf dem Weg zur Klimaneutralität bis 

2045” (Net-Zero Germany – Opportunities and Challenges on the Way to 

Climate Neutrality 2045) which presented necessary development paths, 

obstacles and alternative routes to reach the recent government goal of 

climate policy in the different sectors of the economy, energy generation, 

industry, mobility, buildings, agriculture and banking. 

The tenor of the report as expressed in the summary claims the feasibility 

of climate neutrality by 2045 and makes the point that it would not cost very 

much in net terms because of the 6,000 billions Euro of investment in real 

assets deemed necessary only 1,000 billions are additional but 5,000 billions 

are needed anyway and just will have to be redirected in green technologies 

and measures. Thereby the study concludes that under optimal circumstances 

the entire social cost over the entire period and balanced over all sectors 

could be zero. Which would enable to reach net zero without compromising 

the quality of life and without social and economic hardship. 

While the stated zero-cost conditions already indicate a perspective from 

a very high ground at a high level of abstraction, looking into the sectorspecific 

analysis shows that the zero-cost outcome is not very likely in the 

real world. Below you can find a graph detailing the factual emission 

reductions in previous decades and the necessary reduction for the time 

frames 2019 – 2030 and 2030 – 2045 which shows that particularly the 

energy sector, industry and mobility have to increase the pace in emission 

reductions drastically. 

A closer look at the energy sector that makes up 32 per cent of German 

green house gas emissions shows that the pace of installation of new 

renewable energy (REN) generation capacity has to be almost tripled for the 

coming decade as compared to preceding years from 6,5 GW p.a. to some 

18 GW p.a. which is considerably more than in the years of the German REN 

bonanza 2010 to 2012 (see Graph 2). Also, the transport grid will have to be 

expanded by 12,700 kilometers by 2040 instead of 6,100 in previous 

assumptions and the distribution grid will need strengthening or expansion 

in the order of 400,000 kilometers. Not surprisingly there is more skepticism 

articulated about current developments here. The study points out that the 

energy transformation up to now just used the existing (large) safety margins 

in the system, but that REN deployment is currently insufficient to move 

beyond this point. Consequently, investment in REN and the grid have to be 

increased considerably. 

If this does not happen there is an important risk of rising electricity prices 

due to scarcity and of under coverage of demand that might not in the short 

term be replaceable with imports. Hence the study considers the possibility 

of load shedding in the future. Part of the proposed solution is load management, 

but the discussion about limiting the capacity for e-vehicles charging 

early this year showed, that there is not much acceptance for this kind of 

measures in Germany which indeed look strikingly similar to the load 

shedding that they are intended to avoid. The proposed alternative paths for 

carbon emissions reduction are CCU (Carbon Capture and Utilization) and 

CCS (Carbon Capture and Storage) and the import of green hydrogen. The 

study asserts, that the viable transformation path is still not decided and 

planned. The risk exists, that the energy transformation as it is running now 

will lead to a situation where neither the energy needs of households and 

enterprises can be met nor the climate targets will be reached. 

Another sector that shows contradictions in the study is buildings. While the 

study realistically assumes that in 2050 still some 80 per cent of buildings will 

be of a construction date prior to 2011, it is still supposed that 50 per cent of 

heating will be done with heat pumps and that oil and gas boilers will have 

been phased-out by 2040 and 2050 respectively. No account is made of the 

fact that older buildings are not appropriate for heat pumps and that changing 

this would require extremely cumbersome and expensive reconstruction 

measures. Without these, the assumed gains in efficiency like those that exist 

in electrifying the mobility sector, cannot be realised. The subsequent 

enormous increase in electrical power consumption and particularly the 

extreme peaking effect of this measure in a period of low solar radiation is not 

accounted for either. The obvious and much more viable alternative of 

importing green gases is not taken into consideration for the building sector. 

There also is a macroeconomic weak spot in the reasoning behind the 

proposed net zero path: for some sectors measures are proposed that are 

comparatively labor intensive. What might sound good at first glance and 

could provide some relief for employees e.g. in the automotive industry 

which will face serious pressure on employment due to the much simpler 

manufacturing of electric vehicles compared to vehicles with internal 

combustion engines, does not play well on the macro level. In the context of 

demographic change with a considerably shrinking work force in Germany 

and many European countries it seems ill fated to propose development 

paths and try to generate growth with labor intensive and relatively low 

productivity activities such as energetic refurbishment of buildings or 

massively accelerated deployment of renewable power. 

Not only to prevent load shedding, to avoid high volatility in energy prices 

and to assure true and profound decarbonization of the energy sector, but 

also in terms of labor productivity and effective use of this scarce resource of 

the future, nuclear power appears to be the more reasonable and viable 

instrument to reach net zero. 

7Did you know? 

DID YOU EDITORIAL KNOW? 7 

Average Annual Reduction of Emissions per Sector 

in Million Ton CO 2 -equiv. 

Energy 

Industry 

Mobility 

Buildings 

Agriculture 

0 

0.7 

1.1 

3.3 

5.7 

4.5 

5.1 

3.0 

3.7 

6.3 

7.2 

7.2 

7.2 

7.9 

p 1990–2019 

p 2019–2030 

p 2030–2045 

13.6 

Annual Net Increase in Installed Capacity of REN 

for Electricity Generation in GW 

18.0 

p Annual Capacity Addition in GW 

14.0 

10.9 10.7 

9.3 

5.6 6.6 7.5 8.1 

6.6 6.6 6.2 

6.7 

2010 

2011 

2012 

2013 

2014 

2015 

2016 

2017 

2018 

2019 

2020 

2021 –2030 

2030 –2045 

Source: 

Net-Zero Deutschland – 

Chancen und Herausforderungen 

auf dem 

Weg zur Klimaneutralität 

bis 2045, McKinsey & 

Company, September 

2021 

For further details 

please contact: 

Nicolas Wendler 

KernD 

Robert-Koch-Platz 4 

10115 Berlin 

Germany 

E-mail: presse@ 

KernD.de 

www.KernD.de 

Did you know?


8 

Calendar 

CALENDAR 

2021 

31.10. – 12.11.2021 

COP26 – UN Climate Change Conference. 

Glascow, Scotland, www.ukcop26.org 

01.11. – 05.11.2021 

International Conference on 

Radioactive Waste Management: 

Solutions for a Sustainable Future. 

IAEA, Vienna, Austria, 

https://www.iaea.org/events/ 

international-conference-on-radioactive-wastemanagement-2021 

Online Conference04.11.2021 

Small and Advanced Reactors. 

PMI-Live, 

https://registration.pmi-live.com/ 

tc-events/small-and-advanced-reactor/ 

Online Conference 07.11. – 12.11.2021 

PSA 2021 – International Topical Meeting on 

Probabilistic Safety Assessment and Analysis. 

ANS, Columbus, OH, USA, 

http://psa.ans.org/2021 

Online Conference 15.11. – 16.11.2021 

Nuclear New Builds 2021. 

Prospero Events Group, 

https://www.prosperoevents.com/event/ 

nuclear-new-builds/ 

Hybrid Conference 15.11. – 17.11.2021 

NESTet2021 – Nuclear Education and Training. 

ENS, Brussels, Belgium, 

https://ens.eventsair.com/ 

nuclear-education-and-training 

30.11. – 02.12.2021 

Enlit (former European Utility Week 

and POWERGEN Europe). 

Milano, Italy, 

www.enlit-europe.com 

30.11. – 02.12.2021 

WNE2021 – World Nuclear Exhibition. 

Paris, France, Gifen, 

www.world-nuclear-exhibition.com 

2022 

26.01. – 28.01.2022 

PowerGen International. 

Clarion Events, Dallas, TX, USA, 

www.powergen.com 

06.03. – 10.03.2022 

WM2022 - Waste Management Conference 

X-CD Technologies, Phoenix, AZ, USA 

https://www.wmsym.org/ 

06.03. – 11.03.2022 

NURETH19 – 19 th International Topical 

Meeting on Nuclear Reactor Thermal 

Hydraulics. SCK·CEN, Brussels, Belgium, 

https://www.ans.org/meetings/view-334/ 

24.03. – 25.03.2022 

Nuclear Innovation Conference 2022. 

NRG, Amsterdam, The Netherlands, 

https://www.nuclearinnovationconference.eu/ 

29.03. – 30.03.2022 

KERNTECHNIK 2022. 

Leipzig, Germany, KernD and KTG, 

www.kerntechnik.com 

04.04. – 08.04.2022 

International Conference 

on Geological Repositories. 

Helsinki, Finland, EURAD, 

https://www.ejp-eurad.eu/events/ 

international-conference-geological-repositories 

Postponed to Spring 2022 

4 th CORDEL Regional Workshop – 

Harmonization to support the operation 

and new build of NPPs including SMR. 

Lyon, France, World Nuclear Association, 

https://events.foratom.org 

04.05. – 06.05.2022 

NUWCEM 2022 – 4 th International 

Symposium on Cement-Based Materials 

for Nuclear Wastes. 

Sfen, Avignon, France, 

https://new.sfen.org/evenement/nuwcem-2022 

15.05. – 20.05.2022 

PHYSOR 2022 – International Conference 

on Physics of Reactors 2022. 

ANS, Pittsburgh, PA, USA, 

www.ans.org 

22.05. – 25.05.2022 

NURER 2022 – 7 th International Conference 

on Nuclear and Renewable Energy Resources. 

ANS, Ankara, Turkey, 

www.ans.org 

Postponed to 30.05. – 03.06.2022 

20 th WCNDT – World Conference 

on Non-Destructive Testing. 

Incheon, Korea, The Korean Society 

of Nondestructive Testing, 

https://20thwcndt.com/ 

10.07. – 15.07.2022 

SMiRT 26 – 26 th International Conference on 

Structural Mechanics in Reactor Technology. 

German Society for Non-Destructive Testing, 

Berlin/Potsdam, Germany, 

www.smirt26.com 

04.09. – 09.09.2022 

NUTHOS-13 – 13 th International Topical 

Meeting on Nuclear Reactor Thermal 

Hydraulics, Operation and Safety. 

ANS, Taichung, Taiwan, 

www.ans.org 

NUCLEAR 2021. 

NIA, London, United Kingdom, 

https://events.foratom.org/calendar/ 

nuclear-2021/ 

02.12.2021 

05.04. – 07.04.2022 

GLOBAL 2022 – International Conference 

on Nuclear Fuel Cycle. 

Sfen, Reims, France, 

https://new.sfen.org/evenement/global-2022/ 

This is not a full list and may be subject to change. 

Calendar


First Repatriation of Vitrified 

Reprocessing Waste from Sellafield 

Managing challenges with transport organization 

and radiation protection 


Repatriation from the United Kingdom Reprocessing contracts which the operators of German nuclear 

power plants (EVU) had concluded with the Nuclear Decommissioning Authority (NDA, former BNFL) formed the basis 

for the transport of spent fuel elements to the reprocessing plant at Sellafield, where they were reprocessed. Until 1994, 

the reprocessing of irradiated fuel elements from nuclear power plant operation was mandatory in Germany pursuant 

to § 9 AtG (Atomic Energy Act). Since a national reprocessing concept was dispensed with in 1989, the waste had to be 

transported abroad for reprocessing. From 1994 onwards, an amendment to the Atomic Energy Act made it possible to 

also transport nuclear waste directly to a repository in parallel with reprocessing as a waste management concept. As of 

July 1, 2005, the transfer of irradiated fuel elements to a reprocessing plant is banned. There is an obligation to take 

back the amount of heavy metals produced during reprocessing in accordance with national agreements and the 

corresponding exchange of notes between Germany and the UK. International law states that an amount equivalent to 

the mass of 768 t of heavy metal delivered to Sellafield must be returned to Germany. 

| Fig. 1. 

HLW canister. 

The mass to be returned 

equates to 560 HLW canisters, 

whose return and 

packaging is in turn contractually 

agreed between 

the EVU and the NDA. The 

HLW canister is a waste 

product from the reprocessing 

and takes the form 

of a sealed stainless steel 

cylinder which is filled with 

a glass matrix in which 

dissolved high-level radioactive 

fission products are 

embedded. 

Preparations for the first transport 

from Sellafield 

In preparation for the return of the stream of high-level 

radioactive waste from Sellafield, the design and statutory 

approval of the CASTOR® HAW28M transport and storage 

cask, and the loading planning, were undertaken on the 

basis of the specification of the Sellafield canister and the 

inventory specification in compliance with international 

and national regulations. 

Furthermore, the return of the HLW canisters from the 

UK required the infrastructure for the transport, loading, 

and dispatch to be created, a process which was successfully 

verified by carrying out a cold trial of the cask loading 

and handling operation for the CASTOR® HAW28M cask in 

Sellafield in 2013. 

The first six of a total of 20 empty casks of the design 

CASTOR® HAW28M which were needed – each capable of 

holding 28 canisters – were transported by rail and ship 

from the GNS production plant in Mülheim to Sellafield in 

June 2018. In Sellafield, each was loaded with a total of 

28 HLW canisters in the loading facility of Sellafield 

Limited between December 2018 and November 2019. The 

first return transport from Sellafield to Germany and the 

storage facility of BGZ Gesellschaft für Zwischenlagerung 

mbH at the nuclear power plant site in Biblis (BZB) took 

place in November 2020. 

Challenges with the transport 

organization 

The companies and organizations 

involved in the return transport have 

many years of experience in the 

transport of radioactive materials 

and waste. The means of transport 

employed and the equipment used 

have already been successfully utilized 

several times in comparable projects. 

Until 2011, casks of the type CASTOR® 

HAW28M were already being transported 

by rail from La Hague to 

Gorleben. The same type of cask was 

also used in 2016 to transport vitrified 

waste from Sellafield to the ZWILAG 

in Switzerland. 

Nevertheless, several firsts at the 

same time presented special challenges: 

The transport from Sellafield 

to Biblis was the first repatriation 

from the UK to Germany. For the first 

time in nine years, loaded CASTOR® 

HAW28M casks were transported 

within Germany, and for the first time 

ever, their destination was not the 

central interim storage facility in 

Gorleben, but an interim storage 

facility of a nuclear power plant. 

Coronavirus pandemic 

Whereas the logistical challenges 

were largely fore seeable, manageable, 

and plannable, the coronavirus pandemic 

brought completely new types of complications: The 

transport was originally planned for spring 2020 and all 

necessary preparations had been completed on time. On 

February 27, 2020, the GNS issued the so-called “ Clearance 

for Shipment”. 

But on March 12, 2020, less than a week before loading 

was scheduled to commence in Sellafield, the Federal 

Police regional headquarters in Hanover, which was 

| Fig. 2. 

Cask of the type CASTOR® HAW28M (shown 

in cross section, without shock absorbers). 

9 

FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 

Feature 

First Repatriation of Vitrified Reprocessing Waste from Sellafield ı Marco Wilmsmeier and Michael Köbl


FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 10 

Previous repatriations from France 

Until 2005, spent fuel elements from the operation of German 

nuclear power plants were transported to the United Kingdom and 

France for reprocessing. The radioactive waste produced during 

reprocessing has to be returned to Germany. The GNS Gesellschaft 

für Nuklear-Service mbH has been commissioned by the German 

nuclear power plant operators to prepare and undertake the return 

of this waste to German interim storage facilities. Between 1996 and 

2011, GNS already undertook twelve transports to the central interim 

storage facility in Gorleben, Lower Saxony; these transports involved 

a total of 108 large casks filled with vitrified high-level radioactive 

waste from the reprocessing of German fuel elements in the French 

reprocessing plant in La Hague. The sometimes sizeable protests 

against the transports, extensive discussions on the political level and 

within society, and days of media coverage, resulted in these transports, 

which were generally known as “Castor transports”, becoming 

a symbol of the resistance against nuclear energy. 

responsible for the security of the transport, gave notification 

that the security measures were not justifiable given 

the spread of coronavirus at that time, and hence the 

planning and realization of the transport was suspended 

with immediate effect. 

In the months that followed, extensive coordination 

with all stakeholders, and the police forces involved in 

particular, was necessary to draw up a new transport 

schedule – but still with the proviso that the coronavirus 

risks were manageable, and under much more difficult 

conditions. The pandemic situation led to a new transport 

schedule being agreed in July 2020 for late fall. It was thus 

possible to avoid additional complications resulting from 

the completion of BREXIT at the end of the year, and repeat 

tests which would otherwise have had to be carried out in 

Sellafield on the casks which were already loaded, and the 

shock absorbers which were already mounted. 

In addition to the extensive preparations which had 

already been undertaken for the transport in spring 2020, 

the pandemic brought completely new types of problems: 

A continuous, unbroken hygiene concept had to be 

compiled for the whole transport route from Sellafield 

to the port of Barrow-in-Furness, the sea crossing to 

Nordenham, and the subsequent rail transport to Biblis, 

and agreed with the authorities in both countries. In 

addition to the hygiene measures themselves, which have 

meanwhile become established, the contact between all 

those involved had especially to be kept to a minimum. 

One consequence was therefore that German representatives 

were not allowed to be present at the operations in 

the UK, for example. We had to fall back on local experts 

instead. And the pandemic meant that the police forces in 

particular also had to overcome considerable challenges 

posed by the operational planning along the transport 

route. 

Transport schedule 

On October 2, 2020, GNS was again able to issue the 

“Clearance for Shipment” and this time it was final. Shortly 

after 6 a.m. GMT on October 26, 2020, the first three casks 

left the Sellafield plant heading for the port of Barrow- in- 

Furness. On October 26 and 27, 2020, the casks were 

transferred from the railroad car and onto the transport 

ship, the MV Pacific Grebe. After a sea crossing lasting 

several days, the Pacific Grebe arrived in the port of 

Nordenham on November 2, 2020, under tight security. 

On the very same day, November 2, and the following day, 

| Fig. 3. 

Transferring a CASTOR® HAW28M from the MV Pacific Grebe 

onto a railroad car at the Port of Nordenham. 

November 3, the six casks were again loaded onto railroad 

cars for the last stage of their journey to Biblis. The train 

with its special cars left Nordenham on the evening of 

November 3 and arrived in Biblis without any major 

incidents on the morning of November 4. 

While the transport was in progress, comprehensive 

measurements and tests were carried out to verify the 

proofs which had been provided in advance to obtain the 

authorizations. 

Temperature measurements during the transport 

A condition for the German Federal Institute for Materials 

Research and Testing (BAM) agreeing to the use of the 

transport ship, the MV Pacific Grebe, was proof that heat 

could be safely removed from the closed holds. The 

verification was therefore undertaken by continuously 

recording the surface temperatures of the casks during the 

several days they were at sea. A thermal imaging camera 

was used to verify that the maximum temperature on the 

container surface was 41 °C, and the exhaust air 

temperature was a maximum of 19 °C during the whole 

170 hours of the measurement. 

The safe removal of heat from the holds of the MV 

Pacific Grebe was therefore guaranteed at all times. The 

maximum temperature on the package was below the 

Feature 



| Fig. 4. 

Thermographic image of a CASTOR® HAW28M in the hold 

of the MV Pacific Grebe. 

value calculated in advance and thus significantly below 

the maximum permissible surface temperature of 96 °C. 

Acceleration measurements 

during the various transport stages 

A further measurement requirement originates from the 

9 th amended storage license of Spent Fuel Interim Storage 

Biblis (BZB) regarding the storage of the casks with the 

canisters from Sellafield. To ensure that the leak tightness 

of the sealing barrier for the primary lid complied with the 

specification, it was necessary to verify that the routine 

transport conditions (RBB) were complied with. To this 

end, the acceleration values for each individual cask on 

each stage of the transport were recorded by means of data 

loggers (Moni Log). 

Here as well, it was possible to verify that all stages of 

the transport complied with the RBB, since the acceleration 

of the casks during the transport did not exceed the 

maximum permissible value of 2g in each spatial direction. 

The highest acceleration value on all transport stages was 

0.85 g. 

Radiological Tests 

By far the most comprehensive measuring program 

involved the radiological tests before and during the 

transport. In Sellafield, the ports of Barrow-in-Furness and 

Nordenham, and in Biblis, too, comprehensive radiological 

tests were carried out during the complete transport cycle 

of the six transport and storage casks to verify compliance 

with the requirements of transport legislation and storage 

legislation. 

To safeguard the protection of humans and, as far as the 

long-term protection of human health is concerned, 

the environment against the harmful effects of ionizing 

radiation during the transport and storage of the loaded 

CASTOR® HAW28M, compliance with the relevant limit 

values for the dose rate and surface contamination is also 

checked. The limit values originate from the ADR, 

RID [1, 2] regulations, and the IMDG Code [3]. The 

specifications of § 58 StrlSchV [4] apply to the relocation 

of equipment from nuclear facilities in Germany 

( controlled area), i.e. the (surface) contamination is 

< 0.04 Bq/cm 2 for α-emitters and < 0.4 Bq/cm 2 for 

β/g-emitters. The limit values under storage legislation 

are specified in the technical acceptance conditions of the 

accepting interim storage facility of a nuclear power plant. 

The appropriate limit values which have to be applied to 

the cask or the package (cask with shock absorbers), the 

particular means of transport, and the handling equipment, 

are laid down in the GNS test specifications. 

Execution of the radiological tests 

To document all handling and test steps while the empty 

casks are being transported from Mülheim to Sellafield, 

while they are being loaded and dispatched in Sellafield, 

and also while the loaded casks are being transported from 

Sellafield to Biblis, cask-specific sequence plans including 

test and measurement logs which have to be signed were 

used. 

Prior to carrying out the radiological tests, the calibration 

certificates for the measuring equipment which had 

already been submitted were checked to make sure they 

had not expired. At the start of each radiological test, care 

was taken that the measuring range of the equipment used 

to measure the dose rate was set correctly, and the contamination 

measuring devices were functioning correctly. 

The tests were carried out by qualified radiation protection 

staff. 

Contamination tests: 

All contamination tests were carried out in compliance 

with the limit values specified under transport legislation. 

Protocols which identify representative measuring points, 

and which were laid down on the basis of experience 

gained during the cold trails or previous use of the 

equipment, were used for the measurements. 

The contamination tests performed while the empty 

casks were being transported, and for relocating equipment 

out of the controlled area, served to verify they were contamination 

free according to the limit values under transport 

legislation of < 0.04 Bq/cm 2 for α-emitters or < 0.4 

Bq/cm 2 for β/g-emitters. 

Further contamination tests were carried out on the 

loaded cask and equipment while the cask was being 

loaded and dispatched, and while it was being transported, 

to verify it was contamination free according to the 

limit values under transport legislation of < 0.4 Bq/cm 2 

for α-emitters or < 4 Bq/cm 2 for β/g-emitters. 

While the empty CASTOR® HAW28M casks were 

being transported from the GNS production plant in 

Mülheim to Sellafield, while they were being loaded and 

dispatched in Sellafield, and during the return transport 

to Biblis until the handover to the BZB, the first test 

method used was the wipe test for non-fixed surface 

contamination. A preliminary screening test, if provided 

for, served as an indicative test on the railroad car, for 

example, to prevent contaminated surfaces from being 

14 Screening tests 

29 Wipe tests 

Direct measurements 

at all accessible points 

| Fig. 5. 

Plan of the measuring points for contamination measurements 

on the railroad car before loading/after unloading. 


Feature 




overlooked. Afterward, at each operational site, the 

equipment coming into direct or indirect contact with the 

cask, such as crane traverses, transport and storage racks, 

as well as railroad cars and the decks of the ship, were 

subjected to a direct measurement at all accessible points 

prior to use and checked for fixed contamination. 

While the casks were being loaded and dispatched in 

Sellafield, and during the return transport to Biblis until 

the handover to the BZB, wipe tests were carried out on 

the particular package as well to check for non-fixed 

contamination. 

A total of 270 wipe tests were conducted on the 

casks alone and the contamination measurements were 

assessed. 

Dose rate measurements 

After loading each cask, the g and neutron dose rate was 

measured on the vertical cask at twelve predefined 

measuring points which are highlighted in color and 

spread over the outer surface of the cask. Afterward, the 

average dose rates were calculated taking into account the 

limit values for neutrons of ≤ 250 mSv/h and g+neutrons 

of ≤ 350 mSv/h stipulated in the storage legislation. 

After the cask had been transferred onto a means 

of transport, dose rate measurements were carried out on 

the package as a contact measurement on the basis 

of a predefined plan of the measuring points with 

13 measuring points, and the maximum value was determined. 

The limit value under transport legislation for g+neutrons 

of ≤ 2 mSv/h had to be verified here. 

After the particular cask had been transferred within 

the site from the loading facility to a store to prepare it for 

transport in Sellafield, the g and neutron dose rate was 

measured at ten measuring points on the outer surface of 

the package at a distance of one meter to determine the 

transport index. The transport index is required to rate the 

package category, for the hazardous goods labeling among 

other things. 

Before the loaded casks were transported from the 

dispatching facility in Sellafield in October 2020, further 

dose rate measurements were carried out on at least four 

measuring points in each case at a distance of two meters 

from the outer surface of the transport vehicle after they 

had been transferred onto a means of transport, and after 

changing the mode of transport in Nordenham, to verify 

compliance with the limit value of 100 mSv/h under 

transport legislation. 

While the casks were being loaded and transported, 

540 dose rate measurements (contact, 1 m and 2 m) were 

carried out and assessed. 

Assessment of the measurements 

As expected, the contamination values measured while the 

empty casks were being transported were below the limit 

values under transport legislation, so that this transport 

could be dealt with as a conventional transport. 

The contamination values measured for relocating 

equipment were also below the limit values under 

transport legislation, so that the equipment could be 

returned to its conventional use. 

The contamination measurements on casks and 

equipment during loading, dispatch, return transport until 

the handover to the BZB, showed that the measured values 

were less than 10% of the limit values specified under 

transport legislation of < 0.4 Bq/cm 2 for α-emitters or 

< 4 Bq/cm 2 for β/g-emitters. 

The average g and neutron dose rate was determined 

from the results of the measurement on the loaded casks 

| Fig. 6. 

Dosage measurement at a distance of 2 m from the transport vehicle 

in Nordenham. 

Requirement / result 

Package CASTOR® HAW28M- 

01 02 03 04 05 06 

Proportion of limit value 

(in relation to max. value) 

in % 

Dose rate on surface of cask in acc. 

with storage legislation requirements 

Limit value: g+neutrons < 0.35 μSv/h 0.098 0.099 0.102 0.112 0.108 0.109 32.0 

Dose rate on surface of package in acc. 

with transport legislation requirements 

Limit value: g+neutrons < 2 μSv/h 0.384 0.275 0.342 0.318 0.299 0.383 19.2 

Dose rate at distance of 2 m from outside 

of vehicle in acc. with transport legislation 

Limit value: g+neutrons < 0.1 μSv/h 0.023 0.029 0.025 0.026 0.026 0.028 29.0 

Total activity of cask content in 10 15 Bq 

Maximum permissible: 1,270 × 10 15 Bq 309 325 309 316 324 379 29.8 

| Tab. 1. 

Dose rates and activities of the casks for the transport from Sellafield to Biblis. 

Feature 



required by storage legislation, and recorded. During the 

subsequent check of the calculated dose rates, it was 

confirmed in the presence of experts and GNS that all the 

casks complied with the requirements under storage 

legislation. It was ascertained that the values did not 

exceed 35 % of the maximum permissible limit values. 

The dose rate measurements as a contact measurement 

on the packages showed that the maximum value was 

approx. 20 % of the limit value for g+neutrons of ≤ 2 mSv/h 

under transport legislation. 

The verification of the limit values for the dose rate 

measurements under transport legislation was conducted 

at a distance of two meters from the packages in both 

Sellafield and Nordenham. It was ascertained that the 

maximum value was approx. 30 % of the limit value of 

≤ 100 mSv/h under transport legislation. 

The fact that the values were only a fraction of the limit 

values under transport legislation and storage legislation 

corresponds with the total activity of the HLW canister 

inventory in the respective CASTOR® HAW28M casks 

which were loaded in Sellafield. This is also due to the 

design of the cask shielding and an optimized loading 

plan. 

Conclusion and outlook 

All radiological tests carried out while the empty casks 

were being transported, while the casks were being loaded 

and dispatched in Sellafield, and also while the loaded 

casks were being transported from Sellafield to Biblis until 

the handover to the BZB, prove with the aid of the test 

results determined that during the first HLW return 

transport from the UK to Germany, the limit values under 

transport and storage legislation were safely complied 

with or nowhere near reached, at all locations. The 

measurements of the temperature and the acceleration of 

the casks, which were conducted in addition, likewise 

verified the reliable compliance with all specifications. 

Although the pandemic meant that the measures required 

were much more complex, it was possible to conduct the 

whole transport safely, reliably, and on schedule. The 

experience and insights gained while transporting the 

casks to Biblis form an optimum basis for the two still 

outstanding return transports, both of which involve 

transporting seven federal casks of the type CASTOR® 

HAW28M from Sellafield to the interim storage facilities at 

the Isar and Brokdorf nuclear power plants. 

References 

[1] ADR: European Agreement concerning the International Carriage of Dangerous Goods by Road 

(Accord relatif au transport international des marchandises Dangereuses par Route), in the 

version valid as of January 1, 2021 

[2] RID: Regulation concerning the International Carriage of Dangerous Goods by Rail (Règlement 

concernant le transport International ferroviaire des marchandises Dangereuses), in the version 

valid as of January 1, 2021 

[3] IMDG Code: International Maritime Code for Dangerous Goods, 2020 edition including 

Amendment 40-20 

[4] § 58 StSchV – Radiation Protection Act (Verordnung zum Schutz vor der schädlichen Wirkung 

ionisierender Strahlung – Strahlenschutzverordnung): Verlassen von und Herausbringen aus 

Strahlenschutzbereichen, Artikel 1 V. v. 29.11.2018 BGBl. I S. 2034, 2036 (Nr. 41); 

zuletzt geändert durch Artikel 6 G. v. 20.05.2021 BGBl. I S. 1194 

Authors 

Marco Wilmsmeier 

GNS Gesellschaft für Nuklear-Service mbH, Germany 


Marco Wilmsmeier studied bilaterally at the Gelsenkirchen University of Applied 

Sciences and Sheffield Hallam University, UK, waste management technology and 

environmental engineering. He graduated from Prof. Malcom Denman in Sheffield 

in 2000. Until mid-2001 he was employed as technical operations manager at SCA 

Packaging in Hövelhof before he started as Project Manager for loading service and 

reprocessing transports at GNS. In 2004 he changed to the department of recycling 

of reprocessing waste within the GNS and since then has been responsible as 

Project Manager for HAW recycling from the UK. 

Michael Köbl 

GNS Gesellschaft für Nuklear-Service mbH, Germany 

Michael.Koebl@gns.de 

| Fig. 7. 

Arrival of the MV Pacific Grebe at the Port of Nordenham. Transferring a 

CASTOR® HAW28M from the MV Pacific Grebe onto a railroad car. 

After completing his business administration studies, Michael Köbl started in 2005 

as Head of Internal Communication at Grundig AG in Fürth. In 2008 he joined the 

Public Relations department at Dynamit Nobel AG in Troisdorf, which he was 

leading since 2014. Since 2006 he is employed at GNS in Essen, initially as a press 

and public relations officer and since 2009 as Head of Communications. 

Feature 



14 

INTERVIEW 

“EU Member States Can Choose Their 

Energy Sources and Can Include 

Nuclear in Their Energy Mix as Part of 

Their Effort to Achieve Decarbonisation 

and Carbon Neutrality by 2050.” 


ı Deputy Director-General of DG Energy 

Massimo Garribba 

Deputy Director-General of DG Energy 

Originally from Padova (Italy), Dr Massimo Garribba is a 

qualified electronic engineer with a PhD in Informatics and 

Industrial Electronics. After spending seven years on fusion 

research, he joined the European Commission in 1995, 

originally working on digital issues. In 2004 he moved to the 

Commission’s Directorate-General for Energy, working on 

Euratom coordination and international relations, first as 

Head of Unit, then as Director with a broader remit. In July 

2020, he was promoted to Deputy Director-General 

responsible for the coordination of all aspects of EURATOM 

policy. 

The issue of climate change and the reduction of 

greenhouse gases has become a very prominent 

political priority for the EU, the member states and 

many other nations. At the same time, the EU is 

developing a sustainable investment framework as 

a front-runner among the major economies. Has 

the Commission finally reached a position on 

nuclear and the taxonomy after more than two 

years of controversial debates? 

Not yet. The Taxonomy Regulation reflects a delicate 

compromise on the question of whether or not to include 

nuclear energy in the EU taxonomy. While nuclear energy 

is consistently acknowledged as a low-carbon energy 

source, opinions differ notably on the potential impact on 

other environmental objectives, such 

as the environmental impact of 

nuclear waste. 

The Commission considers that the 

credibility of this assessment is 

crucial. It has requested the Joint 

Research Centre (the Commission’s 

internal scientific service) to draft a 

technical report on the ‘do no 

significant harm’ aspects of nuclear 

energy. 

This is one step in the process. The JRC report was 

reviewed by experts on radiation protection and waste 

management under Article 31 of the Treaty establishing 

the European Atomic Energy Community (Euratom 

Treaty), as well as by experts on environmental impacts 

from the Scientific Committee on Health, Environ mental 

and Emerging Risks. The experts’ reviews are published on 

the Commission’s website. 1 

The Commission is carefully analysing the findings of 

the report, the reviews by the experts, as well as all of the 

other extensive feedback submitted by other interested 

stakeholders. 

The Commission will complete its assessment in prompt 

fashion and will follow up in accordance with the steps laid 

out in the Commission’s Communication on a Strategy 

for Financing the Transition to a Sustain able Economy 2 

published in July 2021. 

As stated in the afore-mentioned Communication, 

the Commission will adopt a complementary Climate 

Taxonomy Delegated Act covering activities not yet covered 

in the first EU Taxonomy Climate Delegated Act, notably 

certain energy sectors, in line with the requirements of the 

Taxonomy Regulation. 

The complementary 

Delegated Act will also 

cover nuclear energy activities, 

subject to and consistent 

with the specific expert review 

process that the Commission 

set out for this purpose. 

The complementary Delegated 

Act will also cover nuclear energy 

activities, subject to and consistent 

with the specific expert review 

process that the Commission set out 

for this purpose. The Commission 

will adopt the complementary 

Delegated Act as soon as possible 

after the end of the specific review 

process focused on nuclear. 

1 EU taxonomy for sustainable activities | European Commission (europa.eu). 

2 https://ec.europa.eu/info/publications/210706-sustainable-finance-strategy_en 

Interview 

“EU Member States Can Choose Their Energy Sources and Can Include Nuclear in Their Energy Mix as Part of Their Effort to Achieve Decarbonisation and Carbon Neutrality by 2050.” ı Massimo Garribba


Setting aside the Commission proposal in a 

complementary Delegated Act, what will in your 

opinion be the consequences for the European 

nuclear industry in either case, i. e. inclusion in or 

exclusion from the taxonomy? 

It will have an influence on the range of (private) financing 

options available to nuclear operators to finance their 

projects. 

An activity that does not qualify as a green economic 

activity under the EU Taxonomy is not necessarily 

unsustainable, given the need to make a ‘substantial 

contribution’ to one of the six objectives and do no 

significant harm (DNSH) to the other five. 

The EU Taxonomy is designed to guide market 

participants in their investment decisions, and it does not 

prohibit investment in any activity. There is no obli gation 

for companies to be Taxonomy- aligned. 

Financial market participants can choose to invest 

in companies that carry out activities that have different 

degrees of environmental performance, including 

activities that do not comply with the EU Taxonomy 

criteria. 

In addition, the EU Taxonomy is a dynamic tool and 

it will evolve as technologies advance and as our 

understanding of solutions progresses. The Taxonomy 

Regulation is open to review every three years and may allow 

the inclusion of activities not covered initially, provided 

technological progress allows for market entry in the near 

future. 

If nuclear in the end would not be considered as a 

sustainable investment, could there be and should 

there be a compensation mechanism in the Euratom 

framework for those member states that want to 

pursue nuclear power for decarbonization, say in 

form of financial assistance? 

EU Member States can choose their energy sources and 

can include nuclear in their energy mix as part of their 

effort to achieve decarbonisation and carbon neutrality by 

2050. 

The European Commission, in line with the 

Euratom Treaty, supports actions to improve the safety of 

nuclear installations, including 

research on safety, security, waste 

management, and assistance on 

nuclear decommis sioning. However, 

it does not provide financial 

support for the construction of 

new nuclear fission power plants. 

Member States opting for nuclear 

energy will of course underline that 

nuclear is a cleaner, more affordable 

and much more reliable source of 

energy than imported fossil fuels. 

Apart from the well-known trenches between the 

ardent opponents and supporters of nuclear power 

among the member states, there are remarkable 

developments going on in countries that have no 

or a low profile on nuclear. Are there Euratom 

policies to eventually support newcomer countries 

like Poland or countries that might endeavour a 

major expansion of their nuclear sector such as 

possibly the Netherlands? 

The EU Treaties leave the choice of the energy mix to the 

individual Member States. 

Based on this and the Euratom Treaty, it is the Member 

State which takes the decision to introduce nuclear power 

in its energy mix and bears the ultimate responsibility for 

its safety and security. 

Once such decision is taken, the Member State must 

fully comply with all relevant provisions of EU rules, i.e. all 

primary and secondary legislation on inter alia nuclear 

safety, radiation protection, management of spent fuel and 

radioactive waste, and non-proliferation. 

The European Commission can certainly advise 

Member States in navigating the different requirements. 

The Commission’s primary role is to ensure that all 

Member States effectively fulfil their obligations. To 

this end, the Commission will readily provide guidance to 

any interested Member State in ensuring compliance 

with EU legislation from the first preparative steps at 

national level. 

Since January, we can observe a major surge in 

energy prices, gas, coal, electricity and the carbon 

price all over Europe. The issue did not pop up in 

the German election campaign, but features 

prominently e.g. in Italy, France and also outside 

the EU in the UK. How might this impact the debate 

on climate policy, energy and nuclear power? 

The rise in wholesale energy prices in the EU since the 

summer is a big issue for the Commission, as it is also 

impacting consumers and companies at this sensitive 

moment of recovery after the Covid-19 pandemic. 

The Commission has published a ‘toolbox’ 3 document 

on 13 October, highlighting the various short and 

medium-term options available to national governments 

to ease the burden on end-users – in particular the most 

vulnerable consumers. 

The surge in prices is driven by an unfortunate 

constellation on the gas market – increased global demand 

at a time of limited supply. But latest indications are that 

the effects will be relatively short-term (with the market 

ex pected to be much more balanced by the spring). 

One key message from the Commission in this 

context is that this should not in any way put the transition 

to clean energy in question. On the contrary, greater 

investment in renewables and energy efficiency measures 

will reduce the EU’s dependence on imported fossil fuels 

and enhance our energy security, and therefore limit the 

chances of such a market spike being repeated. 

Member States opting for 

nuclear energy will of course 

underline that nuclear is a cleaner, 

more affordable and much more 

reliable source of energy than 

imported fossil fuels. And I note 

the declaration to this effect by 

10 Member States in early October, 

which in the context of rising energy prices have 

once again called to include nuclear energy in the EU 

Taxonomy to ensure energy inde pendence and security of 

supply. 

As regards the competitiveness of nuclear energy 

in the longer term, it is dependent on several 

factors, from technological development to policy 

and regulatory frameworks, sustain able supply chains, 

but also market design, business models, financing 

instruments, etc. 

Innovation will certainly be key, if nuclear is to 

make a larger contribution to our carbon- neutral 

future – both in making it competitive and keeping on 

implementing the con tinuous safety improvement 

principle. 

INTERVIEW 15 

3 https://eur-lex.europa.eu/legal-content/EN/TXT/?uri=COM%3A2021%3A660%3AFIN&qid=1634215984101 

Interview 



INTERVIEW 16 

Building a hydrogen ecosystem 

Participation in the Clean Hydrogen Alliance 

500 companies 

2020 

Focus on renewable hydrogen 

6 GW clean hydrogen 

2024 

1000 companies 

2024 

2000 companies 

2050 

40 GW (EU) + 40 GW (non-EU) clean hydrogen 

2030 

would be part of a carbon-free European power system 

largely based on renewables. 

In presenting on 13 October the Commission Toolbox 

of possible measures to address the energy price 

spike, Commissioner Simson underlined that we need to 

ensure a well-functioning and more resilient gas 

market framework. She underlined that gas has a 

role in the transition, but at the same time its 

contribution is bound to change in the long term. It will 

have to become green. And to promote this process, 

the Commission will present before the end of the year a 

comprehensive legislative package to decarbonise our 

gas and hydrogen markets by 2050. As part of this 

package, we will be addressing the issues of security of 

supply and storage. 

Finally, the Commission will launch new actions to 

improve the resilience of critical energy infrastructure to 

new evolving threats. This will include new rules for the 

cybersecurity of cross-border electricity to be published 

next year. 

Emission reduction potential in industries 

min. 9 million tons per year 

2024 

Investment needs in renewable hydrogen electrolysers: 

€ 5-9 billion 

2024 

Next to the price issue and its impact on industrial 

competitiveness there is also the aspect of security 

of supply. Many states are pursuing a coal phaseout, 

Germany and Belgium are phasing out nuclear, 

France, Sweden and Switzerland reduced nuclear 

capacity, the Netherlands end domestic gas production 

and gas storage is running low in Europe 

and particularly Germany for the coming winter. Is 

it time to complement European climate policy 

with a security of energy supply policy beyond 

current Energy Union policies? 

It is true that within the EU we see different policies among 

Member States as regards the role of nuclear power in their 

energy mix. Whereas 

The analyses conducted as part of the 

Commission's 2050 Long-Term Strategy 

confirm that in practically all models and 

under all scenarios nuclear energy would 

be part of a carbon-free European power 

system largely based on renewables. 

min. 90 million tons per year 

2030 

€ 26-44 billion 

2030 

several Member States 

have expressed their 

long-term commitment 

towards nuclear energy 

for ensuring security of 

energy supply and 

meeting climate targets, 

others have decided 

or are considering 

phase-out or cut-down policies of their nuclear 

programmes in the coming decades. 

The analyses conducted as part of the Commission’s 

2050 Long-Term Strategy confirm that in practically 

all models and under all scenarios nuclear energy 

Apart from sustainable finance there are other 

European policies that impact the nuclear sector, 

such as the Energy System Integration and the 

Hydrogen Strategies, the Guidelines on State aid 

for environmental protection and energy and the 

Industrial Strategy. What will we see in Commission 

initiatives here in the next 18 months and 

what will be at stake for the nuclear sector in these 

policies? 

The main energy policy initiatives that the Commission 

will be pursuing in the next 18 months are related to the 

overall ambition of the European Green Deal, following 

on from the proposals to revise energy efficiency and 

renew able energy rules already tabled in July and energy 

infrastructure (TEN-E) where the inter-institutional 

negotiations are very 

much advanced and 

agreement could be 

reached before the end 

of this year. Before the 

end of the year there 

will also be proposals 

to look at decarbonising 

Hydrogen production can 

also be an important 

element for countries 

considering nuclear 

energy in sector coupling. 

the gas market and establishing a hydrogen market, and a 

revision of the Energy Performance of Buildings Directive. 

The Commission will also publish legislative proposals for 

reducing methane emissions. The priority next year will 

primarily be negotiating these proposals with the Member 

States and the European Parliament. 

While recognising that the electricity system of the 

future will be largely based on renewables, the Energy 

System Integration Strategy does not close the door to the 

contribution of other zero emission generation options, 

such as nuclear – recognising the preferences and 

specificities of Member States in this regard. 

Nuclear energy can complement renew able energy 

sources in the integrated energy systems. 

Hydrogen production can also be an important 

element for countries considering nuclear energy in 

sector coupling. 

As mentioned before, the Commission’s over arching 

priority is to ensure that Member States choosing to use 

nuclear energy do so within the applicable Euro pean legal 

framework, meeting the highest standards on nuclear 

safety, on safe and responsible radioactive waste management 

and on radiation protection. 

Interview 



The path towards a European hydrogen eco-system step by step : 

Today - 2024 2025 - 2030 

2030 - 

H 2 

INTERVIEW 17 

From now to 2024, we will 

support the installation of 

at least 6GW of renewable 

hydrogen electrolysers in 

the EU, and the production 

of up to 1 million tonnes of 

renewable hydrogen. 

From 2025 to 2030, 

hydrogen needs to become 

an intrinsic part of our 

integrated energy 

system, with at least 40GW 

of renewable hydrogen 

electrolysers and the 

production of up to 

10 million tonnes of 

renewable 

hydrogen in the EU. 

From 2030 onwards, 

renewable 

hydrogen will be 

deployed at a large 

scale across all 

hard-to-decarbonise 

sectors. 

The most immediate energy project is the Fit for 

55 package on implementing the Green Deal till 

2030. Despite the long and short term decarbonization 

ambitions of the EU and member states we 

see practical policies running in the opposite 

direction such as the German phase-out of nuclear 

power with insufficient compensation by low carbon 

power sources and the new Belgian policy of 

replacing nuclear with gas fired power plants. Will 

we see some kind of Maastricht mechanism on 

climate policy or a European carbon semester in the 

future? 

I can’t immediately see an appetite within the Commission 

– or among EU leaders – to re-open the EU treaties and 

redefine Article 194 of the treaty. 

Already with the National 

Nuclear energy 

can complement 

renewable energy 

sources in the 

integrated energy 

systems. 

Energy and Climate Plans 

(NECP), which were introduced 

in the Clean Energy for 

all Europeans package, the 

EU has introduced a level of 

transparent forward planning 

that has not previously 

been seen. 

The key point here is 

that our absolute priority is to reduce greenhouse gas 

emissions by 55% by 2030. And this is being addressed 

in the Fit for 55 package. The more ambitious targets 

proposed for the new Directives on Energy Efficiency and 

Renewable Energy will require legally-binding commitments 

from Member States – and a monitoring process. 

But there is no one-size-fits-all approach. 

The Commission can also play a role in accompanying 

the EU nuclear industry in improving its competitiveness 

and better integrating the EU energy system of the future, 

by securing the application of the highest safety standards 

and supporting the regulatory processes in EU Member 

States opting for nuclear energy. 

Author 


Head of Media Relations and Political Affairs 

KernD (Kerntechnik Deutschland e.V.) 

nicolas.wendler@kernd.de 

Interview 



18 

DECOMMISSIONING AND WASTE MANAGEMENT 

Investigations of the Tailskin Seal 

During the Retrieval Concept ‘Shield 

Tunnelling with Partial-face Excavation’ 

in the Asse II Mine 

Birte Froebus 

Introduction The Asse II mine is a former salt mine near Remlingen where potash salt was mined from 1909 to 

1925 and rock salt from 1916 to 1964. After salt mining ceased, about 126,000 containers of low- and intermediate- level 

radioactive waste were stored between 1967 and 1978. Due to its high degree of excavation, the mine workings were 

damaged by pressure from the overburden. Saturated access water is penetrating through the cracks that have 

developed. In addition, the stability of the mine is deteriorating. The radionuclides from the present damaged waste 

packages could be transported into the environment via resulting pathways and transport media. 

Due to the situation described above, 

long-term safety can only be guaranteed 

by retrieving the radioactive 

waste according to the current state of 

knowledge, which is why this has 

been stipulated by law since 2013. The 

adopted law ‚Lex Asse‘ creates an 

important legal basis for the retrieval 

of radioactive waste. Through simplified 

procedures and the possibility of 

carrying out work in parallel, the law 

makes it possible to accelerate the 

work. In addition, the public’s right 

to comprehensive information is 

strengthened. Despite all this, it 

should be mentioned that no practical 

experience has been gained so far for 

such retrieval from deep geological 

formations. In the event that the 

deformations caused by the rock 

pressure and the resulting solution 

influx exceed a manageable level and 

the mine has to be abandoned, there 

are explicit emergency measures. In 

this scenario, among other things, a 

targeted counterflooding of the entire 

caverns is initiated – in this case, the 

radioactive waste must remain in the 

mine and could not be retrieved. [1] 

The radioactive waste in the Asse II 

mine is placed in a total of 13 ‘emplacement 

chambers’ (German: Einlagerungskammer 

= ELK) on three 

levels at depths of 511 m, 725 m and 

750 m, which are shown in red in 

Figure 1. ELK 8a/511 is located on the 

511-m-level and ELK 7/725 on the 

725-m-level. The other eleven emplacement 

chambers 1/750, 2/750, 

2/750 Na2, 4/750, 5/750, 6/750, 

7/750, 8/750, 10/750, 11/750 and 

12/750 are located on the 750-mlevel. 

[1] 

In this article, several terms related 

to mining are used to describe the 

mine. The emplacement chambers are 

located on several levels, so to say the 

| Fig. 1. 

Emplacement chambers on different levels in the mine workings of the Asse II mine [1]. 

floors of the mine. Between the 

chambers there are so-called pillars, 

i.e. the walls in the vertical direction, 

and the so-called roofs, i.e. the mighty 

ceilings in the horizontal direction. 

The pillars and the roofs are the 

remaining rock after the chambers 

have been created, which now 

represents the load-bearing structure 

in the mine building. Analogous to 

tunnelling, within the chambers the 

bottom is called the ‚floor‘ and the 

ceiling is called the ‚roof‘. 

Each of the three levels offers a different 

spatial situation, resulting in 

fundamentally different strategies for 

retrieval: 

p ELK 7/725 is still accessible and is 

currently used as a storage facility 

for radioactive waste produced 

during operations. In addition, 

low-level radio active material is 

stored there and the chamber is 

continuously ventilated. Therefore, 

the retrieval of this single 

emplacement chamber is to be 

started as planned. For this 

purpose, a ‚Long-front construction 

method with vertical excavation 

direction‘ is aimed at, which 

is based on mining extraction 

methods. For the recovery of the 

casks, i.e. for un covering, loosening 

and loading; floor-operated 

technology is to be used. Transport, 

on the other hand, is to be 

carried out using decoupled 

ceiling-operated technology. [1] 

p In ELK 8a/511, almost exclusively 

casks with intermediate-level 

radio active contents were stored. 

This is also the reason for the 

special feature of this emplacement 

chamber: a charging 

chamber on the 490-m-level above 

the emplacement chamber, from 

which the casks were lowered 

into the emplacement chamber. 

Assuming that the emplacement 

chamber can be driven on the floor, 

only floor-based technology is to be 

used. [1] 

p The 750-m-level contains the 

majority volume of the radioactive 


Investigations of the Tailskin Seal During the Retrieval Concept ‘Shield Tunnelling with Partial-face Excavation’ in the Asse II Mine ı Birte Froebus


waste, with eleven emplacement 

chambers containing low-level 

radioactive waste (see Figure 2). 

Due to the different rockmechanical 

boundary conditions, the 

chambers are divided into three 

groups here. Here, the roofs in 

particular play a decisive role. In 

the central part of the mine 

buildings lies ELK 2/750 Na2, 

directly below ELK 7/725 and a 

roof of 6 m thickness. This roof is 

considered intact for retrieval, as 

ELK 7/725 will have been cleared 

and backfilled by that time. The 

roofs of the emplacement chambers 

of the chamber group ‚East‘ 

(ELK 1/750, 2/750, 12/750) are 

considered intact, as there are no 

voids above them on the 725-mlevel. 

The remaining seven 

emplacement chambers of the 

chamber group ‚South‘ are closer 

to the overburden of the southern 

flank and probably have unstable 

roofs. In contrast to the 511-m- and 

725-m-levels, there is no final 

planning for the 750-m-level, 

instead there are several different 

approaches. On one hand, there 

are various possibilities to 

approach the chambers individually, 

on the other hand, there is a 

variant to approach the chambers 

one after the other along a curve. 

In the latter variant, only the 

chamber groups ‚South‘ and ‚East‘ 

can be taken into account, and a 

study has already been carried out 

on this issue. [1] In the further 

course of the article, only aspects 

of this retrieval concept will be 

dealt with. 

Retrieval of the chamber 

groups ‚South‘ and ‚East‘ 

(750-m-level) by means of 

‘Shield tunnelling with 

partial-face excavation’ 

In the ‘Study on the Suitability and 

Development Needs of Equipment 

and Tools for Use in the Asse II Mine’ 

in 2015, shield tunnelling with 

partial- face excavation was already 

examined as a possible retrieval 

method. Shield machines are normally 

used in mechanised tunnel 

construction and are characterised by 

the fact that the entire processes 

take place in the protection of an 

enveloping shield. In Figures 3, 4 and 

5, the shield is shown in green, while 

the rear part of the shield, the 

so-called tailskin, is shown in grey. 

Partial-face excavation in particular 

is used for very easily detachable rock 

in stable rock, short heading lengths 

| Fig. 2. 

Ground plan of the construction method in relation to the emplacement chambers of the 750-m-level [2]. 

| Fig. 3. 

Conveying route of the transport container from the tunnel face, to the excavation chamber and to the 

disposal tunnel [2]. 

and non-circular cross-sections. One 

advantage of partial-face excavation 

is the good accessibility to the tunnel 

face. The tunnel face describes the 

material in front of the excavation 

tools of the machine and is located on 

the left in Figures 3 and 4. This 

accessi bility is required when there is 

a high probability of encountering 

obstacles. [3] 

It can be seen in the site plan of 

Figure 2, the emplacement chambers 

of the chamber groups ‚South‘ and 

‚East‘ are located next to each other 

along a curve on the 750-m-level. For 

this reason, five parallel tunnelling 

routes were prioritised in the concept 

‚Shield tunnelling with partial-face 

excavation‘, which encompasses the 

different widths of the emplacement 

chambers and the pillars in between 

in their total width and run along this 

curve in their length. Five partial-face 

excavation machines can drive up the 

drifts with a time delay over different 

distances from east to west and 

completely clear the similarly high 

emplacement chambers by taking into 

account an overprofile. The overprofile 

is necessary to completely 

cover and remove the damaged and 

possibly contaminated inner sides of 

the emplacement chambers. Assembly 

and disassembly caverns must be 

constructed at the beginning and end 

of each driveway. 

A partial-face excavation machine 

therefore alternately passes the pillars 

between the emplacement chambers, 

which are made of salt rock, and the 

contents of the emplacement chambers 

themselves. The contents of the 

emplacement chambers and their 

structure vary from chamber to 

chamber. 

The waste packages were tipped as 

well as stacked horizontally or vertically, 

depending on the storage or 

discharge technique. Salt was blown, 

tipped or dumped to bed and shield 

the radioactive waste packages. Both 

partial and full backfills were produced 

with the crushed salt, also to 

stabilise the surrounding rock. Due to 

convergences as well as post-fractures 

of the surrounding salt rock, the 

solution influxes that occurred over 

the past decades and the resulting 

corrosion phenomena on the containers, 

the conditions within the 

emplacement chambers are unclear. 

The corroded casks themselves may 

DECOMMISSIONING AND WASTE MANAGEMENT 19 





also be deformed, burst as well as 

shattered, and it is assumed that there 

is a matrix of (possibly solidified) 

crushed salt around the casks. [2] 

Partial-face excavation 

machine 

At the tunnel face (left in Figure 3) 

the salt rock of the pillars or the 

packages inside the emplacement 

chambers can be loosened and lifted 

with the help of the tools. A vacuum 

can be created between the tunnel 

face and the machine to prevent 

contamination from being carried 

into the interior of the machine via 

radioactive particles in the air. [2] 

The driving of the emplacement 

chambers and the pillars creates a 

cavity that has to be backfilled behind 

the machine. For this purpose, 

sorel concrete is placed both as 

shotcrete and as extruded concrete 

(see Figure 5). In this way, the rock 

can be stabilised and at the same time 

a structure can be created on which 

the partial-face excavation machine 

can support itself for further travel. 

In this so-called ‚cavity filling‘ (see 

| Fig. 4. 

Rear view of the partial-face excavation machine with cavity filling [2]. 

Figure 4), two tunnel tubes are cut 

out for the period of retrieval. After 

suitable repacking for transport, the 

materials can be removed through the 

larger tunnel tube (disposal tunnel) 

(see also Figure 3), the smaller tube 

serves as access for personnel and for 

the transport of working materials 

( passenger and utility tunnel). After 

successful retrieval and salvage of the 

partial-face excavation machines, the 

two tubes can be backfilled. Both the 

cavity filling and the backfilling of 

the tunnel tubes (transportation 

tunnels) are made of sorel concrete. 

[2] Sorel concrete has already been 

used for the production of flow 

barriers and various backfillings in the 

Asse II mine. The main components 

of sorel concrete are magnesium 

oxide as a binder and crushed salt as 

aggregate, which are mixed with 

magnesium chloride solution. [4] 

Once the tunnel face has been 

completed, the machine can support 

itself against the cavity filling by 

means of thrust cylinders and move 

forward. When advancing, the 

annular gap that becomes free behind 

| Fig. 5. 

Internals and machine equipment of the partial-face excavation machine, e.g. thrust cylinders and 

( inner) tailskin seal [2]. 

the tailskin can be filled with grout 

immediately. At the end of the shield 

shell, an (inner) tailskin seal is provided, 

which seals the interior of the 

machine against the annular gap. [2] 

In Figures 3, 4 and 5, the thrust 

cylinders in the rear part of the 

machine (tailskin) are shown in red, 

the tailskin seal in blue. 

In the course of the in-depth 

investigations, an outer tailskin seal 

was added to the partial-face excavation 

machine designed in the study. 

This seals the steering gap against the 

annular gap, and prevents material 

from the annular gap from getting 

around the shield or up to the tunnel 

face (see Figure 6). This ensures that 

the annular gap is filled conclusively 

under pressure. This is particularly 

important in the area of the pillars 

between the emplacement chambers, 

as this minimises contamination 

carry-over between the chambers and 

the barrier effect of the pillars can be 

maintained to the maximum. 

Differentiated consideration 

of the tailskin seal 

Due to the technical and legal 

framework conditions existing in a 

repository, it is necessary to mirror all 

components and systems of a shield 

machine against the resulting special 

requirements, to evaluate their 

applicability and functioning, and to 

make adjustments if necessary. 

For the investigations of the two 

tailskin seals, a comprehensive catalogue 

was compiled in each case, in 

which various loads play an essential 

role. The mechanical loads on both 

seals are exceptionally high due to the 

square cross-sectional shape of the 

shield machine. This is intensified by 

damage in the roofs, which can lead to 

loosening on the shield skin of the 

machine, and thus also on tailskin 

seals. In addition, chemical stresses 

occur, mainly from the liquid sorel 

concrete and saturated sodium 

chloride solution. The influence of 

ionising radiation on the tailskin seals 

is considered negligible due to the low 

activity inventory. 

Based on this catalogue, conventionally 

used tailskin seals were 

evaluated with regard to their 

suitability for shield tunnelling in the 

Asse II mine. Different aspects have to 

be given priority in the requirements 

for the respective tailskin seal. The 

inner tailskin seal must primarily fulfil 

a barrier effect towards the trailing 

annular gap, usually by building up a 

high pressure difference between the 

areas to be separated. The reason for 




| Fig. 6. 

Schematic of an inner and outer tailskin seal in partial-face excavation machine with description of the 

steering gap and annular gap [own figure]. 

this is the classification of radiation 

protection areas according to §52 of 

the Radiation Protection Ordinance 

(German: Strahlenschutzverordnung 

= StrlSchV), which defines the 

annular gap as a ‚exclusion area‘, 

while the interior of the machine 

represents a ‚controlled area‘ (see 

Figure 6). For the outer tailskin seal 

the focus is on high wear resistance to 

obstacles outside the machine. 

As a result of the investigations, an 

inner tailskin seal was first designed 

which combines various conventional 

design elements. The basis for this is a 

wire brush seal with external spring 

plates and three sealing chambers, 

which is modified with a lubricant 

compression line and enlarged sealing 

chambers. With the help of the 

pressurised tail seal compound, this 

can meet the high demands on the 

barrier effect. In addition, the spring 

plates at the rear of the annular gap, 

in combination with the lubricant 

used, make it easier to start up the 

machine, as brushes in contact with 

the liquid backfilling (grout inside the 

annular gap) would stick together and 

be damaged over time. The lubricant 

itself also protects against the aggressive 

chemical environment in the 

adjacent annular gap. The enlarged 

sealing chambers increase the protection 

of the sealing system with a 

kind of spring effect – both in case of 

post-fractures from the rock mass and 

in case of occurring convergences. 

Subsequently, the outer tailskin 

seal was designed from three rows of 

spring plates, which are also modified 

with lubricant injection lines. The 

arrangement of three seal rows 

ensures sufficient redundancy, which 

is why the lubricant supply was also 

designed redundantly, so that the 

outermost seal row always has an 

associated supply line. In order to 

prevent the spring plates from being 

torn off by obstacles – such as barrel 

bundles from the emplacement 

chambers or rock fragments from the 

surrounding salt rock – their bearing 

was embedded in the shield casing 

and thus protected. The number of 

seals in a row also ensures a certain 

redundancy here. 

Conclusion 

Within the scope of the investigations, 

comprehensive findings were 

obtained for the problems occurring 

during shield tunnelling in the Asse II 

mine. So far, not all components have 

been investigated in a differentiated 

manner. 

In further investigations, for 

example, aspects such as the interactions 

between the shield machine 

and the surrounding salt rock or the 

synergies and disadvantages of the 

parallel tunnel routes could be 

determined. It has been shown that 

the various situations of the partialface 

excavation machine with 

changing framings – such as salt rock, 

already hardened sorel concrete from 

preceding machines or contents of the 

emplacement chambers – represent a 

special challenge. Especially when 

feeding from cavity filling, the 

stability of the lateral walls along the 

travel distances plays an important 

role. Against this background, for 

example, the sequence of the partialface 

excavation machines would have 

to be adjusted so that the first machine 

starts in the direction of travel on the 

left in order to be able to support itself 

against the salt rock in the south – 

especially in the area of the emplacement 

chambers (see Figure 2). The 

grouting (of the annular gap) in the 

area of the emplacement chambers 

could also be considered in more 

in-depth investigations. A further 

aspect would be the compatibility of 

the tunnelling concept with the 

emergency planning for a possible 

flooding of the Asse mine. At present, 

the reduced barrier effect of the 

drilled-through pillars between the 

emplacement chambers as well as the 

volume of the five shield machines 

and their underground infrastructure 

represent significant disadvantages of 

shield tunnelling with partial-face 

excavation. 

The retrieval of radioactive waste 

from the Asse II mine is scheduled to 

begin in 2033, in this sense: Glück 

auf! 

References 

[1] Plan zur Rückholung der radioaktiven Abfälle aus der 

Schachtanlage Asse II – Rückholplan, Verfasser: 

Bundesgesellschaft für Endlagerung mbH Peine/Remlingen, 

Stand: 19.02.2020 

[2] Machbarkeitsstudie für die Methode „Schildvortrieb mit 

Teilflächenabbau“ – Studie zur Eignungsfähigkeit und zum 

Entwicklungsbedarf von Gerätschaften/ Werkzeugen für den 

Einsatz in der Schachtanlage Asse II, Herrenknecht AG im 

Auftrag des Karlsruher Instituts für Technologie (KIT), 

Technologie und Management des Rückbaus kerntechnischer 

Anlagen (TMRK), Schwanau, Karlsruhe; Stand: 13.05.2015 

[3] Maidl, B., Herrenknecht, M., Maidl, U., Wehrmeyer, G. 2011. 

Maschineller Tunnelbau im Schildvortrieb. Ernst & Sohn 

[4] Schachtanlage Asse II: Nachweis der Langzeitbeständigkeit 

für den Sorelbaustoff der Rezeptur A1; ERCOSPLAN, TU BAF 

( IFAC), IFG; Stand: 10.08.2018 

Author 

Birte Froebus 

Research Assistant at 

Karlsruhe Institute of 

Technology, Karlsruhe, 

Germany 

birte.froebus@kit.edu 

Birte Froebus studied civil engineering at the 

Karlsruhe Institute of Technology (KIT), specialising in 

geotechnical engineering. In her master thesis, she 

took the opportunity to combine mechanised tunnel 

construction with the retrieval of radioactive waste, 

for which she was awarded the WiN-Germany 

Prize 2020/21. She then started working as a 

research assistant at the Institute for Technology 

and Management in Construction (TMB) in 

the ’ Deconstruction and Decommissioning of 

Conventional and Nuclear Buildings‘ department. 





22 

ENVIRONMENT AND SAFETY 

Practical Response to a Dirty Bomb 

James Conca 

Introduction The global community has recognized that a class of weapons known as radiation dispersal devices 

(RDDs), or dirty bombs, pose a grave threat to the United States and the European Union. Dirty bombs use conventional 

methods, such as a car bomb, to disperse radioactive materials in a populated economic district to cause great economic 

and social disruption disproportionate to their actual radiological effects and well beyond the physical destruction from 

their conventional bomb components. 

They take many forms, from containers 

of radioactive materials 

wrapped with conventional explosives, 

to aerosolized materials sprayed 

by conventional equipment, to manual 

dispersion of fine powders into the 

environment [1]. Also included are 

radiation-exposure devices (REDs), 

used to expose people to dangerous 

beams or particles of radiation. RDD 

attacks can produce general panic, 

immediate death and long-term 

increases in cancer incidence, longterm 

loss of property use, disruption 

of services, and costly remediation of 

property and facilities. 

The risk is more than just theo retical. 

Several credible designs and 

plans for a dirty bomb attack against 

the United States have been found in 

Al Qaeda records. About 20 years ago, 

two actual dirty bombs were deployed 

by Chechen separatists; one was foiled 

and the other failed. In the mid-2000s, 

38 Alazan shoulder-fired missiles outfitted 

with 137 Cs dirty bomb warheads 

were for sale from a Moldovan arms 

dealer. 

Fortunately, few people will likely 

die from the radiological effects of a 

dirty bomb although tens to hundreds 

could die from the conventional blast. 

Unfortunately, with no precedence, 

we are struggling with generalizations 

| Fig. 1. 

Radiation dispersal devices (RDDs), or dirty 

bombs, are devices that disperse radioactive 

materials by any means possible. The most 

effective means of dispersal from a terrorist 

standpoint is the car bomb with several 

pounds of highly radioactive material that is 

easily dispersable, e.g., 137 CsCl powder in an 

Oklahoma-type ANFO car bomb. (Photo: 

Courtesy of Middle East Intelligence Bulletin) 

and how to prepare and respond to the 

first event. The following structures 

the problem and provides guidelines 

that are fluid and which should be formalized 

by the Department of Homeland 

Security (DHS) in conjunction 

with the Department of Energy (DOE, 

the Nuclear Regulatory Commission 

(NRC), the International Atomic 

Energy Agency (IAEA) and other cognizant 

agencies around the world. 

Rules of Thumb 

The challenge to training anyone 

outside the fields of radiation, radiochemistry 

and nuclear science is to 

provide enough information to be useful 

without creating confusion. First 

responders know all too well how 

dangerous it is to oversimplify. However, 

it is not possible to make all 

responders radiochemists or health 

physicists. So some simplifying concepts 

are in order: 

1) the logistical difficulty in successfully 

carrying out a significant dirty 

bomb attack is roughly the same as 

that of the 9-11 attacks, 

2) the most likely device will be a 

137 CsCl car bomb and the level of 

response and danger to responders 

is of the order of a 20 alarm fire, 

3) defining the hot zone is the most 

important first response and a 

simple alarming dosimeter is the 

most useful piece of equipment for 

a dirty bomb attack, 

4) following protocols, it is difficult to 

obtain a significant radiation dose 

in the first hours of response to an 

attack without actually handling 

radioactive material, 

5) the greater the dispersion, the 

greater the affected area, but the 

lower the dose, 

6) the scene will be a war zone, not a 

superfund or environmental cleanup 

site, 

7) it is possible to quickly triage most 

victims without significant harm to 

the responder, 

8) persons with no significant physical 

injuries should not be significantly 

contaminated, 

9) while not effective against g-radiation, 

the PPE of a firefighter 

(uniform, gloves, goggles and/or air 

purifying respirator) will provide 

complete protection against a- and 

b-radiation, and almost complete 

protection against ingestion and 

inhalation of radioactive material 

dispersed by the attack, including 

g-emitting radioactive material, and 

10) removing clothing and washing 

with soap and water is effective 

at removing radioactive contamination. 

The greater the training and experience 

with radioactive materials, the 

better able the responder will be to 

evaluating the usefulness of such 

generalizations. 

Dirty Bomb Materials 

Radioactive materials are used in 

many fields in almost all countries 

around the world, particularly for 

research, medical, and industrial 

applications. Dozens of radiological 

source producers and suppliers are 

found on six continents, and about a 

billion sources exist worldwide 

although most, like household smoke 

detectors, have such low specific 

activities that they pose no threat. 

With the increase of radioisotope 

applications in nuclear medicine 

diagnostics, therapeutics, sterilization 

and the food irradiation, the 

radio logical source production and 

fabri cation industry is an emerging 

growth industry in several countries, 

parti cularly in areas with depressed 

eco nomies. The rise in the number of 

terrorist acts during the last ten years 

has raised concerns about these 

radiological sources being used in 

RDDs. Because the general public is so 

frightened about anything radioactive, 

panic must be anticipated even 

if there is no likely health threat 

from the radioactive component. The 

degenerate case of a phantom RDD, 

where no radioactive material is used 

but an implication or anonymous tip 

indicates there is, could still cause 

considerable panic with large economic 

consequences, particularly in 

this era of fake news. 

The serious radiological threats 

come from large RDDs containing 


Practical Response to a Dirty Bomb ı James Conca


| Fig. 2. 

The primary uses, specific isotope and activity levels of radiological source materials. Note that the 

largest sources use only 60 Co, 137 Cs or 90 Sr (Courtesy of Greg van Tuyle). 

thousands to hundreds of thousands 

of curies (Ci) of activity. These 

could cause significant and lasting 

health and contamination problems. 

Although many variables determine 

the effectiveness of an RDD attack, 

the key factor is the quantity and type 

of radiological source material that is 

dispersed. Although it has been 

difficult to quantify, globally there are 

about 10,000 sources that exceed 

1000 Ci, and perhaps a thousand that 

exceed 100,000 Ci. 

Briefly, the differences in sources 

relate to their specific activity (the 

type and amount of radiation 

emitted), and its chemical form 

(whether it is a powder, and nonmetal 

solid or a metal). Gamma radiation 

(g) can penetrate great distances 

and, depending upon the energies, 

requires shielding of about 7 inches 

(18 cm) of lead (Pb) or about 3-feet 

(1 meter) of reinforced concrete. Beta 

radiation (b) can only penetrate a 

short distance and the personal 

protective gear of a firefighter can 

block much of the dose. Alpha radiation 

(a) is the least penetrating of all 

and can be stopped by a piece of paper 

or ordinary clothing. 

The most important pathway of 

accumulating dose from a or b sources 

is ingestion, or particularly inhalation, 

where the emitter is directly adjacent 

to tissue for long periods of time. For g 

sources, mere proximity is all that is 

required for significant doses. For 

RDD discussions, isotopes of Pu, Am 

and U are primarily a emitters, 60 Co 

and 137 Cs are g emitters, and 90 Sr is a b 

emitter. 60 Co usually occurs as a metal 

(either pellets or small rods), 137 Cs is 

as a powder, 90 Sr is as a ceramic 

titanate, and Pu, Am and U are various 

oxides, metals, salts and non-metal 

solids. 

Although the public generally 

thinks of Pu and enriched-U when 

hearing the word radioactive, these 

are not considered RDD materials of 

choice because they are primarily a 

emitters, are costly, cannot be 

obtained in large amounts, are welltracked 

and secured, and are more 

useful to terrorists in the production 

of actual nuclear weapons than in 

being wasted in an RDD. In this sense, 

137 CsCl powder is much more effective 

as an RDD material. 

Although inclusion of any radioactivity, 

no matter how small, in a 

dirty bomb will cause disruption at 

some level, the real health and economic 

threat resides in large sources 

| Fig. 3. 

137 CsCl powder, presently used in the irradiation 

industry, is the dirty bomb material of 

choice. 137 Cs is inexpensive (


ENVIRONMENT AND SAFETY 24 

an automobile, working in the fossil 

fuel industry, working in the construction 

industry, and working in the 

nuclear power industry. Almost every 

respondent ranked working in the 

nuclear power industry as either the 

most dangerous, or second most 

dangerous, activity. 

In reality, working in a nuclear 

power plant is the safest job in the 

world. In the last five years, no one has 

died in the nuclear power industry, 

whereas over 2 million Americans 

have died in the other activities: 

smoking > 700,000; consuming 

alcohol > 500,000; driving an automobile 

> 250,000; working in the 

fossil fuel industry > 100,000; 

working in the construction industry 

> 5,000; working in the nuclear 

power industry – 1. This misperception 

of how dangerous radiation is 

constitutes the number one issue 

concerning the effectiveness of a dirty 

bomb because how responders and 

the public respond to a dirty bomb 

attack will determine whether the 

attack is successful or not. Of course, 

there is no comparison between a 

dirty bomb attack and anything within 

our past experience. 

Therefore, short of a multi-year 

national public education initiative on 

radiation, how can we prepare and 

respond to this type of attack? The 

answer is to develop a simplified, 

practical guidance for responding to a 

radiological attack that does not 

depend upon in-depth understanding 

of radiation, that dispels the fear and 

panic that comes from misperceptions, 

and that fits into the NIMS/ICS 

incident command framework so that 

it can actually be implemented during 

a crisis. At the same time, we must 

provide sufficient resources in the 

form of training, documents, web 

sites, and expert consultation that 

allows the responder, if desired, to 

obtain a greater depth of understanding 

over a reasonably short time 

period. 

The Problem for Responders 

Another way to think of the problem 

of providing first responders sufficient 

understanding of radiation is to 

reverse the scenario. Two firefighters 

pull a handline from an engine, open 

the nozzle checking for water flow 

and eliminating air in the line, and 

charging into the burning structure 

make an interior, offensive attack. 

Moments later they emerge stating 

that the fire has been knocked down. A 

passerby observing this activity would 

question the sanity of people rushing 

into a burning structure under a 

canopy of rolling fire just inches above 

their heads. The passerby does 

not know the numerous evaluations 

resulting from great experience that 

occurred in the firefighters’ minds 

prior to making the decision that it 

was relatively safe to enter the structure 

in search of the seat of the fire. 

The passerby would probably not 

have rushed into the structure, and 

probably would have no clue of how to 

find his way to the seat of the fire in 

zero visibility. Just how well firefighters 

could teach these techniques 

in a three-day course to someone not 

in the field and let them loose in 

that situation is open to debate. 

Fortunately, no one would dream of 

trying it. 

Whether a firefighter is teaching a 

scientist how to make an interior fire 

attack, or a health physicist is teaching 

a firefighter how to handle various 

radioactive materials, the challenge is 

for the trainer to instill in the trainee 

not just the knowledge of how to 

function, but also an understanding of 

why to function that way, and most 

importantly, the wisdom that comes 

from experience with risks versus 

benefits of doing it in different ways, 

resulting in good decisions. In the 

classroom, laboratory, and field setting, 

knowledge and understanding 

can be taught by a skilled teacher. 

The student must have trust in the 

teacher’s wisdom while building his 

own bank of wisdom from the outcomes 

of his own decisions, which 

makes hands-on training, regional 

drills, field exercises, and full-scale 

events such as Homeland Security’s 

TOPOFF training, so important. 

Resources for the Responder 

Many resources have appeared, 

particularly online, to fill gaps in 

knowledge on radiological terrorism 

for first responders. 

DOE’s national laboratories provide 

basic radiation training and 

resources, e.g., Sandia National Lab 

www.sandia.gov/mission/homeland/ 

solutions/emergency/index.html, 

and Pacific Northwest National 

Lab www.hammertraining.com/, and 

there are many sites linked to DHS 

(www.dhs.gov/dhspublic/) and various 

private and quasi-private sites 

such as www.homelandresponse. 

org/, and the Responder Knowledge 

Base at http://www.rkb.mipt.org/. 

Sites sponsored by the federal 

government are built specifically to 

serve the needs of emergency 

responders and contain information 

on currently available products, along 

with related information such as 

standards, training, and grants. 

The Responder Knowledge Base is 

anticipating posting the Radiation 

Community Preparedness Resource 

(RadCPR) data base, a comprehensive 

data base on radiation and radiological 

incidents prepared by a team at 

Los Alamos National Laboratory. 

Because these resources consist of 

thousands of pages of information, 

they are best utilized periodically by 

the first responder as ongoing education. 

However, for the heat of the 

moment, in the immediate aftermath 

of a radiological attack, the first 

responder needs a simple set of basic 

principles that are useful but not 

daunting. An excellent guide for a 

simplified approach recently appeared 

as a result of many years of experiments 

at Sandia (S. V. Musolino and 

F. T. Harper in the April 2006 issue of 

Health Physics vol. 90(4), p. 377-385). 

This approach has been incorporated 

into many training programs. 

Training for the Responder 

Many training programs exist for the 

first responder with respect to radiological 

incidents. Because a dirty 

bomb attack is closer to a radwaste 

spill than to any other event, some 

training programs have logically 

adapted the DOE or the NRC equivalent 

of RadWorker II that have been 

used for decades to train radiation 

workers in the nuclear, clean-up and 

disposal industries. However, the first 

responder is not a radiation worker, 

and may not need most of the information 

contained in these trainings. 

What is needed is a combination 

and streamlining of three classes of 

training: RadWorker II, MERRTT, 

and NIMS/ICS. MERRTT (Modular 

Emergency Response Radiological 

Transportation Training Program) is 

the only nationally recognized first 

responder training for handling radiological 

transportation incidents, and 

NIMS/ICS is the general incident 

management system utilized by the 

nation’s response communities. 

At a minimum, any course should 

include equal time in classroom, 

laboratory and field exercises in areas 

including: 

p basic concepts of radiation physics 

and chemistry 

p biological effects of radiation 

p hazard recognition 

p characteristics of RDDs: the source 

and the explosives 

p how dirty bombs are packaged and 

dispersed 




p initial response actions 

p incident control and command 

p handling the walking worried and 

the terrified 

p pre-hospital practices 

p situation board drills 

p radiological instrumentation and 

dosimetry devices 

p Department of Homeland Security 

Guidelines for RDD events and 

what that means to first responders 

p clean-up and ways to mitigate the 

effects of RDDs 

p site forensics and preservation of 

evidence 

p decontamination, disposal and 

documentation 

p when to return to work and living 

spaces after a dirty bomb attack 

What differentiates these courses 

from most is the inclusion of first 

responders on the faculty, including 

fire chiefs, state police detectives, 

National Guard CST personnel, EMTs 

and forensic scientists, as well as the 

radiochemists and health physicists. 

These courses are not necessarily 

meant for advanced radiation event 

responders such as the National 

Guard Civil Support or DOE RAP 

teams that will arrive to assist local 

responders within 12 hours of the 

event, but are meant to provide local 

responders with sufficient information 

to provide incident control 

and command, determine the effected 

areas, address immediate concerns 

such as fire, priority rescue, aid 

citizens and medical personnel, provide 

support to the CST and RAP 

teams after they arrive, and generally 

keep panic to a minimum. This last 

point is critical; if the responders first 

on the scene do not understand what 

a dirty bomb event entails, that 

uncertainty will be communicated to 

the civilians in the area and attempts 

to contain the situation and the 

contamination may fail. Because there 

are over 200,000 first responders in 

the top 100 target areas in the United 

States, many such programs are 

needed. 

So what can a first responder do? 

The following summarizes a simplified 

12-point guidance (variations 

exist and some form will be 

standardized in the near future). 

1. Assume all explosions or areas, 

particularly car/truck explosions, 

are dirty. 

2. If no dose or activity readings are 

available, set up an affected or 

exclusion zone boundary at 500 m 

from ground zero. If readings are 

available, set the full exclusion 

zone (around ground zero) outer 

boundary as about 1 rem/hr 

(10 mSv/hr). This boundary 

will also be the hot zone inner 

boundary. Set the hot zone outer 

boundary as about 0.1 rem/hr 

(1 mSv/hr). Within this zone, 

essential personnel can operate for 

several hours without accumulating 

significant dose. Exact 

adherence may not be feasible 

because of logistical or geometric 

issues and plus or minus a factor 

of 2 can be expected. Set the outer 

boundary of the warm zone 

( affected area) to about 2 mrem/hr 

(20 μSv/hr) depending upon 

operability. Local decisions may 

warrant establishing boundaries at 

2x or 4x background, but these 

may be miles from ground zero. 

3. All personnel in the hot zone 

should wear full PPE (turnout or 

bunker gear) with a particulate full 

face mask and have an updating, 

alarming cumulative dosimeter 

that can be used to track total dose. 

Take any precaution necessary to 

avoid inhaling or ingesting dust 

and particulates. Radioactivity will 

be in particulate form. 

4. When it is determined the situation 

is radiological, immediately 

alert the appropriate secondary 

response teams: CST, RAP and FBI, 

as advised in the unified command 

protocols for your region. If 

necessary in the U.S., call: 

p National Response Center 

1-800-424-8802 

p National Guard CST 

1-800-343-6701 

p FBI (ATF bomb) 

1-888-283-2662 

p DOE (RAP Coordinator) 

1-505-845-4667 

p NRC1-301-816-5100 

p DHS 1-202-727-6161 

p FEMA 1-202-586-8100 

In the European Union, contact 

the IAEA. Specific agencies are 

not yet assembled EU-wide for this 

purpose, although coordinated 

planning is underway. Each 

Member State has its own radiation-monitoring 

system or network 

which usually covers its whole 

territory. The density of radioactivity 

measuring points is 

variable – from several, uniformly 

distributed over the territory to 

hundreds of measuring stations, 

with increased density close to 

nuclear installations. EURDEP 

makes radiological monitoring 

data from most European countries 

available on a routine daily 

basis – and in close to real-time in 

emergencies. To achieve this, 

EU Member States, and other 

European countries which are 

members of EURDEP, send their 

data to EURDEP on a daily basis 

from at least one territorial radiation-monitoring 

network (some 

countries have more than one). 

The EURDEP system makes this 

radiological data available via a 

web page. 

5. Occupancy time outside the hot 

zone but within the warm zone is 

unrestricted for essential personnel 

during initial response (days – 

weeks). Establish Incident Command 

upwind of ground zero at 

the closest point outside the 

affected zone. Have alternative 

positions ready in case of change in 

wind direction. 

6. Evacuate all people from the 

affected area (> 2 mrem/hr) and 

exclude non-essential personnel 

thereafter. Expect self-evacuation 

for large populations of uninjured 

persons and provide them with 

safe designated routes out of the 

affected area (work with building 

managers to establish subterranean 

routes). Try to establish 

quick dose-rate screening, or 

radiological monitors, to determine 

those relatively few needing 

decontamination, but do not 

attempt mass decon of large populations. 

Instead, advise removal of 

external clothing, bag if possible, 

avoid eating, drinking or touching 

facial region, go directly home, 

shower with warm water and soap, 

and do not use hair conditioner, 

hair color, or other fixative hygiene 

products. Local decisions may 

warrant establishing large fire hose 

wash down curtains along decon 

corridors for rapid decon of 

evacuees and equipment, however, 

in large urban settings this will not 

be feasible. 

7. Do not decontaminate vehicles or 

structures during the initial 

response phase. Do not try to 

contain contaminated water, but 

allow, or even encourage, it to 

enter the municipal stormwater 

drainage system. Alert City 

Manager or wastewater treatment 

facility manager for possible 

diversion strategies. 

8. For those heavily contaminated 

persons, e.g., where there is 

obvious surface radioactive 

material or where they are heavily 

injured from the blast, establish 

decon areas and decon corridors 

connecting the hot zone to the 






boundary of the warm zone or 

affected area. Provide those with 

heavy external contamination of 

the upper body with follow-up 

exams to determine possible 

contaminant inhalation or 

ingestion. Countermeasures, e.g., 

Prussian Blue, should be evaluated 

promptly. 

9. Separate persons needing immediate 

medical attention and 

remove outer garments, survey for 

surface contamination, decon if 

necessary and possible, wrap in 

clean blankets in decon zone and 

evacuate. Inform the receiving 

medical facility that the person has 

little or no surface contamination 

or they may deny admittance. 

10. Commence mapping the affected 

area to obtain a rough dose profile 

of the area, marking hot and cold 

spots to assist in avoiding large 

doses during operations, and to 

assess the magnitude of the situation. 

 

11. Essential personnel within the 

affected area should record cumulative 

dose, if possible, and not 

exceed about 5 rem (50 mSv) total 

unless protection of critical infrastructure 

is deemed imperative 

and no alternative exists. Do not 

exceed about 10 rem (100 mSv) 

except to save lives and protect 

critical infrastructure. Note: no 

health effects ever observed for 

doses less than 10 rem. Do not 

exceed about 25 rem (250 mSv) 

unless the responder decides 

voluntarily, and with full 

knowledge of the risks, to save 

large numbers of lives and protect 

critical infrastructure that may 

harm large populations if not 

secured. Do not exceed about 

50 rem (500 mSv). 

12. Sheltering in place is advisable if 

the population is aware of the 

radiological nature ahead of the 

plume, unlikely in most 

cases. Evacuate buildings along 

determined safe routes away from 

the hot zone. Do not shut down 

building ventilation systems. 

Modern ventilation systems will 

filter most radioactive particulates 

and shut down may cause chimney 

effects. 

Once an attack has happened, there 

are limited options for mitigation. 

Spray-on fixatives are being investigated 

to prevent secondary migration 

and to make subsequent clean up 

easier. These may be ideal for heavily 

affected areas such as the immediate 

blast area, and for specific source 

materials such as a-emitters, but the 

best option may result from the fact 

that 137CsCl is so soluble that it can 

be washed off of surfaces with water. 

However, wash down must be done 

quickly and completely, within days 

of the event, to preclude further 

effects such as diffusion into building 

materials, secondary migration, and 

cumulative dose effects. 

Diffusion rates are primarily a 

function of moisture content and 

depend strongly upon weather conditions 

and porosity of the materials. 

If the weather remains dry and sunny, 

little diffusion will occur, but if the 

surfaces become wet (but not enough 

to wash off the Cs), or if surfaces are 

wet during deposition, then significant 

diffusion can occur quickly. On a 

wet surface, 137Cs can diffuse into 

concrete more than a quarter of an 

inch each week, but it would not 

diffuse that much into the granite, 

glass or metal even after several years, 

no matter what the conditions. 

There is considerable debate over 

the wash down approach, but it is 

unlikely any other strategy can be 

implemented rapidly enough to be 

affective. Although a 10 by 10 block 

area in downtown Manhattan has 

approximately one billion square feet 

of surface area, one hundred fire 

hydrants operating for 24 hours 

delivers about one hundred million 

gallons of water, sufficient to wash off 

large areas and wash most of the Cs 

into the stormwater drainage system 

where it will be sufficiently diluted 

and deposited in areas of lesser consequence. 

Alternatively, the wash water 

can be treated at the outflow points 

using inexpensive materials such as 

gabions of zeolitic gravel ($80/ton), 

that are extremely specific for Cs and 

other radionuclides. 

It is essential that the United States 

and the European Union responds 

quickly and efficiently should an RDD 

attack occur in order to minimize the 

long-term effects and deter future 

RDD attacks by showing a high-cost/ 

low-benefit to terrorist groups for such 

weapons. But this can be accomplished 

only if emergency response agencies 

are well trained and have suitable 

plans that can be executed within a 

comprehensive interagency command 

structure. 

Author 

Dr. James Conca 

Senior Scientist 

UFA Ventures, Inc. 

Richland, USA 

jim@ufaventures.com 

Geochemist and Energy scientist, speaker and author 

Dr. James Conca is Senior Scientist for UFA Ventures, 

Inc. in the Tri-Cities, Washington, a Trustee of the 

Herbert M. Parker Foundation, an Adjunct Professor 

at Washington State University in the School of the 

Environment, an Affiliate Scientist at Los Alamos 

National Laboratory and a Science Contributor to 

Forbes on energy and nuclear issues. Conca obtained 

a Ph.D. in Geochemistry from the California Institute 

of Technology in 1985, an MS in Planetary Science in 

1981, and a Bachelors in Geology and Biochemistry 

from Brown University in 1979. 




Equipment Selection Methodology 

of Seismic Probability Safety Assessment 

for Nuclear Power Plant 


1 Introduction A typical seismic probability safety assessment (SPSA) requires geological, structural, and 

systems analysts. In the case of probabilistic seismic hazard analysis (PSHA), geological experts can independently 

perform the analysis, but structural experts performing seismic fragility analysis (SFA). Probability safety assessment 

(PSA) model analysts performing system analysis, cannot perform the analysis independently. Only when the analysis 

is performed through mutual co-operation between the two fields can a reasonable result be achieved. This is a unique 

characteristic of SPSA. 

For SPSA, a seismic equipment list 

(SEL) is first prepared, and reviews of 

the design, construction installation 

data, walkdown and system are 

conducted to analyze structures, 

systems and components (SSCs) that 

have little or no effect on the safety of 

the entire power plant. The final 

selected SSCs are reflected in the 

PSA model to perform SPSA. The 

following are the considerations in 

selecting a general SEL [1]: 

p Identify SSCs that are important to 

safe shutdown from full-power 

PRA models. 

p Identify SSCs from a review of 

seismic evaluation performed for 

the IPEEE (Individual Plant Examination 

for External Events). 

p Identify structures and passive 

components that are important to 

the seismic response. 

p Identify additional SSCs from a 

plant walkdown. 

As can be seen from the above SEL 

selection criteria, equipment is required 

systematically for the safe shutdown 

of a nuclear power plant, and 

critical equipment vulnerable to earthquakes 

is selected through seismic 

response analysis and site surveys. In 

this paper, we propose the equipment 

selection methodology that should be 

essential to SPSA using the characteristics 

of individual plants and internal 

events PSA for existing qualitative and 

ambiguous selection criteria. 

2 Practices for seismic 

equipment selection and 

screening criteria 

In this paper, we review the practice of 

SEL selection and characteristics 

based on cases applied in Korea and 

the United States. In Korea, when an 

SPSA is performed, the SEL is selected 

in consideration of the general points 

mentioned in Chapter 1. After preparing 

the SEL, special screening 

criteria are set, which will be reviewed 

in detail here. Quantitative screening 

means that SSCs, which generally 

establish and meet appropriate 

screening criteria that are expected 

not to affect overall plant safety, are 

excluded from the SPSA model, and 

reflect and analyze only equipment 

and buildings that do not meet the 

screening criteria of the SPSA model. 

For example, there are two 

screening criteria: the first is the 

specific median acceleration capacity 

(A m ), and the second is the high 

confidence of a low probability of 

failure (HCLPF). First, if A m is set as 

the screening criteria, it should not 

contribute to plant safety risk regardless 

of the results of the site’s PSHA, so 

a relatively high value should be set as 

the screening criteria, which can be an 

overly conservative analysis. Moreover, 

there is a weakness in that there 

is no difference in core damage 

frequency (CDF), which is a risk metric 

for SPSA between a power plant with a 

high probability of seismic occurrence 

on the site and a power plant with a 

low probability of seismic occurrence. 

Second, when HCLPF is set as the 

screening criteria, its value is the result 

of the convolution of the hazard curve 

derived by PSHA and fragility of a 

single piece of equipment. As an 

example, if 1.00E-05/yr, the convolution 

value computed in between 

the fragility of specific equipment and 

the site hazard curve is set as the 

screening criterion HCLPF, even if all 

of this equipment directly causes core 

damage, its CDF is 5.00E-07/yr, 

meaning less impact. In other words, 

HCLPF corresponds to a probability of 

equipment failure of 0.05 based on 

95 % reliability; so conservatively, if 

the corresponding earthquake occurrence 

frequency is 1.00E-05/yr, the 

final CDF corresponds to 5.00E-07/yr. 

This will be analyzed in detail in the 

following sections; however, the 

disadvantage is that the SPSA results 

will be optimistic if seismic-induced 

failures are excluded on a single 

screening criterion because equipment 

that is critical to plant safety shutdown 

can be used in various seismic-induced 

initiating events and contributes to the 

CDF of each initi ating event. 

In the case of the United States, 

ASME/ANS RA-S–2008, which can be 

called a PRA standard, describes as 

follows [2]: 

p DEVELOP seismic fragilities for all 

those structures, systems, or components, 

or combination thereof, 

identified by the systems analysis. 

p If screening of high seismic capacity 

components is performed, 

DESCRIBE fully the basis for 

screening and supporting documents. 

For example, it is acceptable 

to apply the guidance given in 

EPRI NP-6041-SL, Rev. 1, and 

NUREG/CR-4334 to screen out 

components with high seismic 

capacity. However, CHOOSE the 

screening level high enough that 

the contribution to core damage 

frequency and large early release 

frequency from the screened-out 

components is not significant. 

This provides a vague criterion that 

fragility analysis should be performed 

to reflect all SSCs in the SPSA model 

and screening can be applied, and this 

case should not have a serious impact 

on both CDF and large early release 

frequency (LERF). Next, guidance 

suggesting the US SPSA methodology 

recommends determining the detailed 

fragility analysis SSCs through the 

steps below [1]. 

p Step 1: Screen Inherently Rugged 

Structures, Systems, and Components. 

p Step 2: Assign Initial, Representative 

Fragility Values for Structures, 

Systems, and Components. 



Equipment Selection Methodology of Seismic Probability Safety Assessment for Nuclear Power Plant ı Junghyun Ryu and Moosung Jae



p Step 3: Create and Quantify an 

Initial Seismic Probability Risk 

Assessment Model. 

p Step 4: Rank the Systems, 

Structures, and Components by 

Importance Measure. 

p Step 5: Perform Detailed Fragility 

Calculations for the Risk- Important 

Structures, Systems, and Components. 

p Step 6: Document the Seismic 

Equipment List Screening Process. 

All equipment except “inherently 

rugged structures, systems, and components” 

is considered in the SPSA 

model, so the evaluation result is quite 

conservative and the SPSA model 

must be established in advance to 

perform this analysis. In addition, 

since there are no specific screening 

criteria for detailed fragility analysis, 

there is a weakness in that the results 

may differ depending on the judgment 

criteria of the performing expert. As 

mentioned above, it is recommended 

that all SSCs should be included in the 

SEL and qualitatively selected and 

removed within the range that does 

not affect the entire CDF or LERF; 

however, no methodo logy can find a 

systematic method for this. 

3 Sensitivity analysis for 

single HCLPF screening 

criteria 

The Nuclear Energy Institute (NEI) 

presents the screening criteria for 

performing fragility calculation [3], 

which is similar to the HCLPF 

screening criteria applied in Korea. To 

confirm the validity of the screening 

criteria presented in this report, we 

performed a sensitivity analysis using 

the top 50 cutsets of SPSA based on 

the data contained in this report. The 

sensitivity analysis consisted of SPSA 

models assuming 4th seismic acceleration 

intervals for convenience. As a 

quantification tool, software using 

minimal cutsets upper bound (MCUB) 

methodology was used, but overestimated 

CDF was derived from high 

seismic acceleration interval and 

quantified using FTeMC [4] software 

with Monte Carlo simulation. The top 

50 cutsets presented in the report 

consist of 22 seismic-induced failures, 

ten random failure events, and four 

human errors events, as shown in 

Tables 1 and 2. These cutsets are also 

considered in the success sequence. 

In this paper, sensitivity analysis 

is carried out on two cases. First, 

we analyzed the changes in CDF 

according to the application of the 

screening criteria using the results of 

the convolution with the single piece 

of equipment; and second, we analyzed 

the changes in CDF when each 

piece of equipment with similar 

HCLPF size is not considered in the 

model. To utilize the results of the 

first, single piece of equipment convolute 

with seismic hazard curve. The 

convolution value was derived using 

PRASSE [5], which is used in the 

quantification analysis of seismicinduced 

initiating events with binary 

decision diagram method and Monte 

Carlo simulation. The results are 

shown in Table 3. 

As Table 2 shows, the higher the 

HCLPF value, the smaller the convolution 

result, but it can be seen that 

there is a slight difference, depending 

on β R and β U . Table 4 shows the 

sensitivity analysis according to two 

screening criteria. 

When screening criteria 5.00E-5/ 

yr is applied, four pieces of equipment 

are excluded from the model, and the 

CDF reduction is 0.5 % based on the 

base model, so there is no significant 

effect. With screening criteria 1.00E- 

4/yr, the change in the overall CDF is 

around 4 %, indicating that applying 

higher screening criteria can have a 

significant impact on the overall results. 

Therefore, the appropriate 

screening criteria was found to be effective 

because they might not 

have a significant impact on the 

entire CDF. 

Second, we examine the amount of 

change in CDF for equipment with 

similar HCLPF. Even with similar 

HCLPF value, the Am value may show 

large difference according to the β 

value, and the effect may be very 

different depending on the seismic 

acceleration intervals determined. In 

Component Description A m β U β R HCLPF 

SEIS-BS-CRANE Seismic failure of polar crane damages reactor vessel causing core damage 1.34 0.3 0.3 0.50 

SEIS-RC-RPV Seismic failure of Reactor Vessel 1.34 0.3 0.3 0.50 

SEIS-RC-RPVINT Seismic failure of Vessel Internals causes ATWS 1.34 0.3 0.3 0.50 

SEIS-SW-FO-TNK Seismic failure of ESW diesel fuel oil tank 1-SW-TK-1 1.25 0.24 0.37 0.45 

SEIS-RH-PUMPS Seismic failure of MD RHR pumps 1.03 0.17 0.31 0.45 

SEIS-EE-FO-TKS Seismic failure of underground EDG Fuel Oil tanks 1-EE-TK-2A/B 1.07 0.3 0.3 0.40 

SEIS-EP-EDG Seismic Failure of the Emergency Diesel Generators (EDG-1, 2, 3) 1.07 0.3 0.3 0.40 

SEIS-EP-LCC Seismic failure of Load Control Centers 1-EP-LCC-1H/J,1H/J-1 0.97 0.24 0.31 0.39 

SEIS-EP-BATT Seismic failure of station batteries 1(2)-EPD-B-1A/B 1.02 0.19 0.42 0.35 

SEIS-EE-DAY-TK Seismic failure of EDG fuel oil day tanks 1.08 0.45 0.24 0.33 

SEIS-EP-XFRMR Seismic failure of 4kv-480v transformers 1(2)-EP-TRAN-1H/J and 1H/J1 0.83 0.24 0.25 0.37 

SEIS-BC-BCHX 

Seismic failure of Bearing Cooling HXs results in flood of the Emergency 

Switchgear Room 

0.77 0.25 0.22 0.35 

SEIS-CC-CCHX Seismic failure of CCW HX results in flood of the Emergency Switchgear Room 0.68 0.32 0.18 0.29 

SEIS-EP-MCC-CV Seismic failure of Motor Control Centers in Cable Vault and Tunnel 0.68 0.32 0.22 0.28 

SEIS-EP-MCC-SB Seismic failure of Motor Control Centers in Service Building 0.69 0.23 0.36 0.26 

SEIS-EE-FO-PMP Seismic failure of the EDG Fuel Oil Transfer Pumps 0.68 0.3 0.3 0.25 

SEIS-FW-TDAFW Seismic failure of Turbine-driven Aux feed water pump 0.68 0.3 0.3 0.25 

SEIS-CS-RWST Seismic failure of Refueling Water Storage Tank 1-CS-TK-1 0.55 0.3 0.2 0.24 

SEIS-BS-TBLDG 

Seismic failure of the Turbine building causes loss of canal(SW) and leads to 

core damage 

0.53 0.42 0.23 0.17 

SEIS-EP-LOOP Seismic failure of offsite power 0.4 0.22 0.22 0.19 

SEIS-CN-ECST Seismic failure of Emergency Condensate Storage Tank 1-CN-TK-1 0.32 0.2 0.07 0.20 

SEIS-EP-AAC-DG Seismic failure of Alternate AC Diesel (batteries) 0.13 0.32 0.24 0.05 

| Tab. 1. 

Equipment fragility data for sensitivity case [3]. 




Component Description Mean Error Factor Distribution Type 

1EEEDG-FR-1EEEG1 EDG-1 Fails to run 1.90E-02 1.26 Log normal 

1EEEDG-TM-1EEEG1 EDG-1 in Test or Maintenance 1.90E-02 1.21 Log normal 

1SACMP-FS-1SAC2C Service Air Compressor 1-SA-C-2 fails to start 4.22E-03 1.12 Log normal 





AACEDG-TM-DG0M Alternate AC diesel in Test or Maintenance 3.07E-02 1.49 Log normal 

HEP-C-1AFW-SLO 

HEP-C-FTSESW-SLO 

HEP-C-FTSIA-SLO 

PROB-RCCA-2 

Operator fails to align alternate AFW source 

during seismic event – low g 

Operator fails to start Emergency SW pump 


Operator fails to start Instrument Air compressor 


Fraction of Control Rods that fail to insert that exceeds 

emergency boration success criteria 

| Tab. 2. 

Random failure and human error probability for sensitivity case [3]. 

2.64E-02 3 Log normal 

2.54E-03 12.8 Log normal 

3.60E-03 3 Log normal 

5.50E-01 1 Uniform 

PROB-RCPSL-182 Probability of 182 gpm RCP seal leak 1.98E-01 1 Uniform 

REC-RHRHX-SLO Operator fails to align RHR during seismic event – low g 3.48E-01 1 Uniform 


Component Case Convolution probability HCLPF 

SEIS-BS-CRANE 

SEIS-RC-RPV Case 1 

3.11E-05 0.50 

SEIS-RC-RPVINT Under 5.00E-05 

3.11E-05 0.50 

| Tab. 3. 

Convolution result for sensitivity case. 

| Tab. 4. 

Sensitivity analysis result for screening criteria. 

addition, the accident sequence varies 

depending on the plant characteristics, 

so the target equipment contributes 

to plant safety which can have 

3.11E-05 0.50 

SEIS-SW-FO-TNK 4.20E-05 0.45 

SEIS-RH-PUMPS 

5.89E-05 0.45 

SEIS-EE-FO-TKS 6.60E-05 0.40 

SEIS-EP-EDG Case 2 

6.60E-05 0.40 

SEIS-EP-LCC Under 1.00E-04 

8.07E-05 0.39 

SEIS-EP-BATT 8.61E-05 0.35 

SEIS-EE-DAY-TK 8.62E-05 0.33 

SEIS-EP-XFRMR 

1.14E-04 0.37 

SEIS-BC-BCHX 1.38E-04 0.35 

SEIS-CC-CCHX 2.08E-04 0.29 

SEIS-EP-MCC-CV 2.16E-04 0.28 

SEIS-EP-MCC-SB 2.26E-04 0.26 

SEIS-EE-FO-PMP 2.33E-04 0.25 

– 

SEIS-FW-TDAFW 2.33E-04 0.25 

SEIS-CS-RWST 3.46E-04 0.24 

SEIS-BS-TBLDG 4.69E-04 0.17 

SEIS-EP-LOOP 6.56E-04 0.19 

SEIS-CN-ECST 9.39E-04 0.20 

SEIS-EP-AAC-DG 5.21E-03 0.05 

Case 

Screening 

Criteria 

# of screening out 

component 

CDF 

Base model - - 6.38E-04/yr - 

-ΔCDF 

(%) 

Case 1 Under 5.00E-05 4 6.35E-04/yr 0.50 % 

Case 2 Under 1.00E-04 10 6.14E-04/yr 3.81 % 

a very different effect on the entire 

CDF. Therefore, screening based on 

single HCLPF criteria will not be able 

to understand all the effects without 

performing various sensitivity analyses 

and may have an optimistic 

effect on the overall results. Table 5 

shows the analysis of the effect on 

CDF that is reduced when four pieces 

of equipment with similar HCLPF 

(battery; emergency diesel generator 

fuel oil day tank; 4kV-480V transformer; 

and bearing cooling heat 

exchanger) are excluded from the 

model. 

As a result, it can be seen that in the 

case of the battery, the reduction rate 

of CDF is very large, cor responding to 

2.25 %, but in the case of the transformer, 

it is very small, corresponding 

to 0.2 %. Therefore, it can be seen that 

if a single HCLPF criterion is applied 

and screened-out, the effect on the 

entire CDF is different for each piece of 

equipment, and thus further analysis 

is required. 

4 Equipment selection 

methodology 

As discussed in the previous section, 

equipment selection for SPSA is not 

being carried out using a rational 

approach, such as by single screening 

criteria or conservatively considering 

all equipment. In this paper, we propose 

a methodology for an equipment 

selection methodology based on all 

three analysis steps of SPSA. 

4.1 Determination of site 

seismic hazard region 

using PSHA 

PSHA is the step of evaluating the 

site-specific seismic probability that is 

the input to the SPSA. Basically, the 

probability of an earthquake is the 

most important step because it directly 




Component Description Am βu βr HCLPF ΔCDF(%) 


SEIS-EP-BATT Seismic failure of station batteries 1.02 0.19 0.42 0.35 -2.25% 

SEIS-EE-DAY-TK Seismic failure of EDG fuel oil day tanks 1.08 0.45 0.24 0.33 -2.22% 

SEIS-EP-XFRMR Seismic failure of 4kv-480v transformers 0.83 0.24 0.25 0.37 -0.20% 

SEIS-BC-BCHX Seismic failure of Bearing Cooling HXs 0.77 0.25 0.22 0.35 -0.96% 

| Tab. 5. 

Sensitivity analysis result for various equipment exclusion. 

Site name Pr (a>SSE) Pr (a>1.0g) Seismic CDF Seismic CDF Proportion (Over 1.0g) 

Seabrook 4.62E-04 3.80E-06 2.99E-05 13 % 

Krsko 4.00E-04 1.40E-06 5.96E-05 2 % 

Limerick 2.29E-04 1.79E-06 5.72E-06 31 % 

IP2 2.12E-04 1.75E-06 8.40E-06 21 % 

Millstone 3 1.82E-04 1.47E-06 8.85E-06 17 % 

Surry 1.32E-04 6.00E-07 2.37E-05 3 % 

Average 2.61E-04 1.79E-06 1.99E-05 14 % 

| Tab. 6. 

Comparison of PSHA and Seismic CDF for reference plant [6, 7]. 

affects the probability of a seismicinduced 

initiating event. In this paper, 

we reviewed the correlation between 

various SPSA cases and PSHA of each 

site. Table 6 shows the specific values 

of the site-specific seismic hazard 

curve and CDF values for each nuclear 

power plant [6, 7]. 

Although an attempt was made 

to review more cases of PSHA 

and seismic core damage frequency 

(SCDF), the published data are 

limited and insufficient to infer an 

accurate correlation; however, the 

trend can be confirmed using published 

data. The values that represent 

the correlation between site SPHA 

and SCDF are the probability of 

exceeding 1.0g and the probability of 

exceeding the safety shutdown earthquake 

(SSE). First, the probability of 

seismic acceleration exceeding 1.0 g 

corresponds to a very large earthquake 

magnitude, so most power 

plants assume that direct core 

damage occurs. However, as a result 

of reviewing the case of power plants 

based on this value, it is difficult to 

represent the characteristics of 

seismic hazards based on Pr(a > 1.0 g), 

since the core damage frequency for 

earthquakes exceeding 1.0g accounts 

for a range of approximately 2 % – 

14 %. 

Next, we reviewed the probability 

of exceeding SSE. Since SSE is the reference 

value for seismic design of 

safety-related SSC, it can be assumed 

that in the event of an earthquake 

below SSE, most equipment can 

maintain its function, which has a 

significant impact on plant safety 

before and after the seismic acceleration 

corresponding to this value. 

When plotting the correlation 

between the probability of occurrence 

| Fig. 1. 

Relationship of Seismic Core Damage Frequency (SCDF) and probability for SSE. 

of SSE and CDF, it can be seen that the 

correlation between the two variables 

is high, as shown in Figure 1. 

Therefore, if this value is broadly 

divided into four types, as shown in 

Table 7, it is possible to classify the 

possibility of earthquake occurrence 

on the site by reoccurrence period. 

In the methodology proposed in 

this paper, the probability of SSE 

occurrence is basically divided into 

four major sections, and we intend to 

use this as the basis for determining 

the equipment groups that need to be 

subjected to fragility analysis. In the 

case of Region A, since earthquakes 

are very likely to occur, it is suggested 

that the number of equipment groups 

that need to be analyzed for fragility is 

most; and in the case of Region D, 

the equipment target for fragility 

ana lysis is least because of the 

lowest probability of earthquakes. To 

examine the adequacy of the four 

distinct values, the data on the 

seismic hazard values of power plants 

operating in the US [8] are examined 

as shown in Figure 2. 

Based on these values, it can be 

seen that the 61 power plants comprise 

7 % in Region A, 21 % in Region 

B, and 26 % in Region C. Lastly, 

Region D occupies 46 %. It can be seen 

that the criteria are valid for classifying 

the overall seismic hazard 

trend of a specific site. 

4.2 Determination of 

equipment group for 

fragility analysis 

The second step of the SPSA is seismic 

fragility analysis. Seismic fragility 

analysis is performed by a structural 

expert, and it is a step to evaluate the 

seismic resistance for the weakest 

failure mode of each piece of 




Category Region name Pr (a>SSE) Recurrence period (yr) 

Region A Very high risk region Over 4.00E-04/yr Under 2,500 year 

Region B High risk region 2.00E-04/yr ∼ 4.00E-04/yr 2,500 year ∼ 5,000 year 

Region C Medium risk region 1.00E-04/yr ∼ 2.00E-04/yr 5,000 year ∼ 10,000 year 

Region D Low risk region Under 1.00E-04/yr Over 10,000 year 

| Tab. 7. 

Categorized site region for PSHA. 


| Fig. 2. 

Hazard curve for US nuclear power plant [8]. 

equipment and to determine the median 

acceleration capacity (A m ); with β R 

and β U representing the variability. 

Seismic fragility analysis requires the 

greatest consumption of time and 

budget among the three steps, as 

structural experts with the particular 

expertise need to perform the analysis, 

as well as review and conduct on-site 

verification of many data such as 

design, construction, and installation 

works. Therefore, the methodology 

proposed in this paper is a method that 

suggests a way to minimize such analysis. 

First, through a survey of 

sufficient data on the fragility of 

general equipment, we examine the 

characteristics of the fragility of equipment 

generally installed in power 

plants. A great deal of research has 

been conducted on the fragility of 

general equipment. Among this 

research, according to the SPID report 

[9], the fragility of SSCs can be largely 

classified into three, and the first 

inherently rugged SSCs generally have 

very high seismic resistance, so there is 

no need to include inherently rugged 

SSCs in the SPSA model. Second, SSCs 

with somewhat high seismic resistance 

are reflected in the SPSA model if the 

impact on SPSA is significant after 

reviewing the magnitude. In the last 

case, SSCs must be considered in the 

SPSA model. In this paper, equipment 

groups are classified into four types 

based on generic fragility data, and 

the results are shown in Table 8. 

Table 8 refers to three reports [1, 

10, 11], and the equipment is categorized 

conservatively based on the lowest 

HCLPF value. The equipment 

group according to HCLPF value has 

the following characteristics: 

p Group I (HCLPF: 0.2g or less): 

Generally, the seismic acceleration 

corresponding to plant SSE is taken 

to 0.2 g, so when an earthquake 

occurs, SSCs will fail with a 

probability of around 5 %. This 

equipment group includes offsite 

power sources, which are generally 

considered the most vulnerable of 

plant installations, fixed electrical 

panels without anchor bolts, 

non-safety class buildings, and 

yard tanks. In this case, the equipment 

corresponding to the most 

vulnerable group should always 

be considered regardless of the 

magnitude of the site seismic 

hazard curve. Typical yard 

tanks include condensate storage 

tank and refueling water storage 

tank. 

p Group II (HCLPF: 0.2g to 0.4g): 

Typically, this equipment group 

comprises active components 

which are operated by power 

source, including fixed electrical 

panels using anchor bolts, relays, 

batteries, small tanks, and 

non-safety grade diesel generators. 

p Group III (HCLPF: 0.4g to 0.6g): 

Generally, this equipment group 

corresponds to the safety-related 

class component; mainly the 

equipment and power sources 

considered for accident mitigation. 

p Group IV (HCLPF: 0.6g or higher): 

Generally, this equipment group 

includes various valves, pipes that 

make up the inherently rugged 

SSCs, including pressurizer, steam 

generators, and equipment installed 

in the main system, such as 

safety injection tanks. The main 

component of this group is equipment 

that directly causes core 

damage. 

After applying these groups to 

individual power plants, there may be 

cases different from the general 

fragility due to a plant’s unique 

design characteristics or constructability. 

Therefore, equipment different 

from the general fragility should be 

included in the analysis through a 

walkdown that is essential for SPSA. 

4.3 Seismic plant response 

analysis (SPRA) 

The final step is the plant seismic 

response analysis, which is divided 

into an analysis of equipment that can 




Group Category Equipment HCLPF 


I Electrical Equipment Offsite power 0.09 

Structures and Tanks Large flat-bottomed tank 0.15 

Electrical Equipment 480 VAC Motor Control Center (MCC) / Switchgear / Bus (unanchored) 0.17 

Electrical Equipment 5 kV+ AC Bus / Switchgear (Unanchored) 0.17 

Electrical Equipment DC Motor Control Center (MCC) / Switchgear / Bus (Unanchored) 0.17 

Electrical Equipment Panel (Electric or Instrument) (Unanchored) 0.17 

Pump & Compressors Diesel-driven pump (non-safety) 0.17 

Structures and Tanks Turbine building 0.17 

Structures and Tanks Other non-safety buildings 0.17 

Piping Buried piping (non-safety pipe) 0.18 

Piping Distributed piping system (welded steel pipe) 0.19 

II Electrical Equipment Cable tray 0.21 

Pump & Compressors Motor-driven pump 0.25 

Pump & Compressors Turbine-driven pump 0.25 

Electrical Equipment 480 VAC Motor Control Center (MCC) / Switchgear / Bus (Anchored) 0.26 

Piping Distributed piping system (cast iron pipe) 0.26 

Structures and Tanks Small tank 0.30 

HX & HVAC Chiller 0.31 

Piping Small LOCA (SLOCA) 0.31 

HX & HVAC Air handling unit (AHU) / Room Cooler / Fan 0.33 

Electrical Equipment Panel (Electric or Instrument) (Anchored) 0.34 

Electrical Equipment Relay Chatter Failure (During Seismic Event) 0.34 

Pump & Compressors Large vertical, centrifugal pump (motor-driven, non-safety) 0.34 

Electrical Equipment Batteries 0.35 

Structures and Tanks Reactor internals and core assembly 0.35 

Piping Buried piping (safety pipe) 0.36 

Electrical Equipment Transformer (Indoor) 0.37 

Structures and Tanks Containment building / drywell 0.38 

Electrical Equipment Emergency Diesel Generator (EDG) 0.40 

III Pump & Compressors Diesel-driven pump (safety) 0.45 

HX & HVAC HVAC duct 0.48 

Pump & Compressors Reactor coolant pump / Recirculation pump 0.50 

Pump & Compressors Compressor 0.51 

Structures and Tanks Reactor building / auxiliary building / control building / fuel building 0.51 

Structures and Tanks Safety diesel generator building 0.51 

Structures and Tanks Other safety buildings 0.51 

Electrical Equipment 4 kV+ AC Bus / Switchgear (Anchored) 0.52 

Piping Medium LOCA (MLOCA) 0.53 

Electrical Equipment Battery Charger 0.55 

Electrical Equipment Inverter 0.55 

Structures and Tanks Reactor pressure vessel (RPV) 0.58 

Valves Large (>10”D) Motor-Operated Valve (MOV) / Damper 0.60 

Valves Small (


cause an initial event and equipment 

that should be used to mitigate safety 

shutdown such as engineering safety 

feature when a seismic-induced initial 

event occurs. Equipment that may 

cause an initial event should be 

included conservatively without a 

screening process, as all of the 

initial events may cause direct core 

damage. However, this equipment 

may be excluded if the frequency of 

occurrence is calculated to be below 

1.00E-07/yr, which corresponds to 

the typical initial event screening 

criteria [12]. Table 9 shows equipment 

that can initiate seismic event 

in general light-water reactor nuclear 

power plant. 

In general, seismic events can 

occur in seismic-induced loss of 

coolant accident, loss of power, loss of 

control, loss of ultimate heat sink, 

main steam line break, and anticipated 

transient without scram. 

Next, when the seismic-induced 

initiating event occurs, an analysis 

of the equipment considered to 

mitigate the event is performed. Like 

all external event analysis, SPSA 

basically uses the event tree and fault 

Seismic initiating event 

Loss of 

Coolant Accident 

(LOCA) 

Loss of Power 

(LOP) 

Loss of Control 

(LOC) 

Loss of Heat Sink 

(LOHS) 

Large 

Medium 

Small 

Small-Small 

Loss of offsite power 

Station black out 

Loss of KV 

Loss of DC power 

Loss of AC power 

Main control room 

Component cooling 

water 

Ultimate heat sink 

Main steam line break (MSLB) 

Anticipate transient without scram (ATWS) 

Building failure 

| Tab. 9. 

Equipment for seismic induced initiating event. 

CASE 

| Tab. 10. 

Sensitivity analysis result for importance analysis of internal events PSA. 

tree used in the internal event PSA. 

Here, non-safety and non-seismic 

equipment that cannot be used conservatively 

is excluded from the 

model. However, although seismicinduced 

initiating events are different 

from internal events, the primary 

heat removal, which is essential for 

mitigation of the accident, must be 

performed by the same equipment 

and procedure. Therefore, in the 

ASME Standard, which is considered 

the standard of the PSA, equipment 

corresponding to the standard can be 

listed based on FV (Fussell-Vesely) 

importance 0.005 higher and RAW 

(Risk Achievement Worth) value of 

two or higher, which are the criteria 

that can be used to classify significant 

basic events. To confirm the application 

of importance value, a sensitivity 

analysis was performed on the 

reference nuclear power plant. In 

Table 10, we show the difference between 

the results of the existing SPSA 

and CDF when only the equipment 

that was evaluated as important in 

internal events was modeled. 

As a result, it is confirmed that the 

results of the sensitivity analysis 

Equipment 

Pressurizer, Steam generator, RCP, RCS 

RCP seal 

Impulse lines 

EDG 

DC panel, Inverter, Battery, Charger 

AC Panel 

Main control board, MCR HVAC 

Pump, Heat exchanger, Piping 

Pump, Heat exchanger, 

Intake structure 

Control rods 

Containment building 

Service building 

Turbine building 

# of basic event 

for SPSA model 

% of 

baseline CDF 

Base model 1,798 100.0 % 

Only FV important basic event 49 97.0 % 

Only RAW important basic event 217 95.9 % 

Intersection FV and RAW 

important basic event 

24 95.5 % 

Union FV or RAW important basic event 242 97.6 % 

showed no significant difference 

between the case of modeling all 

equipment and the case of modeling 

only equipment that was evaluated as 

important in internal events. When 

242 items of equipment, 13 % of the 

total of 1798, were modeled in the 

SPSA model, the CDF was 97.6 % of 

the baseline CDF, and even when 

49 pieces of equipment classified as 

important in basic events based on FV 

are modeled, a result equivalent 

to 97 % of the baseline CDF was 

obtained. In addition, even when only 

24 pieces of equipment were considered, 

a value corresponding to 

95.5 % of the CDF value of the base 

model was derived, indicating that 

the importance of other equipment 

other than this was much lower 

than expected. Therefore, it is 

judged that the equipment selection 

methodology based on the values of 

the internal event FV and RAW is 

much more rational than the 

existing HCLPF-based method, and is 

an efficient method that can 

reflect the characteristics of the 

SPSA model. 

4.4 Summary of the proposed 

equipment selection 

methodology 

The procedure is shown in Figure 3, a 

flow chart of the method of selecting 

equipment for seismic events over the 

three steps described above. 

First, based on the site seismic 

hazard curve obtained as a result of 

PSHA, the SSE value is checked, 

which is the design standard of the 

power plant, and probability value 

exceeding SSE value is checked. 

After that, the region of the site is 

determined by checking the SSE 

reoccurrence period according to 

the SSE excess probability value. 

Through this value, the group of 

equipment that should perform 

fragility analysis is determined. Next, 

based on the results of the PSA 

importance of internal events, basic 

events with an FV value of 0.005 or 

more or RAW value of two or more, 

are listed and described in terms of 

the function of the system. However, 

basic events related to non-seismic 

equipment and human error are 

excluded among the results of the 

PSA importance of internal events. 

Finally, equipment is derived 

through cross-examination between 

the ‐equipment required for the 

function of the mitigating equipment 

derived from an internal event and the 

equipment group subject to fragility 

analysis. 






| Fig. 3. 

Flowchart for the equipment selection methodology. 

5 Case study for the 

equipment selection 

methodology 

In this paper, we compare and 

analyze the seismic PSA results of the 

reference nuclear power plant using 

the proposed methodology. Existing 

PSA results could not be considered, 

as detailed data on the walkdown 

could not be collected. However, even 

if the walkdown is excluded, the 

results are sufficient to demonstrate 

the effectiveness of the methodology 

proposed in this paper. The first step 

is the analysis of site-specific PSHA 

results. The review of the corresponding 

hazard curves indicates 

that the probability of earthquake 

occurrence is relatively low, and the 

probability of exceeding 0.2g on the 

SSE basis corresponds to 1.06E-04/yr, 

corresponding to the Region C 

medium risk region proposed in this 

paper. In terms of the likelihood of 

earthquake occurrence, it is a value 

that is approximately 54 % of the total 

power plants in terms of seismic 

hazard considered by Figure 2. Based 

on this, the plant needs to review 

equipment groups I and II subject to 

fragility analysis. Fragility analysis 

groups I and II include general 

vulnerable equipment including 

offsite power sources and yard tanks, 

as well as general active equipment. 

The next analysis is the plant seismic 

response analysis using internal 

events PSA results; 74 basic events 

were selected with an FV value of 

0.005 or more and 253 basic events 

with a RAW value of two or more. 

Among them, 30 pieces of equipment 

that satisfy both FV and RAW are 

considered, along with 297 basic 

events. Excluding 20 human error 

basic events and 26 non-seismic basic 

events, 251 basic events are considered. 

In addition, except for valves, 

flow elements, flow transmitters, 

radiation transmitters, dampers and 

filters, which are inherently rugged 

SSCs, the total 136 basic events are 

derived. 

The 136 basic events derived can 

be divided into related systems and 

equipment types to achieve the 

following 14 critical system functions. 

Here, if we review the 14 important 

functions, we can confirm easily the 

unique operating characteristics of 

the reference plant. 

p Auxiliary feedwater (AF) supply 

p AF pump room cooling 

p Emergency power supply 

p Component cooling water supply 

p Diesel generator fuel supply 

p Diesel generator room cooling 

p Essential chilled water supply 

p High-pressure injection 

p Low-pressure injection 

p Plant control 

p Ultimate heat sink 

p Reactor containment cooling 

p Safety actuation signal 

p Ultimate heat sink pump room 

cooling 

In the end, cross-examining 14 critical 

systems and functions with fragility 

equipment groups I and II, the results 

shown in Table 11 are obtained. 

The number of equipment items derived 

through cross-examination is a 

total of 23, which is a very small result 

considering the overall equipment in 

nuclear power plants. How ever, when 

reviewing the pre viously analyzed 

SPSA results, it can be confirmed that 

all devices are considered important in 

the existing SPSA model except for the 

three pieces of equipment that initiate 

seismic-induced initiating events, so 

the methodology proposed in this 

paper is very efficient. It can be 

confirmed that it is reasonable. 

6 Conclusion 

In this study, a methodology of the 

equipment selection for SPSA is proposed. 

The single HCLPF screening 

criterion which has been applied for 

SPSA reflects some of the site-specific 

PSHA results but does not reflect the 

plant design characteristics, so if the 

single HCLPF screening criterion is 

applied to the model based on this, an 

optimistic evaluation can be made. 

The methodology proposed in this 

paper has the following advantages: 

p Safety aspects: Single HCLPF 

screening criteria is not applied, 

and all equipment is not reflected 

in the model using general fragility 

data, so realistic and reasonable 

SPSA results are expected. 




Group Category Equipment Mitigation Function Selected equipment 

I Electrical Equipment Offsite power – – 

Structures and Tanks Large flat-bottomed tank Auxiliary feed water supply Condensate storage tank 

Electrical Equipment 480 VAC Motor Control Center (MCC) / 

Switchgear / Bus (unanchored) 

High pressure injection 

Low pressure injection 

Diesel generator fuel supply 

Plant control 

Refueling water storage tank 

Refueling water storage tank 

EDG storage tank 

Load center 

Motor control center 

Electrical Equipment 5 kV+ AC Bus / Switchgear (Unanchored) Plant control Switchgear 

Electrical Equipment 

DC Motor Control Center (MCC) / Switchgear 

/ Bus (Unanchored) 

Plant control 

DC Panel 

Electrical Equipment Panel (Electric or Instrument) (Unanchored) Plant control Electrical panel 

Pump & Compressors Diesel-driven pump (non-safety) – – 

Structures and Tanks Turbine building – – 

Structures and Tanks Other non-safety buildings – – 

Piping Buried piping (non-safety pipe) – – 

Piping Distributed piping system (welded steel pipe) – – 

II Electrical Equipment Cable tray – – 

Pump & Compressors Motor-driven pump Auxiliary feed water supply AF motor driven pump 

Component cooling water supply 

Essential chilled water supply 

Diesel generator fuel supply 

High pressure injection 

Low pressure injection 

Ultimate heat sink 

Component cooling water pump 

Essential chilled water pump 

EDG fuel pump 

HPSI pump 

LPSI pump 

NSCW pump 

Pump & Compressors Turbine-driven pump Auxiliary feed water supply AF turbine driven pump 

Electrical Equipment 480 VAC Motor Control Center (MCC) / 

Switchgear / Bus (Anchored) 

Plant control 

Piping Distributed piping system (cast iron pipe) – – 

Load center 

Motor control center 

Structures and Tanks Small tank Component cooling water supply Component cooling surge tank 

HX & HVAC Chiller Essential chilled water supply Chiller 

Piping Small LOCA (SLOCA) – – 

HX & HVAC Air handling unit (AHU) / Room Cooler / Fan AF pump room cooling AF room AHU 

Diesel generator room cooling 

Reactor containment cooling 

Ultimate heat sink pump room 

cooling 

EDG room AHU 

RCFC fan 

NSCW room AHU 

Electrical Equipment Panel (Electric or Instrument) (Anchored) Plant control Electrical panel 

Electrical Equipment Relay Chatter Failure (During Seismic Event) Plant control Electrical panel 

Pump & Compressors 

Large vertical, centrifugal pump (motor-driven, 

non-safety) 

– – 

Electrical Equipment Batteries Plant control Electrical panel 

Structures and Tanks Reactor internals and core assembly – – 

Piping Buried piping (safety pipe) – – 

Electrical Equipment Transformer (Indoor) – – 

Structures and Tanks Containment building / drywell – – 

Electrical Equipment Emergency Diesel Generator (EDG) Emergency power supply Emergency diesel generator 


| Tab. 11. 

Selected equipment for fragility analysis. 

p Economics aspects: As the largest 

portion of the required manpower 

for SPSA is the detailed fragility 

analysis work, the methodology 

proposed in this paper is economically 

beneficial as the number of 

pieces of equipment subject to 

SPSA decreases. 

The equipment selection methodology 

proposed in this paper requires 

analysis of all three parts of 

SPSA, unlike the previously proposed 

methodology. This means that selecting 

equipment through consideration 

of only one part, such as PSHA, may 

lead to incorrect results. SPSA has its 

own uncertainty, so if one factor 

affects several steps, the uncertainty 

becomes very large. Therefore, the 

analysis of equipment not essential for 

SPSA can increase this uncertainty, so 

it can be said that it is necessary for a 

much more realistic analysis that is 

not considered in advance. 

In recent years, SPSA has evolved 

from a single unit criterion evaluation 

to an evaluation of multiple units 

operating at one site. In this case, the 

number of equipment items handled 

by the SPSA increases, resulting in 





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unnecessary model enlargement. Therefore, even in such 

case, it is necessary to select equipment that has a significant 

influence on seismic events, and applying the equipment 

selection methodology proposed in this paper can 

contribute to the simplification of the SPSA model and the 

accuracy of the analysis of the quantitative results, and 

minimize unnecessary analysis. Therefore, it is expected 

that uncertainty errors will be minimized. 

References 

[1] Seismic Probabilistic Risk Assessment Implementation Guide. USA: Electric Power Research Institute; 

2013, 3002000709. 

[2] Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear 

Power Plant Applications. USA: The American Society of Mechanical Engineers; 2009, ASME/ANS 

Ra-SA-2009. 

[3] White Paper: Criterion for Capacity-based Selection of SSCs for Performing Fragility Analysis in a 

Seismic Risk-based Evaluation. USA: Nuclear Energy Institute; 2012. 

[4] FTeMC Quick Guide Fault Tree Top Event Probability Evaluation Using Monte Carlo Simulation. 

Korea: Korea Atomic Energy Research Institute; 2017, KAERI-ISA- memo-FTeMC-01, Rev. 1. 

[5] Development and Validation of the Seismic Probabilistic Safety Assessment Software PRASSE. 

Korea: Korea Atomic Energy Research Institute; 2012, KAERI/TR-4649. 

[6] Probabilistic safety assessment for seismic events. IAEA; 1993, TECDOC-724. 

[7] M Vermaut, Ph Monette P Shah R.D Campbell, Methodology and results of the seismic probabilistic 

safety assessment of Krško Nuclear Power Plant, Nuclear Engineering and Design, 

2 May 1998, Volume 182, Issue 1, Pages 59-72 

[8] Risk Assessment of Operational Events Handbook Volume 2 – External Events. USA: Nuclear 

Regulatory Commission; 2008, Revision 1.01 

[9] Seismic Evaluation Guidance Screening, Prioritization and Implementation Details (SPID) for the 

Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. USA: Electric Power 

Research Institute; 2013, 1025287 

[10] A Methodology for Analyzing Precursors to Earthquake- Initiated and Fire-Initiated Accident Sequences. 

USA: Nuclear Regulatory Commission; 1998, NUREG/CR-6544. 

[11] Surry Seismic Probabilistic Risk Assessment Pilot Plant Review. USA: Electric Power Research Institute; 

2010, 1020756. 

[12] Identification of External Hazards for Analysis in Probabilistic Risk Assessment. USA: Electric 

Power Research Institute; 2011, 1022997. 

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ISSN 1431-5254 

Junghyun Ryu is a doctoral student in the Department of Nuclear Engineering at 

Hanyang University. He works in FNC Technology in South Korea. 

Prof Dr Moosung Jae 

Professor at Hanyang University, South Korea 

jae@hanyang.ac.kr 

Moosung Jae is Professor of Department of Nuclear Engineering, Hanyang 

University, Seoul, Korea. He is now Director of MUR Risk Assessment Group 

( MURRG Center) at HYU. 

Prior to joining Hanyang University, Seoul, Korea, he had worked for as a senior 

researcher for Korea Atomic Energy Research Institute (KAERI). 

He received B.S. and M.S. in Nuclear Engineering fromSeoul National University and 

Ph.D. in Nuclear Engineering from University of California, Los Angeles, USA (1992). 

Now he is a member of INSAG (International Nuclear Safety Group), IAEA as well as 

a member of the National Academy of Engineering of KOREA. 




Experimental Study of Convective 

Heat Transfer Through Fuel Pins of a 

Nuclear Power Plant 

Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus 

Cylindrical heat sources are used to model the fuel pins in the reactor core of a nuclear power plant and also in the 

spent fuel storage. This research was aimed at studying convection through slender cylinders arranged in a square 

array, placed inside a ventilated enclosure. Four cylindrical heat sources were manufactured, each having L/D ratio of 

6.1. These heat sources were placed inside an enclosure which had one inlet and three outlets. At first, optimum 

configuration of outlets for heat transfer was found by keeping the inlet speed of air and heat flux constant. This 

optimum outlet configuration was then used to study the effect of changing heat flux on heat transfer. Speed of air was 

kept constant at 0.7 m/s throughout this study. The heat flux was changed, ranging from 79.83 W/m 2 to 513.59 W/m 2 

and Rayleigh number changed accordingly from 2.093x10 9 to 8.575x10 9 . It was observed that by increasing Rayleigh 

number, Nusselt number decreased along with the heat transfer. 

1 Introduction 

Convective heat transfer is one of the 

important areas of research in heat 

transfer. Efficient heat transfer from 

heat sources has been an important 

topic of research over the past few 

years. Cylindrical heat sources are 

widely used in many industries 

such as building, solar and nuclear 

industry. In nuclear industry, fuel pins 

inside reactor core and spent fuel 

storage can be modeled as cylindrical 

heat sources with convective fluid 

flowing around them. Due to such 

vast applications and importance, 

researchers around the world have 

studied heat transfer through cylindrical 

heat sources both experimentally 

and computationally to solve issues 

related to temperature control. 

There are many configurations 

in which convective heat transfer 

through cylindrical heat sources can 

be studied. The medium through 

which the heat is transferred, could be 

finite or infinite, the heat sources 

could be single or multiple, and could 

be ranged in a horizontal, vertical or 

inclined orientation, the convective 

fluid could be air, water or any other 

suitable fluid, the type of convection 

could be free, forced or mixed convection 

and the enclosure could be 

ventilated or non-ventilated. All these 

different types of configurations 

have been studied previously. Popiel 

[1] reviewed heat transfer through 

slender cylinders in which he used the 

data put forth by Cebeci [2] and 

differentiated between heat transfer 

through thick and slender cylinders. 

This is important since thick cylinders 

can be approximated as flat plates and 

thin cylinders cannot. Griffiths and 

Davis [3] studied convection in 1922, 

where they studied vertical cylinders 

in isothermal condition. They did 

experiments on various cylinders, 

keeping their diameters constant and 

varying the length of the cylinders 

and determined the average Nusselt 

numbers. Later on, Morgan [4] used 

that data to find two correlations, 

which are valid over their respective 

ranges of dimensionless number. Fujii 

et al. [5] did experiments on isothermal 

vertical cylinders, using three 

different fluids i.e. water (Pr = 5), 

mobiltherm oil (Pr = 100) and spindle 

oil (Pr = 100). Separate correlations 

were then developed for the local 

Nusselt numbers for each of the fluids 

used. Jarral and Campo [6] carried 

out an experimental study, using 

isoflux boundary condition and determined 

local Nusselt numbers in terms 

of Rayleigh number for air. They did 

their experimentation using cylinders 

with three different slenderness 

ratios. Ali Riaz et al. [7] performed 

experiments on vertical cylinders and 

horizontal cylinders to find the heat 

transfer coefficients. The Nusselt 

number was observed to decrease 

from the bottom of the cylinder, 

towards the top, up to a certain point 

after which it started to increase. The 

reason behind this is that thermal 

boundary layer thickness increases 

from bottom to top of the cylinder. In 

the horizontal configuration, however, 

the local Nusselt number was 

least at the outlet and maximum at the 

inlet. Arshad et al. [8] performed 

experiments with natural convection 

at high Rayleigh numbers using nine 

cylinders in a 3x3 array placed 

vertically and enclosed inside an 

enclosure. It was observed that 

surface temperatures increased up to 

a specific point and then decreased, 

this is attributed to mixing, which 

results in increase in heat transfer. 

K. Hata et al. [9] studied heat transfer 

through natural convection in laminar 

region, using vertical rods placed in a 

7x7 array placed in liquid sodium. 

The effect of pitch-to-diameter (P/D) 

ratio, array size, bundle geometry and 

Rayleigh number on heat transfer, 

was observed by calculating Nusselt 

number under different conditions. 

Yuji Isahai and Naozo Hattori [10] 

worked on a numerical study of heat 

transfer through natural convection in 

a heated rod bundle, that was placed 

vertically in an equilateral triangle 

configuration inside an enclosure. 

They used a total of 19 cylinders in 

their study in a hexagonal formation, 

with the center cylinder dedicated for 

instrumentation. Five different P/D 

ratios were studied which varied 

between minimum of 1.1 to maximum 

of 7.0. Abdul Jabbar Khalifa and Zaid 

Ali [11] studied heat transfer through 

natural convection in single and 

in multiple cylinders. For multiple 

cylinder configuration, they used nine 

cylinders in a 3 x 3 array, out of which 

only three cylinders were heated. A 

square array configuration was used 

for cylinders having a P/D ratio of 2. 

The fluid used for heat transfer was 

water. Heat flux was varied and its 

effect was studied. K. Tehseen et al. 

[12] performed a numerical study, 

using ANSYS, to find the effects of 

different orientations of core on heat 

transfer along bare circular tubes and 

tube bundles. They found that the 

heat transfer was directly proportional 

to the heated length and 

inversely proportional to the inside 

diameter of the tube. 

37 

RESEARCH AND INNOVATION 


Experimental Study of Convective Heat Transfer Through Fuel Pins of a Nuclear Power Plant ı Atif Mehmood, Ajmal Shah, Mazhar Iqbal, Ali Riaz, Muhammad Ahsan Kaleem and Abdul Quddus


RESEARCH AND INNOVATION 38 

Convection heat transfer is affected 

by geometry parameters of enclosure 

such as position and area of vents. 

There has been limited research 

considering such parameters and, 

therefore, this found the motivation 

behind the present experimental 

study. In this present work, cylindrical 

heat sources were used, which can be 

used to model fuel pins inside a 

nuclear reactor or spent fuel inside 

spent fuel storage. Heat transfer 

through a 2x2 array of heat sources in 

square configurations was studied. 

Heat transfer between heat sources 

and air was tabulated and compared 

in different arrangements of the 

outlets present on the enclosure. The 

results were then plotted showing the 

relation between Nusselt number and 

modified Rayleigh number. 

2 Mathematical Modeling 

The equations used for the present 

system were: 

Q in = P(2.1) 

Q out = Q in = V × I(2.2) 

Due to small thickness of aluminum 

tube, all the heat from the heat source 

is conducted to the outer surface. 

Q out = Q conduction =Q convection + Q radiation 

(2.3) 

Q convection = h avg A s (T s,avg – T b )(2.4) 

Where, h avg is the average heat 

transfer coefficient. 

And 

For calculating the local heat transfer 

coefficients, the following formula 

was used. 

Q convection = h x A s (T x – T b )(2.5) 

Heat losses through radiation can be 

calculated by: 

(2.6) 

ε is 0.04 for Aluminum and σ is 

5.67x10 -8 W/m 2 K 4 . Due to small 

emissivity of Aluminum and a very low 

value of σ, radiation can be neglected 

and it can be assumed that all the heat 

transfer from the outer surface of heat 

sources is via con vection. 

The formulae for Nusselt number 

and modified Rayleigh number are: 

(2.7) 

 

(2.8) 

 

(a) 

| Fig. 1. 

Thermocouple locations (a) front view of a heat source (b) Top view of the enclosure. 

3 Experimental Setup 

Four identical sources each having 

a diameter of 50.8 mm and a 

slenderness ratio of 6.1 were manufactured. 

Each of these heat sources 

had a rated power of 1000 W. Outer 

surface of heat sources was made of 

Aluminum. Each of the heat sources 

was equipped with four K-type 

thermocouples, the positions shown 

in Figure 1. In Figure 1(b), the stars 

show the thermocouple facing 

the flow channel and the triangles 

represent the thermocouples on the 

opposite side of the flow channel. 

The heat sources were placed 

inside a wooden enclosure of dimensions 

20”x20”x20”. There were three 

outlets of same size (4”x2”) in the 

enclosure, one at the top and one 

each on right top and left top of the 

enclosure. For the measurement of 

the ambient temperature, a thermocouple 

was installed near inner 

boundary of the wooden enclosure. 

To convert the data provided by the 

thermocouples into temperatures, 

PANGU data acquisition system was 

| Fig. 2. 

Experimental setup. 

(b) 

used. All these components of the 

experimental setup are shown in 

Figure 2. 

In this research work, data 

acquisition system was used which 

had a calibration uncertainty of 1 °C 

and data scatter was observed to be 

5.1 °C. Total uncertainty in measurement 

of temperature was calculated to 

be 4.8 %. Uncertainty in voltage 

was 1 % with a full scale reading of 

600 V and in current, the uncertainty 

was 1.5 % with a full scale reading of 

10 A. Anemometer that was used 

for the measurement of velocity of air 

had an uncertainty of 0.03 m/s. 

Also, different dimensions that 

were measured e.g., diameter of the 

aluminum pipes, length of aluminum 

pipes and size of inlet and outlet vents 

had an uncertainty of about 2 mm. 

Each experiment was repeated thrice 

so that the results are reported with 

precision and accuracy. Mean values 

were shown in the graphs and error 

bars were displayed, representing 

standard deviation in the results. 

4 Results and Discussions 

In the experimentation phase, the 

outlet configuration was first 

optimized to give best heat transfer 

and then the further experimentation 

was carried out using different heat 

fluxes. 

4.1 Optimization of Outlets 

Configuration 

Five cases were studied, each case 

differing from the other on the basis of 

locations of the outlets’ opening. 

In the natural convection case, top 

outlet was opened and inlet at the 

bottom was opened, but the fan was 

remained switched off. Air flowed 

freely and naturally from the bottom 

and exited through the top outlet. In 

other four cases, bottom inlet was 

opened with the fan operating at a 




(a) 

(c) 

| Fig. 3. 

Variation of Nu with outlet configuration for (a) cylinder 1, (b) cylinder 2, (c) cylinder 3, (d) cylinder 4. 

constant speed of 0.7 m/s. The 

corresponding outlets were opened 

for each case and temperatures were 

measured which were then used to 

find the values of heat transfer 

coefficients for each case. 

In Figure 3, the near outlet refers 

to the outlet which is closest to that 

particular cylinder and far outlet is the 

outlet which is at a distance from that 

cylinder. For example, for cylinders 

1 and 2, near outlet is the left outlet 

and far outlet is the right outlet as 

shown in Figure 1. 

It is evident from the plots above, 

that for any given cylinder, the heat 

transfer is the least in case of natural 

convection and improved for forced 

convection as more outlets were then 

opened. Heat transfer is maximum 

when all the three outlets were 

opened. This gave an optimum configuration 

of outlets with respect to 

heat transfer. 

(b) 

(d) 

value of 8.575x10 9 at heat flux of 

513.59 W/m 2 . The results for these 

five cases are shown in Figure 4. 

It is clear form the plotted points 

that the heat transfer dropped as the 

(a) 

flux and modified Rayliegh number 

was increased. Nusselt number did 

not change much as the modified 

Rayleigh number increased from 

2.09x10 9 to 5.79x10 9 . Beyond that 

point, with increase in the modified 

Rayleigh number, there is a certain 

decrease in the values of Nusselt 

number. The reason behind this 

behaviour is attributed to the fact that 

the outlets were not large enough to 

allow heat transfer away from the 

cylinders and the temperatures were 

seen to go higher as the heat flux and 

modified Rayleigh number of the 

heaters was increased. 

5 Conclusions 

Heat sources were manufactured 

and arranged in a square array 

inside a ventilated enclosure with 

one inlet and three outlets at 

three different walls of the enclosure. 

Experiments were carried out for 

five different outlet configurations at 

constant heat flux and constant 

speed of air form the inlet to 

determine the optimum configuration 

of the outlets for heat transfer. 

After that, experiments were carried 

out for five different heat fluxes 

ranging from 79.83 W/m 2 to 

513.59 W/m 2 and the results were 

plotted in the form of dimensionless 

numbers. 

(b) 


4.2 Effect of Heat Flux 

on Heat Transfer 

A total of five different heat fluxes 

were used to study their effect on heat 

transfer. Heat fluxes were changed 

from 79.83 W/m 2 to 513.59 W/m 2 . 

The data was also translated in the 

form of dimensionless numbers. Due 

to change in heat fluxes, modified 

Rayleigh number, Ra* changed from a 

minimum value of 2.093x10 9 at heat 

flux of 79.83 W/m 2 to a maximum 

(c) 

(d) 

| Fig. 4. 

Variation of Nu with modified Ra for (a) cylinder 1, (b) cylinder 2, (c) cylinder 3, (d) cylinder 4. 





The results of the experiments can 

be summarized as following: 

p Heat transfer improved by almost 

25 % in the cases with forced 

convection i.e., when the fan was 

switched on and all three outlets 

were opened as compared to the 

natural convection case when the 

fan was off. 

p Heat transfer from the cylinders 

seemed to remain almost constant 

with increase in heat flux and the 

modified Rayleigh number up to a 

certain range and then decreased 

as the modified Rayleigh number 

was increased further. 

p The reason could be attributed to 

the fact that the size of the outlets 

was small (4” x 2”) as compared 

to the size of the enclosure 

(20”x20”x20”). 

Nomenclature 

h 

Q 

A 

Heat transfer coefficient 

Heat Supplied 

Area 

q” Heat Flux 

σ 

β 

k 

Ra* 

V 

I 

P 

ε 

α 

ν 

Nu 

Stefan-Boltzmann constant 

Volumetric thermal expansion coefficient 

Thermal conductivity 

Modified Rayleigh Number 

Voltage supplied 

Current 

Power supplied 

Emissivity 

Thermal diffusivity 

Momentum diffusivity 

Nusselt Number 

Subscripts 

x Local value 

avg 

s 

b 

Average value 

Value at surface 

Bulk 

References 

[1] C. O. Popiel, “Free Convection Heat Transfer from Vertical 

Slender Cylinders: A Review,” Heat Transfer Engineering, 

vol. 29:6, pp. 521-536, 2008. 

[2] T. Cebeci, “ Laminar-Free-Convective-Heat Transfer from 

the Outer Surface of a Vertical Slender Circular Cylinder,” 

Proc. 5th International Heat Transfer Conference, vol. 3, 

pp. 15-19, 1974. 

[3] E. Griffiths and A. Davis, “The transmission of heat by 

radiation and convection,” Food Investigation Board, 

Special Report, vol. 9, 1922. 

[4] V. T. Morgan, “The Overall Convective Heat Transfer from 

Smooth Circular Cylinders,” Advances in Heat Transfer, 

vol. 11, pp. 199-264, 1975. 

[5] T. Fujii, M. Takeuchi, M. Fujii, K. Suzaki and H. Uehara, 

“ Experiments on natural-convection heat transfer from 

the outer surface of a vertical cylinder to liquids,” 

International Journal of Heat and Mass Transfer, vol. 13, 

no. 5, pp. 753-770, 1970. 

[6] S. Jarall and A. Campo, “Experimental Study of Natural 

Convection from Electrically Heated Vertical Cylinders 

Immersed in Air,” Experimental Heat Transfer, vol. 18, 

no. 3, pp. 127-134, 2005. 

[7] A. Riaz, A. Shah, A. Basit and M. Iqbal, “Experimental 

Study of Laminar Natural Convection Heat Transfer from 

Slender Circular Cylinder in Air Quiescent Medium,” in 

Proceedings of 2019 16th International Bhurban 

Conference on Applied Sciences & Technology (IBCAST), 

Islamabad, 2019. 

[8] M. Arshad, M. Inayat and I. Chughtai, “Experimental 

study of natural convection heat transfer from an 

enclosed assembly of thin vertical cylinders,” Applied 

Thermal Engineering, vol. 31, no. 1, pp. 20-27, 2011. 

[9] K. Hata, K. Fukuda and T. Mizuuchi, “Laminar natural 

convection heat transfer from vertical 7x7 rod bundles 

in liquid sodium,” Journal of Nuclear Engineering and 

Radiation Science, 2018. 

[10] N. Hattori and Y. Isahai, “A Numerical Study of Natural 

Convection in a Vertical Cylinder Bundle,” Heat Transfer- 

Asian Research, vol. 32, no. 4, 2003. 

[11] A. J. N. Khalifa and Z. A. Hussien, “Natural convection 

heat transfer from a single and multiple heated thin 

cylinders in water,” Heat Mass Transfer, vol. 51, 

pp. 1579-1586, 2015. 

[12] K. Tehseen, K. Qureshi, M. Basit, R. Nawaz, W. Siddique 

and R. Khan, “Computational Heat Transfer Analysis of 

Tubes and Tube Bundles with Supercritical Water as 

Coolant,” ATW-INTERNATIONAL JOURNAL FOR NUCLEAR 

POWER, vol. 65, no. 11-12, pp. 588-594, 2020. 

Authors 

Atif Mehmood 

Lecturer, Department of 

Mechanical Engineering, 

Pakistan Institute of 

Engineering and Applied 

Sciences (PIEAS), Nilore, 

Islamabad 

atifmehmood@ 

pieas.edu.pk 

Lead author of this article, Mr. Atif Mehmood, is a 

lecturer at Department of Mechanical Engineering, 

Pakistan Institute of Engineering and Applied Sciences 

(PIEAS), Islamabad. He completed his bachelors in 

mechanical engineering from National University of 

Sciences and Technology (NUST), Islamabad before 

joining PIEAS for post-graduate studies. He has an 

interest in thermal hydraulics as well as design 

engineering. 

Ajmal Shah 

Professor, Department of Mechanical Engineering, 


(PIEAS), Nilore, Islamabad 

ajmal@pieas.edu.pk 

Mazhar Iqbal 

Lecturer, Department of Mechanical Engineering, 



mazhariqbal@ 

pieas.edu.pk 

Ali Riaz 

Post-graduate fellow, Department of Mechanical 

Engineering, Pakistan Institute of Engineering and 

Applied Sciences (PIEAS), Nilore, Islamabad 

aliriaz3.60@gmail.com 

Muhammad Ahsan Kaleem 

Lecturer, Department of Mechanical Engineering, 



ahsankaleem@pieas.edu.pk 

Abdul Quddus 

PhD scholar, Department of Mechanical Engineering, 



engquddus613@yahoo.com 




Preliminary CFD Analysis of Innovative 

Decay Heat Removal System for the 

European Sodium Fast Reactor Concept 

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, 

Janos Bodi and Enrico Girardi 

A major safety design objective for sodium cooled fast reactors is the practical elimination of the prolonged failure 

of the decay heat removal function. In order to achieve this objective, innovative decay heat removal system concepts 

are being developed. This paper discusses the preliminary CFD analysis of the design of an Innovative decay heat 

removal system implemented in the reactor pit which replaces the safety vessel of the European Sodium Fast Reactor. 

The efficiency of the oil cooling system is evaluated. A representative part of the reactor pit structure is modelled. The 

heat path is from the reactor vessel wall through a gap, an insulation layer into a concrete structure. The heat sinks are 

the oil cooling system and the ambient. A simplified case is used for verification with an analytical solution. It is shown 

that the oil cooling system is capable to remove the heat from the structure and keep the temperature there below 

70 degrees Celsius. At nominal conditions about 3 MW have to be removed and at a top oil cooling temperature of 

approximately 200 degrees. 



Sodium-cooled Fast Reactors (SFRs) 

are among the most advanced concept 

of the Generation-IV reactor systems 

[1] with extensive design, construction, 

operation and decommissioning 

experience. Several experimental, prototype 

and power reactors have been 

built in Europe and elsewhere [2]. 

Currently, there are two com mercial 

operating SFRs in Russia as well as 

experimental reactors ope rating in 

China and India. SFRs have several 

attractive features which allow them to 

meet the Generation-IV International 

Forum (GIF) objectives for future 

nuclear energy systems in terms of 

safety, sustainability, economics and 

proliferation re sistance. The fast neutron 

spectrum in connection with low 

neutron absorption and low moderation 

characteristic leads to enhanced 

neutron economy, which allows in 

addition to efficient fission power 

generation, the breeding of nuclear 

fuel and burning of trans uranic and 

minor actinides. In addition, the outstanding 

heat transfer property of 

sodium enables the design of a compact 

reactor core and high- performance, 

low-pressure heat transfer 

system. The SFR technology is in 

particular suitable to meet the twin 

missions of sustainability and improved 

economics both by efficient 

resource utilization and the effective 

management of plutonium and other 

minor actinides in a closed fuel cycle 

and by efficient power production 

using innovative power conversion 

systems. 

The current SFR technology R&D 

focuses on innovations to improve 

plant economics by enhancing reliability 

and reduction of capital cost as 

well as to enhance safety up to the 

level of the requirements for the 

Generation-IV plants. The safety goals 

for Generation IV reactors require 

excelling in safety and reliability; the 

reduction of the likelihood and degree 

of reactor core damage; and the 

elimination of the need for offsite 

emergency response. 

In order to meet these goals innovative 

safety concepts are needed in 

all areas of safety design of the 

fundamental safety functions. Beside 

reliable and robust reactivity control 

mechanism, highly reliable heat removal 

systems for assuring adequate 

cooling of safety relevant components 

and structures and effective options 

for dealing with severe accidents need 

to be implemented. 

Within the FP7 Euratom project 

CP-ESFR [3], the concept of the 

European Sodium-cooled Fast Reactor 

(ESFR) has been proposed and is 

currently being further improved 

within the Euratom Horizon-2020 

project ESFR-SMART [4] where new 

safety provisions have been proposed 

considering safety objectives envisaged 

for Generation-IV reactors and the 

update of European and international 

safety frameworks [5], taking into 

account the Fukushima accident. The 

proposed measures aim at enhancing 

safety and improving the robustness 

of the safety demonstration. The 

ESFR- SMART project aims to improve 

the reactor safety to the level of the 

requirements for the Generation-IV 

reactors, and make a proposal for 

new safety options, based on both 

present and previous projects experiences. 

In line with these objectives, 

several innovative approaches to the 

safety design of the ESFR concepts 

are being investigated. One of the 

safety innovations concerns the 

replacement of the safety vessel with 

reactor pit which is able to provide 

improved operational and safety 

functions of both the decay hear 

removal system and the containment 

function. With this innovation, the 

aim is to discard the safety vessel and 

to keep its safety functions, i.e., containment 

of the primary sodium in 

case of the reactor vessel leak without 

reduction of the primary sodium free 

level below the intermediate heat 

exchanger windows, by modifying the 

reactor pit geometry and by using the 

metallic liner on the reactor pit 

surface. In addition, this approach 

provides options to implement a new 

decay heat removal option within the 

reactor pit. With the implementation 

of this decay heat removal system, an 

additional safety provision is provided 

which could allow to practically 

eliminate a major safety design 

concern for sodium cooled fast 

reactors that of the prolonged failure 

of the decay heat removal function. 

This paper discusses the preliminary 

CFD analysis of the design of 

the innovative decay heat removal 

system implemented in the reactor pit 

which consists of oil and water based 

loops that provides capability to 

remove the entire decay heat of the 

reactor. Furthermore, the paper 

explains the overall design of the 

reactor pit including the preliminary 


Preliminary CFD Analysis of Innovative Decay Heat Removal System for the European Sodium Fast Reactor Concept ı 

Aleksander Grah, Haileyesus Tsige-Tamirat, Joel Guidez, Antoine Gerschenfeld, Konstantin Mikityuk, Janos Bodi and Enrico Girardi



| Fig. 1. 

ESFR concept plant layout. 

parametric calculations of temperature 

distributions for various design 

scenarios. The efficiency of the oil 

cooling system is evaluated. A representative 

part of the reactor pit 

structure is modelled. The heat path is 

from the reactor vessel wall through a 

gap, an insulation layer into a concrete 

structure. The heat sinks are the 

oil cooling system and the ambient. 

The oil cooling system is situated at 

the metal liner which is mounted 

along the insulation layer. In the gap. 

A simplified case is used for verification 

with an analytical solution. 

Furthermore, it is shown that the oil 

cooling system is capable to remove 

the heat from the structure and keep 

the temperature there below 70 °C. At 

nominal conditions about 3 MW have 

to be removed and at a top cooling 

temperature of approximately 200 °C. 

2 ESFR system layout 

2.1 ESFR design approach 

The ESFR concept has been developed 

within the FP7 CP-ESFR project (2009 

to 2013), considering past European 

experience in SFR technology in 

particular in the French SPX2 project 

and in European Fast Reactor project 

which involved a wider European 

cooperation. For both projects, the 

operational experience of the Superphenix 

reactor SPX [6][7] provided 

valuable inputs to improve safety, 

reliability, in service inspection and 

repair and economics. 

The ESFR concept is a large pool 

type industrial Sodium Fast Reactor of 

1,500 MWe/3,600 MW th . The design 

objectives for ESFR include simplification 

of structures, improved In-service 

inspection and repair capabilities, 

reduction of risks related to sodium 

fires and to the water/sodium 

reaction, improved fuel maintenance, 

core catcher with the capability for a 

whole core discharge and improved 

robustness against external hazards 

[5]. The ESFR core is composed of 

two zones of inner and outer fuel 

assemblies, and 3 rows of reflectors. 

There are two independent control 

rod assembly systems with additional 

passive reactivity insertion mechanisms. 

The core design aims at a fuel 

management scheme with a flexible 

breeding and minor actinide burning 

strategy. 

The ESFR concept plant layout is 

sketched in Figure 1. It is based on 

options already considered in previous 

and existing pool sodium 

fast reactors, with several potential 

improvements regarding safety, inspection 

and manufacturing. A 

particular attention is also given to 

compactness. The reactor vessel is 

cooled with sodium (submerged weir) 

and is surrounded by a reactor pit 

which replaces the safety vessel 

including all its function for normal 

operation and accident conditions. 

The reactor vault can be inspected for 

maintenance. 

The secondary system comprises 

six 600 MWth parallel and independent 

sodium loops, each connected 

to an intermediate heat 

exchanger (IHX) located in the reactor 

vessel. Each loop (see Figure 1) 

includes one Mechanical Secondary 

Pump (MSP), modular Steam Generators 

(SG) and one Sodium Dump 

Vessel (SDV). In addition to the 

normal power removal, the secondary 

system provides the DHR function in 

case of primary pumps trip. 

The main design objective for the 

Decay Heat Removal (DHR) system is 

to practically eliminate the prolonged 

loss of the decay heat removal 

function. In order to achieve this 

objective, three independent DHR 

systems have been implemented (see 

Figure 1). The first DHR system, 

DHRS-1, is provided by sodium/air 

heat exchangers connected to each 

IHX. These loops replace the Direct 

Reactor Cooling (DRC) System in the 

previous design and have various 

advantages such as avoiding additional 

roof penetrations and maintaining 

cold column in the IHXs [5]. The 

second DHR system, DHRS-2, is 

provided by cooling the steam generator 

modules by air in natural or forced 

convection through hatch openings, 

as is done in the Phenix reactor, providing 

the heat sink for the secondary 

loop. The third DHR system, DHRS-3, 

is implemented in the reactor pit with 

two independent cooling circuits, one 

with oil heat exchanger brazed on the 

liner and one with water inside the 

concrete. The water loop is capable to 

maintain the whole pit at temperatures 

below 70 °C. Due to the removal 

of the safety vessel, the DHRS-3 which 

is directly attached to the liner is 

expected to be more efficient and to 

assure a large part of the Decay Heat 

Removal close to 100 %. 

The safety analyses being currently 

performed indicate that with the 

current DHR concept the demonstration 

of practical elimination of a 

prolonged loss of the DHR function is 

feasible. 

2.2 Description of the reactor 

pit design 

One of the main innovative approaches 

to the safety design of the ESFR 

concept concerns the replacement of 

the safety vessel with reactor pit that 

aims to improve the operational and 

safety functions of both the decay heat 

removal systems and the containment. 

All existing SFRs have a safety 

vessel (see an example of Super phenix 

safety vessel Figure 2) around the 

main vessel. The function of this safety 

vessel is to confine the primary sodium 

in case of the main vessel leakage, so 

as to avoid lowering of the primary 

sodium free level below the inlet 

windows of the intermediate heat 

exchangers and thus to provide an 

efficient natural convection through 

the core. In case of the main vessel 

leakage, the reactor is not recoverable 

and the core must be unloaded. Due to 

the need to wait for reduction of the 

residual power of the assemblies, this 

handling could take a significant 

duration (i.e. higher than one year) 

especially in the ESFR-SMART design 

without external sodium storage. The 

safety vessel must therefore remain 

filled with sodium for a long time. The 





| Fig. 2. 

Arrival of the safety vessel inside the reactor 

pit of Superphenix. 

potential danger in these conditions is 

that the reactor pit is not designed to 

withstand a sodium leak from this 

safety vessel. Moreover, this safety 

vessel leak would also lead to interruption 

of the core cooling by natural 

convection, leading to a very difficult 

overall situation. 

A number of measures have been 

taken to prevent leakage of the safety 

vessel: slight overpressure between 

the two vessels to detect a possible 

leak and choice of different materials 

to avoid a common failure mode on 

corrosion. It is recalled that it is a 

problem of corrosion on welds that 

led to the leakage of the Superphenix 

storage drum vessel, and that this leak 

was taken up by the safety vessel of 

this storage drum [3]. 

The scenarios of vessel leakage are 

diverse, from corrosion leakage to 

leakage on a severe accident with 

mechanical energy release. This leads 

to high uncertainties in the temperatures 

and leakage rates, which 

make it difficult to demonstrate the 

safety vessel mechanical strength 

against the corresponding thermal 

shocks. Moreover, the French licensing 

authority requires considering the 

double leakages in order to verify that 

the situation does not lead to cliff-edge 

effect in term of radiological releases 

in the environment. This demonstration 

was required after the SPX 

external sodium storage leak. It is 

particularly required if the core 

unloading is high. 

The safety vessel is a proven option, 

especially demonstrated during the 

incident at the Superphénix storage 

drum [6], and is adopted in all existing 

fast reactors. However, the evolution 

of safety standards leads us to look at 

other options where its functions 

could be directly taken over by a 

reactor pit capable of withstanding a 

sodium leak, and thus a long-term 

mitigation situation. It was an option 

that had already been looked at in the 

EFR project with a vessel anchored in 

the pit, which option was later abandoned 

for reasons of feasibility and 

design difficulties. 

The proposed design of the reactor 

pit is composed of the following 

domains (see Figure 3, Figure 4 and 

Figure 5): 

p A mixed concrete/metal structure 

with a water cooling system inside 

the concrete supports the thick 

metal slab to which the reactor 

vessel is attached. Together with 

the reactor roof, it provides a 

sealed containment which must 

keep its integrity in all the cases of 

normal or accidental operations 

[1]. 

p Inside the concrete/metal structure, 

blocks of insulating materials 

(non-reactive with sodium) are 

installed. Alumina is selected as 

reference material for the insulation 

blocks. A conventional insulation 

layer could be considered in 

future to increase insulation effects 

(outside the scope of the paper). 

A metallic liner is placed on the 

surface of the insulation blocks. The 

gap between the reactor vessel and 

the liner must be small enough 

(350 mm was chosen) to avoid 

decrease of the primary sodium free 

level below the intermediate heat 

exchanger (IHX) windows in case of 

sodium leakage from the reactor 

vessel. During normal operation, the 

primary sodium free level is 1,350 mm 

below the roof. In case of primary 

sodium leak about 300 m 3 of sodium 

will leave the reactor vessel to fill the 

gap and the new equilibrium free level 

of the primary sodium will be about 

3,070 mm below the reactor roof. 

| Fig. 4. 

Drawing of the reactor pit design. 

With this level of sodium inside the 

primary circuit, there is still a 

1,090 mm sodium level above the 

upper IHX openings, which allows a 

sodium inflow into the IHX and a good 

natural convection and core cooling. If 

the sodium temperature decrease to 

180 °C , the volume of remaining 

sodium will decrease of about 205 m 3 

due to the change in the density, It 

would cause the sodium level to go 

down to around 50 mm above the IHX 

entry bottom. And the natural convection 

remains possible. 

p The oil cooling system is installed 

next to or even inside the liner 

(Figure 3). 

| Fig. 3. 

Detail of the reactor pit design (1 reactor vessel, 2 liner, 3 insulation, 

4 concrete/metal structure, 5 oil cooling system (DHRS-3.1), 6 water 

cooling system (DHRS-3.2, 7 gap)). 







(a) 

| Fig. 5. 

Drawings of the reactor pit cooling systems: (a) oil and (b) water. 

p Finally, a special concrete with 

alumina (aluminous concrete) 

which could withstand, without 

significant chemical reaction with 

sodium, a leakage of the liner 

could be used between the liner 

and the insulation (blocks of 

alumina). 

Two independent active cooling 

systems are proposed in the reactor pit 

(we use the acronyms DHRS-3 for the 

combination of these two systems): 

p The oil cooling system (DHRS-3.1) 

close to the liner (Figure 5a). The 

oil under forced convection can 

remove the heat transferred by 

radiation from the reactor vessel at 

high temperature. Conversely to 

water, the adopted synthetic oil is 

resistant to high temperatures 

above 300 °C and reacts with 

sodium without producing hydrogen. 

As an example the commercial 

oil called ”Therminol SP“ [8] can 

be used in normal operation at 

temperatures up to 315 °C. 

p The water cooling system 

(DHRS-3.2) for the concrete 

cooling is installed in the concrete 

(Figure 5b) and aims at maintaining 

the concrete temperature 

under 70 °C in all possible situations, 

even if the oil system is lost. 

Both oil and water circuits work 

during normal operation and have to 

maintain the concrete temperature 

below 70 °C. This margin is intended 

(b) 

to ensure the concrete integrity and to 

protect it from thermal degradation. 

During the reactor shutdown the oil 

system alone has to be able after few 

days to remove all the decay heat 

generated by the fuel. In case of the 

reactor vessel leak and loss of the oil 

system, the water system should be 

able to remove the decay heat 

generated by the core and to maintain 

the concrete below 70 °C. 

2.3 Main design-basis 

scenarios 

The reactor pit must be designed for 

the following three main scenarios: 

p Scenario 1: Normal operation: 

The main vessel temperature is at 

about 400 °C. The operation of the 

oil cooling system is sufficient to 

maintain the correct thermal 

conditions in the pit (i.e. less than 

70 °C for the concrete of the mixed 

structures). 

p Scenario 2: Operation in exceptional 

decay heat removal 

regime: The safety studies should 

take into account exceptional 

situations of successive losses of 

decay heat removal systems. In this 

case, in exceptional situations of 

Categories 3 and 4, the reactor 

vessel is allowed to reach temperature 

of 650 °C. The two cooling 

systems (oil and water) must make 

it possible to maintain the concrete 

temperature below 70 °C while 

playing an important role in the 

decay heat removal (in this study 

only the oil cooling system is taken 

into account; further publication is 

following). 

p Scenario 3: Operation in accident 

situation of sodium leakage: 

In a situation of little leak, 

vigorous sodium cooling is possible 

with the redundant and available 

DHRS, to bring the sodium to a 

temperature corresponding to the 

handling temperature (180 °C). 

Therefore, the maximum temperature 

of the sodium in the gap 

should not exceed 200 °C. The 

demonstration of the oil cooling 

system availability in case of 

reactor vessel leakage is difficult 

and we assume as hypothesis that 

the oil cooling system is no longer 

available. The operation of the 

water cooling system alone must 

be sufficient to maintain the 

concrete temperature below 70 °C 

(further publication). 

NB: But other studies need later to 

be performed, taking in account a 

leakage (or combined) due to a high 

thermal transient (e.g., due to failure 

of DHR system) and if the failure is 

due to the severe accident. In this case 

the sodium temperature will be higher 

3 Computational fluid 

dynamics analysis 

3.1 CFD model 

The objective is the development of a 

simple Computational Fluid Dynamics 

(CFD) model to compute the steadystate 

heat transfer from the reactor 

vessel through the reactor pit. The 

computations are performed with 

ANSYS CFX which is a parallelized 

high-performance CFD software tool 

[9]. The steady-state model is based 

on finite volume technique applied to 

solve the Navier-Stokes equations. To 

achieve results with low computation 

time, the reactor pit is divided into 

several sections and the CFD analysis 

is performed for every section. The 

drawing of the geometry for one 

section, the “elementary cell” of the 

reactor pit structure, is shown in 

Figure 6. For the calculation example, 

the oil cooling system is installed 

inside the liner of the special wavy 

shape (see Figure 7). 

The following simplifications and 

assumptions are applied at this stage 

of computations: 

p The gap between the reactor vessel 

and the liner is considered as 

vacuum for first CFD computations 

to minimize the CPU time. 





| Fig. 6. 

Drawing of an “elementary cell” of the reactor pit model. 

| Fig. 7. 

The CFD model of an “elementary cell” of the reactor pit model. 

p The material of the insulation layer 

is assumed to be glass wool for the 

current study. 

p Only steady state computations are 

performed. 

Fast solution is important to be able to 

perform the large amounts of 

parametric studies. The resulting 

CFD model geometry is shown in 

Figure 6. The domains for the model 

are as follows (l is the thermal 

conductivity): 

p The outer surface of the reactor 

vessel wall is shown on the lefthand 

side (red). 

p The gas gap (white); l = 

0.026 W m -1 K -1 . However, for the 

computations nearly-vacuum is 

assumed here as worst case and the 

heat transport parameters (density 

and specific heat) are set to very 

low values and the solution of the 

flow is switched off. 

p The stainless steel liner of the wavy 

shape is proposed (blue); l = 

60.5 W m -1 K -1 with the pipes of the 

oil cooling system inside (dark 

blue). 

p The insulation layer (yellow); l = 

0.04 W m -1 K -1 . 

p The concrete structure (grey); l = 

1.4 W m -1 K -1 . 

p The right-hand side surface interfaces 

the environment (black). 

The boundary conditions for the CFD 

model in Figure 6 are as follows: 

p Fixed temperature of the reactor 

vessel wall (red surface at the left) 

for different axial levels (T vw = 

400 °C, 500 °C, 600 °C and 700 °C). 

This is intended to represent the 

different levels of the heat source 

inside the vessel. 

p The heat sink is the environment 

outside of the concrete structure 

(black surface at the right). As a 

first approximation the following 

data is taken for the heat transfer: 

T a = 50 °C, k = 6 W m -2 K -1 , where 

k is the heat exchange coefficient. 

For later calculations, the heat 

sinks will include the water cooling 

system. 

p The other outer surfaces are 

defined as “symmetric”, i.e. 

adiabatic. 

Calculations were made first 

without taking into account the oil 

cooling system in order to see if 

one could reach, with the only 

heat removal to the environment, 

required temperature level in the 

concrete domain. 

3.1 Analytical solution 

for verification 

The current CFD model without 

local heat sinks (the cooling systems) 

is convenient for verification by 

com parison to an analytical solution. 

This requires one-dimensional 

model of the heat transfer by radiation 

in the gap and by heat conduction 

in the solids. The heat 

path is from the hot reactor vessel 

wall towards the outer concrete 

surface facing the environment. 

The heat flow due to radiation in the 

gap between a hot surface (index 1) 

and a cold surface (index 2) can be 

written as: 

(1) 

Hereby T is the (absolute) temperature 

and σ = 5.67⋅10 -8 W m -2 K -4 is the 

Stefan-Boltzmann constant. Assuming 

emissivity of stainless steel for both 

surfaces (ε1 = ε2 = ε = 0.4, a typical 

value for stainless steel) and full 

visibility (F 12 = 1) and for A 1 ≈ A 2 = A 

the temperature of the cold surface 

yields: 

(2) 

Heat conduction through a cylindrical 

solid wall in direction of the radius r 

can be written as 

 

(2) 

Hereby l is the thermal conductivity 

of the solid and A(r) = 2πrh the 

cylindrical surface at the radius r and 

for a segment of the height h. The 

integration between the radii r 1 and r 2 

with the corresponding temperatures 

T 1 and T 2 yields the temperature at the 

outer radius 

 

(1) 

3.3 CFD results without 

the cooling system 

For this computation the oil cooling 

channels in the liner in Figure 6 

are not taken into account. The liner 

is considered to be flat (equal 

thickness) and without pipes. The 

temperature along the heat path 

in the centre of the domains is 

shown in Figure 8. In Figure 8a, 

the constant-temperature boundary 

condition is applied at the primary 

vessel wall (T vw = 400 °C, 500 °C, 

600 °C and 700 °C, red surface in 

Figure 7). In the nearly-vacuum 

domain of the gap the heat transfer 

takes place by means of radiation and 

the temperature is almost identical to 

the boundary condition. The temperature 

of the metal liner is also very 

close to the temperature of the vessel 

(since the cooling system is not 

modelled in this case). As expected, 

the main temperature drop takes 

place in the insulation layer. However, 

the temperature in the concrete is 

mostly above the required limit of 

70 °C as shown in Figure 8b. Furthermore, 

in Figure 8a and b the CFD 

results are compared with the analytical 

solutions of Eq. (2) and Eq. (4). 

The maximum deviation of the 

temperature is less than 5 °C. 







(a) 

| Fig. 8. 

The temperature along the heat path (in x direction); computation (CFD) and analytical solution (Ana); a) the whole length; b) the concrete domain. 

The dotted red line denotes the temperature limit of 70°C. 

| Fig. 9. 

Temperature field, T vw = 700 °C, T cc = 200 °C. 

3.4 CFD results with 

the cooling system 

For this computation a constant 

average temperature T cc at the oil 

cooling channel walls (Figures 5 and 

6) is set as boundary condition. The 

aim is to understand the interaction 

between the reactor vessel wall 

temperature, the oil cooling system 

temperature, and the maximum 

concrete temperature. The final goal is 

to estimate the power removed by the 

oil cooling system. 

In Figure 9 the temperature field 

for the computed elementary cell is 

depicted. The hot reactor vessel with 

the constant temperature of the vessel 

wall equal to T vw = 700 °C is on the 

left-hand side. The temperature of the 

oil cooling channel walls is set to T cc = 

200 °C. Most of the heat transfer takes 

place between the vessel wall and the 

cooling channel. The maximum 

concrete temperature is slightly below 

70 °C. 

In Figure 10 the temperature is 

shown along the main heat path (from 

the hot main vessel wall at the left 

towards the ambient on the right). In 

Figure 10a the temperature is shown 

along the x axis (symmetry centre line 

of the elementary cell) and along a 

(b) 

line cutting the centre of one of 

the cooling channels. The case 

corresponds to the temperature field 

in Figure 9. As expected, the hottest 

location of the liner is at the centre 

line (maximum distance to the cooling 

channels). The temperature minimum 

is defined by the cooling channel temperature 

(T cc = 200 °C). In Figure 10b 

and 10c the temperature is given for 

four different vessel wall temperatures. 

For all cases, the concrete 

temperature remains below the limit 

(70 °C) denoted by the red dotted 

line. The case with T vw = 700 °C corresponds 

to Figure 10a and to the 

temperature field shown in Figure 9. 

In Figure 11a the maximum concrete 

temperature T c is shown versus 

the cooling channel temperature for 

different vessel wall temperatures. 

The case discussed in Figure 9 and in 

10a corresponds to the point T vw = 

700 °C and T cc = 200 °C which is 

slightly below the dotted red line. 

Consequently, according to this model 

the temperature of the cooling channel 

wall has to be maintained below 

200 °C to keep the concrete temperatures 

below the limit. For lower vessel 

wall temperatures the cooling channel 

wall temperature can be higher. 

(a) (b) (c) 

| Fig. 10. 

Temperature profile along the x-axis for a cooling channel temperature of T cc = 200°C. 





(a) 

| Fig. 11. 

a) The maximum concrete temperature and b) Power removed by the oil cooling system at differ-ent temperatures of the reactor vessel and the oil cooling 

channel wall. 

Parametric calculations of the 

power removed by the oil cooling 

system at different vessel temperatures 

are shown in Figure 11b. They 

are evaluated with grey body factors 

of vessel and liner equal to 0.4 that 

is a typical value for stainless steel. 

If necessary this value could be 

increased by liner surface processing 

to increase its emissivity and the 

associated decay heat removal capacity. 

Calculations were performed 

for various oil temperatures (Figure 

11b). An average temperature of 

about 200 °C is proposed for operation 

which is far below the operating range 

of the “Therminol SP” oil (315 °C). 

The power removed by radiation 

from the reactor vessel at 400 °C 

(~nominal core inlet temperature) is 

about 3 kW/m 2 .The surface of the 

reactor vessel, radiating towards 

the oil cooling system, is about 

1,050 m 2 . So at nominal operation, 

approximately 3 MW will have to be 

removed in order to maintain the 

oil at an average temperature of 

approximately 200 °C. 

In exceptional regime of decay 

heat removal (situations in category 3 

and 4) the system can then remove 

(the main vessel being at 650 °C), 

a power of about 15 MW. This 

value doesn’t take into account the 

exchanges by gas convection between 

vessel and liner, and could be also 

increased by special processing of the 

liner surface to increase its emissivity 

coefficient. The value of 15 MW corresponds 

to the decay heat power 

level after about three days. 

For Scenario 2 (See Section “Main 

design-basis scenarios”) we assume 

that the average oil temperature (and 

therefore the liner) is at ~200 °C. For 

Scenario 3 we assume that the other 

DHRS guarantee that the primary 

sodium is at about 200 °C and the oil 

system is out of operation, therefore 

the liner will be also at ~200 °C. Thus, 

the conducted preliminary analysis 

based on the use of the oil cooling 

system alone is potentially applicable 

to all three scenarios. Nevertheless, 

the necessity of the water cooling 

system will be analysed at the next 

phase of the (more detailed) analysis. 

NB: thermal exchange have been 

calculated only with radiative and 

conduction effects. The exchanges by 

convection have not been taken in 

account and should be estimated later. 

4 Conclusion 

Within the Euratom H-2020 ESFR- 

SMART project, various design 

options have been proposed to 

improve the safety of the ESFR 

concept to the level of the requirements 

for the Generation-IV reactors 

based on both present and previous 

projects experiences. One of the main 

innovative approaches to the safety 

design of the ESFR concepts concerns 

the re-placement of the safety vessel 

with reactor pit in order to provide 

improved operational and safety 

functions of both the decay hear 

removal system and the containment. 

The focus of the present paper is the 

preliminary CFD analyses of the 

innovative decay heat removal design 

implemented in the reactor pit. 

Two active (forced-convection) 

cooling systems are proposed: an oil 

cooling system close to the metallic 

liner and a water cooling system in 

the reactor pit concrete structure. 

The main goals of the cooling systems 

are: 

1) to keep the reactor vessel at 

acceptable temperature level and 

therefore provide its integrity and 

2) to maintain the concrete temperature 

below 70 °C in the three 

scenarios of operation envisaged 

(b) 

(normal operation, decay heat 

removal without and with primary 

sodium leak). 

The conducted CFD analysis in the 

present paper is concentrated on the 

evaluation of the oil cooling system. It 

considers vessel wall temperatures up 

to 700 °C. The main heat transfer 

mechanism from the wall towards a 

metal liner at the other side of the gap 

is assumed to be thermal radiation. 

It can be shown, in the scenarios 

studied, with the oil cooling system 

designed to keep the liner temperature 

at around 200 °C, that this system 

alone can fulfill the goals to safely 

remove the residual heat of about 

3 MW in nominal conditions and to 

maintain the temperature of the 

concrete of the reactor pit below 70 °C 

in the two accidental situations. 

The water cooling system seems only 

a supplementary safety protection 

against any pit structure concrete 

damage and will be subject of further 

publication. 

Acknowledgement 

The research leading to these results 

has received funding from the 

Euratom research and training 

programme 2014-2018 under grant 

agreement No 754501. 

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(2017). 

[7] Guidez, J., Gerschenfeld, A., Grah, A., Tsige-Tamirat, H., 

Mikityuk, K., Bodi, J., Girardi, E., “European Sodium Fast 

Reactor: innovative design of reactor pit aiming at suppression 

of safety vessel”, ICAPP 2019, May 12 – 15, Juan-les-pins, 

France. 

[8] Therminol® SP Heat Transfer Fluid: 

https://www.therminol.com/products/Therminol-SP 

[9] ANSYS CFX: https://www.ansys.com/products/fluids/ansys-cfx 

Authors 

Aleksander Grah 

JRC, Petten, Netherlands 

aleksander.grah 

@ec.europa.eu 

Dr. Aleksander Grah is Scientific Officer at the 

European Commission Joint Research Centre, Petten, 

The Netherlands. Currently, he is working in the 

field of Safety of advanced reactors applying Computational 

Fluid Mechanics (CFD). 

Haileyesus 

Tsige-Tamirat 

JRC, Petten, Netherlands 

haileyesus.tsige-tamirat 

@ec.europa.eu 

Antoine Gerschenfeld 

CEA, Paris-Saclay, France 

antoine.gerschenfeld 

@cea.fr 

Antoine Gerschenfeld is a research engineer at the 

Fluid Mechanics Section of the French Commissariat 

à l’Energie Atomique, specializing in the thermalhydraulics 

of Generation 4 nuclear reactors. He is also 

an assistant professor at Ecole Polytechnique and a 

member of the Framatome scientific advisory board. 

Konstantin Mikityuk 

PSI, Villigen AG, 

Switzerland 

konstantin.mikityuk 

@psi.ch 

In charge of Advanced Nuclear Systems group at 

Paul Scherrer Institute, Switzerland. Main interest: 

research related to design and safety assessment of 

Generation-IV fast-spectrum nuclear reactors 

Janos Bodi 

PSI, Villigen AG, 

Switzerland 

janos.bodi@psi.ch 

Dr. H. Tsige-Tamirat is Scientific Officer at the 

European Commission Joint Research Centre, Petten, 

The Netherlands. Currently, he is working in the field 

of Safety of advanced reactors. 

Joel Guidez 

CEA, Paris-Saclay, France 

joel.guidez@cea.fr 

Janos Bodi is a nuclear engineering professional. 

His main professional activities are related to 

neutronic and thermal-hydraulic analysis of 

Sodium-cooled Fast Reactors. Furthermore, he 

contributes to the reactor design development 

of Generation IV reactor concepts. 

Enrico Girardi 

EDF, Paris-Saclay, France 

enrico.girardi@edf.fr 

Joel Guidez retired in March 2020 while remaining 

scientific advisor to the CEA and working in ESFR 

SMART European project. Before this he was member 

of the operational committee of the office of the High 

Commissioner, honorary President of the ST7 SFEN 

section, member of the RGN editorial board, 

representative of France at the RSWG of GIF, president 

of the GCFS – French safety advisory group – tripartite 

CEA/EDF/AREVA, and scientific manager of the GEN 

IV segment at CEA. 

Dr. Enrico Girardi is currently an Expert on Reactor 

Physics, Modelling & Simulation at Eléctricité de 

France R&D Safety and Fuel Cycle group. His main 

research fields are on both PWR, SMR and SFR reactor 

physics, and in particular on neutronic modelling. In 

the last 15 years he has contributed to several reactor 

calculation chain development at EDF and has been 

leading the EDF R&D “Generation IV reactor” project 

from 2015 to 2018. 





Study on Verification of SPACE Code 

Based on an MSGTR Experiment at the 

ATLAS-PAFS Facility 

Kyungho Nam 

A Multiple Steam Generator Tube Rupture (MSGTR) accident is defined in Korea as an accident in which more than 

five U-tubes of a steam generator break down. To understand the thermal hydraulic phenomena of a MSGTR accident, 

an experimental study was conducted by KAERI. The experiment was conducted to simulate the transient phenomena 

caused by the rupture of five U-tubes, and to validate the heat removal capacity of the Passive Auxiliary Feedwater 

System (PAFS) during the transient. 

In this paper, an MSGTR experiment 

at the ATLAS-PAFS test facility was 

simulated using the SPACE code to 

verify the prediction capability of this 

code for multiple failure accident, 

which is involved in the design extension 

condition. The overall system 

transient behavior obtained using 

SPACE code showed similar trends 

with the experimental results in terms 

of factors such as the system pressure, 

mass flow rate, and collapsed water 

level on the component. Additionally, 

a sensitivity analysis was conducted 

using the experimental correlation 

for PAFS which is included in the 

wall condensation model as an 

option in SPACE code. As a sensitivity 

calculation results, it is also recommended 

that the PAFS model in 

SPACE code be applied to obtain more 

accurate prediction results about the 

PAFS operation by performing a 

safety analysis of the APR+ nuclear 

power plant. 


1.1 Background 

Following the Fukushima nuclear 

disaster in 2011 and based on the 

lessons learned from the accident, 

there have been many changes to the 

relevant safety design criteria and/or 

regulations around the world. The 

Nuclear Safety and Security Commission 

(NSSC) in Korea has required 

plant-specific accident management 

plans, which extend beyond design 

basis accidents to include severe 

accidents. The revised regulation in 

Korea has determined a list of multiple 

failure accidents that must be considered 

for any accident management 

plan [1]. Multiple failure accidents 

should be considered for Design 

Extension Conditions, which is 

defined by the International Atomic 

Energy Agency (IAEA) Specific Safety 

Requirement [2, 3]. 

The Multiple Steam Generator 

Tube Rupture (MSGTR) accident is 

selected as one of the multiple failure 

accidents by the Korean NSSC, and it 

is defined as an accident in which 

more than five U-tubes of a steam 

generator rupture. In a MSGTR 

accident, the pressurized primary 

coolant leaks into the secondary 

system and thus exposes radioactive 

material. Therefore, it is important 

that the extent of the leak is limited 

and that the pressure drop across the 

break be kept as low as possible to 

reduce the radioactive release. 

Compared to single tube rupture, 

MSGTR causes quicker depressurization 

of the reactor coolant system 

(RCS) and places greater demand on 

the inventory makeup process. 

To elucidate the thermal hydraulic 

process of the MSGTR accident, 

an experimental study using the 

Advanced Test Loop for Accident 

Simulation (ATLAS) facility was 

conducted by the Korea Atomic 

Energy Research Institute (KAERI) 

[4]. The experiment simulated the 

rupture of five steam generator tubes, 

and the results showed that the 

Passive Auxiliary Feedwater System 

(PAFS) adopted in the Advanced 

Power Reactor Plus (APR+) had 

sufficient cooling capacity to mitigate 

the accident. The PAFS is one of the 

advanced safety features of a passive 

cooling system that allow it to 

replace a conventional active Auxiliary 

Feed-water System (AFWS) [5]. In a 

typical Nuclear Power Plant (NPP), 

a motor-driven or turbine-driven 

auxiliary feedwater is supplied after 

the wide-range water level of a steam 

generator is decreased below the low 

steam generator level. However, to 

confirm the cooling capability of 

PAFS compared to that of AFWS, an 

experimental scenario was conducted 

in which the PAFS was supplied to 

an intact SG instead of auxiliary 

feedwater. The PAFS is a passive 

system capable of condensing the 

steam generated in a steam generator 

and feeding the condensed water to 

the steam generator using gravity. 

For current safety analyses of 

Korean nuclear power plants, thermalhydraulic 

safety analysis codes 

supplied by foreign vendors such 

as Westinghouse and Combustion 

Engineering have been used. The 

Ministry Of Trade, Industry and 

Energy (MOTIE) in Korea launched 

the ‘Nu-Tech 2012’ project to improve 

the competitiveness of the Korean 

nuclear industry in 2006, and the 

Korean nuclear industry developed 

the SPACE (Safety and Performance 

Analysis CodE for nuclear power 

plants) code [6]. This code was 

approved by the Korean NSSC in 

2017, and it will replace outdated 

vendor-supplied codes and will be 

used for safety analyses of operating 

nuclear power plants in Korea as well 

as the design of advanced reactors. 

While the SPACE code will mainly be 

used in safety analyses, the code with 

best-estimate capabilities will be able 

to cover performance analysis as well. 

The programming language for the 

SPACE code is C++, and this 

code adopts the advanced physical 

modeling of two-phase flows, namely 

two-fluid three-field models which 

comprise gas, continuous liquid, and 

droplet fields. 

According to the General Safety 

Requirement (GSR) of IAEA, any 

calculation method and computer 

codes used in safety analysis must 

undergo verification and validation 

[7]. Further, verification calculations 

of system tests or integral tests are 

used to validate the general consistency 

of the revision [8]. Therefore, 

a verification calculation based on 

integral effect experiments is needed 

to improve the reliability of the prediction 

results of the SPACE code. 



Study on Verification of SPACE Code Based on an MSGTR Experiment at the ATLAS-PAFS Facility ı Kyungho Nam



In particular, the multiple failure 

accident condition is the new safety 

design criteria, so the verification 

process should be performed. To this 

end, the first objective of this study is 

to verify a multiple failure accident 

calculation capability of the SPACE 

code, while the second objective is to 

confirm the transient phenomena of 

MSGTR and the cooling effect of PAFS 

during MSGTR with respect to the 

first objective, as mentioned above. 

The heat loss phenomenon is a 

measure of the total heat transfer of 

heat from either conduction, convection, 

radiation, or any combination 

of these. Newton’s law of cooling 

states that the rate of heat loss of an 

object is directly proportional to the 

difference in the temperature between 

the object and its surroundings. Under 

the experimental conditions of high 

temperature and high pressure, the 

heat loss is particularly likely to 

increase because of the temperature 

difference between the experiment 

component and the surrounding 

atmosphere. This physical phenomenon 

can affect the heat transfer 

experiment, and it plays an important 

role in the performance of the system. 

The heat loss is a function of area in 

accordance with the convective heat 

transfer equation. According to the 

design document, the ATLAS facility 

has a relatively large surface area to 

volume ratio which is in accordance 

with the design characteristic [9]. For 

this reason, additional work to 

confirm the heat loss effect on the 

ATLAS-PAFS facility was also conducted 

to confirm the heat loss effect 

on the system behavior of the integral 

test facility. 

1.2 A brief description 

of ATLAS-PAFS facility 

As mentioned previously, KAERI has 

been operating an integral effect test 

facility, ‘ATLAS’, for the transient and 

accident simulation of a Pressurized 

Water Reactor (PWR) [10]. The 

reference plant of ATLAS is the 

APR1400, which has been developed 

by the Korean nuclear industry. 

ATLAS has the same two-loop features 

as the APR1400, and it is designed 

using the scaling method to simulate 

various test scenarios as realistically 

as possible [9, 10]. This test facility 

also includes design features of the 

OPR1000, which is Korean standard 

NPP, such as a cold-leg injection mode 

for safety injection and a low pressure 

safety injection mode. 

As mentioned above, the PAFS is 

one of the advanced passive safety 

systems adopted in the APR+, and an 

experimental program is currently 

underway at KAERI to validate the 

cooling and operational performance 

of the PAFS [11]. The main objective 

of the ATLAS-PAFS integral effect test 

is to investigate the thermal hydraulic 

behavior in the primary and 

secondary systems of the ATLAS 

during a transient at which PAFS is 

actuated. The PAFS facility is 

described in further detail below. 

2 Description of 

ATLAS-PAFS model 

for MSGTR scenario 

using SPACE code 

2.1 Experiment scenario 

To validate the prediction capability 

of the SPACE code, the experimental 

information provided by KAERI was 

utilized [4]. The target scenario for 

the experiment is a MSGTR with a 

PAFS operation occurrence and 

asymmetric cooling. To initiate the 

MSGTR transient, first, the break 

valve at the SGTR simulation pipe line 

is opened. By opening the break valve, 

the primary system inventory was 

discharged from the hot side of the 

lower plenum to the upper location of 

the steam generator secondary hot 

side. Next, the primary system began 

to be depressurized and the secondary 

side water level of the steam generator 

increased. At the same time as the 

HSGL signal occurrence, reactor trip 

occurred, and the core power started 

to decrease following the decay curve. 

For the transient calculation, the 

decay power curve was considered 

along with the tables for time versus 

power. The main feedwater isolation 

valves (MFIVs) and the main steam 

isolation valves (MSIVs) for two steam 

generators were closed after delay 

times. The main steam safety valves 

(MSSVs) on the steam line opened 

due to the pressure increase of SG-1, 

and these valves were kept in cyclic 

operation of opening and closing to 

protect the primary and secondary 

systems from over-pressurization. The 

accident caused the depressurization 

of RCS and resulted in a Low PZR 

Pressure (LPP) signal. Further, the 

Safety Injection Pumps (SIPs) began 

after delay times. It was assumed that 

only one safety injection pump per 

train was operated for the experiment 

scenario. In accordance with this 

assumption, SIP-1 and SIP-3 were 

available. The injection flow rate was 

applied using pressure – mass flow 

curve based on experiment data. To 

simulate an accident management 

measure based on the cooling performance 

of PAFS during a MSGTR, 

the PAFS was supplied to an intact 

SG-2 instead of auxiliary feedwater 

after the water level of the SG-2 fell 

below the PAFS operation set point 

due to the decay power. It was also 

assumed that the auxiliary feed 

water system would not work for the 

assessment of PAFS cooling capability. 

After the initiation of the PAFS, the 

decay heat was removed from the RCS 

by the natural convection of the PAFS. 

Finally, the whole system was cooled 

down in a stable manner with the 

successful operation of SIPs, MSSVs, 

and PAFS. 

2.2 Brief overview of model 

information 

For the experimental simulation, 

the ATLAS-PAFS test facility was 

modeled using SPACE code as shown 

in Figure 1. In the calculation, the 

decay power was imposed in 

accordance with the experiment, and 

the operation logics of the safety 

systems, such as safety injection and 

PAFS, were reflected. The geometrical 

and material information of the 

components and pipe lines in the 

ATLAS-PAFS test facility was also 

reflected [9, 12]. The reactor vessel 

was separated to simulate the core, 

bypass flow, reactor vessel lower 

plenum, and the reactor vessel upper 

head. The core model included the top 

and bottom inactive core regions, 

average channel, and hot channel. 

The safety injection system had four 

independent trains and a direct vessel 

injection (DVI) mode. Each train of 

the safety injection system consisted 

of safety injection pumps (SIPs). Injection 

lines could be aligned to either 

reactor vessel down-comer for DVI 

injection. The pressurizer was modeled 

as a single component with 

10 vertical nodes. The lower part was 

connected to hot leg through a surge 

line separated into five nodes. The hot 

legs and cold legs were modeled with 

four cells each, while the intermediate 

legs were modeled with five nodes. 

The steam generator included five 

nodes for the evaporator and two 

nodes for the economizer. The main 

steam safety valves were modeled into 

three separate groups; each group 

was operated on a different set point 

of pressure in the steam generator 

dome. 

2.3 SGTR simulation facility 

The geometry of the ATLAS facility, 

which was composed of a SGTR 

simulation pipe and connected with 





| Fig. 1. 

Modeling diagram of ATLAS and SGTR simulation pipe using SPACE code. 

a PAFS facility, was modeled as 

shown in Figure 1. In the ATLAS 

facility with the SGTR simulation 

pipe, the primary system inventory 

was discharged from the hot side 

of the lower plenum to the upper 

location of the SG-1 secondary 

hot-side to simulate a MSGTR 

accident. It is composed of a break 

simulation valve, an orifice flow 

meter, an orifice, and break nozzles. 

The ‘PIPE’ component option of SPACE 

code was used to model the upstream 

pipe, which was part of the SGTR 

simulation pipe; this pipe section was 

geometrically divided into 10 nodes. 

The break nozzle was installed to 

simulate a five-tube rupture with a 

non-choking orifice. This tube section 

is modeled using the ‘CELL’ component 

option of SPACE code. The 

input of the inner diameter is 

1.756 mm and the total flow area is 

the summation of the five-tube area. 

An orifice with a 1.68 mm hole was 

installed at the end of the break 

nozzles to simulate the choking flow 

condition at tube rupture, and the 

break nozzles were designed to 

maintain the equivalent pressure drop 

in the case of the non-choking flow 

condition. The mass flow rate through 

the SGTR simulation pipe was 

measured using an orifice flow meter. 

These orifice and orifice flow meter 

sections were modeled using the 

‘FACE’ component option of SPACE 

code. The experiment began by 

opening an initiation valve to simulate 

a MSGTR on the SG-1. This valve was 

modeled as the ‘TRIP VALV’ component 

option of SPACE code, and it 

was opened at the start of the transient 

calculation. The MSGTR occurs 

on the tubes of SG-1, which are 

connected to the SGTR simulation 

pipe; five break nozzles are opened in 

total [4]. 

2.4 Passive Auxiliary 

Feedwater System (PAFS) 

The steam supply and return water 

line connected the PCHX to the SG-2 

of the ATLAS [12]. Therefore, as 

shown in Figure 2, the PAFS was 

modeled by adding junctions at the 

main steam line and at the economizer 

nozzle as the inlet and outlet of 

the PAFS, respectively. The steam 

supply line and the return water line 

were divided into 24 nodes and 

31 nodes, respectively. The diameters 

of the steam supply line and the return 

water line were about 0.04 m and 

0.03 m, respectively. The PAFS 

operation valve was connected to the 

return water line and the feed water 

line, and it was modeled as a trip 

valve. This valve open signal was 

synchronized to the instant that the 

collapsed water level in the steam 

generator reached the set point of the 

low steam generator level. Once the 

valve was open, the latched option 

made it impossible for the valve to 

close again. The end of the return line 

was connected to the bottom nozzle 

of the steam generator economizer 

volume. The PCHX was the most 

important component of PAFS, and it 

was filled with condensate water and 

the return water line on steady state 

condition. The condensation tube of 

PCHX was modeled with 24 nodes as 

shown in Figure 3. The length of the 

horizontal nodes was about 0.23 m 

and the horizontal part of the 

PCHX was modeled as 1.806 m. An 

inclination of 3° was applied to the 

horizontal tube region while an 

inclination of 41.2° was applied to one 

inlet node and one outlet node to 

simulate a U-shaped bend. These 

design values were determined to 

prevent the condensation-induced 

water hammer inside the tube of 

PCHX [13]. The area of the PCHX pipe 

component was about 22.35 cm 2 , 

which equals the summation of the 





| Fig. 2. 

Connection nodding diagram of PAFS to ATLAS steam generator. 

| Fig. 3. 

Nodding diagram of PCHX. 

three-tube area. The connected 

heat structures were modeled as a 

cylindrical shape. The inner and outer 

coordinates were the inner and outer 

radii of the tube, respectively. The 

number of tubes was used as an input 

for equivalent heat transferring area. 

The top and bottom headers of PCHX, 

which both play roles in preventing 

the vibration of the PCHX tube in 

the PCCT, were modeled as cell 

components. The Passive Condensate 

Cooling Tank (PCCT) of PAFS was 

designed as a rectangular pool. When 

the PAFS was actuated, the heat 

transfer from the PCHX caused the 

pool water in the PCCT to evaporate, 

after which the steam flowed through 

the upper pipe on the top. This upper 

pipe was connected to the upper cells 

of the PCCT. Finally, the ‘TFBC’ 

component, which was connected to 

the upper pipe, played a role in maintaining 

the atmospheric pressure. The 

water pool of the PCCT served as a 

heat sink. The core decay heat was 

transferred through the condensation 

of steam inside the tubes. Therefore, 

the extracted heat increased the 

temperature of the pool water. It is 

expected that rigorous boiling 

occurred at the tube outside the 

surface, and a strong buoyancy flow 

was also expected. In order to simulate 

natural convection by buoyancy flow 

in the PCCT, the 3D option of SPACE 

code was applied to model the PCCT 

facility as illustrated in Figure 4. This 

PCCT model was divided into 

182 cells, where the y direction and 

the z direction were respectively 

divided into 13 and 14 cells. The 




| Fig. 4. 

Nodding diagram of PCCT using ‘3D’ option of SPACE code. 

cooling water was filled up to nine 

nodes of the z direction, and the upper 

nodes were in atmosphere condition. 

The initial cooling water level in PCCT 

was about 3.8 m while the water 

temperature was 28.8 °C. 

3 Results and discussions 

3.1 Steady-state results 

Before using the SPACE model for 

transient analyses, a consistent set of 

parameters must be obtained by the 

steady-state initialization process. 

Table 1 lists the initial conditions of 

the experiment, calculated results of 

the SPACE code, and difference. The 

initial core power generated by the 

heater rods was 1.627 MW, and the 

heat loss of the primary piping into 

the atmosphere was estimated to 

be about 97.1 kW based on the 

information of the initial experiment 

condition. During the initial conditions, 

the major thermal hydraulic 

parameters, including the system 

pressure, fluid temperature, and mass 

flow rate, were reasonably consistent 

with the experiment condition. 

The calculation time started from 

-1000.0 s to 0.0 s and all design 

parameters converged after -500.0 s. 

Parameter Experiment SPACE code Difference (%) 

Primary system 

Core power (MW) 1.627 1.627 0.00 

Heat loss (kW) 97.1 97.2 0.40 

Pressurizer (PZR) pressure (MPa) 15.52 15.50 0.13 

PZR level (m) 3.71 3.71 0.00 

Cold leg flow rate (kg/s) 1.9728 1.941 1.61 

Core inlet temp. (°C) 292.0 289.4 0.89 

Core outlet temp. (°C) 327.5 325.2 0.70 

Secondary system 

Steam flow rate (kg/s) 0.4019 0.4153 3.33 

Feed water flow rate (kg/s) 0.4209 0.4194 0.36 

Feed water temp. (°C) 233.3 233.3 0.00 

Steam dome pressure (MPa) 7.83 7.81 0.20 

Steam temp. (°C) 295.7 294.2 0.51 

SG collapsed water level (m) 4.97 4.97 0.00 

Passive Auxiliary Feedwater System (PAFS) 

Initial PCCT level (m) 3.8 3.8 0.00 

Initial PCCT temp. (°C) 28.8 28.8 0.00 

| Tab. 1. 

Comparison results between initial condition of experiment and steady state results of SPACE code. 

After checking the steady state condition, 

the transient calculation 

started at 0.0 s of the steady state 

condition. 

3.2 Transient analysis results 

According to the agreement of the 

ATLAS Domestic Standard Problem-05, 

the experimental data should be 

confidential. Therefore, all of the 

experiment results in this paper 

including the time frame (t*) were 

divided by an arbitrary value and 

plotted on the non-dimensional axis. 

3.2.1 Sequence of transient 

calculation result 

Table 2 summarizes the measured 

and calculated sequences of a 

MSGTR-PAFS test. The transient was 

initiated with the MSGTR in normal 

condition. To initiate the MSGTR 

transient, the break valve was opened 

at 0.0 s with the pressurizer heater off 

according to the experiment scenario. 

The break flow contributed to the 

steam generator overfill, and the 

collapsed water level of the SG-1 

reached a set point of HSGL signal, 

leading to the generation of the HSGL 

signal. This signal was generated at a 

later time in the calculation case than 

it was in the experiment case. The 

reactor trip occurred simultaneously 

due to the HSGL signal, and the MSIVs 

and MFIVs were also closed after their 

respective delay times. After the 

MSIVs were closed, the pressure in 

steam generator was controlled by the 

cyclic operation of MSSVs. According 

to the experimental assumptions, the 

heater power which simulates the 

decay power had the ANS-73 decay 

curve applied for the experimental 

condition, and the heater power 

started after the reactor trip. For the 

calculation, time verse decay power 

data was applied in accordance with 

the experiment information [4]. The 

time at which the MSSVs first opened 

in the experiment and the calculation 

results were the same after the HSGL 

signal was triggered. During this 

steam generator overfill period, the 

pressure in the primary system 

decreased substantially. As a result, 

when the primary system pressure 

decreased to the set point of the LPP 

trip, the LPP signal was triggered, and 

the SIP injection was initiated after 

the delay time. The collapsed water 

level of SG-2 decreased due to the 

decay power, then reached a set point 

of PAFS operation. Finally, the PAFS 

operation valve automatically opened 

and passive cooling using PAFS 

was initiated. 





Event Experiment (t = t*) SPACE code (t = t*) Remarks 


MSGTR initiation 0.0000 0.0000 PZR heater off at same time as experiment scenario 

HSGL signal 0.0011 0.0024 SG collapsed water level > set-point 

Reactor trip 0.0011 0.0024 

MSIV close 0.0015 0.0028 

MFIV close 0.0018 0.0031 

3.2.2 Primary system behaviors 

The decay power was one of the important 

factors in this analysis, and 

the calculation results were consistent 

with the experimental result, as 

shown in Figure 5. Figure 6 shows 

the measured and calculated break 

flow, which were normalized by the 

maximum value of the experiment. 

The break flow rate largely depended 

on the pressure difference between 

the primary and the secondary 

systems. Therefore, as shown in 

Figure 6, the peak flow rate occurred 

at the beginning of transient. After the 

occurrence of peak flow, the break 

flow in the experiment case oscillated 

and gradually decreased. On the other 

hand, the calculation result shows 

that the break flow was maintained 

constantly. The break flow was deeply 

related to the difference between 

the pressures of the primary and 

secondary systems. After the pressure 

Coincident with HSGL 

Decay power start 0.0023 0.0036 Delay times after reactor trip 

MSSV first opening 0.0026 0.0033 SG pressure > set-point 

LPP signal 0.0248 0.0232 PZR pressure < set-point 

SIP injection start 0.0276 0.0260 Delay times after LPP signal 

PAFS operation start 0.7204 0.7173 SG collapsed water level < set-point 

| Tab. 2. 

Sequence of transient analysis result. 

of the primary system was reduced by 

the break, the pressure of the PZR 

was predicted to be lower than the 

experiment results (t* < 0.2 in 

Figure 7). The pressure of the primary 

system was relatively lower than that 

obtained in the experiment results, 

thus indicating a lower break flow 

rate. Due to the lower break flow rate, 

the pressure of PZR was maintained 

somewhat higher than the experiment 

result, which resulted in the break 

| Fig. 5. 

Comparison result of core power. 

| Fig. 6. 

Comparison result of MSGTR break flow rate. 

| Fig. 7. 

Comparison result of system pressure. 

| Fig. 8. 

Comparison result of SIP injection rate. 




| Fig. 9. 

Comparison result of SG collapsed water level. 

| Fig. 10. 

Comparison result of PAFS line mass flow rate. 


| Fig. 11. 

Comparison result of PCCT water level. 

| Fig. 12. 

Comparison result of fluid temperature at PAFS line after PAFS operation. 

flow rate being maintained (t* > 0.2 

in Figure 7). At the moment the 

break occurred, the pressure in the 

pressurizer began to decrease rapidly 

due to the loss of coolant through the 

break point. The depressurization 

rates of the pressurizer obtained from 

the experiment and the calculation 

results were almost the same, as 

shown in Figure 7. The pressure in the 

pressurizer continually decreased and 

reached the set point of the LPP signal. 

As the LPP signal was generated, 

the SIP injection was initiated, as 

shown in Figure 8. During the 

beginning phase of SIP injection, the 

calculation result of the SIP flow 

rate was slightly higher than the 

experimental result. This was 

attributed to that fact that the PZR 

pressure according to the calculation 

result remained lower than that 

obtained in the experiment result. 

As the accident progressed, the SIP 

injection rate obtained in the calculation 

result was consistent with the 

experiment result. When the RCS 

pressure reached the saturation 

pressure due to the decay power, 

plateau pressure was observed. In 

addition, the water injected by the SIP 

contributed to the primary system 

pressure. Thus, as shown in Figure 7, 

the primary system pressure was 

maintained. 

3.2.3 Secondary system 

behaviors 

After the MSIVs were closed by the 

HSGL signal, the pressure in the steam 

generator was maintained within the 

range of the opening and closing set 

points of the MSSVs. The transient 

behavior of pressure in the steam 

generator shows that the calculation 

result is consistent with the experimental 

results. Following the PAFS 

operation, the pressure in an intact 

SG-2 drastically decreased due to the 

cooling by PAFS; the SG-2 depressurization 

trend is similar to the core 

cooling rate trend. 

Figure 9 shows a comparison of 

the SG collapsed water levels in the 

experiment and in the SPACE calculation. 

The water level of broken SG-1 

increased and reached the full water 

level. By contrast, that of an intact 

SG-2 decreased rapidly and reached a 

set point of the PAFS operation. As the 

PAFS signal was triggered, the PAFS 

operation valve automatically opened, 

and the main steam from the SG-2 

flowed into the steam supply line. As 

shown in Figure 10, the mass flow 

peaked, and then natural circulation 

flow was formed. The mass flow rate 

in the case of SPACE calculation was 

consistent with the experimental 

result. The main steam from the steam 

supply line flowed into the condensation 

tubes, and the condensate 

circulated through the return water 

line to the economizer of the 

steam generator. After the PAFS was 

actuated, the water level also 

increased due to the thermal 

expansion, as shown in Figure 11. 

3.2.4 Wall condensation heat 

transfer model for PCHX 

The wall condensation heat transfer 

rate in PCHX is one of the dominant 

factors for determining the PAFS 

cooling capability. Therefore, the 

many precedent studies related 

to condensation in PCHX were 





performed. The wall condensation 

models were incorporated into the 

SPACE code heat transfer package. For 

pure steam condensation, like the 

problem addressed in this paper, the 

same model used in RELAP5/MOD3.3 

was selected as a default model. The 

maximum value among Nusselt’s 

[14], Shah’s [15], and Chato’s [16] 

correlations is used to consider the 

geometric and turbulent effects. The 

experimental correlation for PAFS is 

also included in the wall condensation 

model as an option in SPACE code 

[17]. Based on the calculation results 

mentioned in chapter 3.2.1, this 

model was applied. In this section, 

the default model and experiment 

correlation model for PAFS were 

compared to confirm the prediction 

ability of SPACE code for PCHX 

cooling performance. Chen’s model, 

which is the default option in SPACE 

code, is applied to the outside of PCHX 

[18]. 

The fluid temperature after PAFS 

operation is shown in Figure 12. The 

calculation result shows that the fluid 

temperature on the return water line 

is higher than the experiment, and 

that the default option case is highest. 

That means that the calculation 

results which were obtained using the 

default option and the PAFS model 

underestimated cooling performance. 

However, the calculation using the 

PAFS model was more accurate than 

the default option. 

4 Conclusions 

In this study, a MSGTR experiment 

with the PAFS operation performed by 

KAERI was simulated using the SPACE 

code. This study focused on verifying 

the prediction capability of the SPACE 

code for MSGTR accidents, which is 

one of the multiple failure accidents, 

and to evaluate the cooling capacity 

of the PAFS during such an accident. 

The calculation results showed that 

the overall system transient results 

using the SPACE code showed 

comparatively similar trends with the 

experimental results in terms of the 

system pressure, mass flow rate, and 

collapsed water level in components. 

Therefore, it was concluded that the 

SPACE code reasonably predicted the 

experiment. And, the default model, 

which uses the maximum value 

among Nusselt’s, Shah’s, and Chato’s 

and the experiment correlation 

model for PAFS in SPACE code were 

compared to confirm the prediction 

ability of SPACE code for PCHX 

cooling performance. The calculation 

result using the PAFS model was more 

accurately estimated than the default 

option. 

Based on the present calculation 

results, the following conclusions 

were obtained in this study. First, the 

experiment and calculation results 

showed that the cooling capability of 

PAFS is sufficient to replace the active 

auxiliary feedwater system during 

MSGTR transient. Additionally, it is 

recommended that the PAFS model in 

SPACE code be applied for more 

accurate prediction results for PAFS 

operation by performing a safety 

analysis of an APR+ nuclear power 

plant. 

Acknowledgements 

This work was performed within the 

program of the fifth ATLAS Domestic 

Standard Problem (DSP-05), which 

was organized by the Korea Atomic 

Energy Research Institute (KAERI) in 

collaboration with the Korea Institute 

of Nuclear Safety (KINS) under the 

national nuclear R&D program funded 

by the Ministry of Education 

(MOE) of the Korean government. 

The authors are also grateful to the 

fifth ATLAS DSP-05 program participants: 

KAERI for the experimental data 

and to the council of the fifth DSP- 

05 program for providing the opportunity 

to publish the results. 

Acronyms and Abbreviations 

NPP 

HSGL 

LPP 

MSIV 

MFIV 

Nuclear Power Plant 

High Steam Generator Level 

Low Pressurizer Pressure 

Main Steam Isolation Valve 

Main Feed-water Isolation Valve 

MSSV Main Steam Safety Valve 

SIP 

PAFS 

PCCT 

Safety Injection Pump 

Passive Auxiliary Feedwater System 

Passive Condensate Cooling Tank 

PCHX Passive Condensation Heat Exchanger 

HL 

CL 

IL 

SG 

Hot Leg 

Cold Leg 

Intermediate Leg 

Steam Generator 

References 

[1] Korean NSSC, Regulations on the Scope of Accident 

Management and the detailed criteria of Accident 

Management Capability Evaluation, Notification No. 2017- 

34 of the Nuclear Safety and Security Commission in Korea, 

Rev.1 (2017). 

[2] IAEA, Safety of Nuclear Power Plants: Design, IAEA Specific 

Safety Requirements No. SSR-2/1, Rev.1 (2016). 

[3] J.H. Koh et al., Current status of design extension conditions 

technology development for prevention of severe accident 

in nuclear power plants, 13th International Conference on 

probabilistic Safety Assessment and Management (PSAM 

13) (2016). 

[4] KAERI, Experimental Study on the Multiple Steam Generator 

Tube Rupture with Passive Auxiliary Feedwater System 

Operation, KAERI/TR-8010/2020 (2020). 

[5] J. Cheon et al., The Development of a Passive Auxiliary 

Feedwater System in APR+, ICAPP2010, San Diego, USA 

(2010). 

[6] S.J. Ha et al., Development of the SPACE code for nuclear 

power plants, Nuclear Engineering and Technology 43, 

(2011) 45-62. 

[7] IAEA, Safety Assessment for Facilities and Activities, IAEA 

Safety Standards Series No. GSR Part 4, Rev.1 (2016). 

[8] Petruzzi, A., D’Auria, F., Thermal-Hydraulic System Codes in 

Nuclear Reactor Safety and Qualification Procedures, 

Science and Technology of Nuclear Installations, Vol. 2008, 

Article ID 460795, (2008) p.16. 

[9] KAERI, Description report of ATLAS facility and 

instrumentation, KAERI/TR-7218/2018 (2018). 

[10] Y.S. Kim et al., Commissioning of the ATLAS thermalhydraulic 

integral test facility, Annals of Nuclear Energy, 

Vol. 35, (2008) pp. 1791-1799. 

[11] K.H. Kang et al., Separate and integral effect tests for 

validation of cooling and operational performance of the 

APR+ Passive Auxiliary Feedwater System, Nuclear 

Engineering and Technology 44, No.6, (2012). 

[12] KAERI, Description report of ATLAS-PAFS Facility and 

Instrumentation, KAERI/TR-5545/2014 (2014). 

[13] B.U. Bae et al., Design of condensation heat exchanger for 

the PAFS(Passive Auxiliary Feedwater System) of APR+ 

( Advanced Power Reactor Plus)”, Annals of Nuclear Energy, 

Vol.46, (2012) 134-143. 

[14] Nusselt, W.A., The surface Condensation of Water Vapor, 

Zieschrift Ver. Deut. Ing., 60, (1916) p.541. 

[15] Shah, M.M., A general correlation for heat transfer during 

film condensation inside pipes, Int. J. Heat Mass Transfer, 

22, (1979) p.547. 

[16] Chato, J.C., Laminar Condensation inside Horizontal and 

Inclined Tubes, American society of heating, refrigeration 

and air conditioning engineering journal, 4, (1962) p.52. 

[17] B.U. Bae et al., Evaluation of mechanistic wall condensation 

models for horizontal heat exchanger in PAFS (Passive 

Auxiliary Feedwater System), Annals of Nuclear Energy, 

Vol.107, (2017) 53-61. 

[18] K.Y. Choi et al., Development of a wall-to-fluid heat transfer 

package for the SPACE code, Nuclear Engineering and 

Technology, Vol.41, No.9 (2009). 

Author 

Kyungho Nam 

Associate 

research engineer 

Korea Hydro & 

Nuclear Power Co., LTD. 

Central Research Institute 

Safety Analysis Group 

Deajeon, Republic of 

Korea 

Kyungho Nam. Associate research engineer at Korea 

Hydro & Nuclear Power Central Research Institute 

(KHNP CRI). Majored in nuclear engineering and have 

a master’s degree 4 years of experience in nuclear 

safety analysis, and the current work is also safety 

analysis using safety analysis code and containment 

P/T analysis. 




Top 

IAEA Presents New Platform 

on Small Modular Reactors 

and Their Applications 

(iaea) The IAEA presented its newly 

established Platform on Small Modular 

Reactors (SMRs) and their Applications, 

aimed at supporting countries 

worldwide in the development and 

deployment of this emerging nuclear 

power technology, during an event on 

the margins of the of the 65 th IAEA 

General Conference. 

With more than 70 SMR designs 

under development in 17 countries 

and the first SMR units already in 

operation in Russia, SMRs and their 

smaller cousins, microreactors (MRs), 

are forecast to play an increasingly 

important role in helping the global 

energy transition to net zero. Still, the 

technology, its safety and economic 

competitiveness must be de monstrated 

before SMRs can be more widely 

deployed, panellists agreed. 

The IAEA’s new Platform on SMRs 

and their Applications will assist 

countries in addressing these and 

related challenges. Using as a reference 

the Nuclear Energy Series publication, 

Technology Roadmap for Small 

Modular Reactor Deployment, the 

Platform provides experts with a 

one-stop shop to access the IAEA’s full 

array of support and expertise on 

SMRs, from technology development 

and deployment (including nonelectric 

applications) to nuclear safety, 

security and safeguards. 

“High standards of nuclear safety, 

security and non-proliferation must be 

ensured for SMR deployment,” Mikhail 

Chudakov, IAEA Deputy Director 

General and Head of the Department 

of Nuclear Energy said in his opening 

remarks. “But beyond this, it is 

generally recognized that if SMRs are 

going to be successful, they will need 

to be economically com petitive with 

respect to other clean energy alternatives. 

Achieving that will require 

accelerating their technological development 

and readiness level.” 

The IAEA has in place several 

activities related to SMRs to support 

its Member States through cooperation 

in SMR design, development and 

deployment and to serve as a hub for 

sharing SMR regulatory knowledge 

and experience. Although the IAEA 

safety standards can generally be 

applied to SMRs, global experts from 

the SMR Regulators’ Forum are 

working on a tailor-made solution to 

help national authorities regulate this 

new class of nuclear power reactors, 

which are expected to generate up to 

300 megawatts (electrical) (MW(e)) 

of power depending on their design. 

Participants spoke about challenges 

facing SMR development and 

deployment and how the new Platform 

could be used to help countries 

address them. 

“Potential topics for medium to 

long term activities include supply 

chain development, the development 

of industrial codes and standards, and 

suitable deployment strategies,” said 

Marco Ricotti, a professor of nuclear 

engineering at Italy’s Politecnico 

di Milano, who chairs the IAEA’s 

Technical Working Group for Small 

and Medium-sized or Modular 

Reactors (TWG-SMR). 

“Based on discussions within the 

(SMR Regulators’) Forum and also 

within our own regulatory forums at 

work, we feel strongly that it is not 

realistic in the near term to develop 

detailed guidance for every technology,” 

said Marcel de Vos, who works 

on new reactor licensing at the 

Canadian Nuclear Safety Commission. 

“The pragmatic approach in our view 

is that we need to work with what we 

have and make calculated improvements 

as experiences are gathered 

and gained.” 

“There is a lot of work ahead of us, 

but we are working efficiently and in 

the right direction with Member 

States,” said Lydie Evrad, Deputy 

Director General and Head of the 

Department of Nuclear Safety and 

Nuclear Security. 

“The Platform is a very powerful 

interdepartmental mechanism, 

bringing together expertise from 

across the organization on SMRs,” 

said Stefano Monti, Chair of the 

Platform Implementation Team and 

Head of the IAEA’s Nuclear Power 

Technology Development Section. 

| www.iaea.org 

Foratom: Limited Attention 

to Nuclear in Commission’s 

Energy Price Communication 

(foratom) FORATOM would have 

liked to see today’s communication 

from the Commission pay closer 

attention to the role which low-carbon 

and dispatchable nuclear can play in 

mitigating the current energy crisis. 

By including European nuclear in its 

toolkit of measures to tackle energy 

prices, it would have a unique opportunity 

of limiting its dependence on 

carbon intensive natural gas imports, 

thereby reducing its exposure to 

wholesale price fluctuations and its 

carbon footprint. 

“As highlighted in the communication, 

the current price increases are 

being driven by higher natural gas 

prices on the global market”, states 

Yves Desbazeille, FORATOM Director 

General. “Therefore, as the EU moves 

to increase its share of variable renewables, 

it is essential that EU policy 

supports other low-carbon European 

sources to ensure reduced dependency 

on imports.” 

The Communication also highlights 

the effects which lower availability 

of renewables has had on the 

market, leading to supply constraints. 

Because nuclear can provide both 

baseload and dispatchable electricity, 

it acts as a perfect counterbalance at 

times when renewables are unavailable. 

As noted in the Communication, 

nuclear currently accounts for 

around 25% of the electricity mix in 

the EU. 

With industrial activity ramping up 

post COVID, this has led to an increase 

in demand for energy. “It would be a 

mistake to treat this as a short-term 

issue. It is clear that demand for 

electricity is expected to increase 

dramatically in the push to decarbonise 

Europe’s economy” adds 

Mr. Desbazeille. “Therefore, the EU 

needs to already be putting solutions 

in place today to ensure that it is able 

to generate enough low-carbon 

electricity in Europe to meet growing 

demand. This means supporting the 

development of nuclear energy”. 

The Communication also makes 

reference to the sustainable finance 

taxonomy, reiterating the point that a 

complementary Delegated Act (CDA) 

‘will cover nuclear energy subject to 

and consistent with the results of the 

specific review process underway in 

accordance with the EU Taxonomy 

Regulation’. As this review is now 

complete, and the experts have overall 

concluded that nuclear is taxonomy 

compliant, we urge the Commission to 

urgently publish the CDA to avoid 

nuclear being unfairly penalised. 

| www.foratom.org 

NEI’s New Ad Campaign 

“See the Light” Embraces a 

Carbon-Free Future 

(nei) What’s New? The Nuclear Energy 

Institute launched its “See the Light” 

ad campaign, which speaks to the 

critical role nuclear energy plays in our 

carbon-free future. With eye-catching 

bursts of light illumi nating a dark 

room, the ad emphasizes that a 

brighter future must be powered by a 

diverse energy system championed by 

wind, solar and nuclear. 

57 

NEWS 

News


58 

NEWS 

Fast Facts: 

p The new ad addresses broader 

issues around the need to increase 

investments in carbon-free technologies, 

while also committing to 

more nuclear energy alongside 

wind and solar. 

p The “See the Light” campaign will 

appear primarily inside the 

Beltway on digital and social media 

platforms. The campaign’s paid 

media strategy aims to share the 

ad’s compelling visuals and 

connect how we power our daily 

lives and address the climate crisis. 

p NEI’s creative advertising and 

media buying are consolidated 

under Bully Pulpit Interactive 

(BPI). 

What Maria Korsnick, president and 

chief executive officer of NEI, has to 

say: “Our new ‘See the Light’ campaign 

demonstrates that we are at a pivotal 

moment – thinking differently about 

the best way to reach a carbon-free 

future. To make this future a reality, it 

will require policymakers and the 

private sector to increase investments 

and support policies that position 

nuclear as the backbone of our energy 

future. Nuclear energy, paired with 

wind and solar, can be the source 

that powers us to a brighter future – 

helping us meet the challenges of 

electricity production, job creation 

and decarbonizing our economy.” 

Big Picture: The ad campaign 

mirrors increasing attention for 

solutions to address the climate crisis 

along with the funding necessary 

to meet climate goals. Policymakers 

and the private sector are taking 

steps to invest in that future – making 

nuclear energy a key element of these 

efforts. 

This year, the Biden administration 

proposed a record-high budget 

proposal of $1.9 billion for nuclear 

programs and has pledged to bring 

carbon emissions from electricity 

generation close to zero by 2035. 

Congress has also introduced several 

proposals to support nuclear plants, 

demonstrating the strong bipartisan 

support for nuclear as a source of 

reliable carbon-free power that can 

power the grid while also generating 

clean hydrogen and providing 

well-paying, long term jobs. And in 

the private sector, utilities are making 

bold commitments to reach net zero 

carbon emissions by 2050 or sooner – 

viewing nuclear energy as a key 

element to reaching their commitments. 

The ad can be found on NEI’s 

website and YouTube, as well as our 

social media channels under the 

hashtag #SeeTheLight. 

| www.nei.org 

“More Uranium Development 

Needed to Meet Demands of 

Growing Nuclear Fleet” – 

World Nuclear Association 

launches the Nuclear Fuel 

Report 2021 

(wna) World Nuclear Association 

launched the 2021 edition of The 

Nuclear Fuel Report, concluding that 

the positive trend in nuclear 

Operating Results July 2021 

Plant name Country Nominal 

capacity 

Type 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 

Energy generated, gross 

[MWh] 

Month Year Since 

commissioning 

Time availability 

[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Month Year Month Year 

OL1 Olkiluoto BWR FI 920 890 744 675 009 4 274 075 281 308 792 100.00 92.43 100.00 91.22 98.62 91.33 


KCB Borssele 1) PWR NL 512 484 735 348 976 2 020 948 174 089 745 94.00 78.85 94.03 90.41 91.35 77.57 

KKB 1 Beznau 7) PWR CH 380 365 744 277 593 1 808 997 135 020 383 100.00 93.89 100.00 93.47 98.14 93.51 


KKG Gösgen 7) PWR CH 1060 1010 744 774 841 4 448 202 335 334 791 100.00 94.67 99.91 82.38 98.25 82.49 

CNT-I Trillo PWR ES 1066 1003 744 779 046 4 063 296 268 087 144 100.00 76.69 100.00 75.98 97.32 74.41 

Dukovany B1 PWR CZ 500 473 744 359 263 1 998 118 121 642 557 100.00 80.36 100.00 79.62 96.58 78.56 




Temelin B1 PWR CZ 1082 1032 744 803 159 4 000 384 133 571 674 100.00 73.34 99.96 72.53 99.77 72.68 

Temelin B2 PWR CZ 1086 1036 0 0 4 559 781 130 148 685 0 82.07 0 81.74 0 82.54 

Doel 1 PWR BE 467 445 744 342 458 2 042 436 142 056 277 100.00 85.94 99.92 85.28 98.60 86.14 

Doel 2 PWR BE 467 445 681 304 498 1 956 341 140 566 406 91.55 83.12 88.79 82.20 87.34 82.19 

Doel 3 PWR BE 1056 1006 744 763 114 5 335 922 276 548 232 100.00 99.45 100.00 99.09 96.29 98.72 

Doel 4 PWR BE 1086 1038 744 799 060 5 539 882 282 899 871 100.00 100.00 100.00 100.00 97.42 98.92 

Tihange 1 PWR BE 1009 962 744 736 020 5 101 732 312 968 707 100.00 100.00 99.77 99.62 98.35 99.62 

Tihange 2 PWR BE 1055 1008 744 762 548 4 118 827 269 822 643 100.00 81.81 99.34 77.32 97.98 77.32 

Tihange 3 PWR BE 1089 1038 744 790 300 5 460 436 292 114 095 100.00 100.00 99.88 99.96 98.02 99.09 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 


[MWh] 


[%] 

Energy availability Energy utilisation 

[%] *) [%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

GKN-II Neckarwestheim 1,2,4) DWR 1400 1310 539 727 270 6 063 720 357 415 264 72.38 86.84 72.38 86.82 69.93 85.33 

KBR Brokdorf DWR 1480 1410 744 1 040 747 7 032 265 378 295 594 100.00 100.00 99.64 99.92 94.21 93.19 

KKE Emsland DWR 1406 1335 744 1 007 106 6 474 751 375 485 452 100.00 92.15 100.00 92.02 96.20 90.56 

KKI-2 Isar DWR 1485 1410 744 1 058 590 7 357 319 384 786 362 100.00 100.00 99.97 99.99 95.39 97.06 

KRB C Gundremmingen SWR 1344 1288 744 983 282 6 807 442 357 285 208 100.00 100.00 100.00 99.65 97.52 98.85 

KWG Grohnde DWR 1430 1360 744 1 011 906 6 057 299 404 817 648 100.00 87.44 99.97 87.11 94.54 82.81 

News


gene rating capacity projections that 

began in the previous (2019) report 

con tinues. In the period, two newcomer 

countries – Belarus and the 

United Arab Emirates – connected 

their first reactors to the grid; further 

reactors were commissioned in China, 

India, Pakistan and Russia; construction 

of new reactors was 

launched in China, Iran and Turkey; 

and many other countries are considering 

either expanding existing 

nuclear programmes (e.g. Argentina, 

Brazil, Bulgaria, Netherlands, 

Romania, South Africa) or building 

their first reactors (e.g. Egypt, Poland, 

Uzbekistan). The nuclear generation 

capacity is expected to grow by 2.6 % 

annually, reaching 615 GWe by 2040 

in the Reference Scenario. 

The report found that world 

uranium production dropped considerably 

from 63,207 tonnes of 

uranium (tU) in 2016 to 47,731 tU in 

2020. Unfavourable market conditions, 

compounded by the Covid-19 

pandemic, led to a sharp decrease in 

Operating Results August 2021 

investment in the development of new 

and existing mines. The currently 

depressed uranium market has caused 

not only a sharp decrease in uranium 

exploration activities (by 77 % from 

$2.12 billion in 2014 to nearly $483 

million in 2018) but also the curtailment 

of uranium production at 

existing mines, with more than 

20,500 tonnes of annual production 

being idled. Uranium production 

volumes at existing mines are projected 

to remain fairly stable until the 

late 2020s, then decreasing by more 

than half from 2030 to 2040 

Intense development of new projects 

will be needed in the current 

decade to avoid potential supply 

disruptions. A number of projects at 

very advanced stages of development 

are waiting for an improved supplydemand 

market situation in order to 

commence uranium production. 

According to the report, there will 

have to be a doubling in the development 

pipeline for new projects by 

2040. There are more than adequate 

project extensions, uranium resources 

and other projects in the pipeline to 

accomplish this need, but it is essential 

for the market to send the signals 

needed to launch the development of 

these projects.Beyond mining, the 

report found that: 

p In the conversion sector, near-term 

reactor requirements in UF6 will 

be largely covered by commercial 

inventories. By 2023, global conversion 

production is expected to 

meet requirements due to the 

ramp-up and restart of existing 

facilities. Nevertheless, in the longrun 

more conversion capacity will 

be needed. 

p In the enrichment sector, excess 

capacity is currently used for 

underfeeding and tails re-enrichment, 

bringing in approximately 

6,000 – 8,000 tU in support of the 

undersupplied uranium market. 

This will largely be reduced over 

time, as enrichment requirements 

rise due to nuclear generating 

capacity growth. 

*) Net-based values 

(Czech and Swiss nuclear 

power plants 

gross-based) 

1) Refueling 

2) Inspection 

3) Repair 

4) Stretch-out-operation 

5) Stretch-in-operation 

6) Hereof traction supply 

7) Incl. steam supply 

BWR: Boiling 

Water Reactor 

PWR: Pressurised Water 

Reactor 

Source: VGB 

59 

NEWS 

Plant name Country Nominal 

capacity 

Type 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 


[MWh] 

Month Year Since 

commissioning 


[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Month Year Month Year 



KCB Borssele PWR NL 512 484 744 371 004 2 391 952 174 460 749 100.00 81.55 100.00 90.37 97.31 80.09 


KKB 2 Beznau 1,2,7) PWR CH 380 365 131 48 079 1 970 778 142 347 079 17.61 89.49 17.37 89.46 16.58 88.90 

KKG Gösgen 7) PWR CH 1060 1010 744 775 908 5 224 110 336 110 699 100.00 95.35 99.82 84.61 98.39 84.52 

CNT-I Trillo PWR ES 1066 1003 744 779 119 4 842 415 268 866 263 100.00 79.67 100.00 79.05 97.35 77.33 







Doel 1 PWR BE 467 445 744 342 593 2 385 029 142 398 870 100.00 87.73 99.92 87.14 99.24 87.81 

Doel 2 PWR BE 467 445 744 342 520 2 298 860 140 908 925 100.00 85.27 99.98 84.47 98.34 84.25 

Doel 3 PWR BE 1056 1006 644 594 195 5 930 116 277 142 427 86.52 97.80 86.14 97.44 74.58 95.64 

Doel 4 PWR BE 1086 1038 744 796 782 6 336 664 283 696 653 100.00 100.00 100.00 100.00 97.12 98.69 

Tihange 1 PWR BE 1009 962 744 737 243 5 838 975 313 705 950 100.00 100.00 99.84 99.64 98.31 99.45 

Tihange 2 PWR BE 1055 1008 744 749 533 4 868 359 270 572 176 100.00 84.13 96.93 79.82 96.23 79.73 

Tihange 3 PWR BE 1089 1038 744 789 968 6 250 404 292 904 062 100.00 100.00 99.85 99.95 97.98 98.95 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 


[MWh] 


[%] 

Energy availability Energy utilisation 

[%] *) [%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

GKN-II Neckarwestheim DWR 1400 1310 736 1 010 380 7 074 100 358 425 644 100.00 88.52 100.00 88.50 97.21 86.84 

KBR Brokdorf DWR 1480 1410 744 1 038 035 8 070 300 379 333 629 100.00 100.00 98.65 99.76 94.05 93.30 

KKE Emsland DWR 1406 1335 744 1 008 876 7 483 627 376 494 328 100.00 93.15 100.00 93.04 96.39 91.30 

KKI-2 Isar DWR 1485 1410 744 1 072 168 8 429 487 385 858 530 100.00 100.00 100.00 99.99 96.67 97.01 

KRB C Gundremmingen SWR 1344 1288 744 985 716 7 793 158 358 270 924 100.00 100.00 100.00 99.70 97.78 98.72 

KWG Grohnde DWR 1430 1360 744 1 005 707 7 063 006 405 823 355 100.00 89.04 100.00 88.76 93.89 84.23 

News


60 

NEWS 

p In the fuel fabrication sector, competition 

may become more intense 

from both the commercial and 

technological perspective, due to 

increased interest in developments 

of advanced fuels (e.g. for nonlight 

water reactors). Nuclear fuel 

demand increasing in Asia and 

decreasing in the West may cause 

fuel vendors to move from a 

regional to a more global market 

approach. 

Sama Bilbao y León, Director General 

of World Nuclear Association, said 

following the launch of the Nuclear 

Fuel Report 2021:“Given its unique 

combination of attributes – reliability, 

affordability, low-carbon and universal 

deployability – it is clear that 

nuclear energy will play an even larger 

role in the electricity and energy 

systems of tomorrow. The Nuclear 

Fuel Report makes it clear that 

sufficient uranium resources exist to 

meet the expected growth, but 

uranium markets need to rebalance to 

incentivise investment in uranium 

mining to support the expansion of 

the global nuclear fleet.” 

| www.world-nuclear.org 

Company News 

Environmentalists Praise 

World’s Only Floating NPP 

for High Efficiency and 

Eco-friendliness 

(rosatom) ROSATOM’s floating 

nuclear power plant in the city of 

Pevek of Russia’s Chukotka Autonomous 

Okrug has been visited by a 

public expedition for the very first 

time. Led by Alexey Yekidin, a 

leading researcher at the Institute of 

Indus trial Ecology of the Ural Branch 

of the Russian Academy of Sciences, 

the expedition united ecologists, 

academics, and representatives of 

public associations. 

Participants were tasked with 

collecting and analysing data on the 

environmental and radiation safety of 

the floating nuclear power plant, as 

well as assessing the plant and its 

overall operation and communicating 

their findings to the public. 

The environmentalists carried out 

measure ments both at the station 

itself and in the surrounding area, as 

well as in the city of Pevek. 

Their findings showed that 

background radiation in both the 

vicinity of the floating nuclear power 

plant and in the city of Pevek arise 

exclusively from natural sources: i.e., 

from natural radionuclides and 

cosmic radiation, and that the average 

value of said radiation does not exceed 

0.12 μSv/h in either location. 

“More than 20 measurements were 

taken in the FNPP’s industrial area, as 

well as in the surrounding area and in 

the city of Pevek, and no artificial 

radionuclides were found at the 

surveyed sites. It has, therefore, been 

concluded that the operation of the 

floating nuclear power plant does not 

negatively impact the region’s radioecological 

situation,” said Aleksey 

Yekidin. 

Alan Khasiev, chairman of the 

coordinating council of the interregional 

public environmental movement 

“Oka,” said: “Our programme 

has been in operation since 2010. 

During this time, we have carried out 

44 full-scale environmental monitoring 

expeditions to most of the 

Russian-designed NPPs operating in 

both Russia and abroad, as well as to 

those under construction. The ecological 

expedition to the floating 

nuclear power plant and around Pevek 

is the latest stage of this programme.” 

Green Party member and biologist 

Larisa Kosyuk said: “In the context of 

the carbon emissions regulation introduced 

by the EU, the FNPP project can 

serve as an example of green technologies 

in the energy sector. Such 

power plants will be especially useful 

in the regions of Russia’s Far North 

and the Far East, where there are no 

hydrological resources of energy, and 

the delivery of fuel – such as coal and 

oil products – is expensive. According 

to the International Energy Agency, 

nuclear energy is the second largest 

source of energy in the world, 

accounting for 10% of the world’s 

electricity generation.” 

Kirill Toropov, deputy director of 

the local branch of Rosenergoatom 

JSC, told the expedition participants 

about the changes that have taken 

place in the town since the floating 

nuclear power plant has been connected 

to the grid: “Not only did the 

arrival of the floating nuclear power 

plant introduce an additional source of 

energy, it opened a new chapter in the 

development of SMR technologies in 

the region’s energy sector and in that 

of the country as a whole. Since its 

commissioning, the FNPP has already 

established itself as a reliable and 

innovative source of thermal and 

electric energy. One cannot help bu 

notice its positive contribution to improving 

the environmental situation, 

both in Pevek itself (with a 30 % 

reduction in soot emissions from the 

Chaunskaya CHPP) and in the 

surrounding bodies of water (as 

evinced by the restoration of flora and 

fauna in the Chaunskaya Bay and the 

return of seals and other species of 

marine animals).” 

Ivan Leyushkin, head of the Pevek 

municipal district, noted the expedition’s 

importance for informing the 

region’s residents about the safety of 

nuclear generation. Leyushkin also 

spoke about the crucial role that the 

NPP is playing in the region’s development: 

“Since the floating nuclear 

power plant has started operation, 

107 million roubles’ worth of socially 

significant projects have been 

implemented in the Pevek district. 

Our cooperation with Rosatom will 

continue in the future. In September 

of this year, a cooperation agreement 

was signed between the governor of 

the Chukotka Autonomous Okrug 

Roman Kopin and the ROSATOM’s 

Director General Alexey Likhachev.” 

Facts 

Equipped with two KLT-40S reactors, 

the FNPP is a source of 70 MW’s worth 

of electrical energy and 50 Gcal/h’s 

worth of thermal energy. It comprises 

the floating power unit (FPU) “Akademik 

Lomonosov” and onshore infrastructure 

designed to supply heat and 

electricity from the FPU to consumers. 

Since its commissioning, the FNPP 

has carried out continuous industrial 

environmental monitoring, as well as 

monitoring of air, soil cover, sea water, 

bottom sediments, aquatic biological 

resources, as well as atmospheric 

emissions and the management of 

waste from production and consumption. 

The cost of the operation of the 

plant’s environmental protection 

system amounted to more than 

17 million roubles in 2020. The NPP’s 

operation also made it possible to 

prevent more than 300 thousand 

tonnes of atmospheric CO 2 equivalent 

emissions in 2019 and 2020. 

| www.rosatom.ru 

Market data 

(All information is supplied without 

guarantee.) 

Nuclear Fuel Supply 

Market Data 

Information in current (nominal) 

U.S.-$. No inflation adjustment of 

prices on a base year. Separative work 

data for the formerly “secondary 

market”. 

News


Uranium 

Prize range: Spot market [USD*/lb(US) U 3O 8] 

140.00 

120.00 

) 1 

Uranium prize range: Spot market [USD*/lb(US) U 3O 8] 

140.00 

) 1 

120.00 

61 

100.00 

100.00 

80.00 

80.00 

60.00 

40.00 

20.00 

Yearly average prices in real USD, base: US prices (1982 to1984) * 

60.00 

40.00 

20.00 

NEWS 

0.00 

1980 

1985 

1990 

1995 

2000 

2005 

2010 

2015 

2020 

2021 

Year 

* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2021 

* Actual nominal USD prices, not real prices referring to a base year. Year 

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2021 

| Uranium spot market prices from 1980 to 2021 and from 2009 to 2021. The price range is shown. 

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown. 

Separative work: Spot market price range [USD*/kg UTA] 

Conversion: Spot conversion price range [USD*/kgU] 

180.00 

26.00 

) 1 ) 1 

160.00 

140.00 

0.00 

24.00 

22.00 

20.00 

Jan. 2009 

Jan. 2010 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 

Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Jan. 2021 

Jan. 2022 

120.00 

18.00 

16.00 

100.00 

14.00 

80.00 

12.00 

10.00 

60.00 

8.00 

40.00 

6.00 

20.00 

4.00 

2.00 

0.00 

0.00 

Jan. 2009 

Jan. 2010 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 


Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Jan. 2021 

Jan. 2022 


Jan. 2009 

Jan. 2010 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 


Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Jan. 2021 

Jan. 2022 


| Separative work and conversion market price ranges from 2009 to 2021. The price range is shown. 

)1 

In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps. 

* Actual nominal USD prices, not real prices referring to a base year 

Sources: Energy Intelligence, Nukem; Bilder/Figures: atw 2021 

Uranium prices [US-$/lb U 3 O 8 ; 1 lb = 

453.53 g; 1 lb U 3 O 8 = 0.385 kg U]. 

Conversion prices [US-$/kg U], 

Separative work [US-$/SWU (Separative 

work unit)]. 

2017 

p Uranium: 19.25–26.50 

p Conversion: 4.50–6.75 

p Separative work: 39.00–50.00 

2018 

p Uranium: 21.75–29.20 



2019 

p Uranium: 23.90–29.10 



2020 

January to March 2020 

p Uranium: 24.10–27.40 



April 2020 

p Uranium: 27.50–34.00 



May 2020 

p Uranium: 33.50–34.50 



June 2020 

p Uranium: 33.00–33.50 



July 2020 

p Uranium: 32.50–33.20 



August 2020 

p Uranium: 30.50–32.25 



September 2020 

p Uranium: 29.90–30.75 



October 2020 

p Uranium: 28.90–30.20 



November 2020 

p Uranium: 28.75–30.25 



December 2020 

p Uranium: 29.50–30.40 



2021 

January 2021 

p Uranium: 29.50–30.50 



February 2021 

p Uranium: 28.75–29.10 



March 2021 

p Uranium: 27.25–31.00 



April 2021 

p Uranium: 28.40–31.00 



May 2021 

p Uranium: 29.15–31.35 



June 2021 

p Uranium:31.00–32.50 



July 2021 

p Uranium:32.20–32.50 



August 2021 

p Uranium:32.20–36.00 



| Source: Energy Intelligence 

www.energyintel.com 

News


62 

For the Sake of Nuclear Safety and Security, 

it’s Time for Rogue Actors to Leave the Stage 

NUCLEAR TODAY 

John Shepherd is 

editor-in-chief of the online 

publication 

New Energy 360 & World 

Battery News. 

Sources: 

Rafael Grossi on Iran: 

https://bit.ly/3E3BZzC 

University of Bristol 

report: 

https://bit.ly/3Ei04mv 

As I sat down to write this article, the UK’s newly installed foreign minister, Liz Truss, announced an upcoming 

meeting with her Iranian counterpart on the sidelines of the UN General Assembly in New York. 

Truss said she would tell Iran’s foreign minister, Amir 

Abdollahian, that his country should return “to the nuclear 

deal negotiating table before it is too late” – a phrase so 

often recycled by world leaders on the issue. 

Sadly, the words ‘too late’ have, in the case of Iran and 

its in-out-in-out adherence to international agreements, 

lost all meaning. 

To refresh our minds, Truss was referring to the Joint 

Comprehensive Plan of Action (JCPoA), an agreement on 

the Iranian nuclear programme supposedly reached in 

Vienna, SIX years ago, between Iran and the five permanent 

members of the UN Security Council – China, France, 

Russia, UK, US – plus Germany (the P5+1), together with 

the EU. 

Well, political leaders have since come and gone in 

many of the concerned nations and the Iran nuclear deal 

has ricocheted up and down the scale of geopolitical 

priorities. But one thing has not changed: Iran’s alternating 

acceptance and defiance of what it chooses to ‘like’ about 

the deal. This pattern of behaviour has gone on for so long 

that it defies belief. 

The revolving door of Iran’s diplomatic dance with the 

nuclear energy community is still turning. As this article 

was written, the director-general of the International 

Atomic Energy Agency (IAEA), Rafael Grossi, was holding 

talks with Mohammad Eslami, Iran’s vice-president and 

head of the Atomic Energy Organization of the Islamic 

Republic of Iran (AEOI). 

Under normal circumstances, it would be uplifting to 

hear of such high-level talks, were it not for the fact that 

Mr Grossi is the third IAEA chief in my own humble career 

reporting on matters nuclear – and Iran was a recalcitrant 

player long before. 

I remember interviewing a former IAEA chief, 

Mohamed ElBaradei, at his office in Vienna about the 

Iran question some 20 years ago. I asked then what I 

ask again now… What have all these years of talking 

achieved? 

I suppose keeping a lid on potentially more terrifying 

nuclear behaviour by Iran is an achievement of sorts, but is 

that the best the world can hope for? 

And of course, Iran’s actions continue to embolden 

others, such as North Korea, who believe they have the 

right to benefit from nuclear energy without guaranteeing 

the peaceful nature of its use in their hands. 

Those of us who are of a ‘certain age’ remember the 

1995 deal between the US, Japan and South Korea, which 

was to establish the Korean Peninsula Energy Development 

Organization. That deal garnered financing to 

support the construction of two light water reactors in 

North Korea, in return for the state to freeze its illicit 

plutonium weapons programme. It was yet another 

ultimately doomed venture. 

Fast forward to this year, and we once again have the 

IAEA “reiterating previous calls” for North Korea to comply 

with its obligations, including full and effective implementation 

of Non-Proliferation Treaty safeguards. 

The current head of the IAEA said North Korea’s 

secretive nuclear activities are (once again) “a cause for 

serious concern… and deeply troubling”. 

It seems even the special relationship that former US 

President Donald Trump said he had established with the 

North Korean leader counted for nothing after all – not 

even after those “love letters” Kim Jong-un penned to the 

president. How fickle – and how unsurprising! 

So how do we redouble our efforts and work for a world 

in which all those who wish to use atomic power for 

peaceful purposes adhere to the requirements to ensure 

real nuclear safety and security? 

I’m reminded of a quote attributed to former UK prime 

minister Sir Winston Churchill: “Success is not final, failure 

is not fatal: it is the courage to continue that counts.” 

Courage will indeed be needed and so will 

determination – and readers will know that our industry 

has determination, expertise and zeal in abundance. 

To demonstrate, I was interested to read of new 

cooperation on pioneering radiation mapping research 

carried out by researchers from the University of Bristol 

and Ukrainian researchers and engineers at the Chernobyl 

Nuclear Power Plant. 

The university team was given privileged access to the 

now infamous control room of the plant’s fourth reactor, 

where they deployed specially developed radiation 

mapping and scanning sensors. These were also deployed 

inside the New Safe Confinement, the protective structure 

erected to cover the remains of the destroyed reactor and 

the original ‘sarcophagus’, which was hastily constructed 

in the aftermath of the 1986 accident. 

The aim of that visit was to further explore the value of 

autonomous and semi-autonomous radiation mapping 

systems in high-radiation environments. The teams say 

that in deploying these systems in the exclusion zone and 

at the plant, researchers were able to better define the 

location and amount of residual radiological hazards. 

This work encapsulates the vision and drive that is the 

hallmark of research supported by the nuclear industry – 

harnessing advances in technology to address the legacy of 

Chernobyl (and Fukushima), thereby enhancing expertise 

and knowhow to strengthen technical capabilities for the 

years ahead. 

We must look to the future because I venture to suggest 

that is where nuclear’s best years lie, provided rogue actors 

are not allowed to continue with clandestine activities – 

which will cause immense harm to the reputation of the 

nuclear industry as a whole. 

Looking ahead, peaceful nuclear activities should also 

go beyond the generation of electricity, such as innovations 

in using heat from nuclear power plants for heating 

or cooling, or as an energy source towards the production 

of fresh water, hydrogen or other products. 

To return to the Churchill quotation, having the courage 

to continue with support and investments for nuclear 

technologies will indeed improve lives and combat the 

worse effects of climate change. 

However, world leaders and nuclear industry leaders 

need to exclude access to nuclear technologies – once and 

for all – to those who refuse to be demonstrably transparent 

about their activities. Those bad actors represent 

the greatest risk to nuclear safety and security and the 

biggest threat to the nuclear energy industry’s future. 

Nuclear Today 

For the Sake of Nuclear Safety and Security, it’s Time for Rogue Actors to Leave the Stage ı John Shepherd

TIMES ARE CHANGING 

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