atw - International Journal for Nuclear Power | 02.2020


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atw Vol. 65 (2020) | Issue 2 ı February

EU “Green Deal” – Just Not with Nuclear Energy?

The “Bologna Process”, “Lisbon Strategy”, “Europe 2020” are three examples of initiatives and programmes of the

European Union, which have been initiated significantly or even leadingly by the EU Commission in recent decades to

provide impulses for the further development of the European community. The new EU Commission, which took office

on 1 December 2019 for a five-year term with President Ursula von der Leyen, presented its key concept, the “European

Green Deal”, a roadmap for shaping the Community's economy and society in a sustainable manner. EU President

Ursula von der Leyen explained: “The European Green Deal is our new growth strategy – for growth that brings us more

than it costs us...”. EU Vice-President Frans Timmermanns, who will be in charge of implementing the programme,

added: “We are in a climate and environmental emergency. With the European Green Deal we can contribute to the

health and well-being of our citizens by changing our economic model from the ground up...”.

In terms of content, the Green Deal presented should lead

to measures that promote the efficient use of resources. In

addition, the economy should be transformed into a clean

and cycle-oriented system, biodiversity should be

preserved and pollution reduced. The limitation of climate

change is central. In addition, the Green Deal should cover

all economic sectors, namely transport, energy, agriculture

and buildings, as well as the steel, cement, information

and telecommunications sectors, and the textile and

chemical industries.

The “Green Deal” is explained in more detail in a

24-page document presented by the EU Commission in its

first presentation. The following topics can be made for the

energy sector:

Timeline: In March 2020, the Commission will present

a draft for a European climate law, which aims to achieve

climate neutrality by 2050 and is to be incorporated into

the legislation of the community states.

EU climate targets and emissions trading system:

A revision of the EU energy tax directive is to be proposed.

Environmental aspects are to be given priority and the

European Parliament and the Council are to be given the

possibility to adopt proposals for this framework by qualified

majority under the ordinary legislative procedure.

National Energy and Climate Plans: Revised energy and

climate plans of the EU Member States are to be submitted

by them in the short term. The EU Commission will evaluate

these plans to determine whether the level of targets is

sufficient. The results of the assessment will be included in

the process of raising the EU climate targets for 2030. To

this end, the relevant regulations are also to be reviewed

and revised if necessary by mid-2021.

Energy efficiency and market integration: Priority

should be given to energy efficiency in all measures. While

maintaining technological neutrality, the European energy

market should be fully integrated, networked and digitised.

Transport: In the transport sector, climate neutrality

requires a 90 % reduction in relevant emissions by 2050.

To this end, the Commission is to adopt a strategy for

s ustainable and intelligent mobility in 2020, covering all

emission sources. Electric mobility, including the associated

infrastructure, will be of great importance.

Innovation and financing: The Commission proposes a

target of 25 % of the EU budget for climate action and will

present a European Sustainable Investment Plan to mobilise

up to € 1,000 billion over the next 10 years. Innovations for

climate action under Horizon Europe will account for 35 %

of the budget. In addition, up to € 2,000 billion in investment

is to be mobilised from citizens and industry.

The EU Commission's very clear formulations of its

objectives up to this point bear a fundamental guiding

principle of the Community: to formulate and achieve

objectives openly together.

Discussions are currently underway in the committees

on the details of possible individual measures for achieving

these goals. With regard to the importance of nuclear

energy, however, familiar patterns of action of individuals

are emerging, which tend to lack technological openness

and the freedom and openness in the energy mix for the

indivi dual member states, i.e. also the choice of the nuclear


In order to assess the nuclear energy option, it is

certainly interesting to take stock of its significance for the

EU, in figures.

Nuclear energy in the EU today stands for:

p 26 % of total electricity production,

p 50 % of low-emission production,

p 1,100,000 jobs


p an annual GDP of more than 500 billion euros.

With total emissions of climate-impacting gases amounting

to around 12 g CO 2 -equivalent, nuclear energy, together

with wind, is also the lowest emission energy source of

power generation. In the energy system, nuclear energy is

characterised by high availability of nuclear power plants

with a large potential for flexible feed-in, which is essential

for the integration of the volatile sources of renewable

energy. As a further primary energy source, nuclear energy

broadens the basis of an energy mix that is as broadly based

as possible, the uranium is geographically widely available

and the mass of nuclear fuel that has to be moved for use in

nuclear power plants is low – 1 kg of nuclear fuel for reactor

use corresponds to about 150,000 to 200,000 kg of hard

coal units.

Nuclear energy not only offers advantages for a lowcarbon

energy system, but also supports sustainability by

securing and creating urgently needed jobs – and this

against the background of global competition, which the

EU must also face up to with this new package of “Green

Deal” measures.

Openness and equality, even if there are different views

or assessments of individual technologies, not dogma,

remain fundamental for forward-looking decisions in a

common EU. These principles must also not stop at nuclear

energy, which is seen in EU member states as a pillar of the

future energy mix in electricity generation, in some cases

even as the mainstay.

Nuclear energy must therefore remain recognised as an

instrument for environmental protection in the EU and

must be promoted at this level of sustainability on a par

with other technologies.

Christopher Weßelmann

– Editor in Chief –




EU “Green Deal” – Just Not with Nuclear Energy?

atw Vol. 65 (2020) | Issue 2 ı February

EU-„Green Deal“ – nur nicht mit der Kernenergie?


„Bologna-Prozess“, „Lissabon-Strategie“, „Europa 2020“ sind drei Beispiele für Initiativen und Programme der

Europäischen Union, die wesentlich oder auch führend in den vergangenen Jahrzehnten durch die EU-Kommission initiiert

wurden, um Impulse für die Weiterentwicklung der europäischen Gemeinschaft zu liefern. Die am 1. Dezember 2019 neu für

eine fünfjährige Amtszeit angetretene EU-Kommission mit der Präsidentin Ursula von der Leyen präsentierte schon wenige

Tage später ihr Kernkonzept, den „European Green Deal“, eine Roadmap, um Wirtschaft und Gesellschaft der Gemeinschaft

nachhaltig zu gestalten. EU-Präsidentin Ursula von der Leyen erläuterte dazu: „Der European Green Deal ist unsere neue

Wachstumsstrategie – für ein Wachstum, das uns mehr bringt, als es uns kostet ...“. EU-Vizepräsident Frans Timmermanns, der

das Programm federführend umsetzen soll, fügte hinzu: „Wir befinden uns in einem Klima- und Umweltnotstand. Mit dem

European Green Deal können wir zu Gesundheit und Wohlergehen unserer Bürgerinnen und Bürger beitragen, indem wir

unser Wirtschaftsmodell von Grund auf verändern ...“.

Inhaltlich soll der präsentierte Green Deal zu Maßnahmen

führen, die einen effizienten Umgang mit Ressourcen fördern.

Die Wirtschaft soll sich zudem zu einem sauberen und

kreislauforientierten System wandeln, Biodiversität soll

erhalten und Schadstoffbelastung reduziert werden. Zentral

ist die Begrenzung des Klimawandels. Zudem soll der Green

Deal alle Wirtschaftsbereiche erfassen, namentlich Verkehr,

Energie, Landwirtschaft und Gebäude sowie den Stahl-,

Zement-, Informations- und Telekommunikationssektor wie

auch die Textil- und Chemieindustrie.

Etwas detaillierter erläutert ist der „Green Deal“ in einem

24-seitigen Dokument, das die EU-Kommission mit ihrer

ersten Präsentation vorlegte. Für den Energiesektor lassen

sich dazu folgende Punkte festhalten:

Terminierung: Im März 2020 wird die Kommission den Entwurf

für ein Europäisches Klimagesetz vorlegen, das Klimaneutralität

für das Jahr 2050 zum Ziel hat und in die Gesetzgebung

der Gemeinschaftsstaaten aufgenommen werden soll.

Klimaziele der EU und Emissionshandelssystem: Eine

Überarbeitung der EU-Energiesteuerrichtlinie soll vorgeschlagen

werden. Umweltaspekte sollen dabei im Vordergrund

stehen und Europäisches Parlament sowie Rat sollen

die Möglichkeit erhalten, für diesen Rahmen Vorschläge im

Rahmen des ordentlichen Gesetzgebungsverfahrens mit

qualifizierter Mehrheit anzunehmen.

National Energy and Climate Plans: Überarbeitete Energieund

Klimapläne der EU-Mitgliedstaaten sollen kurzfristig von

diesen vorgelegt werden. Diese wird die EU-Kommission

dahingehend bewerten, ob das Niveau der Ziele ausreichend

ist. Ergebnisse der Bewertung werden in den Prozess der

Erhöhung der EU-Klimaziele für 2030 einfließen. Dazu sollen

bis Mitte 2021 auch die einschlägigen Vorschriften geprüft und

ggf. überarbeitet werden.

Energieeffizienz und Marktintegration: Vorrang bei allen

Maßnahmen ist der Energieeffizienz einzuräumen. Unter

Wahrung der technologischen Neutralität soll der euro päische

Energiemarkt vollständig integriert, vernetzt und digitalisiert


Transport: Im Transportsektor ist für Klimaneutralität eine

Reduktion der relevanten Emissionen bis 2050 in einem Umfang

von 90 % erforderlich. In 2020 soll dazu von der Kommission

eine Strategie für eine nachhaltige und intelligente Mobilität

verabschiedet werden, die alle Emissionsquellen betrifft. Große

Bedeutung werden Elektromobilität einschließlich der zugehörigen

Infrastruktur besitzen.

Innovation und Finanzierung: Die Kommission schlägt das

Ziel eines Anteils von 25 % des EU-Haushaltes für Klimaschutzmaßnahmen

vor und wird einen Europäischen Plan für

nachhaltige Investitionen vorlegen, der in den kommenden

10 Jahren bis zu 1.000 Milliarden Euro mobilisieren soll. Innovationen

für Klimaschutzmaßnahmen im Rahmen von Horizon

Europe sollen 35 % des Budgets umfassen. Darüber hinaus

sollen bis zu 2.000 Milliarden Euro an Investitionen bei den

Bürgern und der Industrie mobilisiert werden.

Die bis dahin in ihren Zielen sehr eindeutigen Formulierungen

der EU-Kommission tragen grundsätzlich einen

Leitgedanken der Gemeinschaft: Ziele gemeinsam offen zu

formulieren und zu erreichen.

Im Detail der möglichen einzelnen Maßnahmen für die

Zielerreichung laufen die Diskussionen in den Gremien

derzeit. Mit Blick auf die Bedeutung der Kernenergie zeichnen

sich allerdings bekannte Handlungsmuster Einzelner ab, die

Technologieoffenheit und die Freiheit und Offenheit bei der

Ausgestaltung des Energiemixes für die einzelnen Mitgliedsstaaten,

also auch die Wahl der Option Kernenergie, eher

missen lassen.

Für eine Beurteilung der Option Kernenergie ist sicherlich

eine Bestandsaufnahme ihrer Bedeutung für die EU, in Zahlen,

von Interesse.

Kernenergie in der EU steht heute für:

p 26 % der gesamten Stromerzeugung,

p 50 % der emissionsarmen Erzeugung,

p 1.100.000 Arbeitsplätze


p ein jährlich erwirtschaftetes BIP von mehr als 500 Milliarden


Mit ganzheitlichen Emissionen klimawirksamer Gase in Höhe

von rund 12 g CO 2 -Äquivalent ist die Kernenergie zudem

gemeinsam mit Wind die emissionsärmste Form in der Stromerzeugung

überhaupt. Im Energiesystem zeichnet sich die

Kernenergie aus durch hohe Verfügbarkeit der Kernkraftwerke

mit einem großen Potenzial für flexible Einspeisung,

welches für die Integration der volatilen Quellen Erneuerbarer

unabdingbar ist. Als weiterer Primär energieträger

erweitert die Kernenergie die Basis eines möglichst breit

aufgestellten Energiemixes, der Energierohstoff Uran ist geografisch

weiträumig verfügbar und die Masse an Kernbrennstoff,

die für den Einsatz in Kernkraftwerken bewegt werden

muss, ist niedrig – 1 kg Kernbrennstoff für den Reaktoreinsatz

entspricht etwa 150.000 bis 200.000 kg Steinkohleeinheiten.

Kernenergie bietet dabei nicht nur Vorteile für ein

kohlenstoffarmes Energiesystem sondern unterstützt auch

die Nachhaltigkeit durch die Sicherung und Schaffung von

dringend benötigten Arbeitsplätzen – und dies vor dem

Hinter grund des weltweiten Wettbewerbs, dem sich die EU

auch mit diesem neuen Maßnahmenpaket des „Green Deal“

stellen muss.

Offenheit und Gleichheit, auch bei unterschiedlichen

Ansichten oder Bewertungen von einzelnen Technologien,

nicht Dogmentreue sind für zukunftsweisende Entscheidungen

in einer gemeinsamen EU weiterhin von grundlegender

Bedeutung. Diese Prinzipien dürfen auch nicht vor

der Kernenergie halt machen, die in Mitgliedsstaaten der EU

als teils sogar tragende Säule des zukünftigen Energiemixes in

der Stromerzeugung gesehen wird.

Kernenergie muss daher als Instrument für Umweltschonung

in der EU anerkannt bleiben und gleichranging mit

anderen Technologien auf diesem Nachhaltigkeitsniveau

gefördert werden.

Christopher Weßelmann

– Chefredakteur –


EU “Green Deal” – Just Not with Nuclear Energy?

Kommunikation und

Training für Kerntechnik

Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?

Wählen Sie aus folgenden Themen: Dozent/in Termin/e Ort

3 Atom-, Vertrags- und Exportrecht

Atomrecht – Ihr Weg durch Genehmigungs- und


RA Dr. Christian Raetzke 18.02.2020 Berlin

Atomrecht – Das Recht der radioaktiven Abfälle RA Dr. Christian Raetzke 10.03.2020 Berlin

Export kerntechnischer Produkte und Dienstleistungen –

Chanchen und Regularien

RA Kay Höft M.A. (BWL) 17.06.2020 Berlin

Atomrecht – Was Sie wissen müssen

3 Kommunikation und Politik

RA Dr. Christian Raetzke

Akos Frank LL. M.

11.11.2020 Berlin

Public Hearing Workshop –

Öffentliche Anhörungen erfolgreich meistern

Dr. Nikolai A. Behr 10.11. - 11.11.2020 Berlin

3 Rückbau und Strahlenschutz

In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:

3 Nuclear English

Das Strahlenschutzrecht und

seine praktische Umsetzung

Stilllegung und Rückbau in Recht und Praxis

Dr. Maria Poetsch

RA Dr. Christian Raetzke

Dr. Stefan Kirsch

RA Dr. Christian Raetzke

17.03. - 18.03.2020

16.06. - 17.06.2020

29.10. - 30.10.2020


23.09. - 24.09.2020 Berlin

English for the Nuclear Industry Angela Lloyd 01.04. - 02.04.2020 Berlin

3 Wissenstransfer und Veränderungsmanagement

Erfolgreicher Wissenstransfer in der Kerntechnik –

Methoden und praktische Anwendung

Dr. Tanja-Vera Herking

Dr. Christien Zedler

24.03. - 25.03.2020 Berlin

Haben wir Ihr Interesse geweckt? 3 Rufen Sie uns an: +49 30 498555-30


INFORUM Verlags- und Verwaltungs gesellschaft mbH ı Robert-Koch-Platz 4 ı 10115 Berlin

Petra Dinter-Tumtzak ı Fon +49 30 498555-30 ı Fax +49 30 498555-18 ı

Die INFORUM-Seminare können je nach

Inhalt ggf. als Beitrag zur Aktualisierung

der Fachkunde geeignet sein.

atw Vol. 65 (2020) | Issue 2 ı February


Issue 2 | 2020





EU “Green Deal” – Just Not with Nuclear Energy? E/G 63

Inside Nuclear with NucNet

Medical Radioisotopes / Why Changes are Needed

to ‘Unstable’ Supply Chain 68

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69

Calendar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70

Feature | Major Trends in Energy Policy and Nuclear Power

Highlights of the World Nuclear Performance Report 2019 71

Spotlight on Nuclear Law

New Ways of Public Participation

in Nuclear Licensing Procedures G 74

Energy Policy, Economy and Law

An Integrated Approach

to Risk Informed Decision Management 76

Environment and Safety

Design and Implementation of Embedded System

for Nuclear Materials Cask in Nuclear Newcomers 81

Research and Application of Nuclear Safety Culture

Improvement Management for NPPs in China 87

Design Principles for Nuclear and Operational Safety

of HTR NPPs – a Review G 94

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident

in Bushehr VVER-1000/V446 Nuclear Power Plant 98

Research and Innovation

Experimental Study of Thermal Neutron Reflection Coefficient

for two-layered Reflectors 105


Workshop on the “Safety of Extended Dry Storage

of Spent Nuclear Fuel” – SEDS 2019 109

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .112

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .113


Vogtle Unit 4 Containment Vessel

©2019 Georgia Power Company


Unit 3 Low Pressure Turbine

©2019 Georgia Power Company

Nuclear Today

Climate of Opinion Frowns on Germany

as Nuclear Exit Continues 118

Imprint 92



= German

= English/German

Insert: AiNT – Aus- und Fortbildungsprogramm 2020


atw Vol. 65 (2020) | Issue 2 ı February


Major Trends in Energy Policy

and Nuclear Power



71 Highlights of the

World Nuclear Performance Report 2019

Jonathan Cobb

Spotlight on Nuclear Law

74 New Ways of Public Participation in Nuclear Licensing Procedures

Neue Wege der Öffentlichkeitsbeteiligung in atomrechtlichen Verfahren

Tobias Leidinger

Energy Policy, Economy and Law

76 An Integrated Approach to Risk Informed Decision Management

Howard Chapman, Maria Cormack, Caroline Pyke,

John-Patrick Richardson and Reuben Holmes

Environment and Safety

81 Design and Implementation of Embedded System

for Nuclear Materials Cask in Nuclear Newcomers

M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

87 Research and Application of Nuclear Safety Culture Improvement

Management for NPPs in China

Xiaozhao Xu, Jun Guo and Sujia Li

94 Design Principles for Nuclear and Operational Safety

of HTR NPPs – a Review

Konstruktionsprinzipien zur nuklearen und betrieblichen Sicherheit

von HTR-KKW – ein Review

Urban Cleve


atw Vol. 65 (2020) | Issue 2 ı February



Medical Radioisotopes / Why Changes

are Needed to ‘Unstable’ Supply Chain

Ageing production facilities and low prices of technetium-99m have contributed to a lack of

production capacity, which has made supply unreliable. In a new report, the Nuclear Energy Agency

(NEA) proposes policy changes that could solve the problem. David Dalton looks at the background

and the challenges.

What is Technetium-99m?

Technetium-99m (Tc-99m) is an essential product for

health systems that is used in 85 % of nuclear medicine

diagnostic scans performed worldwide, or around

30 million patient examinations a year, making it the most

commonly used medical isotope. It is essential for accurate

diagnoses of diseases such as cancer, heart disease and

neurological disorders including dementia and movement

disorders. It is also the most common diagnostic radioisotope,

estimated to be used in approximately 85 % of all

NM diagnostic scans worldwide.

The production of Tc-99m is a complex process which

includes irradiation of uranium targets in nuclear research

reactors to produce molybdenum-99 (Mo-99), extraction

of Mo-99 from targets in specialised processing facilities,

production of Tc-99m generators – a device used to extract

Tc-99m from a decaying sample of Mo-99 – and shipment

to hospitals.

But the supply chain is complicated. Neither Mo-99 or

Tc-99m can be stored for very long. Mo-99 has a half-life of

66 hours, that is, its radioactivity decreases by half in

66 hours, and the half-life of Tc-99m is only six hours.

Given this complexity, supply has often been unreliable

over the past decade due to unexpected shutdowns and

extended maintenance periods at some of the facilities

(the research reactors) that produce Mo-99, many of

which are relatively old. These shutdowns have created

global shortages. In particular in 2009-10, a series of

unexpected outages at reactors led to a global supply crisis

and a severe shortage of Tc-99m.

four research reactors (in Belgium, the Czech Republic,

the Netherlands and Poland) supplies two processors

(in Belgium and the Netherlands). The problem is that

some reactor operators are captive to local processors and

have little choice but to continue supply even at prices

that are too low, while government funding sustains their


What is the problem

with radioisotope supply?

Supply of Tc-99m is a “just-in-time” activity – it has to

be delivered as it is needed – requiring continuous

production in a complicated and aging supply chain that

combines a mix of governmental and commercial entities.

Governments control the availability of enriched uranium

required for medical isotope production and also largely

control legislation governing how much health care

providers (doctors and hospitals) charge for nuclear

medicine diagnostic scans. The central steps of the supply

chain, including processing and generator manufacturing,

are mainly commercial. Processors and generator manufacturers

wield market power, while supply continues to

be supported by government funding of some processors

and of nuclear research reactors that perform irradiation.

The resulting inability by reactor operators to increase

prices sufficiently for full cost recovery, combined with

insufficient reserve capacity (in the event of a reactor

outage, for example) at various steps of the supply chain,

leave security of supply vulnerable and the market

economically unsustainable.

How is Technetium-99m produced?

To prepare doses for patient scans, specialised pharmacies,

called nuclear pharmacies, elute Tc-99m daily from Mo-99

containers. These containers are called Tc-99m generators

and their manufacturers require regulatory approval to

sell them. Pharmaceutical companies manufacture and

sell Tc-99m generators commercially. They buy Mo-99 in

bulk from processing entities that transform irradiated

uranium into a Mo-99 liquid used to fill Tc-99m generators.

These processors procure uranium as a raw material and

contract with nuclear research reactors that perform

irradiation services.

What is the role of nuclear reactors?

Nuclear research reactors perform the primary irradiation

services. Most irradiations – the process by which an object

is exposed to radiation – are performed by reactors close to

processor facilities. In some cases (Argentina, Australia

and South Africa), the reactor and the processor are

co-located within the same organisational structure and

the single local reactor is the sole irradiator for the

processing facility. If the reactor is out of operation

for a period, the processor cannot operate and if the processor

is out of operation, the output from the reactor

cannot be processed. In Europe, an informal network of

Is the NEA proposing solutions?

The NEA says funding for the commercial production of

Tc-99m by governments of producing countries should

stop. This could help solve continuing supply problems.

What the NEA wants to see is “full cost recovery” for reactor

operators. The report suggests that the main barriers to

this are in the structure of the supply chain, the cost

structure and funding of nuclear research reactors and the

resulting behaviour of others in the supply chain.

The central problem is that the current structure of

the supply chain for medical radioisotopes leaves some

participants – notably the primary producers at research

reactors – unable to increase the prices of their services to

levels that would cover their costs.

The discontinuation of government funding would

compel producers to increase prices. This could, in the

short-term, destabilise supply and would therefore need to

be accompanied, at least temporarily, by measures to help

ensure that price increases are passed on through the

supply chain. One way to achieve this would be to increase

price transparency and encourage supply chain

participants to comply with commitments to increase

prices. A temporary price floor could help ensure that

producers are able to make up for the reduction of

government funding through additional revenue.

Inside Nuclear with NucNet

Medical Radioisotopes / Why Changes are Needed to ‘Unstable’ Supply Chain

atw Vol. 65 (2020) | Issue 2 ı February

The report also proposes the establishment of a

commodities trading platform that could make prices

more responsive to supply and demand and help ensure

production capacity is available. Alternatively, governments

could maintain funding of production but have

end-user countries bear the costs in proportion to the share

of total supply they consume. Governments could also aim

to reduce the reliance on the current supply chain through

substituting Tc-99m with alternative isotopes or diagnostic

methods, or by investing in alternative means of producing

Mo-99/Tc-99m. However, the latter two options could be


What happens next?

The NEA is calling on governments of producer and

end-user countries to co-ordinate their efforts and evaluate

each option in more depth. It says a more detailed study of

reactor and processor production costs is needed, along

with details of the level of current government funding of

producers, and the magnitude of price increases that

would be necessary to achieve full-cost recovery. In 2017

the NEA said the supply chain should be sufficient until at

least 2022, but the situation still requires careful and

well-considered planning for the foreseeable future. “No

single option can be recommended as the preferred

solution to current issues with the reliability of supply and

each option has a number of strengths and weaknesses,”

the report concludes.



The Independent Global Nuclear News Agency Editor

responsible for this story: David Dalton

Avenue des Arts 56 2/C

1000 Bruxelles



Did you know...?

Comprehensive Study of Economic and Social Costs

of the Nuclear Phase-out in Germany 2011-2017

The National Bureau of Economic Research in Cambridge,

Massachusetts, published the paper “The Private and External Costs

of Germany’s Nuclear Phase-out” by Stephen Jarvis, Olivier

Deschenes and Akshaya Jha in its NBER Working Paper series in

December 2019. The paper uses hourly plant level data and pollution

monitoring data to analyze the impact of the original plant closures

in 2011 and the subsequent ones till the end of 2017 not only on

aggregate electricity prices and carbon emissions but also to

estimate the effects on electricity production costs and local air

pollution. To compare the real phase-out with a hypothetical no

phase-out scenario the authors developed a machine learning

framework that combines the hourly power plant data with

information on electricity demand, local weather conditions,

electricity prices, fuel prices and plant characteristics.

The overall results of the effort confirm the results of other studies

that nuclear electricity production in Germany was primarily replaced

by increased fossil fuel production from coal and gas fired plants. The

paper also shows that the cost of electricity production in Germany

increased and that global and local pollution from electricity

generation increased substantially. The overall social cost of the

phase-out to German producers and consumers is estimated at

12 billion dollar per year on average (2017 USD). The majority – over

70 percent – of these costs is due to the increase in local air pollution

resulting from the shift from nuclear to fossil generation. In the

graphs below some of the numerical results of the study are

presented that compare the phase-out with the calculated no

phase-out scenario.

For further details

please contact:

Nicolas Wendler


Robert-Koch-Platz 4

10115 Berlin


E-mail: presse@

Estimated Impact of the Nuclear Phase-out on the Operating Profits

of Nuclear and Fossil Power Plants, on Wholesale Electricity Prices and Electricity Production Costs

(Annualized Averages from March 2011 to December 2017)

p Profits

60 %

p Production Costs

63.60 %

30 %

0 %

-30 %

30.10 %

23.20 %

17.00 %

4.00 %

8.10 %

2.50 %

-0.80 %

-33.90 %

-37.90 %

Nuclear Lignite Hard Coal Gas Oil

12.70 %

3.90 %

Wholesale Electricity Prices/

Overall Production Costs


“The Private and

External Costs of

Germany’s Nuclear


Stephen Jarvis,

Olivier Deschenes,

Akshaya Jha,

NBER Working Paper

No. 26598

Did you know...?

atw Vol. 65 (2020) | Issue 2 ı February





19.02. – 21.02.2020

International Power Summit 2020. Hamburg,

Germany, Arena International,


TotalDECOM – International Conference. London,


02.03. – 03.03.2020

Forum Kerntechnik. Berlin, Germany, VdTÜV & GRS,

02.03. – 06.03.2020

International Workshop on Developing a

National Framework for Managing the Response

to Nuclear Security Events. Madrid, Spain, IAEA,

08.03. – 12.03.2020

WM Symposia – WM2019. Phoenix, AZ, USA,

08.03. – 13.03.2020

IYNC2020 – The International Youth Nuclear

Congress. Sydney, Australia, IYNC,

15.03. – 19.03.2020

ICAPP2020 – International Congress on Advances

in Nuclear Power Plants. Abu-Dhabi, UAE, Khalifa


18.03. – 20.03.2020

12. Expertentreffen Strahlenschutz. Bayreuth,

Germany, TÜV SÜD,

22.03. – 26.03.2020

RRFM – European Research Reactor Conference.

Helsinki, Finland, European Nuclear Society,

25.03. – 27.03.2020

H2020 McSAFE Training Course. Eggenstein-

Leopoldshafen, Germany, Karlsruhe Institute of

Technology (KIT),

29.03. – 02.04.2020

PHYSOR2020 — International Conference on

Physics of Reactors 2020. Cambridge, United

Kingdom, Nuclear Energy Group,

31.03. – 02.04.2020

4 th CORDEL Regional Workshop on

Harmonization to support the Operation and

New Build fo NPPs including SMRs. Lyon, France,


30.03. – 01.04.2020

INDEX International Nuclear Digital Experience.

Paris, France, SFEN Société Française d’Energie


31.03. – 03.04.2020

ATH'2020 – International Topical Meeting on

Advances in Thermal Hydraulics. Paris, France,

Société Francaise d’Energie Nucléaire (SFEN),

08.04. – 09.04.2020

International SMR & Advanced Reactor Summit

2020. Atlanta, GA, USA, Nuclear Energy Insider,

19.04. – 24.04.2020

International Conference on Individual

Monitoring. Budapest, Hungary, EUROSAFE,

20.04. – 22.04.2020

World Nuclear Fuel Cycle 2020. Stockholm,

Sweden, WNA World Nuclear Association,

05.05. – 06.05.2020


Berlin, Germany, KernD and KTG,

10.05. – 15.05.2020

ICG-EAC Annual Meeting 2020. Helsinki, Finland,


11.05. – 15.05.2020

International Conference on Operational Safety

of Nuclear Power Plants. Beijing, China, IAEA,

12.05. – 13.05.2020

INSC — International Nuclear Supply Chain

Symposium. Munich, Germany, TÜV SÜD,

12.05. – 14.05.2020

KELI – Conference for Electrical Engineering, I&C

and IT in generation plants. Bremen, Germany,

VGB PowerTech,


Nuclear Solutions Exhibition. Warrington, UK,

Industrial Exhibition,

17.05. – 22.05.2020

BEPU2020– Best Estimate Plus Uncertainty International

Conference, Giardini Naxos. Sicily, Italy,


18.05. – 22.05.2020

SNA+MC2020 – Joint International Conference on

Supercomputing in Nuclear Applications + Monte

Carlo 2020, Makuhari Messe. Chiba, Japan, Atomic

Energy Society of Japan,

20.05. – 22.05.2020

Nuclear Energy Assembly. Washington, D.C., USA,


31.05. – 03.06.2020

13 th International Conference of the Croatian

Nuclear Society. Zadar, Croatia, Croatian Nuclear


06.06. – 12.06.2020

ATALANTE 2020. Montpellier, France, CEA,

07.06. – 12.06.2020

Plutonium Futures. Montpellier, France, CEA,

08.06. – 10.06.2020

8 th Asia Nuclear Business Platform. Yogyakarta,

Indonesia, Nuclear Business Platform,

08.06. – 12.06.2020

20 th WCNDT – World Conference on

Non-Destructive Testing. Seoul, Korea, EPRI,

15.06. – 19.06.2020

International Conference on Nuclear Knowledge

Management and Human Resources Development:

Challenges and Opportunities. Moscow,

Russian Federation, IAEA,

15.06. – 20.07.2020

WNU Summer Institute 2020. Japan, World Nuclear


02.08. – 06.08.2020

ICONE 28 – 28 th International Conference on

Nuclear Engineering. Disneyland Hotel, Anaheim,


01.09. – 04.09.2020

IGORR – Standard Cooperation Event in the International

Group on Research Reactors Conference.

Kazan, Russian Federation, IAEA,

09.09. – 10.09.2020

VGB Congress 2020 – 100 Years VGB. Essen,

Germany, VGB PowerTech e.V.,

09.09. – 11.09.2020

World Nuclear Association Symposium 2020.

London, United Kingdom, WNA World Nuclear


16.09. – 18.09.2020

3 rd International Conference on Concrete

Sustainability. Prague, Czech Republic, fib,

16.09. – 18.09.2020

International Nuclear Reactor Materials

Reliability Conference and Exhibition.

New Orleans, Louisiana, USA, EPRI,

28.09. – 01.10.2020

NPC 2020 International Conference on Nuclear

Plant Chemistry. Antibes, France, SFEN Société

Française d’Energie Nucléaire,

28.09. – 02.10.2020

Jahrestagung 2020 – Fachverband Strahlenschutz

und Entsorgung. Aachen, Germany, Fachverband

für Strahlenschutz,

07.10. – 08.10.2020

3 rd India Nuclear Business Platform. Mumbai,

India, Nuclear Business Platform,

12.10. – 17.10.2020

FEC 2020 – 28 th IAEA Fusion Energy Conference.

Nice, France, IAEA,

21.10. – 23.10.2020

2 nd Africa Nuclear Business Platform.

Accra, Ghana, Nuclear Business Platform,

26.10. – 30.10.2020

NuMat 2020 – 6 th Nuclear Materials Conference.

Gent, Belgium, IAEA,

09.11. – 13.11.2020

International Conference on Radiation Safety:

Improving Radiation Protection in Practice.

Vienna, Austria, IAEA,

24.11. – 26.11.2020

ICOND 2020 – 9 th International Conference on

Nuclear Decommissioning. Aachen, Germany,


07.12. – 10.12.2020

SAMMI 2020 – Specialist Workshop on Advanced

Measurement Method and Instrumentation

for enhancing Severe Accident Management in

an NPP addressing Emergency, Stabilization and

Long-term Recovery Phases. Fukushima, Japan,


17.12. – 18.12.2020

ICNESPP 2020 – 14. International Conference on

Nuclear Engineering Systems and Power Plants.

Kuala Lumpur, Malaysia, WASET,

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


atw Vol. 65 (2020) | Issue 2 ı February

Highlights of the World Nuclear

Performance Report 2019

Jonathan Cobb

The world’s nuclear plants continue to perform excellently. Growth is strong; but for the industry to reach the

Harmony goal of supplying at least 25 % of the world’s electricity before 2050, much greater commitment from

policymakers will be required.

The need for the reliable, predictable and clean electricity

generated by nuclear has never been greater and, worldwide,

that is reflected in the growing number of new build

programmes underway.

However, a number of factors – both internal and

external – are creating profound challenges for nuclear

power in some of its most mature markets.

Nuclear reactors generated a total of 2563 TWh of

electricity in 2018, up from 2503 TWh in 2017. This was

the sixth successive year that nuclear generation has risen,

with output 217 TWh higher than in 2012 (Figure 1).

Nuclear generation increased in Asia, East Europe &

Russia, North America, South America and West & Central

Europe. Generation fell in Africa, which has only two

reactors operating, both in South Africa.

In 2018 the peak total net capacity of nuclear power in

operation reached 402 GWe, up from 394 GWe in 2017.

The end of year capacity for 2018 was 397 GWe, up from

393 GWe in 2017 (Figure 2).

Over 2019 six reactors with a combined generating

capacity of 5178 MWe were added to the grid, while nine

units were permanently shut down. Based on provisional

figures global nuclear generating capacity stood at

391 GWe at the end of 2019.

Construction was started in 2019 on three new power

reactors: unit 2 of the Kursk II plant in Russia; unit 1 of

China’s Zhangzhou plant; and unit 2 of Iran’s Bushehr


Of the 442 reactors that were operable at the end of

2019, over half were in the USA and Europe where, despite

the vital importance of nuclear to achieving sustainable

energy goals, reactor retirements continue to outpace

capacity additions.

In 2018 the global average capacity factor was 79.8 %,

down from 81.1 % in 2017 (Figure 3). Despite the small

reduction, this maintains the high level of performance

seen since 2000 following the substantial improvement

over the preceding years. In general, a high capacity

factor is a reflection of good operational performance.

However, there is an increasing trend in some

countries for nuclear reactors to operate in a loadfollowing

mode to accommodate variable wind and

solar generation, which reduces the overall capacity


There was a substantial improvement in capacity

factors from the mid 1970s through to the end of the

1990s, which since has been maintained. Whereas nearly

half of all reactors had capacity factors under 70 %,

the share is now less than one-quarter. In 1978 only 5 %

of reactors achieved a capacity factor higher than 90 %,

compared to 33 % of reactors in 2018 (Figure 4). Capacity

factors in 2018 are broadly similar to the previous five

years, and reflect the consistently high capacity factors

seen over the past 20 years.


Source: World Nuclear Association and IAEA Power Reactor Information Service (PRIS)









West & Central Europe

South America

North America

East Europe & Russia








Source: World Nuclear Association, IAEA PRIS







| Fig. 1.

Nuclear electricity production 1970 to 2018.











Not operating












| Fig. 2.

Nuclear generation capacity operable (net) 1971 to 2018.












Source: World Nuclear Association, IAEA PRIS


















There is no significant age-related trend in nuclear

reactor performance. The mean capacity factor for reactors

over the last five years shows little variation with age

(Figure 5). In 2019 five reactors reached the milestone

of 50 years of operation: Tarapur 1 and 2 in India, Nine

Mile Point 1 and R.E. Ginna in the US and Beznau 1 in


The continued good operation of reactors is an

indication of the potential for longer operations. In the US



| Fig. 3.

Global average capacity factor 1970 to 2018.






















Highlights of the World Nuclear Performance Report 2019 ı Jonathan Cobb

atw Vol. 65 (2020) | Issue 2 ı February


Number of reactors























1978 1988 1998 2008 2009 2010 2011 2012 2013 2014 2015 2016 2017 2018

Source: World Nuclear Association, IAEA PRIS

| Fig. 4.

Long-term trends in capacity factors 1978 to 2018.

Source: World Nuclear Association, IAEA PRIS

Sum of reference unit power (MWe)








Turkey Point units 3 and 4 became the first reactors to be

issued with licences authorizing them to operate for up to

80 years.

Most reactors under construction today started

construction in the last nine years (Figure 6). A small

number of reactors have been formally under construction

for a longer period, but may have had their construction

suspended. For Mochovce 3&4 in Slovakia, where first

concrete was poured in 1987, construction was suspended

between 1991 and 2008. Start-up of the first unit is now

expected in 2020.

Over the course of nuclear energy’s 66 years of

commercial operation reactor designs have evolved. One

characteristic of that evolution has been an overall increase

in reactor capacity, particularly over the first thirty years of

reactor development.

Reactor start-ups are predominantly taking place in

non-OECD countries, demonstrating the importance of

nuclear energy in growing economies.

Permanent shutdown Operable Under construction













Source: World Nuclear Association, IAEA PRIS



















Reactor construction start date

| Fig. 6.

Operational status of reactors with construction starts since 1983.

West & Central Europe

South America

North America

East Europe & Russia





| Fig. 7.

Capacity of first grid connection 1954 to 2018.



















































0 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49

Source: World Nuclear Association, IAEA PRIS

Age of reactor (years)

| Fig. 5.

Mean capacity factor 2014-2018 by age of reactor 2014 to 2018.

The evolution of reactor start-ups in different regions is

shown in Figure 7. The majority of reactor capacity built

between 1970 and 1990 was in West and Central Europe

and in North America. Since that period the majority of

reactor start-ups have been in Asia, with grid connections

in East Europe and Russia also contributing to new global


There is growing demand for electricity, and that

electricity must be cleanly generated. The world’s

population continues to grow, the economic and societal

aspirations of developing countries are undimmed and

demand grows as modern society produces ever-more uses

of electricity.

Nuclear energy can meet this growing demand,

providing clean and reliable supplies of electricity.

In May 2019, the International Energy Agency (IEA)

published its report, “Nuclear Power in a Clean Energy

System”. The vital role for nuclear energy was set out by

IEA Director General Fatih Birol, who said; “Without an

important contribution from nuclear power, the global

energy transition will be that much harder.”

The IEA report made it clear that nuclear can make a

significant contribution to achieving sustainable energy

goals and enhancing energy security. However, urgent

action is needed to ensure that this significant contribution

can be made.

Fatih Birol said; “Policy makers hold the key to nuclear

power’s future. Electricity market design must value the

environmental and energy security attributes of nuclear

power and other clean energy sources.”

These conclusions were echoed by the OECD

Nuclear Energy Agency’s (NEA) report, “The Costs

of Decarbonisation”, which observed that; “Decarbonizing

the electricity sector in a cost-effective manner while

maintaining security of supply requires the rapid

deployment of all available low-carbon technologies.”

To achieve this would require policymakers to

recognize and allocate the system costs to the technologies

that cause them and to encourage new investment in

all low-carbon technologies by providing stability for

investors. The overall conclusion of the NEA analysis was

that the most effective way to achieve deep decarbonization

of the electricity generation mix was to have a high

proportion of electricity supplied by nuclear power.

This conclusion echoes that reached in the Intergovernmental

Panel on Climate Change (IPCC) report on

Global Warming of 1.5 °C, published in 2018. This report

evaluated 85 scenarios that would achieve the goal of

limiting global warming to 1.5 °C.

On average, these scenarios would see nuclear

generation increasing by around two and a half times by

2050. In a representative scenario, where societal and


Highlights of the World Nuclear Performance Report 2019 ı Jonathan Cobb

atw Vol. 65 (2020) | Issue 2 ı February

technological developments follow current patterns,

nuclear generation increases over five-fold.

It is evident that unless nuclear energy is a significant

part of the global response to climate change it is highly

unlikely we will be able to achieve a full decarbonization of

our generation mix, and even if it were possible the costs

would be exorbitant.

Over the last two years the call for action on climate

change has become louder and more urgent. Some have

questioned whether nuclear energy can be deployed

quickly enough to tackle climate change in time. The fact is

that nuclear energy is making a major contribution to

avoiding climate change today, with more than 10 % of the

world’s electricity supplied by nuclear generation.

One of the most effective actions to be taken to avoid

greenhouse gas emissions is to ensure those reactors

continue to operate to their full potential. The average age

of the nuclear fleet is around 30 years. This year, five

reactors have achieved fifty years of operation and reactors

today are seeking approval for 60 or even 80 years of

operation. Many of our current reactors have the potential

to still be part of a fully decarbonized generation mix in


More than 50 reactors are under construction, and half

of those are expected to start generating electricity over

the next two years.

Using nuclear avoids carbon dioxide emissions, as it

reduces our dependence on coal. By 2025, the reactors

under construction today will avoid the emission of

450 million tonnes of carbon dioxide each year – adding to

the already two billion tonnes of CO 2 avoided by the

existing fleet. This is equivalent to the combined annual

CO 2 emissions of Japan, Germany and Australia.

Where reactors are decommissioned over the next

30 years, new reactors should be constructed to replace

them. As well as ensuring the continuation of the benefits

of nuclear generation, construction and commissioning

of replacement reactors will ensure that key skills are

retained and local communities continue to have

employment opportunities.

But can nuclear generation be expanded fast enough to

combat climate change? During the rapid expansion of

nuclear generation in France in the 1980s and 1990s, most

reactors were built in six to seven years. In recent years

in China, nuclear reactors have been frequently

constructed in around five years. In 2018, the global

median construction time was longer, eight-and-a-half

years, primarily because of the high proportion of first of a

kind reactors starting in 2018.

A commitment to a substantial expansion of nuclear

generation would deliver the benefits of series construction,

including faster and lower cost construction.

The IPCC’s 1.5 °C report states that global greenhouse

gas emissions need to start to decline almost immediately.

Reactors under construction and the continued operation

of existing reactors can contribute to this goal. But to

achieve the further reductions that will be necessary from

2025, and net zero emissions by 2050, decisions to invest

in new nuclear build will need to accelerate urgently.

The nuclear industry’s Harmony goal is for nuclear

generation to supply 25 % of the world’s electricity before

2050. This would require at least 1000 GWe of new nuclear

build. To achieve this, new nuclear capacity added each

year would need to accelerate from the current 10 GWe to

around 35 GWe for the period 2030-2050. Those countries

operating nuclear power plants should commit to continue

to do so and those countries with recent experience of new

nuclear build should commit to a rapid expansion of

their construction programmes to deliver significant new

nuclear construction from 2025.

Beyond 2025 more countries will be able to contribute

to achieving our Harmony goal. More new nuclear

generation will be needed to bring economic growth, as

developed countries continue their efforts to decarbonize

their generation mixes and developing countries

endeavour to meet demand for electricity driven by

growing populations and industrial expansion essential to

modern life.

If we are to be serious about climate change we should

also be serious about the solutions. Transitioning to a

low-carbon economy that meets the energy needs of the

global community presents a daunting task. But it is a

challenge that must be met, and one that can only be met

by using the full potential of nuclear energy.


Dr Jonathan Cobb

Senior Communication Manager

World Nuclear Association

Tower House, 10 Southampton Street

London WC2E 7HA, UK



Highlights of the World Nuclear Performance Report 2019 ı Jonathan Cobb

atw Vol. 65 (2020) | Issue 2 ı February


Neue Wege der Öffentlichkeitsbeteiligung in atomrechtlichen


Tobias Leidinger


Die Beteiligung der Öffentlichkeit in atomrechtlichen Genehmigungsverfahren – z. B. für die Erlangung einer

Genehmigung zum Rückbau eines Reaktors – ist obligatorisch und in der Atomrechtlichen Verfahrensordnung (AtVfV)

im Einzelnen verbindlich geregelt. Neben diesem Standard-Repertoire gewinnen informale Beteiligungsformate in der

Praxis atomrechtlicher Genehmigungsverfahren zunehmend an Bedeutung. Dazu gehören z. B. Bürgerforen im Vorfeld

der Antragstellung oder auch die Einbindung von Beteiligungsgruppen während des Genehmigungsverfahrens. Unter

rechtlichen Gesichtspunkten stellt sich damit die Frage, wie sich die informalen Formate zu den förmlichen Beteiligungsvorgaben

der AtVfV verhalten.


Rechtliche Vorgaben und Freiräume

bei der Verfahrensgestaltung

1 Vorgaben der AtVfV

Nach der AtVfV sind die Vorgaben für die Beteiligung

der Öffentlichkeit in atomrechtlichen Genehmigungsverfahren

klar bestimmt: Der öffentlichen Bekanntmachung

des Vorhabens (§§ 4, 5) folgt die Auslegung

des Antrags samt Unterlagen (Sicherheitsbericht, Kurzbeschreibung

und UVP-Bericht) (§ 6). Innerhalb der

zweimonatigen Auslegungsfrist können Einwendungen

erhoben werden (§ 7), die – soweit sie für die Zulassung

relevant sind – anschließend in einem nicht öffentlichen

Erörterungstermin erläutert und erörtert werden können

(§ 8). Daran schließt sich die eigentliche Prüfphase –

regelmäßig unter Einbeziehung von externen, behördlich

beauftragten Sachverständigen – in Bezug auf die Genehmigungsvoraussetzungen

an, die mit der abschließenden

Entscheidung der Genehmigungsbehörde endet (§ 15).

2 Freiräume jenseits der AtVfV

Jenseits dieser zwingenden Vorgaben bestehen in Bezug

auf die Verfahrensgestaltung ergänzende Freiräume:

Das gilt sowohl für den Zeitraum vor der Antragsstellung

(Frühe Öffentlichkeitsbeteiligung) als auch danach

( begleitende Öffentlichkeitsarbeit).

Bereits vor der Antragsstellung (und damit vor Beginn

des förmlichen Verfahrens) kann der Vorhabenträger –

so wie in der Bestimmung über die frühe, informale

Öffentlichkeitsbeteiligung in § 25 Abs. 3 Verwaltungsverfahrensgesetz

(VwVfG) als Option vorgesehen – die

Öffentlichkeit über die Ziele, Mittel und Auswirkungen

seines Vorhabens unterrichten und auch inhaltlich einbinden.

Insoweit handelt es sich um eine „Soll-Vorgabe“,

d. h. es besteht keine Pflicht, diesen Weg zu beschreiten.

Die frühe Öffentlichkeitsbeteiligung nach § 25 Abs. 3

VwVfG zielt darauf, das Vorhaben zu optimieren, Transparenz

zu schaffen und die Akzeptanz der späteren

Genehmigungsentscheidung zu fördern. Denn hier geht es

nicht allein um frühzeitige Information, sondern um

einen echten Diskurs (Gelegenheit zur Äußerung und

Erörterung) und die Berücksichtigung der daraus

gewonnenen Erkenntnisse im Rahmen des an schließenden

förmlichen Verfahrens. Zur Konkretisierung der nach

§ 25 Abs. 3 VwVfG eröffneten frühen Öffentlichkeitsbeteiligung

steht mit der VDI-Richtlinie „Frühe Öffentlichkeitsbeteiligung

bei Industrie- und Infrastruktur projekten”

(VDI 7000) seit 2015 ein hilfreiches Instrument zur

Verfügung. Die VDI 7000 wurde als „Management-

Leitfaden” entwickelt, um Vorhabenträger konkret bei der

Vorbereitung und Durchführung früher Öffentlichkeitsbeteiligung

zu unter stützen. Zentrales Anliegen der

Richtlinie ist es, durch die frühe Beteiligung der

Öffentlichkeit Vertrauen in Akteure und Prozesse

aufzubauen, die im weiteren Verfahren helfen,

das Vorhaben insgesamt einfacher und effizienter

umzu setzen. Dabei können die Maßgaben der VDI-

Richtlinie flexibel eingesetzt werden – je nach Vorhaben

und Anforderungen –, um unterschiedliche Ansprüche

und Inhalte zu bedienen.

Das Ergebnis einer vor Antragstellung durchgeführten

frühen Öffentlichkeitsbeteiligung soll der betroffenen

Öffentlichkeit und der Behörde nach § 25 Abs. 3 VwVfG

spätestens mit der Antragstellung, im Übrigen unver züglich

mitgeteilt werden. Das geschieht in der Praxis regelmäßig

durch einen informativen Bericht des Antragsstellers zu

den durchgeführten Veranstaltungen und eingesetzten

Formaten, den der Antragssteller seinem förmlichen

Genehmigungsantrag beifügt und zugleich über das

Internet der Öffentlichkeit zur Verfügung stellt.

Freiräume für die Öffentlichkeitsbeteiligung bestehen

aber auch nach Beginn des förmlichen Genehmigungsverfahrens.

Dabei können unterschiedliche Wege

be schritten werden: Zum einem kann die Öffentlichkeit

auch jetzt – parallel zum förmlichen Verfahren – wiederkehrend

über den Fortgang der Planung und die

Konkretisierung einzelner Projektschritte informiert und

eingebunden werden. Das kann – wie im Rahmen der

frühen Öffent lichkeitsbeteiligung – mittels verschiedener

Formate, z. B. in Bürgerforen, durch Newsletter, Info-

Veranstaltungen oder „Tage der Offenen Tür“, erfolgen.

Zum anderen kann eine „Beteiligungsgruppe“ gebildet

werden, die sich aus interessierten „Stakeholdern“ verschiedener

Interessengruppen zusammensetzt. Sie bildet

ein begleitendes „ Gesprächsforum“, das z. B. unter

Beteiligung eines externen Moderators wiederkehrend

zusammentrifft, um bestimmte Aspekte des Vorhabens

vertieft zu erörtern. Dabei unterstützt der Antragssteller

dieses Beteiligungsformat durch qualifizierte Informationen,

die schriftlich oder durch seine Fachleute für die

Beteiligungsgruppe zur Verfügung gestellt werden.

II Zum Verhältnis paralleler Öffentlichkeitsbeteiligungsverfahren

Werden förmliche und informale Öffentlichkeitsbeteiligung

in Bezug auf ein Genehmigungsvorhaben

gleichzeitig durchgeführt, stellt sich unter rechtlichen

Aspekten die Frage nach ihrem Verhältnis zueinander: Im

Grundsatz sind beide Ebenen und Vorgänge unabhängig

voneinander. Das bedeutet insbesondere, dass Fehler im

förmlichen Verfahren nicht unter Verweis auf Vorgänge

oder Informationen im informalen Beteiligungsverfahren

„ausgeglichen“ oder „ungeschehen“ gemacht werden

können. Die Vorgaben des förmlichen Verfahrens nach der

AtVfV sind also strikt einzuhalten. Werden sie gleichwohl

verletzt, entscheiden die gesetzlichen Vorgaben in den

Spotlight on Nuclear Law

New Ways of Public Participation in Nuclear Licensing Procedures ı Tobias Leidinger

atw Vol. 65 (2020) | Issue 2 ı February

§§ 44-46 VwVfG über die „Fehlerfolgen“, d. h. darüber, ob

die Genehmigung dann als „nichtig“, oder ein Fehler als

„heilbar“ oder „unbeachtlich“ zu bewerten ist, so dass die

Behördenentscheidung im Ergebnis Bestand hat. Insoweit

existiert eine facettenreiche Kasuistik in der Rechtsprechung.

Um Fehler auszuschließen und das förmliche Verfahren

nicht „angreifbar“ zu machen, ist es eine besondere

Herausforderung, im Rahmen der informalen Öffentlichkeitsbeteiligung

sicherzustellen, dass der Umgang mit

Informationen „fair“ und „transparent“ erfolgt: Zum einen

sollte gewährleistet sein, dass „Dritte“, die sich im

förmlichen Verfahren beteiligen wollen, in Bezug auf

Informationen nicht schlechter gestellt werden als

diejenigen, die auch informal eingebunden werden. Das

lässt sich z. B. durch die Bereitstellung der Informationen

auf der Homepage des Antragsstellers einrichten.

Besondere Vorsicht ist zum anderen auch in Bezug auf

sicherungs relevante Informationen erforderlich: Geht es

um SEWD-relevante (Störmaßnahmen oder sonstige Einwirkungen

Dritter) Vorgänge sind die Vorgaben des

Sicherheits überprüfungsgesetzes (SÜG) strikt zu wahren.

Infor mationen, die inhaltlich die Anforderungen der Kennzeichnung

VS-NfD oder VS erfüllen, dürfen weder im

Rahmen der förmlichen noch der informalen Beteiligung

bekannt werden. Das begrenzt auch die jeweiligen

Diskussionen oder Erörterungen in der Sache – egal auf

welcher Ebene. Das dient letztlich dem Grundrechtsschutz

aller Beteiligten, der nicht mehr gewährleistet wäre,

wenn sensible Daten über potentielle Szenarien und

erfor derliche Schutzmaßnahmen öffentlich erörtert



III Fazit

Es besteht ein weiter Rahmen für die Öffentlichkeitsbeteiligung

in atomrechtlichen Genehmigungsverfahren:

Neben den zwingend einzuhaltenden förmlichen Vorgaben

der AtVfV bestehen Freiräume ergänzend für neue Wege,

um die Öffentlichkeit vor und/oder während des

Genehmigungsverfahrens auch informal zu informieren

und einzubinden. Auch wenn beide Beteiligungsebenen

rechtlich betrachtet unabhängig voneinander bestehen,

dienen sie letztlich dem gleichen Ziel: Möglichst verständlich

zu informieren, Kritik und Anregungen einzubeziehen

und damit Vertrauen sowie die Akzeptanz in

Bezug auf das Vorhaben zu fördern.


Prof. Dr. Tobias Leidinger

Rechtsanwalt und Fachanwalt für Verwaltungsrecht

Luther Rechtsanwaltsgesellschaft

Graf-Adolf-Platz 15

40213 Düsseldorf

Spotlight on Nuclear Law

 New Ways of Public Participation in Nuclear Licensing Procedures ı Tobias Leidinger

atw Vol. 65 (2020) | Issue 2 ı February



An Integrated Approach to

Risk Informed Decision Management

Howard Chapman, Maria Cormack, Caroline Pyke, John-Patrick Richardson and Reuben Holmes

The nuclear industry presents a unique combination of challenges in the planning and deployment of both large and

small projects.

| Fig. 1.

Hierarchy of Controls, (by the National Institute of Occupational Safety and

Health) [1].

The requirements of the regulatory

framework, diverse stakeholders, cost

effectiveness of investment, and the

management of actual and perceived

risks all contribute to the complexity

of decisionmaking. Pragmatic decisions

must be made to balance all of

these and any other factors.

The best solution to solve a

problem today might not be the best

solution tomorrow. The challenge is to

understand uncertainty from the

decision-making process and demonstrate

that decisions are made transparently.

This paper examines a solution to

decision-making in the nuclear industry

to help prevent lack of stakeholder

buy-in due to the complexity of the

problem. The method encourages

communication with all stakeholders

before during and after the decision-making

process and conveys the

output in a simple and highly visual

way that satisfies all their different

interests and points of view.

provide a simple visual display of

complex information allowing key

decision points to be compared and

contrasted. It allows several different

metrics under consideration to be

examined on a level playing field to

provide transparent, timely and accurate

decisions to be reached. Integral

to CompariCube® is an intuitive

graphical output designed to allow

stakeholders to interrogate and

examine the basis of the decision.

Through engagement with the

client and key stakeholders time

dependent risk profiles are established

over a number of metrics (such as

safety, cost, security, sustain ability

and environment). These are presented

as blocks in a cube. The chosen

solution is the one that minimises the

risk over time, with the solution that

has the smallest integral over the 3D



Nuclear engineered solutions traditionally

follow a standard hierarchical

methodology to safety starting with

elimination of the hazard wherever

possible, followed by reduction, isolation,

followed by control, Personal

Protective Equipment (PPE) and

discipline, with reliance upon PPE and

procedures being the weakest and

therefore least favourable hazard

management strategy as shown in

Figure 1.

CompariCube® is a registered

trade mark of National Nuclear Laboratory

Ltd 2016.

An alternative approach may place

early reliance upon the use of less

favourable hazard management

strategy control, for a relatively short

duration of time. Overall it may be

acceptable to be at the lower end of

the standard hierarchical safety

approach, if the resultant overall

integral of risk for the whole project is

assessed to be less.

This is exemplified in Figure 2

which shows a predictive risk profile

typically involved in achieving

safer, sooner and cheaper pragmatic

solutions. The overall risk for each

approach is expressed as the area

under each of the individual two


Historically, radical options may

be considered at early stages in the

optioneering process, but are often

relegated without further adequate

in-depth analysis. When comparing

options to find a solution to a problem,

traditionally only the highly engineered

solutions are considered. This

may not provide the lowest risk option

in aggregate, and may unintentionally

increase the total risk of the project

over its lifetime.

NNL aimed to create a holistic and

flexible approach to risk reduction,

which accounts for the entire lifetime

of the project and reduces overall risk


The National Nuclear Laboratory

(NNL) has developed CompariCube®.

This software tool and accompanying

process has been used for intelligent

strategic decision- making when faced

with complex challenges, and can be

used to support short-term investment

for long-term savings.

CompariCube® allows the analyses

of comparative data and metrics to

| Fig. 2.

Comparison of Two Options.

Energy Policy, Economy and Law

An Integrated Approach to Risk Informed Decision Management ı Howard Chapman, Maria Cormack, Caroline Pyke, John-Patrick Richardson and Reuben Holmes

atw Vol. 65 (2020) | Issue 2 ı February

by challenging conventions around

the hierarchy of controls. The challenge

was to develop a methodology

that enables early evaluation of

various options and their acceptability

over the whole project lifetime,

accounting for all conceivable project

risks. NNL’s CompariCube® was developed

to overturn this historical

approach and offer a new way to allow

all options to be equally evaluated.

Evaluating options

for holistic risk reduction

While the term “risk” as used thus far

is in the context of safety, the concept

can be broadened to accommodate a

wider definition in terms of project

risk. Project risks encompass a broad

range of factors including cost,

environmental risks, regulatory requirements,

affordability, sustainability,

deliverability and many others


Such complex and high value

investment decisions as encountered

in the nuclear industry require high

levels of stakeholder engagement and

acceptance, across a broad range of

parties. Stakeholder groups will

have varying degrees of specialised

knowledge and each will prioritise

different interests (an illustration

of such stakeholders is shown in

Figure 3). A key challenge for

CompariCube® is to incorporate

effective and transparent communication

across all stakeholders with

varying degrees of knowledge and

different interests.

To accomplish this, CompariCube®

makes use of a simple visualisation

interface, to allow users to examine

the visual representation of project

risk in the form of a three-dimensional

“risk cube” for each option, which can

be manipulated into different views.

This ability to intuitively represent

the risk curves shown in Figure 4

is key to the utility of CompariCube®

as a communication and stakeholder

engagement tool, as well as a decision

facilitation tool.


The user is able to define all the axes,

by setting risk levels, deciding the key

metrics of importance to the project,

and by defining the time duration and


When the user defines the metrics,

they add a set of questions they have

designed to capture the issues pertinent

to each metric. Each question has

a user-defined set of answers.

Not all aspects of a project will be

rated as equally important. As such

| Fig. 3.

Illustration of Stakeholders with Different Knowledge and Interests.

| Fig. 4.

Illustration of the Comparicube® Output Concept.

CompariCube® offers the ability to

weight each metric and each question

according to its importance to

the decision-making process. This

flexibility is an essential part of the

decision- making process.

Figure 5 shows a schematic

diagram of how the user defines

the inputs along each of the three

axes. The metrics axis shows how

the metrics may have different

weightings, represented by differently

sized circles. The questions can also

be weighted according to their relative


Handling uncertainty

The ”Compariline” decision line

technique is a unique approach

developed by NNL in support of

CompariCube® to model uncertainty

from highly qualitative data. NNL

con ducted a literature review of [2]

to [8] to consider the modelling

of uncertainty with limited hard

data. Expert judgement around uncertainty

was generally applied

under such circumstances. However,

this typically requires not only a

good understanding of the area

of interest but also of the concept of

uncertainty. The development of

Compariline is based on an adaptation

of semantic differential type questions

commonly used in survey sampling

to estimate levels of agreement

to a given statement (for instance

from strongly agree to strongly


Decision makers identify themselves

as either “Expert”, “Knowledgeable”

or “Naïve” in their confidence/

experience/authority around the particular

question being asked. By identifying

the individuals according to

their knowledgeability, CompariCube®

weights responses;

1 Expert = 2 Knowledgeable = 4 Naive

When responses from all individuals

have been collated, they will make

up the Compariline, as shown in

Figure 6.


Energy Policy, Economy and Law

An Integrated Approach to Risk Informed Decision Management ı Howard Chapman, Maria Cormack, Caroline Pyke, John-Patrick Richardson and Reuben Holmes

atw Vol. 65 (2020) | Issue 2 ı February


| Fig. 5.

Schematic Diagram of CompariCube® Case Construction.

It is possible for the output data

from the Compariline to be translated

into a beta probability distribution.

The beta distribution has an upper

and lower bound and it has shape

parameters that allow for it to represent

a broad range of distributions,

from a bounded normal distribution

to a heavily skewed distribution.

Figure 7 presents four beta distributions

with differing shape and scale

parameters. The beta distribution is

considered to be most adaptable

towards the types of distributions

arising from CompariCube® questions.

Key: X Expert X Knowledgeable X Naïve

| Fig. 6.

Decision Line with Weighted Scoring.

| Fig. 7.

Beta Probability Distributions.

The current CompariCube® graphical

output is a 3D bar chart, similar

to that shown in Figure 4. Future development

of CompariCube® will include

the adaptation of graphical output

to include Error bars (similar to

that presented in Figure 8a), or with

upper and lower bound profile (similar

to that presented in Figure 8b, or

Figure 8c when applied to the 3D

graphical output).

Application – Radiometric

monitoring system

improvement decision

CompariCube® has successfully been

used on a number of different complex

investment and development,

high capital expenditure decisions in

the nuclear industry.

One specific case study example

involved the use of CompariCube® by

the Project Team to assist with the

complex decision-making process to

help choose between the partial

replacement of a Radiometrics Surveillance

Systems (RSS) versus complete

replacement of the RSS in a Post

Irradiation Examination (PIE) facility.

The RSS had been installed and

operational for more than 25 years,

with radiometric instruments being

added and removed over this time to

suit plant operations. One of the early

challenges facing the project team

was to consider the relative benefits

and dis-benefits of the partial replacement

of the RSS at a cost of circa £ 1 m

versus complete replacement of the

RSS at a cost of circa £ 5 m.

CompariCube® was utilised by the

project team to assist with this complex

decision-making process, with

key stakeholders. A set of six common

key metrics was identified which

included Safety, Cost, Deliverability,

Regulatory acceptability, Substantiation,

and Functionality which could

be used to compare the options on an

equivalent basis. A set of 16 detailed

questions was created to allow investigation

of each key metric which

ultimately allowed the preferred

option to be selected.

The CompariCube® study concluded

that a complete replacement of

the RSS at a cost of circa £ 5 m was the

preferred option as shown in Figure 9.

CompariCube® provided results

which were easy and intuitive to

understand and communicate, uncertainties

to be captured and sensitivities

explored in real-time. The output

from CompariCube® allowed interrogation

of underpinning information

and provided an auditable record of

all input data.

Application – Fuel sampling

programme options

assessment study

In order to identify a suitable waste

management solution for a fuel

sampling programme, an assessment

study was required to explore and

prioritise the options available for the

arising fuel remnants and associated

wastes. Any potential solution needed

to allow the immediate customer

requirement to be delivered, and to

also be acceptable to the various other

stakeholders involved.

CompariCube® was utilised by

the project team to assist with this

complex decision-making process,

with two workshops held with key

stakeholders. The aim of the first

workshop was to generate the options

for management of the waste by

defining an option set. Participants

were split into sub-groups to facilitate

focused brainstorming and provide

definition to each generated option.

Each option presented different technical

characteristics; requirements in

terms of investment and planning of

facility time; and technical and

engineering challenges with respect

to sampling, analysis, and waste

disposability. Four options were


This workshop also defined the

information requirements that would

Energy Policy, Economy and Law

An Integrated Approach to Risk Informed Decision Management ı Howard Chapman, Maria Cormack, Caroline Pyke, John-Patrick Richardson and Reuben Holmes

atw Vol. 65 (2020) | Issue 2 ı February

a) Bar chart with error bars b) Box-whisker type plot c) Continuous variable with upper and lower bound

| Fig. 7.

Beta Probability Distributions.

be needed for each option generated,

ahead of the second workshop, which

focussed on the option evaluation.

Members of the project team

designed the CompariCube® study in

advance of the second workshop. A set

of five common key metrics was

identified which could be used to

compare the options on an equivalent

basis. In order to best capture the

relative merits of the four options, a

large set of detailed questions was

created by which to evaluate each

option. The study, which was presented

and completed during the

workshop, evaluated a total of one

hundred and thirty-six questions

across the five metrics.

The evaluation and prioritisation

stage resulted in a list of options

ordered by acceptability to stakeholders

(as shown in Figure 10), with

the yellow colour used to indicate

‘ acceptable to the customer/other

stakeholders’ and the dark orange

colour used to indicate ‘not meeting

customer requirements’. An Uncertainty

Index was also made available

for each option. This prioritised

options list was subsequently used

to inform the waste management

strategy for the fuel sampling programme

going forwards.

relevance to stakeholders on a country-specific

basis, allowing public engagement

activities to be tailored

within member states. A future use

of CompariCube® is proposed for

creating public-engagement specific

studies based on the concept of the

materiality matrix. The benefits of

using CompariCube® for such a purpose

would be in producing a clear

visible output allowing stakeholder

issues and priorities to be readily


CompariCube® could also be incorporated

into and support other types

of public engagement techniques,

such as the ‘Hybrid Forum’ [10], and

the ‘Backcasting’ technique used

in a proposed social sustainability

framework for energy infrastructure

decisions [11].

The Hybrid Forum concept has

previously been used to make a decision

on the best flooding mitigation

strategy for a town in the UK, which

involved “experts” and “laypersons”

working together to find a solution

[12]. The principles underpinning

Hybrid Forums see all stakeholders as

equals who have valuable expertise

and knowledge, they facilitate the cocreation

of new knowledge between

“experts” and “laypersons” and they

work on the basis that all issues are

not known in advance of the forums.

Issues instead emerge through

dialogue and can lead to unforeseen

solutions to problems and establish

partnerships between stakeholders

that previously held opposing

positions. It is proposed that

CompariCube® could incorporate

input from various stakeholders, no

matter their area of expertise, and is

flexible enough to include new metrics

as they emerge and make the output

understandable to a wide range of

stakeholders allowing the co-creation

of knowledge and understanding

between “experts” and “laypersons”,

where everyone’s input adds value to

the decision-making process, and the


Public engagement

As illustrated in Figure 3, the local

community are a key stakeholder in

the decision-making process. The

Corporate Social Responsibility Group

within Finnish nuclear power company

(Teollisuuden Voima Oyj) (TVO)

uses a Materiality Matrix tool, which

is used to identify the aspects of social

responsibility with the greatest

relevance for the company’s stakeholders

and business operations [9].

This tool provides valuable insight

into areas which hold the most

| Fig. 9.

CompariCube® Graphical output showing preferred Option.

Energy Policy, Economy and Law

An Integrated Approach to Risk Informed Decision Management ı Howard Chapman, Maria Cormack, Caroline Pyke, John-Patrick Richardson and Reuben Holmes

atw Vol. 65 (2020) | Issue 2 ı February


| Fig. 10.

CompariCube® Overall Assessment Cube.

end result is reached in an open and

transparent way. Recent research at

the NNL has looked to bring this

technique into the nuclear sector [13]


The ‘Backcasting’ technique involves

communities working together

to develop a series of “future energy

scenarios”, and in turn work

backwards to put in place the steps

that are needed to get them to their

desired scenario. It is argued that

CompariCube® could play a role

in the decision-making process for

communities to choose their preferred

future energy scenario option.


Howard Chapman would like to thank

Dr Colette Grundy Head of Regulation,

Advanced Nuclear Technology, Business

Energy and Industrial Strategy

(BEIS), seconded from the Nuclear

Innovation Research Office (NIRO).

Colette retains a role as NNL

Laboratory Fellow in nuclear regulation

and was involved in the early

conceptualisation and development of



[1] “Hierarchy of Controls”. U.S. National Institute for

Occupational Safety and Health. Retrieved 2017-01-31.,”


[2] Y. Ben-Haim and M. Demertzis, “Decision Making in Times of

Uncertainty: An Info-Gap Perspective (De Nederlandsche

Bank Working Party Paper No. 487),” 26 November 2015.

[Online]. Available:

[Accessed 16 March 2018].

[3] R. Schapire, “COS511: Theoretical Machine Learning,” 1 May

2014. [Online]. Available:


[Accessed 20 March 2018].

[4] J.-S. R. Jang, C.-T. Sun and E. Mizutani, Neuro-Fuzzy and Soft

Computing: A Computational Approach to Learning Machine

Intelligence, Michigan: Prentice Hall, 1996, 1997.

[5] O. T. A. Henningsen, maxLik: A package for maximum

likelihood estimation in R, Computational Statistics, vol. 26,

pp. 443-458, 2011.

[6] B. Bolker, “Maximum likelihood estimation and analysis with

the bbmle package,” 2017. [Online]. Available: https:// [Accessed

16 March 2018].

[7] A. B. Collier, “Fitting a model by maximum likelihood,”

18 August 2013. [Online]. Available:

[Accessed 16 March 2018].

[8] W. Li, “Appendix B,” in Risk Assessment of Power Systems:

Models, Methods and Applications, Wiley, 2014.

[9] Teollisuuden Voima Oyj, “Materiality Analysis and

Responsibility Aspects,” [Online]. Available:


[Accessed 30 September 2019].

[10] M. Callon, P. Lascoumes and Y. Barthe, Acting in an Uncertain

World: An Essay on Technical Democracy, MIT Press, 2009.

[11] J. Whitton, I. M. Parry, M. Akiyoshi and W. Lawless,

“ Conceptualizing a Social Sustainability Framework for

Energy Infrastructure Decisions,” Energy Research & Social

Science, vol. 8, pp. 127-138, 2015.

[12] S. J. Whatmore and C. Landstrom, “Flood Apprentices:

An Exercise in Making Things Public,” Economy and Society,

vol. 40, no. 4, pp. 582-610, 2011.

[13] University of Manchester, “Beyond Consultation: Hybrid

Forums for the Development of Nuclear Energy,” 17 July

2018. [Online]. Available: https://www.mub.eps.manchester.

[Accessed 30 September 2019].

[14] Times & Star, “Volunteers are needed for nuclear think-tank,”

18 September 2019. [Online]. Available:


[Accessed 30 September 2019].


Howard Chapman

Maria Cormack

Caroline Pyke

John-Patrick Richardson

Reuben Holmes

National Nuclear Laboratory


Central Laboratory, Sellafield,

Seascale, Cumbria, CA20 1PG

United Kingdom

National Nuclear Laboratory

Limited (reg. office)

Chadwick House

Birchwood Park

Warrington, Cheshire WA3 6AE

United Kingdom

Energy Policy, Economy and Law

An Integrated Approach to Risk Informed Decision Management ı Howard Chapman, Maria Cormack, Caroline Pyke, John-Patrick Richardson and Reuben Holmes

atw Vol. 65 (2020) | Issue 2 ı February

Design and Implementation of

Embedded System for Nuclear Materials

Cask in Nuclear Newcomers

M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

Nuclear newcomer countries facing a number of key challenges in infrastructure development, e.g. they have no

Intelligent Transportation Systems. Therefore, one of the challenges is the safety and security of nuclear materials

during transportation, storing and disposing. Where, nuclear and radiological terrorism continues to be a worldwide

concern as the nature of security threats evolves. This paper tries to solve that challenge by design and implement of an

embedded system for nuclear materials cask. This system is suitable for developing countries, where it is cost effective

and it uses the existing infrastructure. By using GPS, GSM/GPRS and microcontroller, the embedded system will enable

the responsible bodies to remotely and continuously; tracking, monitoring and inspection of nuclear materials casks;

during transportation, storing and disposing. The ORIGEN code is used to calculate the thermal and radioactivity loads

of the cask. The application of this system allows the rapid intervention of the concerned bodies, which will prevent

many accidents, in particular those caused by terrorists, like stealing or dispersing of nuclear materials.

1 Introduction

Recent advancements in nuclear

fission technology towards Small

Modular Reactor systems, arising

principally from their lower projected

construction costs makes them

applicable for a small investment.

These benefits have led many to

predict that the number of such units

will increase rapidly in developing

countries. In addition, developing

countries made the decision to embark

on a nuclear power program to

enhance security of energy supply by

diversification of energy resources,

reduce electric power production cost

and inhibit greenhouse gas emissions.

Therefore, Nuclear Materials (NM)

inventories are predicted to increase

rapidly in developing countries.

Knowing that, the threat of nuclear

terrorism remains one of the greatest

challenges to international security,

beside the weak infrastructure of

developing countries. The NM will be

mostly vulnerable to terrorism,

especially in transportations. Therefore,

additional measures are required

to militate against this risk. One of

these measures is the continuous

monitoring of nuclear materials

casks/packages; during transportation,

storing and disposing. Advancements

in microelectronics, wireless

tech nology and encryption can be

achieve that continuous surveillance,

by integration the modern microcontrollers

with sensors and wireless

communication techniques. The

continuing monitoring of the NM can

counter the terrorism threats by

informing the first responders (e.g.

Police and fire fighter) to not only

know the position of the incident, but

also the nature and severity of the

accident before approaching the scene

of the event, which allowing prompt

response. Also, continues monitoring

can be enhanced the safety and

security of NM.

If there are challenges for advanced

countries in the facing of nuclear

terrorism, the challenges for developing

countries are greater. Therefore,

this paper used the existing infrastructure

in these countries to build

an Embedded System (ES) that can be

used for continuous monitoring and

surveillance of NM, which will help

nuclear newcomers to counter nuclear


2 Related work

For nuclear terrorism countering, the

continuous monitoring system like as


CommBox and RAMM systems

technology [1] can be used as in

advanced countries. Regarding to the

lone wolves threats, where recently

the world has been suffering from.

The most lone wolves harmful attacks

were by trucks. Admittedly, the

destruction will be severely increased

if the truck was loaded with NM.

Therefore, the NM is Vulnerable to the

lone wolves threats especially during

transportations. The ARG-US and

RAMM systems are vulnerable to

counter this kind of terrorism. Therefore,

to overcome this vulnerability

the works in [2, 3] proposed a new

design approach for Intelligent Transportation

Systems (ITS) based internet

of things to counter the lone wolf,

just before the attacks done by trucks.

For developing countries like it is the

case of most nuclear newcomers,

where they have not an ITS nor privet

satellite communication like


or Iridium (for two way satellite

communications) [4], the system in

[5] can be used. In this system, a

customized Global System for Mobile

communication (GSM) module is

designed for wireless radiation monitoring

through Short Messaging

Service (SMS). This module is able to

receive serial data from radiation

monitoring devices such as survey

meter or area monitor and transmit

the data as text SMS to a host server. It

provides two-way communication for

data transmission, status query, and

configuration setup. Integration of

this module with a radiation monitoring

device will create mobile and

wireless radiation monitoring system

with prompt emergency alert at high

level radiation. But, this system absent

the tracking of the NM. Therefore, in

this paper, the proposed system used

the global satellite communication for

NM tracking as shown in the following

sections. The ES can be attached to

the NM casks.

3 Proposed embedded

system design and


The proposed system is an ES consists

of a microcontroller with onboard

GPS and GSM modules, sensors,

application software, a database

server and web page, Figure 1. The

ES monitors critical parameters,

including the status of seals, movement

of object, and environmental

conditions of the NM cask in real time.

Also, it provides an instant warning or

alarm messages (i.e. SMS), when



Environment and Safety

Design and Implementation of Embedded System for Nuclear Materials Cask in Nuclear Newcomers ı M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

atw Vol. 65 (2020) | Issue 2 ı February


| Fig. 1.

Embedded system block diagram.

preset thresholds for the sensors are

exceeded. The information collected

by the system is transmitted to a dedicated

central database server that can

be accessed by authorized users across

the responsible bodies via a secured

network. The ES allows the tracking

and inspecting of the casks throughout

their life cycles in storage, transportation,

and disposal. The software

provides easy-to-use graphical interfaces

that allow access to all vital

information once the security and

privilege requirements are met.

3.1 Sensor modules

As a prototype, the ES sensors include

safety sensors (e.g. radiation and

temperature), security sensor (e.g.

the status of seals), and driver v iolation

detector (e.g. the speed of the

truck). In this paper, the Evolutionary/European

Power Reactor (EPR)

Spent Fuel (SF) is selected as a hypothetical

source for NM.

p Safety sensors are used to indicate

the radiation and temperatures

levels statues of the NM cask. The

ORIGEN [6] computer code is used

to calculate the thermal and radioactivity

loads of the cask which will

be used to determine the sensor

threshold level. The ORIGEN computer

code flowchart is shown in

Figure 2. The preparation details

of the ORIGEN input file based EPR

fuel are stated in [7]. The radiation

and temperature sensors threshold

level are determined as follows.

1. Radiation: As will be proven later,

when EPR SF (5 % enriched) is

placed in a real cask system, the

dose rate on the external surface

of the cask will be lower than

1,000 mrem/hour. Therefore, the

ES prototype will be used the

PIN diodes to detect the increasing

in gamma level, where the

1,000 mrem/hour is sufficient to

excite the PIN diodes. Any gamma

detector PIN diode circuit consists

of a low noise amplifier and comparator,

Figure 3. The photodiode

circuit stated in [8] was used for

the gamma ray detection. The

advantage of using a photodiode is

its small sensitive area; therefore,

it is suitable to the high dose rate of

the cask and it is not affected by the

low background rate due to cosmic


2. Temperature: EPR SF can be

loaded into MPC-24 baskets. Using

the stated equation [7] of Peak

Cladding Temperature (PCT) given

Decay Heat (DH), when the DH is

1.050 kW/assembly, the error free

PCT is 307.12 °C in normal condition

operations. The 24 PWR SF

assemblies storage cask system

with a burn-up of 55 Giga Watt

Day/Metric Ton Uranium (GWD/

MTU) and 25.2 kW DH load, the

normal temperature for long-term

events (e.g. onsite and offsite

transportations, and storage) are

302 °C, 64 °C and 67 °C for PCT,

overpack outer surface and air

outlet; respectively [9]. Therefore,

for the EPR SF, the normal temperatures

are 307 °C, 69 °C and 72 °C

for PCT, overpack outer surface

and air outlet; respectively. The

normal temperature limits for

overpack outer surface and air

outlet are 98 °C and 72 °C; respectively.

For prototype, the circuit

used two digital temperature

sensors, where their positions are

in overpack outer surface (near the

top air outlet) and air outlet, the

temperature alarm SMS will

delivered to the control unit (or to

an emergency specified telephone

number) if the temperature

exceeds 98 °C and 72 °C for overpack

outer surface and air outlet;


p Status of Seals: The seal sensor can

be located under one or two of the

seal bolts of the cask overpack. The

seal sensor is a short circuit wire

warped around the bolt of the

cask overpack. When the bolt is

loosened, the short circuit wire will

open the circuit. Therefore, the microcontroller

trigger an alarm, the

alarm is broadcasted by SMS to the

responsible bodies.

3.2 Online cask monitoring

and tracking

The designed and implemented ES is

used for receiving location data from

satellites (via GPS module) and

monitoring data (via sensors), then

transmitting the received data to the

desired web servers using a General

Packet Radio Services (GPRS) connection

(via GSM module).

| Fig. 2.

ORIGEN computer code flowchart.

| Fig. 3.

Gamma detector PIN diode circuit block diagram.

Environment and Safety

Design and Implementation of Embedded System for Nuclear Materials Cask in Nuclear Newcomers ı M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

atw Vol. 65 (2020) | Issue 2 ı February

| Fig. 4.

Embedded system frame construction.

| Fig. 5.

Embedded system operation flowchart.

p GPS Module: The ES used the

recommended minimum specific

GPS/Transit data ($GPRMC)

frame. This frame contains information

about the locations of the

cask and the cask speed over

ground. The speed can be used as a

driver violation, if it exceeds a

predefined value (e.g. 80 km/

hour). Also, it can be used as a

motion detector for the cask in

storage, if greater than zero km/


p GSM Module: The ES used GSM

and GPRS international communications

standard to provide wireless

communications capabilities.

The sending of the SMS messages

are the functions of the GSM module.

The connection of the ES to the

internet is through the mobile operators


p Web servers: The server functions

are receiving data from the ES, securely

storing it, and serving this

information on demand to the user.

There are two servers. The first

is for secret data, e.g. the cask

monitoring data, while the second

server is for tracking data.

3.3 Microcontroller

The microcontroller used in ES

is a Programmable System-On-Chip

Cypress chip. The chip includes CPU

core, configurable blocks of analogous

and digital logic, and programmable

interconnects. This architecture

allows the user to create customized

peripheral configurations for each


3.4 Proposed Frame Format

Data is sent to the main servers as

frame format. All data are grouped in

a frame with a special format as shown

in Figure 4. Frame fields contain; cask

identification number (ID), cask

tracking location, seal status, and cask

monitoring sensor data. The microcontroller

takes the location data from

the GPS module and put it in its field

in the frame.

3.5 Proposed Embedded

System Operation

The ES operation methodology is

shown in Figure 5. When the ES

starts, it reads the sensors statues and

sends theses data for monitoring and

tracking servers by GPRS. In addition,

if any one of the sensor values exceeds

the limit, the ES sends instantaneous

SMS to the predefined telephone

number; and the monitoring and

tracking are instantaneous. For power

saving, in normal operations (i.e. radiation

level, T1 and T2 temperatures

lower than limits, and the seal is not

opened) the system is programmable

to wait a time between each reading

process (e.g. in a casks storage site,

the waiting time will be about ten


4 Results

The ORIGEN computer code simulation

results of the EPR SF radiation

source terms and the practical results

of the ES operation will be stated in

the next subsections.

4.1 Gamma Source Terms


Radiation source terms of SF are

photons and neutrons. In this paper,

the photon source of the EPR SF is

calculated using the ORIGEN code

based on the EPR parameters, where

the photons are the source term of

gamma. The EPR SF photon source

decay of the activation products,

actinides and daughters, and fission

products are calculated, Figure 6 (a).

As shown, the main gamma source

term is the fission products photons.

The radioactive characteristic of the

EPR SF has previously been calculated,

but for a burnup and enrichment

of 60 GWD/MTU and 4 % [10],

respectively. To make sure that our

calculations of the gamma sources

(i.e. 5 % enriched EPR) correspond to

that calculations (i.e. 4 % enriched

EPR), we compared our results with

the reference results, Figure 6 (b)

shows the photon sources decay comparison

of the results.

From Figure 6 (b), beyond five

years of SF cooling, the differences

between the two curves are small. For

example, at the cooling value of

20 years, it is found that the percentage

difference is about (0.078393 %,

providing that, the values of the fluxto-dose

conversion coefficients for

(a) 5 % enriched fuel

(b) 5 % enriched and 4 % enriched fuels

| Fig. 6.

EPR spent fuel gamma source decay.


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Design and Implementation of Embedded System for Nuclear Materials Cask in Nuclear Newcomers ı M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

atw Vol. 65 (2020) | Issue 2 ı February


(a) Normal values

(a) Tracking through Google maps web page

(b) Abnormal values

| Fig. 7.

Cask monitoring server screenshot.

(b) Tracking through the Traccar modern platform

| Fig. 8.

Online cask tracking.

photons given by ANSI/ANS-6.1.1-

1977 are about 20 % larger than the

version 1991, and ICRP74 coefficients

at the energies of interest. Therefore,

the cask shielding calculations in

[11] can be applied to our work. This

means that, when EPR SF (5 %

enriched) is placed in a real cask

system, the dose rate on the external

surface of the cask will be lower than

1,000 mrem/hour, which satisfied the

U.S. Nuclear Regulatory Commission

requirements [12].

4.2 Online Cask Monitoring

The online NM cask monitoring data

are given by accessing the server of

the responsible body. The cask

monitoring data are cask ID, seal

status, location (north and east), radiation

status, overpack outer surface

temperature, and air outlet temperature.

As a prototype, the cask

seal status is all right or opened,

while, cask radiation and temperature

status are all right (i.e. lower than

threshold limit) or over limit (i.e.

larger than threshold limit), Figure 7

(a, b).

4.3 Online cask tracking

There are two methods for NM cask

tracking, where the truck’s location

is given through; Google maps web

page or Traccar Modern Platform


4.3.1 Tracking through Google


To track the truck’s location through

the Google maps, the authorized user

will copy the longitude and latitude

received from a cask monitoring

server to a Google maps web page

to view the truck’s location on

Google maps. For example, the web

address shown in Figure 7 is (https:


31.309676); the user should copy

this address to any internet browser

to locate the truck in Google map,

Figure 8 (a). This method reduces the

code complexity and cost of the ES.

4.3.2 Tracking through

the Traccar platform

In this method, the ES used the free

and open source Traccar system

provided by Traccar Ltd [13]. Traccar

supports more protocols and device

models. It includes a fully featured

web interface for desktop and mobile

layouts. With Traccar, the NM cask

can be viewed in real-time with no

delay, by the ES GPS module. Traccar

has various mapping options, including

road maps and satellite

imagery, Figure 8 (b). The cost of a

single user account on a shared

Traccar server (for 5 devices + address

information in status and reports) is

$20.00/month. While the cost of the

own tracking server (for 50 devices +

address information in status and

reports) is $100.00/month. These

subscriptions include all features

provided by Traccar platform like

Geofencing, except SMS alerting

which need a supplement subscription.

4.4 SMS warnings and alerts

The warning SMS about driver harsh

driving violations (like speeding…)

can be sent to a predefined telephone

number without any delay by Traccar

system. In addition, for minimizing

the incident consequences, the alerts

about danger states of the cask (e.g.

temperature and/or radiation level

exceeding the limit values, seal

opened…) can be sent directly to a

predefined telephone numbers (like

police and/or nuclear safety staff), to

insure the fast response and rapid


5 Discussion

In the next subsections, some

problems that will face the applications

of the ES are mentioned.

Also, the solutions are stated.

5.1 Ionizing radiation effects

on electronic circuits

In normal operation of the NM cask,

the dose rate on the external surface

of the package must be lower than

1,000 mrem/hour (=1000 mrad/

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Design and Implementation of Embedded System for Nuclear Materials Cask in Nuclear Newcomers ı M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

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atw Vol. 65 (2020) | Issue 2 ı February


hour, =0.01 gray/hour). Therefore,

the electronic components of the ES

will be not affected. The ionizing

radiation effect on GPS and microcontrollers

modules will be shown


Renaudie et al. [14] deduced that

the commercial of – the – shelf GPS

receivers operate faultless up to

accumulated ionizing doses of more

than 9 krad (air). This is acceptable

for cask in normal operation.

The experimental results of a

collected irradiation environment of

gamma rays and neutron on MCS96

microcontroller were presented by

Xiao-Ming et al. [15]. The influence of

the synergistic effect are:

(1) the static power supply current

begins to increase at the total

ionizing dose (TID) of 8.0 krad (Si)

in the single gamma ray irradiation

environment, while in the mixed

irradiation environment it begins

to increase at the TID of 2.3 krad

(Si) and the neutron fluence of

7.51011 n/cm 2 ,

(2) when the microcontroller fails

to run, the neutron fluence is

approximately 1.21012 n/cm 2 and

the TID is 3.7 krad (Si),

(3) when the internal clock generator

fails to provide a clock signal, the

TID is 46.6 krad (Si) in the single

gamma ray irradiation environment,

while the TID is 17.7 krad

(Si) and the neutron fluence is

5.81012 n/cm 2 in the combined

irradiation environment.

The results shown that, the microcontroller

does not fail until the

TID exposure accumulates up to

11.3 krad (Si), and performs normally

even when the neutron fluence is up

to 3.01013 n/cm 2 . Therefore, the

microcontroller performs normally in

the ES radiation environment, where

the dose must be lower than

1000 mrad/hour.

5.2 GSM network losing

The physical protection of stored SF

and the geologic repository requirements

are stated in [12] insure the

continuous surveillance of the storage

and the repository sites. This means,

the electrical and the communication

systems in the site must be maintained.

Depending on these requirements,

the store and the repository

sites must have more than one

communication networks (e.g. two

to three GSM networks, which can

be used by the ES). Therefore, the

unavailability of the communication

networks is too rarely. Finally, the

using of the Iridium satellite (e.g.

RockBLOCK 9602) transceiver [16]

models is another option for the ES.

Where, RockBLOCK 9602 allows

sending and receiving short messages

from anywhere on earth, providing a

clear view of the sky. It works far beyond

the reach of Wi-Fi and GSM networks.

It works in the middle of any desert

and ocean. The interface of the

RockBLOCK to the ES board is easy,

Figure 1, with a serial interface and

can be operated with a three-wire

connection, which are used to transmit,

receive and ground signals. The

module can be read out using the AT

command interface. The main drawback

of the RockBLOCK is its cost,

where the costs are, 249 $/module

price, 20 $/activation fee, and 19 $/

monthly fee and usage rating:

1.17 $/1KB.

5.3 Data security

For maximizing the cask data security,

the proposed data frame format (in

Figure 4) designed and implemented

according to our private construction.

Therefore, it can be reformatted every

some time based on our security

constrains. In addition, we can secure

data based on Advanced Encryption

Standard methodology.

6 Conclusion

Advancements in microelectronics,

wireless technology and encryption

have opened opportunities that

previously were not available to the

nuclear sector. The ES tracking and

monitoring system is enhancing the

safety and security; reducing the need

for manned surveillance; providing

real-time access to status and event

data; and providing overall cost effectiveness.

The ES precise monitoring

and tracking of the nuclear materials

can perform the terrorists countering,

provided that additional terrorism

countermeasure like, the speed

response and rapid intervention of

the security bodies. The ES is suitable

to nuclear newcomers, where most of

them are developing countries.


Authors wish to acknowledge the

Professor Ezzat A. Eisawy for his

strong support. They are thankful to

Eng. Nagdy for his cooperation in the

laboratory work. Also, they want to

thank Eng. Emile Rushdie for his

precious discussion.


[1] Y.Y. Liu, K.E. Sanders, and J.M. Shuler, “Advances in tracking

and monitoring transport and storage of nuclear material,”

IAEA-CN-244- 186, 2016.

[2] Hassan F. Morsi, M. I. Youssef, and G. F. Sultan, “Novel design

based internet of things to counter lone wolf, part-A: Nice

attack,” Proceedings of the International Conference on

Advanced Intelligent Systems and Informatics, Egypt, 2017,

Advances in Intelligent Systems and Computing, vol. 639, pp.

875-884, Springer, doi: 10.1007/978-3-319- 64861-3_82

[3] H. F. Morsi, M. I. Youssef, and G. F. Sultan, “Novel design

based internet of things to counter lone wolf, part-B: Berlin

attack,” Japan-Africa Conference on Electronics, Communications

and Computers (JAC-ECC), Egypt, 2017, pp. 164-169,

IEEE, doi: 10.1109/JEC-ECC.2017.8305802

[4] U.S. Department of Energy Office of Environmental Management,

“TRANSCOM fact sheet: transportation tracking and

communication system”, 2009.

[5] Nur Aira Abd Rahman, Noor Hisyam Ibrahim, Lojius Lombigit,

Azraf Azman, Zainudin Jaafar, Nor Arymaswati Abdullah, and

Glam Hadzir Patai Mohamad, “GSM module for wireless

radiation monitoring system via SMS”, IOP Conf. Series: Materials

Science and Engineering, vol. 298, paper no. 012040,

2018, doi:10.1088/1757-899X/298/1/012040

[6] INVAP, “ORNL/ORIGEN Version 2.1”, 2004.

[7] M. I. Youssef, G. F. Sultan, and Hassan F. Morsi, “Cooling

period calculation of evolutionary power reactor spent fuel

for dry management safety”, Nuclear and Radiation Safety

Journal, vol. 2, no. 70, pp. 17-21, 2016, UDC


[8] Burkhard Kainka, “Improved radiation meter counter for

alpha, beta and gamma radiation”, Elektor, issue11,

pp. 20-25, 2011.

[9] Ju-Chan Lee, Kyung-Sik Bang, Ki-Seog Seo, and Ho-Dong

Kim, “Thermal analysis of a storage cask for 24 spent PWR

fuel assemblies”, 14th International Symposium on the

Packaging and Transportation of Radioactive Materials

(PATRAM 2004), Germany, 2004.

[10] Markku Anttila, “Radioactive characteristics of the spent fuel

of the finnish nuclear power plants”, Posiva working report

2005-71, 2005.

[11] Anssu Ranta-aho, “Review of the radiation protection

calculations for the encapsulation plant”, Posiva Working

Report 2008-63, 2008.

[12] Nuclear Regulatory Commission, “U.S. nuclear regulatory

commission regulations: title 10, code of federal regulations”,

Parts 71, 63, 72, and 73, 2010.

[13] Traccar Ltd, “GPS tracking software - free and open source

system - Traccar”,

[14] C. Renaudie, M. Markgraf, O. Montenbruck, and M. Garcia,

“Radiation testing of commercial-off-the-shelf GPS

technology for use on low earth orbit satellites,” 9 th European

Conference on Radiation and Its Effects on Components and

Systems, France, 2007, pp. 1-8. IEEE, doi: 10.1109/


[15] Jin Xiao-Ming, Fan Ru-Yu, Chen Wei, Lin Dong-Sheng, Yang

Shan-Chao, Bai Xiao-Yan, Liu Yan, Guo Xiao-Qiang, and

Wang Gui-Zhen, “Synergistic effects of neutron and gamma

ray irradiation of a commercial CHMOS microcontroller”,

Chinese Physics B, paper no. 066104, vol. 19, no. 6, 2010.

[16] Proprietary and Confidential Information, “Iridium 9602 SBD

transceiver developer’s guide, revision 6.0”, Iridium

Communications Inc., 2010.


M. I. Youssef

Faculty of Engineering

Al Azhar University

Cairo, Egypt

M. Elzorkany

National Telecommunication


Cairo, Egypt

G. F. Sultan


Egyptian Nuclear and Radiological

Regulatory Authority

Cairo, Egypt

Environment and Safety

Design and Implementation of Embedded System for Nuclear Materials Cask in Nuclear Newcomers ı M. I. Youssef, M. Elzorkany G. F. Sultan and Hassan F. Morsi

atw Vol. 65 (2020) | Issue 2 ı February

Research and Application of

Nuclear Safety Culture Improvement

Management for NPPs in China

Xiaozhao Xu, Jun Guo and Sujia Li

The traditional nuclear safety culture improvement work in is mainly about propagandize, training, and behavior

observation to instill the concept. The Lack of systematic evaluating and closed-loop management makes it difficult to

ensure the effectiveness. Based on these, the nuclear safety culture improvement management research work was

carried out. This article proposes a nuclear safety culture dynamic improvement model and some practical applications

has been carried out based on the model. Firstly, a nuclear safety culture standard that can reflect the international

advanced experience and the characteristics of Chinese culture is developed; Secondly, a continuous improvement of

nuclear safety culture evaluation methods and mechanisms is established, and the nuclear safety culture evaluation

management system is designed and developed with the whole process of the data acquisition, storage, analysis,

processing, and feedback; Finally, a comprehensive nuclear safety culture quantitative evaluation model combining

Back Propagation (BP) neural network and Analytic Hierarchy Process (AHP)-Fuzzy comprehensive evaluation method

is designed and applied based on the use of evaluation data and the fuzzy mathematical theory, data validation shows

that this model can be used for evaluating the comprehensive grade of nuclear safety culture in NPPs, and providing

basis for NPPs and corporate to monitor the nuclear safety culture level.


1 Preface

From the typical events in the

domestic nuclear power industry, the

operation events of the unit shut down

due to the failure of employees to

comply the procedures have occurred

occasionally [1], and the recurrence

of events caused by the failure to

implement the corrective action

measures required by the operating

experience feedback has also

emerged. These events demonstrate

the importance of nuclear safety

culture to nuclear safety [2].

Strengthening nuclear safety by

raising the nuclear safety culture

level is a common consensus in the

nuclear industry.

It was found that the traditional

nuclear safety culture promotion work

mainly instills the nuclear safety

culture concept through publicity and

training [3]. In recent years, and some

nuclear power plants have used international

advanced experience to

carry out activities such as personnel

behavior observation and coaching,

the combination of the theory and

behavior practice has been greatly

improved compared to the initial

one-way infusion. Nevertheless, it was

found that this model is not enough to

ensure the continuity and effectiveness

of nuclear safety culture enhancement

[4], lack of systematic

evaluation and closed-loop management

of nuclear safety culture level, it

is difficult to accurately identify the

culture weakness and take corresponding

improvement measures of

action [5].

Therefore, it is necessary to carry

out research work on nuclear safety

culture improvement management in

NPPs. Based on the nuclear safety

culture dynamic improvement model,

this article has carried out related

research and application work in the

development of nuclear safety culture

standards, nuclear safety culture

evaluation methods and application

of evaluation data.

2 Research on nuclear

safety culture


management of NPPs

2.1 Nuclear safety culture

dynamic improvement


In three-level cultural theory, Edgar

H. Schein [6] found culture includes

three levels of underlying and visible

basic assumptions, values, and

behaviors, they are integrated and

interrelated. If we continue to

strengthen cultural values and change

individual behavior through various

actions, we can guide individuals’

basic assumptions and values to

change in the desired direction.

International Atomic Energy

Agency (IAEA) proposed the relevant

requirements [7] for the organization

to enhance the nuclear safety culture,

and defined the nuclear safety

culture commitments of policy level,

managers and individuals. The

advantage is that the classification

and the responsibilities of each level is

clear and specific, but there is no

description of the relationship, and

there is no clear driving force to

enhance nuclear safety culture and

lack of evaluation.

The definition of nuclear safety

culture [8] by the World Nuclear

Operators Association (WANO) in

2006 clearly shows the relationship

between employees and leaders in the

promotion of nuclear safety culture,

and points out the role of leaders in

the promotion of nuclear safety


Based on the research and analysis

of the above-mentioned theory,

the nuclear safety culture dynamic

improvement model for NPPs is proposed,

as shown in Figure 1.

The model clarifies the role and

location of the corporate and NPPs in

the promotion of nuclear safety

culture, the corporate should issue

unified, clear, layered, and highstandard

nuclear safety culture

common language and put forward

the requirements for implementing

the nuclear safety culture enhancement,

the common language will be

widely publicized through training,

publicity and other activities to

deepen understanding, finally, the

requirements will be reflected in the

behavior of the on-site personnel.

The leaders of the nuclear power

plants play a vital leading role in the

process of enhancing the nuclear

safety culture, they are decisive forces.

They not only set an example by themselves,

but also act as a model for

practicing nuclear safety culture.

Leaders should conduct observation

Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February


| Fig. 1.

Nuclear safety culture dynamic improvement model for NPPs.

and coaching on the behavior of

employees [9], including contractor

employees, to improve the nuclear

safety culture level of employees.

Carrying out the nuclear safety culture

enhancing activities is not to increase

the management system, but to incorporate

the nuclear safety culture

requirements into the management

measures of NPPs. Finally, the corporate

should organize Nuclear Safety

Culture Assessment (NSCA) regularly,

use relevant means to understand the

nuclear safety culture status and

weakness of NPPs, so as to achieve

continuous improvement of the

nuclear safety culture level.

Current work about nuclear safety

culture enhancement has basically

met the requirements of NPPs in the

promotion of nuclear safety culture

which mentioned in the model, such

as training, publicity, behavior observation

and management measures

implementation [10]. The regulatory

requirements for corporate in the

model are key issues that need to be

addressed. This article will introduce

relevant research and application

work around these key issues, including

nuclear safety culture standards,

nuclear safety culture evaluation

methods and management systems,

and nuclear safety culture comprehensive

quantitative evaluation


2.2 Development of nuclear

safety culture standard

Establishing a unified nuclear safety

culture standard is an essential

element for the organization to promote

the nuclear safety culture. By

clarifying the basic requirement and

behavior criterion, all levels of

individuals in the organization can

improve the nuclear safety culture

level in accordance with the unified

goals and requirements.

In the development process of

nuclear safety culture standards, the

related requirements put forward in

the Nuclear Safety Culture Policy

Statement were considered, the

“Healthy Nuclear Safety Culture

Traits” [11] that issued by Institute of

Nuclear Power Operations (INPO) of

U.S. and WANO were also studied,

these will be taken care of during the

standard development process. Based

on this, the following “Ten Principles

of Excellence Nuclear Safety Culture”

(referred to as “Ten Principles”) were


In the development process of the

“Ten Principles”, two principles are

basically based on the INPO traits, and

the other principles are integrated

and supplemented according to the

above requirements, in particular,

some new attributes have been added


Ten Principles

such as “reflecting long-term performance”,

“avoiding organizational

complacency”, “being sensitive to

change” and “reporting truthfully to

regulators”. The newly developed

“Ten Principles” includes 10 principles,

46 attributes and 237 behavior

examples, these are the requirements

and reference practices to carry out

nuclear safety culture construction for

NPPs. Table 1 shows the comparison

of the NSC Ten principles and INPO/

WANO Ten Traits.

3 Development and

application of nuclear

safety culture evaluation


Nuclear safety culture evaluation can

be used to test and verify the nuclear

safety culture promotion effect of the

NPPs [12]. By identifying the nuclear

safety culture weakness, it is possible

to develop and implement improvement

actions in a targeted manner to

improve the nuclear safety culture

level continuously [13].

3.1 Design and application


In order to manage and utilize various

nuclear safety culture evaluation data

effectively, and to provide a basis for

subsequent data analysis, the Nuclear

Safety Culture Evaluation Management

System (NSCEMS) was designed

and developed.

NSCEMS mainly collects the data

of the NSCA and questionnaire

survey, and stores, processes and

analyzes the relevant data through

the data management module. The

system can evaluate the nuclear

safety culture status of corporate

and NPPs, and can realize multidimensional

evaluation data analysis

and trend analysis to identify common

problems and downgrade trends,

and transform related issues into

improved actions.

NSCEMS consists of three subsystems

and modules, NSCA system,

NSC questionnaire system and NSC

evaluation management module, the

workflow of the system is shown in

Figure 2.


Ten Traits


Ten Traits

Principles/Traits 10 10 10

Attitudes 46 40 40

Behavior examples 237 0 217

Posters 10 0 0

| Tab. 1.

Comparison of the NSC Ten principles and INPO/WANO Ten Traits.

Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February

NSCA system is mainly used for the

collection, processing and analysis of

relevant data on the NSCA, including

data obtained from personnel interviews

and behavior observations, and

provides a basis for the NSCA conclusions.

NSC questionnaire system is

mainly used for the collection, processing

and analysis of the questionnaire

survey data, involving the conclusions

of the questions and related

departments and post information,

etc., and provides a basis for comprehensively

grasping the acceptance

and implementation effects of the

NPPs on the ten principles. NSC

evaluation management module is a

sub-module of the NPP peer review

data management platform. It is

mainly used to unify the relevant data

of on-site assessment and questionnaire

survey and NSCA conclusions,

and realize the comprehensive processing

and analysis of nuclear safety

culture evaluation data.

There are three types of data and

information involved in the NSCEMS,

including questionnaire data, on-site

assessment data and NSCA conclusions.

Integrate the positive and

negative attributes obtained from the

analysis of the three types of data,

focus on the common problems

reflected by them, and comprehensively

derive the positive and negative

attributes that need attention. Table 2

shows an analysis case of the common

nuclear safety culture problems.

Based on the results of the comprehensive

analysis, focus on and

feedback negative attributes, and find

relevant facts and supporting evidence

in the three types of data, and conduct

the root cause analysis. Corporate and

NPPs can develop corrective actions

to improve the nuclear safety culture


4 Design and application

of the comprehensive

nuclear safety culture

quantitative evaluation


The nuclear safety culture level has

always been a qualitative concept, not

a quantitative concept [14]. In the

past, the assessment of the nuclear

safety culture level mainly stayed on

the basis of subjective or expert


The current NSCA mainly uses

questionnaires [15], on-site interviews

and other methods to obtain

employees’ views, attitudes and

opinions on nuclear safety culture

[16]. Through the positive, negative

and neutral evaluation data, the

| Fig. 2.

NSCEMS workflow.

overall situation of nuclear safety

culture and the weakness are

proposed. This method basically

realized the systematic evaluation of

the nuclear safety culture. Although

some quantitative data were initially

borrowed in the evaluation process,

the evaluation conclusions are still

qualitative, and it is impossible to

visually give the overall nuclear safety

culture status and what kind of the

nuclear safety culture level of the NPP.

The NSC comprehensive quantitative

evaluation model is used to solve

this problem. Considering the multilevel

nature of nuclear safety culture

and the fact that NPP is a complex

open system with many qualitative

factors, this article adopts BP neural

network and AHP-Fuzzy to carry out

Questionnaire data


WE.5- Alternate Process

for Raising Concerns

LA.5- Provide resources

WE.1- Respect is Evident

LA.1- Strategic

Commitment to Safety

LA.6- Incentives, Sanctions

and Rewards

NSCA Site Interview

data analysis

| Tab. 2.

Analysis case of the common nuclear safety culture problems.

research and design of a comprehensive

quantitative evaluation model for

nuclear safety culture comprehensive


4.1 Nuclear safety culture

quantitative level design

American psychologist Abraham

Maslow put forward the theory of

demand hierarchy in “Human Incentive

Theory” in 1943. Based on this theory

and combined the definition

of nuclear safety culture [17], the

nuclear safety culture is divided into

seven stages to correspond to the seven

nuclear safety culture levels, specifically

includes instinctive response

stage, passive management stage,

active management stage, employee

participation stage, team mutual

NSCA conclusions


Main negative


PI.2- Evaluation LA.5- Provide resources LA.5- Provide resources

LA.7- Change


LA.5- Provide resources

NS.3- Risk control

throughout the whole

work process

NS.5- High quality


CO.2- Bases for Decisions WE.1- Respect is Evident

PI.2- Evaluation

LO.5- Training

NS.5- High quality procedures

LO.2- Operating Experience

LO.5- Training

LO.3- Conduct


PI.2- Evaluation

LO.5- Training

WE.1- Respect is Evident

NS.5- High quality



Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February






Vector value

(V i )

Main characteristics

of the organization

1 [0, 50) 25 Individuals at all levels in the organization lack nuclear safety awareness, and their safety

behaviors are based on their own instinctive reactions.

2 [50, 60) 55 The source power on nuclear safety mainly comes from the requirements of regulators and

superiors. Individuals believe that nuclear safety is the responsibility of the leaders.

3 [60, 70) 65 The management has a certain understanding of the importance of nuclear safety, the

organization has defined the nuclear safety responsibilities and authority of individuals at

all levels, and enhances individuals’ nuclear safety awareness by improving the quality of

procedures and organizing training.

4 [70, 80) 75 Individuals understand their nuclear safety responsibilities and actively improve their safety

skills and safety awareness. Most line employees are willing to work with management to

improve and enhance the NSC.

5 [80, 90) 85 The organization recognizes nuclear safety as a collective responsibility, focusing on communication,

recognizing the value of all individuals, and recognizing that respect for employees

is important for nuclear safety. Free flow of information in the organization, management

level and employees work together to improve the NSC.

6 [90, 95) 92 Individuals at all levels in the organization have a strong NSC concept, basically forming a

team value with nuclear safety is emphasized over competing priorities, and continuously

improving the NSC level through continuous learning, training, and self-improvement.

7 [95, 100] 97 The organization has reached a highly self-disciplined NSC level. The NSC concept has been

integrated into every employee in the organization. The organization is full of trust and

respect. From management level to individuals, it pays close attention to nuclear safety.

NSC stage









Team mutual






| Tab. 3.

Nuclear Safety Culture Level Comparison Table.

assistance stage, continuous improvement

stage and high self- discipline

stage, Table 3 shows the nuclear safety

culture level com parison.

In addition to the main features

and stages, the assignment range and

vector value of each nuclear safety

culture level are also included, and

these two types of data are mainly

determined based on the experience

of the expert group.

4.2 A quantitative evaluation

method of nuclear safety

culture based on BP neural


BP neural network is a multilayer feed

forward network, which is trained

according to the error back propagation

algorithm style, it was found that

the BP neural network can learn and

store a large number of I/O mapping

relations, without prior mathematical

equation describing the mapping

relations, and it is very suitable for

processing a non-linear information

processing requirements [18]. The

topology structure of BP neural

network model includes input layer,

hidden layer and output layer, as

shown in Figure 3.

According to the characteristics of

the BP neural network, the design

process of the NSC quantitative rating

| Fig. 3.

The topology structure of BP neural network model.

| Fig. 5.

BP neural network error curve.

| Fig. 4.

Algorithm flow chart of NSC quantization neural network.

Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February

model, determine the 46 NSC

attribute evaluation score for the

input data, 7 NSC levels as the output

of the network model, the model of

relationship between the input and

output for complex nonlinear model,

and set a hidden layer.

The learning rule of BP neural

network is to use the gradient descent

method to continuously adjust the

weight and threshold of the network

through back propagation, so that

the squared error of the network is

minimized. In this algorithm, there

are 16 initial training samples,

including 7 ideal data samples, 7

fault- tolerant data samples, and 2

actual data samples. The algorithm

flow is shown in Figure 4.

The network achieves convergence

in step 87 and the actual output value

of the network satisfies the error

requirement through the learning of

the training samples, as shown in

Figure 5 for details.


Matrix E is the error value after

iterative calculation, it can be seen

that the actual output value of the

neural network is basically consistent

with the expected output value. The

model can be used for the evaluation

of nuclear safety culture quantitative


4.3 Nuclear safety culture

quantitative evaluation

method based on AHP-

Fuzzy comprehensive

evaluation method

AHP is a multi-objective decisionmaking

method combining qualitative

and quantitative analysis. It was found

that the method determines the

weight coefficient of each index by

decomposing the decision problem

into a hierarchical structure [19].

Fuzzy comprehensive evaluation

method is a method of making comprehensive

decision-making on things

subject to various factors by using

fuzzy mathematics and fuzzy statistics

in a fuzzy environment. Combining

the two methods, the main factors

affecting the NSC are established to

form an orderly hierarchical level

index [20]. The AHP method is used

to calculate the relative importance

degree between each level of indicators.

Finally, the fuzzy comprehensive

Importance level

| Tab. 4.

Nuclear Safety Culture Level Comparison.

evaluation method is used to calculate

the final nuclear safety culture level.

The nuclear safety culture quantitative

rating steps based on the AHP-

Fuzzy comprehensive evaluation

method are as follows.

1) Establish an evaluation indicator

set, the nuclear safety culture

primary and secondary indicator

systems are completely based

on the framework of NSC ten

principles. The primary indicators

are 10 principles, and the secondary

indicators are a number of

attributes for each principle, for a

total of 46.

2) Determine the weight of each level

of indicators, the AHP method is

used to determine the weight set

of the primary and secondary

indicators of nuclear safety culture.

The relative importance of each

evaluation index is judged by

the discriminant matrix method.

Table 4 is the scale of the pairwise

index of each level.

In the process of using the discriminant

matrix method, the specific

evaluation results are determined

after expert discussion and have

certain authority. Table 5 shows the

discriminant matrix and weight of the

primary indicators, and the discriminant

matrix and weight of the secondary

indicators are calculated in the

same way.

According to the above method,

the weight set of the primary and

secondary indicators can be calculated,

wherein W is a primary indicator

weight set, and W i is a secondary

indicator weight set under the

principle i.

3) Fuzzy comprehensive evaluation,

based on the data points generated

by the nuclear safety culture onsite

assessment, all the attributes

of the coverage nuclear safety

culture evaluation of these data

can be evaluated by the weighted



The element r ij (row i and column

j) in the matrix R indicates the

membership degree of the evaluation

indicator from the factor u i to

the v j level, and combines W with

the evaluation matrix R to obtain

the evaluation result vector B of

the secondary indicators.


| Tab. 5.

Discriminant matrix and weight of the primary indicators for NSC quantitative evaluation.

C ij Assignment

Two elements (i, j) are equally important 1

Element i is slightly more important than element j 3

Element i is significantly more important than element j 5

Element i is strongly important than element j 7

Element i is extremely important than element j 9

Intermediate value between the above adjacent judgments 2,4,6,8

Element i is compared with element j and is opposite to the above judgment result

A A 1 A 2 A 3 A 4 A 5 A 6 A 7 A 8 A 9 A 10 Weight

A 1 1 2 4 1 3 2 1/2 5 5 7 0.169

A 2 1/2 1 3 1/2 2 1 1/3 4 4 5 0.106

A 3 1/4 1/3 1 1/4 1/2 1/3 1/5 2 2 3 0.046

A 4 1 2 4 1 3 2 1/2 5 5 7 0.169

A 5 1/3 1/2 2 1/3 1 1/2 1/4 3 3 4 0.069

A 6 1/2 1 3 1/2 2 1 1/3 4 4 5 0.106

A 7 2 3 5 2 4 3 1 6 6 7 0.252

A 8 1/5 1/4 1/2 1/5 1/3 1/4 1/6 1 1 2 0.030

A 9 1/5 1/4 1/2 1/5 1/3 1/4 1/6 1 1 2 0.030

A 10 1/7 1/5 1/3 1/7 1/4 1/5 1/7 1/2 1/2 1 0.021

1/C ij


Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February


Model calculation result

Where bi is calculated from the

column j of W and R, which indicates

the membership degree of

the evaluation indicator to the NSC

level V j (j=1, 2, …, 7). After the

above calculation, the results of the

indicator evaluation of B 1 , B 2 ,…,

B 10 are obtained through fuzzy

comprehensive evaluation.


4) NSC level calculation, after obtaining

the fuzzy comprehensive

evaluation vector B, the final

nuclear safety culture level is calculated

based on the fuzzy comprehensive

evaluation vector and

the NSC level vector V, where

V=(V 1 ,V 2 ,V 3 ,V 4 ,V 5 ,V 6 ,V 7 ) T .

G = B • V(5)

BP Neural network model measurement level (a) > AHP-Fuzzy

model measurement level (b)

BP Neural network model measurement level (a) = AHP-Fuzzy

model measurement level (b)

BP Neural network model measurement level (a) < AHP-Fuzzy

model measurement level (b)

| Tab. 6.

NSC Quantitative Grade Criterion in NPPs.

According to the calculation result

of G, the range of assignment of

each grade is compared, and the

final nuclear safety culture level is


4.4 Application of Quantitative

Evaluation Method of

Nuclear Safety Culture

Considering that the above quantitative

evaluation models are based on

fuzzy theory, in the actual application

process, we will comprehensively

consider the calculation results of the

BP neural network and AHP-Fuzzy

models, and determine the final

nuclear safety culture quantification

based on the criteria shown in

Table 6.

According to the above method,

the NSCA results of three NPPs have

been measured by using the nuclear

safety culture comprehensive

Comprehensive evaluation result (C)




quantitative evaluation model, and

the specific calculation results are

shown in Table 7.

Based on the analysis of the model

verification results, the following

conclusions can be known.

1) From the conclusion of the nuclear

safety culture comprehensive

evaluation level, the test NPPs are

basically in the third and fourth

level of nuclear safety culture, that

is to say, they are basically in

the active management stage or

employee participation stage. This

shows that the design of the comprehensive

quantitative evaluation

model is basically reasonable and


2) It can be seen from Table 7 that the

number of negative nuclear safety

culture conclusions does not show

a significant proportional trend to

the nuclear safety culture comprehensive

evaluation level. The

reason is that the NSCA conclusions

refer to the on-site assessment

data points, but more based

on the evaluation of the collected

cases or facts to make judgments.

The quantitative evaluation model

is based on the judgments of all

the evaluation data of NPPs, which

can objectively reflect the overall

nuclear safety culture level of



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ISSN 1431-5254

Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February

Plant A Plant B Plant C

NSCA conclusion

3) The quantitative rating and NSCA

can complement each other. NSCA

pays attention to some specific

points, and the quantitative evaluation

model provides the overall

nuclear safety culture trend of


4) Since there are not many assessments

based on the new nuclear

safety culture standards, the horizontal

comparison of the nuclear

safety culture level between NPPs

can be initially realized. After the

data is accumulated, it can be

applied to the vertical comparison

of the nuclear safety culture level,

the nuclear safety culture trend

will be identified timely, the corrective

actions will be taken and to

improve nuclear safety culture continuously.

5 Conclusions

In view of the current problems of lack

of continuity and effectiveness in the

nuclear safety culture improvement

work of nuclear power plants, this

article can provide solutions by conducting

research and application work

of nuclear safety culture improvement

management of NPPs. The nuclear

safety culture dynamic improvement

model has the foundation of theory

and practice. The nuclear safety

culture standard not only reflects the

international advanced practices but

also reflects its own experience, and

can provide guidance for the NPPs to

carry out the nuclear safety culture

promotion work.

Based on the continuous improvement

of nuclear safety culture evaluation

technology, nuclear safety culture

evaluation can be effectively carried

out to identify the nuclear safety

culture weakness of NPPs. At the same

time, through the establishment of

nuclear safety culture evaluation

management system and its supporting

data analysis mechanism, it can

help corporate and NPPs to mine common

problems and urgent problems

from various evaluation data.

The nuclear safety culture comprehensive

quantitative evaluation model

has realized the secondary utilization

of the evaluation data, and solved the

problem that there are no effective

means to evaluate the overall nuclear

safety culture level of NPPs. Combined

with the nuclear safety culture evaluation

method, the model can be used to

monitor the nuclear safety culture

trend for corporate and NPPs.

At present, the nuclear power

industry have given full attention to

nuclear safety culture. The research

results of nuclear safety culture

management research of nuclear

power plants have broad application

prospects in China. For example,

government regulatory agency and

utilities can apply the relevant results

of this article to conduct nuclear safety

culture monitoring, evaluation and

comprehensive quantitative rating,

so as to achieve comprehensive monitoring

of the nuclear safety culture

level. For NPPs, this achievement can

also be used to identify weakness and

improve the nuclear safety culture

level continuously.


4 positive observations

5 general observations

6 negative observations

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Xiaozhao Xu

Senior Engineer

Assessment Technology Supervisor

of Research Institute of Nuclear

Power Operation

Jun Guo

Senior Engineer

Assessment Technology Director of

Research Institute of Nuclear Power


Sujia Li

Professor of Engineering

Vice President of Research Institute

of Nuclear Power Operation

China National Nuclear

Corporation (CNNC)

No.1021 Minzu Street, East lake

High-tech Development Zone

Wuhan City, Hubei Province, China

4 positive observations

4 general observations

3 negative observations

BP neural network model calculation results Level 4 Level 4 Level 3

AHP-Fuzzy model calculation results Level 4 (78.84) Level 5 (81.79) Level 4 (74.23)

Comprehensive quantitative evaluation level Level 4 Level 4 Level 3

| Tab. 7.

Calculation and application of NSC comprehensive quantitative evaluation model.


Environment and Safety

Research and Application of Nuclear Safety Culture Improvement Management for NPPs in China ı Xiaozhao Xu, Jun Guo and Sujia Li

atw Vol. 65 (2020) | Issue 2 ı February


Konstruktionsprinzipien zur nuklearen

und betrieblichen Sicherheit von HTR-

KKW – ein Review

Urban Cleve

1 Ziele der HTR-Entwicklung. Bereits zu Beginn seiner Tätigkeit in 1956 als Leiter der Reaktorentwicklung

bei BBC sah es Prof. Dr. Rudolf Schulten als seine Aufgabe, ein inhärent sicheres Kernkraftwerk zu

entwickeln. Keine noch so schwierig zu beherrschende nukleare oder betriebliche Störung durfte zu einem „GAU“

führen. Szenarien, die zu einer Verunsicherung der Bevölkerung führen könnten und damit die Akzeptanz von KKW

erschweren, ja verhindern könnten, sollten durch nuklear-physikalische Maßnahmen und Konstruktionen unmöglich

sein. Ziel war der „katastrophenfreie“ Kernreaktor [1].

Restrisiken soll es nicht geben, sie sind

grundsätzlich auszuschließen. Er

erdachte die Kugel als betriebs sicheres

Brennelement. Hohe Temperaturen

sollten möglich sein, daher Graphit

mit seiner Festigkeit bis zu 3.000 °C

als wesentliches Bauelement für die

Brennelemente, für den Reaktorkern

und als Moderator. Kühlung des

Reaktorbettes durch ein inertes Gas

wie Helium im geschlossenen Kreislauf

innerhalb eines Druckbehälters.

Dies sind bis heute die wichtigsten

Bauelemente eines HTR. Es waren

geradezu visionäre Überlegungen

[2, 3].

Gebaut nach diesen Ideen wurden

das 15 MW el AVR-Versuchskernkraftwerk

in Jülich [2] und das THTR-

300 MW el -Demonstrationskernkraftwerk

der VEW in Hamm/Uentrop-


2 Sicherheitsanforderungen

an zukünftige


In einer Besprechung auf Vorschlag

des BMBF-Referates 722 „Energie“

erläuterte Prof. Dr. K. Kugeler [3, 4]

sicherheitstechnische Anforderungen

an (V)HTR-KKW, die über die derzeit

nach Fukushima geforderten Anforderungen

der RSK [5] hinausgehen.

Alle in einem Bericht der Reaktor-

Sicherheitskommission (RSK) erwähnten

Kriterien lassen sich mit

einem HTR realisieren.

Im Einzelnen sind dies:

p Erdbebenauslegung und Bodendynamik;

p Hochwasserauslegung;

p Weitere externe Ereignisse wie

extreme Wetterbedingungen, Flugzeugabsturz,

Cyberangriff, Pandemie;

p Kombinationswirkungen von externen


p vollständiger Ausfall der Stromversorgung;

Weiter sind Anforderungen, beschrieben

unter „Konkrete Maßnahmen“,

soweit diese für einen HTR überhaupt

in Frage kommen, und die beschriebenen

Schadensszenarien konstruktiv

und planungstechnisch realisierbar

und gelten als grundlegende Anforderungen

an die Sicherheit, werden

also in die Sicherheitsberichte aufgenommen.

Darüber hinaus werden die folgenden

zusätzlichen Forderungen erfüllt

[3, 4].

p Berstsicherer Primärgaseinschluss,

auch bei Terrorangriffen und Sabotage

von innen und außen;

p Selbsttätige Nachwärmeabfuhr;

p Coreauslegung unempfindlich

gegen Reaktivitätsstörungen;

p Core unempfindlich gegen Lufteinbruch;

p „Zero-Emissionskonzept“ auch bei


p Keine radioaktiv ver-/bestrahlte

oder kontaminierte Teile außerhalb

des KKW, kein Transport

dieser Teile über die Straße oder

Schiene zwingend erforderlich;

Eine Notkühlung für Brennelemente

und ein Abklingbecken mit Kühlwasserversorgung

ist nicht erforderlich,

da abgezogene Kugelelemente keine

Nachwärmeproduktion haben.

3 Erfahrungen aus dem

Betrieb des 15 MW el


in Jülich

Die beim Betrieb des AVR gewonnenen

positiven und negativen

Erfahrung werden in dem beschriebenen

neuen Konzept berücksichtigt.

Als grundlegende Erfahrungen

sind anzusehen [13, 19]:

p Der zweimalige Nachweis der inhärenten

Sicherheit durch einen

simulierten GAU; /3/;

p Die geringe Bruchrate bei der

Umwälzung der Brennelemente;

p Die ausgezeichnete Stabilität der


p Die einwandfreie Funktion der

Abschalt- und Regelstäbe;

p Die Möglichkeit, Reparaturen an

wichtigen Komponenten, Gebläse,

Beschickungsanlage, z. T. während

des Betriebes durchführen zu

können, ohne dass das Personal

einer zu hohen Strahlendosis ausgesetzt


p Die unerwartet geringe Menge an


Negativ war der Schaden am Dampferzeuger

durch eine undichte

Schweißnaht. Diese Störung nach

INES 1 war von Anfang an eingeplant

worden. Die zur Behebung eines

Schadens erforderlichen konstruktiven

und betrieblichen Maßnahmen

waren getroffen. Das Schadensereignis

lief wie in den Betriebsgenehmigungen

und Betriebsvorschriften

festgelegt ab und der

Schaden wurde behoben. Negativ war

die lange Stillstandszeit des Reaktors.

Eventuelle Auswirkungen eines

solchen Dampf/Wassereinbruchs waren

lange vor Inbetriebnahme des

AVR von mehreren renommierten

wissenschaftlichen Instituten und den

Genehmigungsbehörden untersucht

worden. Auch wurde experimentell

das Verhalten heißer Brennelemente

in einem Kugelbett bei plötzlicher

Abkühlung durch Wasser überprüft.

In einem Bericht [9] wird dies

nicht berücksichtigt und kann

möglicherweise auf die zeitliche

Differenz zwischen Bericht in 2006

und der erteilten Betriebsgenehmigung

seitens der RSK und dem

TÜV im Jahre 1964 basieren. Ohne die

erfolgten positiven Untersuchungen

wäre eine Betriebsgenehmigung für

den AVR nicht erteilt worden. Die im

Bericht zusammengefassten Informationen

sind seit 40-50 Jahren

bekannt, also kein neuer Gedanke.

Environment and Safety

Design Principles for Nuclear and Operational Safety of HTR NPPs – a Review ı Urban Cleve

atw Vol. 65 (2020) | Issue 2 ı February

4 Erfahrungen mit

dem 300 MW el -THTR-


in Hamm-Uentrop/


Die Entscheidung, nach dem AVR mit

einer Leistung von 15 MW zu einem

KKW mit einer Leistung von 300 MW

überzugehen, und das noch, bevor

Betriebsergebnisse des AVR vorlagen,

war ein extrem mutiger Schritt. Es

war das Ziel, nachzuweisen, dass ein

HTR-KKW mit konventionellen Kraftwerken

gleicher Größe im Netzbetrieb

eingesetzt werden kann, und hierzu

war diese Entscheidung notwendig

und vor allem auch aus heutiger Sicht


Das Grundkonzept des THTR-300

musste gegenüber dem AVR bei

mehreren wichtigen Konstruktionen

geändert werden:

p Spannbetondruckbehälter anstelle

zweier Stahlbehälter;

p Helium Primärgasdruck 40 bar

gegenüber 10 bar;

p Änderung der BE-Abzugsvorrichtung;

p Keine doppelt ummantelten


p Kühlgasströmung von oben nach


p Abschalt- und Regelstäbe in den


Leider etwas spät wurde bei einer

Nachberechnung des Cores erkannt,

dass wegen des wesentlich größeren

Coredurchmessers und der höheren

Nachwärmeproduktion der Reaktor

nach Abschaltung nicht kaltgefahren

werden konnte. Das Erst-Konzept

musste also geändert werden. Es

wurden zwei Vorschläge besprochen.

Die Abschaltstäbe sollen direkt in das

Kugelbett eingefahren werden, die

Regelstäbe verbleiben im Grahitreflektor

oder alternativ ein Ringcore

mit Abschalt- und Regelstäben in den

Graphiteinbauten. Da noch keine

Erfahrungen über das Verhalten der

Graphiteinbauten aus dem AVR

vor lagen, fiel die Entscheidung zugunsten

der Lösung mit Einfahren der

Stäbe in das Brennelementbett. Die

Gefährdung der Brennelemente durch

Bruch und das mögliche Verbiegen

der Stäbe wurden bewusst in Kauf

genommen und als das geringere

Risiko angesehen.

Der befürchtete Kugelbruch ist

beim Betrieb des THTR eingetreten.

Er ist so hoch, dass dieses Konstruktionsmerkmal

bei weiteren HTR nicht

mehr verwendet werden kann.

Weiter kam es zu Problemen mit

der Abzugseinheit für die Brennelemente,

auch hierdurch kann

| THTR Thorium-Hochtemperatur-Reaktor bei Hamm-Uentrop.

zusätzlicher Bruch eingetreten sein.

Die Abzugseinheit des AVR ist wesentlich

besser und soll bei künftigen Anlagen

unverändert eingebaut werden.

Beide Erfahrungen hatten keinerlei

Einfluss auf die nukleare Sicherheit

der Anlage, sie führte aber zu nicht

unerheblichen betrieblichen Herausforderungen.

Die positiven Erfahrungen aber

sind, dass Ziele, Erkenntnisse und

Erfahrungen, die mit dem THTR-300

erreicht werden sollten, erfolgreich

realisiert wurden.

Dies sind:

p Ein HTR-Kernkraftwerk ist genau

so gut regelbar wie ein konventionelles


p Ein Frequenzregelbetrieb ist sehr

gut realisierbar;

p Vom Netz geforderte Leistungsschwankungen

können problemlos

nachgefahren werden;

p Der Betrieb mit Zwischenüberhitzung,

bislang einmalig mit

einen KKW realisiert, war uneingeschränkt

möglich, mit einem

thermodynamischen Wirkungsgrad,

der genau so gut ist, wie bei

konventionellen Kraftwerken.

p Alle Komponenten, d. h. vor allem

die Gebläse, die Abschalt- und

Regelstäbe, die Brennelement-

Beschickungs- und -Umwälzanlage,

die Helium-Gaskreisläufe

arbeiteten trotz Leistungsvergrößerung

genau so zuverlässig

wie beim AVR;

p Der Sekundärteil mit konventioneller

Stromerzeugung arbeitet

absolut betriebssicher;

p Der Spannbetonbehälter ist bei

Stilllegung der Anlage das beste,

einfachste, sicherste und preiswerteste


Diese umfassenden Erfahrungen

ermöglichen den Bau neuer HTR-

Groß-Kernkraftwerke [8; 14 – 19].

5 Die Sicherheit der


Von entscheidender Bedeutung für die

Sicherheit der HTR-Technik war und

ist die Entwicklung der Graphitkugeln

mit eingepressten, in drei Hülllagen

umschlossenen Coated Particles. Hierbei

werden UO 2 + ThO 2 oder UC +

ThC als Brutbrennstoffe oder jedwede

weitere Brennstoff- Partikel kombi nation

[6] mit einem Kern- Durchmesser

von 0,5 – 0,7 mm von drei gasdichten

Lagen aus pyro lytischem Kohlenstoff

beschichtet. Deren Durchmesser beträgt

ca. 0,9 mm. Etwa 15.000 bis

30.000 dieser Partikel werden in

den Graphit der Kugeln eingepresst.

Messungen haben gezeigt, dass diese

sehr harten Schichten bis 1.600 °C

gegen den Austritt von Spaltpro dukten

gasdicht bleiben. Man nennt diese

Coated Particles wegen ihrer Härte

auch Panzerkörner. Sie sind so hart,

dass sie auch bei einem Kugelbruch

nicht beschädigt werden, was der

Kugel bruch im THTR und die dennoch

geringe Aktivität des Primärheliums

zeigen. Diese dreifache Beschichtung

sind die ersten drei Sicherheitsbarrieren

gegen den Austritt von

Spaltprodukten in das Primärgas. So

konnte beim AVR die Aktivität des

Primär- Heliumgases innerhalb des

Reaktorbehälters von der zunächst angenommenen

Aktivität von 10 7 Curie

auf gemessenen 760 Curie gesenkt

werden. Ein Wert, der auch bei einer


Environment and Safety

Design Principles for Nuclear and Operational Safety of HTR NPPs – a Review ı Urban Cleve

atw Vol. 65 (2020) | Issue 2 ı February


Totalemission in die Umgebung nicht

zu einer zu hohen Belastung geführt


Beim THTR betrug die Aktivität

1x10 7 Bq bei 47.000 m 3 Primär gasvolumen.

Auch hier wäre bei einer

Totalemission keine Evakuierung der

Umgebung erforderlich geworden.

Abgebrannte und abgezogenen

Brennelemente müssen sicher gelagert

werden, um den Nicht verbreitungsvertrag/Non-Profileration


(NPT) für nukleare Brennstoffe einhalten

zu können [6, 16]. Bei der

großen Zahl von Brennelementen mit

variierendem Gehalt von Uran,

Thorium und dem bei der Verbrennung

von U-238 entstehende spaltbaren

Plutonium, sowie den strahlenden


und borhaltigen Kugeln ist eine

Markierung oder gar Nummerierung

nicht möglich. Jedes einzelne Element

wird aber gemessen. Der Plutoniumgehalt

hängt von der Höhe des

Abbrands des U-238 ab, je höher der

Abbrand, umso geringerer Rest von

Plutonium. / 16/ Diese Zusammenhänge

sind im Detail erforscht und

beschrieben von D.L. Moses [6].

Die jahrzehnte lange Lagerung der

Kugeln aus dem AVR und dem THTR

in Jülich und Ahaus beweisen, dass

eine gesicherte und sichere Lagerung

dieser Elemente einfach und problemlos

möglich ist.

Brennelemente haben den alles

entscheidenden Einfluss auf die

Sicherheit jedes Kernkraftwerkes.

Einen solch hohen Sicherheitsstand

und einfache Handhabung hat

kein anderes Brennelement.

6 Sicherheitsmaßnahmen

für die Gesamtanlage

Die Konstruktion der Gesamtanlage

erfolgt nach den in Kap. 2 festgelegten


Ausgenommen hiervon sind alle

sekundären Anlagen, wie Stromerzeugung,


und alle anderen verfahrenstechnischen

Anlagenbereiche. Der Sekundärteil,

also die Stromerzeugung, war

bereits beim THTR-300 nicht Teil des

atomrechtlichen Genehmigungsverfahrens.

Der Betrieb hat gezeigt, dass

durch den Wasserdampf, der im

Primärgaskreislauf liegende Dampferzeuger

erzeugt wird, keine Radioaktivität

in den Sekundärteil übertragen

wurde. Die Turbogruppe

konnte verkauft werden und war

über viele Jahre anschließend weiter

in Betrieb. Dies war nur möglich,

da sie während des nuklearen Betriebes

nicht kon taminiert worden ist.

Die He-He- Primärgaswärmetauscher

arbeiten eher mit noch höherer

Sicherheit gegen Spaltproduktdurchbruch.

Ehrgeiziges Ziel der Sicherheitsplanung

ist das „Zero-Emissions-


Auch im schlimmsten möglichen

Störfall soll und darf keine unzulässig

hohe radioaktive Strahlung oder

Kontamination der Umgebung

möglich sein.

Nuklear-physikalisch gilt, dass die

Anlage inhärent sicher ist [1, 3, 16].

Betrieblich werden folgende baulichen

Maßnahmen getroffen:

p Erdbebensicheres Fundament in

maximaler Stärke eines etwa zu

erwartenden Erdbebens, mindestens

Stärke 6;

p Über dem Fundament wird ein

Bunker mit starken Betonwänden

errichtet. Dieser ist sturm- und

wasserfest und damit luft-, gasund

wasserdicht auszulegen;

p Auf dem Fundament steht die

Stützkonstruktion für den Spannbetonbehälter,

diese trägt den


p Im Bunker werden alle Komponenten

bearbeitet oder endgelagert,

die radioaktiv strahlen oder

kontaminiert sind;

p Die Sicherheitseinrichtungen;

p Dies sind:

p Die Abzugseinheiten für Brennelemente,

diese liegen innerhalb

der Stützkonstruktion für

den SBB;

p Die Be-Schnellabzugsanlage;

p Eine Werkstatt mit Dekontamination

der zur Reparatur

vorgesehenen Kom ponenten;

p Das Lager für abgebrannte

Brennelemente und bei Schnellabzug;

p Die Notstromeinrichtungen

und Batterie;

p Alle im Störfall erforderlichen

Hilfsanlagen, auch die mobilen;

p Die Um- und Abluftreinigungsanlagen

und Filter;

p Um bzw. über den gesamten

nuklearen Teil wird ein Containment

errichtet, dessen Volumen so

groß und druckfest ist, dass das

gesamte Primärgasvolumen des

SBB aufgenommen werden kann.

p Brennelement-Schnellabzug;

Im äußersten Notfall, bspw. bei

Gefahr kriegerischer Handlungen,

oder wenn keine der übrigen

Sicherheitsmaßnahmen einsetzbar

sein sollten, kann das Core durch

Kugelabzug von Hand und deren

Lagerung in speziellen Behältern

in relativ kurzer Zeit von allen

Kugeln entleert werden. Positiv ist,

dass keine Nachwärmeproduktion

erfolgt, die Behälter also nicht

gekühlt werden müssen, und

dass mehrere Abzugseinheiten

vorhanden sind.

p Die Notsteuerstelle:

Es werden 2 Notsteuerstellen vorgesehen,

die 1. In der Warte, also

im Sekundärbereich, die 2. im


7 Die Konstruktion



Sicherheitsrelevante Komponenten


p Der Spannbetonbehälter mit

Linerkühlsystem, Isolierung und


Der Spannbetonbehälter ist nach

den dreifachgasdichten Hüllen

der Coated Particles die vierte

Sicherheitsbarriere gegen den

Austritt von Spaltprodukten. Er ist

gleichzeitig das Bio-Schild.

Versuche in eine 1:20 Modell mit

warmem Wasser haben nachgewiesen,

dass ein Spannbetonbehälter

nicht längere Zeit aufreißen

kann. Nach einer Druckentlastung

bei eingetretener Undichtigkeit

ziehen die Spannkabel

den Beton so zusammen, dass er

wieder gasdicht ist. Der Bruch des

Test-SBB fand erst bei 5-fachem

Überdruck gegenüber Auslegungsdruck

statt, einer Druckerhöhung,

die praktisch nicht eintreten


Im Betrieb kann der SBB nur durch

zu hohe Temperaturen gefährdet

werden, der Betrieb der Linerkühlung

muss also gewährleistet

sein. Weiter erhält der SBB einen

speziellen Beton mit höherer Festigkeit

und verbesserter Wärmeleitfähigkeit

nach außen.

Die elektrischen Antriebe der

Kühlwasserpumpen werden mittels

Notstromdieselanlagen und

Batterien im Notfall abgesichert;

zusätzlich können mobile Versorgungsanlagen

eingesetzt werden,

sodass eine Kühlung gesichert


Letztlich kann der Druck im SBB

durch Absenken des Druckes des

Primär-Heliums und Abpumpen in

das Heliumlager druckentlastet


p Die Primärgasgebläse

Eingebaut werden mehrere Gebläse,

bspw. sechs. Sie haben die

Environment and Safety

Design Principles for Nuclear and Operational Safety of HTR NPPs – a Review ı Urban Cleve

atw Vol. 65 (2020) | Issue 2 ı February

Aufgabe, die im Core produzierte

Wärme im geschlossenen Primärgaskreislauf

über die He-He-

Wärmetauscher an das Sekundär-

Helium zu übertragen, das die dort

produzierte Wärme an die folgenden

Anlagen im Sekundärteil

abgibt. Im Störfall sollen die Gebläse

betriebstüchtig sein, um die

im Core produzierte Nachwärme

nach außen abführen zu können.

Sie haben daher eine Sicherheitsfunktion,

wobei der Betrieb eines

Gebläses ausreicht, um die Nachwärme

abführen zu können [7].

Aus diesem Grunde müssen die

Antriebe der Gebläse mittels Notstromanlage

und Batterien abgesichert

sein. Wenn beide ausfallen

sollten, bleibt genügend Zeit

[16] um die Motoren mittels eines

mobilen Hilfsaggregates auch von

Hand betätigen zu können [7].

p Die Abschalt- und Regeleinrichtungen:

Alle Abschalt- und Regelstäbe

befinden sich im Reflektor. Sie

fallen bei Stromausfall durch

Schwerkraft durch Auslösen der

Kupplung in die Reflektoren ein.

Bei der großen Zahl genügt es,

wenn ca. 1/3 der Stäbe ausgelöst

werden, um den Reaktor abzuschalten.

p Der Brennelementschnellabzug:

Der Brennelementschnellabzug

ermöglicht es, vor allem wenn

mehrere Abzugseinheiten vorhanden

sind, per Schwerkraft die

Kugeln in relativ kurzer Zeit abzuziehen,

das Core also zu entleeren,

und die Kugelelemente im Lager zu


p Das Brennelementlager:

Vom Brennelementlager kann

keinerlei Gefahren für die Anlage

ausgehen. Ein gekühltes „Abklingbecken“

ist nicht erforderlich. Es

liegen langjährige Erfahrungen mit

der Lagerung der AVR- und der

THTR-Brennelemente vor.

p Das Containment:

Dies ist die 5. und letzte Barriere

gegen den Austritt von Radioaktivität

in die Umgebung;

p Instrumentierung und


Die zentrale Warte befindet sich im

Sekundärteil, hier ist auch die

1. nukleare Notsteuerstelle untergebracht.

8 Beherrschung extremer

Einwirkungen von außen

a. Kriegerische Ereignisse,

Cyberangriff, Pandemie:

Maßnahmen: Entleerung des

Cores über Schnellabzug; Damit

kann das KKW keine Gefahr mehr

für die Umgebung darstellen.

Abpumpen des Heliumgases in das


b. Flugzeugabsturz, Raketen/

Droh nenangriff von außen:

Schadensfolge: Containment wird

durchschlagen, der Spannbetonbehälter

mit 6 – 8 m dicken vorgespannten

Betonwänden wird

nicht durchschlagen. Gebläse

und/oder Abschalt-Regelstäbe

werden beschädigt. Wegen des

Einbaus von Rückhaltevorrichtungen/Dichtungen

für das Primärgas

in den Behälterdurchdringungen

dieser Komponenten bleibt der

SBB gasdicht.

Keine nuklearen Schadensfolgen,

keine Kontamination der Umgebung.

Bei geringer Undichtigkeit

des SBB kann das Containment

provisorisch drucklos abgedichtet

werden, ohne dass die Gefahr

einer zu hohen Strahlenbelastung

des Personals besteht.

Ferner: Entleeren des Cores;

Abpumpen des He-Primärgases;

c. Explosion durch Sabotage

innerhalb des Containments:

Folgen: Schutzbehälter wird

durchschlagen, Spannbetonbehälter

bleibt dicht, keine Kontamination

der Umgebung.

d. Explosion durch Sabotage

innerhalb des Bunkers:

Folge: keine unmittelbare

Beschädigung des SBB, keine

Kontamination der Umgebung.

e. Hochwasser, Sturm, Tsunami,

extreme Wetterlagen

Folgen: Bunker bleibt dicht,

keinerlei Folgen.

9 Schlussbetrachtung:

Die inhärente Sicherheit eines HTR-

Reaktors ist die Basis für alle Sicherheitsanalysen.

Haupt-Planungskriterium für die

Sicherheit von Kernkraftwerken ist

die Sicherheit der gesamten Anlage.

Die Sicherheit aller eingesetzten

Komponenten und der Gesamtkonstruktion

ist von entscheidender


Sicherheitskriterien müssen gegenüber

Wirtschaftlichkeitsfragen absoluten

Vorrang haben.

Wichtig ist, dass denkbare Störungen

nur langsam ablaufen, dadurch

ist genügend Zeit, die richtigen

Maßnahmen zur Minderung eines

Schadens einzuleiten.

Mit den beschriebenen Konstruktionsprinzipien

ist das von Schulten

angestrebte „Zero-Emissionskonzept

auch bei Betriebsstörungen“ für KKW


Ein derartig hoher Sicherheitsstandard

kann von keinem der derzeit

in Betrieb oder Planung befindlichen

KKW erreicht werden.

Alle Konstruktionsprinzipien sind


Es gilt:

„Der sicherste Reaktor ist auch der

wirtschaftlichste Reaktor“.


1. Kurt Kugeler: „Gibt es den katastrophenfreien Kernreaktor?“

Physikalische Blätter 57 (2001) Nr.11.

2. Festschrift: „50 Jahre AVR“ 2009;

3. Urban Cleve: „Die inhärente Sicherheit der HTR-Kernkraftwerke

mit Kugeln als Brennelemente“. 2012.

4. Kurt Kugeler: „Aspekte der VHTR-Entwicklung“.

Besprechungsvorlage KIT-KARLSRUHE Dez: 2011.

5. RSK Arbeitsgruppe RS I 3: “Erste Überlegungen zu

Konsequenzen aus Fukushima“. RS I 3 13042/9.

6. David L. Moses: „ Nuclear Safeguards Considerations for

Pebble Bed Reactors (PBRs)“. Paper Nr. 185 HTR-Conference

Prague 2010.

7. W.Rehm und W. Jahm: “Thermodynamisches Sicherheitsverhalten

des HTR bei Coraufheizunfällen“. BWK Bd. 39

(1987) Nr. 10.

8. Horst Bieber: „Hochtemperatur-Reaktor in Hamm Störfallaber

bei wem?“ DIE ZEIT (1986/24).

9. Rainer Moormann: „A safety re-evaluation of the pebble bed

reactor operation and its consequences for future HTRconcepts“.

FZ-Jülich, Jül-4275.

10. W. Krämer: “ Die Angst der Woche/ Warum wir uns vor den

falschen Dingen fürchten”. ISBN 978-3-492-05486-7 2011

11. VDI-Gesellschaft Energietechnik: “AVR – 20 Jahre Betrieb”.

VDI Berichte 729, VDI-Verlag, 1989.

12. Urban Cleve: “Verpaßte Entwicklung im Kernkraftwerksbau”.

FAZ 22. 7.2008.

13. Urban Cleve: „Die Technik der Hochtemperaturreaktoren“.

Atw 12/2009.

14. Urban Cleve: „Technik und künftige Einsatzmöglichkeiten

nuklearer Hochtemperaturreaktoren“. Fusion Heft 1 2011.

15. Urban Cleve: „A Technology Ready for Today“. 21st Century

Science & Technology; 2010.

16. Urban Cleve, Klaus Knizia, Kurt Kugeler: “The Technology of

High Temperature Reactors”. ICAPP-Congress Nice 2011.

17. Urban Cleve: “Die Technologie des Hochtemperaturreaktors

und nukleare Hochtemperaturtechnik zur Erzeugung flüssiger

Brennstoffe, von Wasserstoff und elektrischer Energie“. Atw


18. Urban Cleve: „The Technology of High Temperature Reactors

and Production of Nuclear Heat“. University of Cracow,


19. Urban Cleve: “Nuclear High Temperature Power Station with

Pebble Bed Reactor”. KTG Dresden, 24. März 2012.

20. Urban Cleve: “Breeding of 232Uranium using 232Thorium

with a Pebble Bed Reactor”.


Dr.-Ing. Urban Cleve

Ex. CTO/HA-Leiter Technik

of BBC/Krupp Reaktorbau GmbH,


Hohenfriedbergerstr. 4

44141 Dortmund, Germany


Environment and Safety

Design Principles for Nuclear and Operational Safety of HTR NPPs – a Review ı Urban Cleve

atw Vol. 65 (2020) | Issue 2 ı February


Probabilistic Analysis of Loss of Offsite

Power (LOOP) Accident in Bushehr

VVER-1000/V446 Nuclear Power Plant

Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

The aim of this study is to present the level-1 of Probabilistic Safety Assessment (PSA) analysis of the Loss of Offsite

Power (LOOP) in Bushehr VVER-1000/V446 Nuclear Power Plant (NPP) using the Hands-On Integrated Reliability

Evaluations (SAPHIRE) software. PSA is a very suitable method for determining scenarios of accidents and estimating

the risk of a power plant. LOOP is one of the beyond design basis accidents that can lead to melting of the reactor core

and dangerous environmental consequences. Therefore, the study of this accident and its consequences is very

important in nuclear power plant. For this purpose, the event tree and fault tree analysis of LOOP event is considered by

SAPHIRE code and compared with the Bushehr NPP Final Safety Analysis Reports (FSAR). The total frequency of LOOP

event that would lead to core damage is 3.40e-6 per year.


Nuclear power plants have a lot of

equipment, similar to other industrial

plants, whose performance depends

on electric power. Various equipment

for monitoring and controlling the

operation of units, equipment in

safety systems, ventilation systems,

pumps, lighting and other equipment

are examples of this.

Supply of offsite power plays major

role for safety of Nuclear Power Plants

(NPPs). Loss of Offsite Power (LOOP)

event is an important contributor to

the total residual risk at NPPs. The

availability of Alternating Current

(AC) electrical power to NPPs is thus

essential for safe operations and

accident recovery [3]. When the plant

loses offsite power (connections to the

external grid), the LOOP event occurs.

In this case, on-site power can be

provide by emergency diesel generators.

The LOOP is a transient accident.

After this accident, the reactor's scram

is required and core melting occurs

when the emergency electrical supply

system fails to supply the power of the

safety systems.

PSA is a very suitable method for

determining scenarios of accidents

and estimating the risk of a power

plant. LOOP plays an important role

in melting the reactor core and

its complications, the probabilistic

analysis of this event is very necessary.

Few studies have investigated the

Probabilistic analysis on LOOP in

different NPPs. Cepin considered

Assessment of Loss of Offsite Power

Initiating Event Frequency [1]. The

loss of offsite power frequency is

considered and the results compared

to the generic results. Jiao and et al

investigated Analysis of Loss of Offsite

Power Events at China’s Nuclear

Power Plants [5]. The analysis in this

paper would provide the increasing in

reliability of the offsite power system.

Statistical Analysis of Loss of Offsite

Power Events is performed by

Volkanovski and et al [8]. In this

study, the LOOP frequencies obtained

for the French and German nuclear

power plants during critical operation.

The reliability of offsite power of

nuclear power plants in evolving

power systems investigated by

Henneaux and et al [4]. In this

investigation the Factors affecting on

LOOP frequency were identified.

Faghihi and et al considered the

Level-1 probability safety assessment

of the Iranian heavy water reactor

using SAPHIRE software [2]. In part

of this study, LOOP event and its role

in core damage is investigated. An

Approach to Estimate SBO Risks in

Multi-unit Nuclear Power Plants with

a Shared Alternate AC Power Source is

performed by Jung and et al [6].

They developed a suitable method to

evaluate accurately the amounts of

risks, core damage frequencies and

site risks, resulting from a station

blackout event.

This paper presents results of the

level-1 of PSA analysis for a LOOP

scenario in a Bushehr-1 VVER-1000

Nuclear Power Plant (BNPP). The

initiating event (IE) with complete

loss of AC power, belongs to the typical

beyond design basis accidents (BDBA)

for which the time of plant survivability

without severe fuel damage

depends solely on built-in safety


For PSA analysis of LOOP, it should

be noted that the initiating event and

relative event tree must be determined,

and subsequently, the failure

analysis of the safety systems is done

by fault tree analysis. The event tree

and fault tree analysis of LOOP event

is considered by SAPHIRE code [7].

The fault trees determine the top

events occurrence probability by

determining the minimal cut sets of

basic events for top events. The

probability of fault tree is applied to

calculate the probability of sequences

of event tree. These sequences could

be determined the frequencies of core

damage states (CDS) and core

successful states (CSS). LOOP event

data are extracted from the Bushehr

NPP Final Safety Analysis Reports

(FSAR) [3]. The SAPHIRE code results

compared with FSAR results.

The estimation of total core

damage frequency (CDF) value was

performed with using mean values of

IE frequency, mean values of the

reliability indices for elements, mean

values of the common cause failures

(CCF) model parameters and mean

value of operator error probabilities.

Methods and Materials

Bushehr-1 VVER-1000 is a pressurized

water reactor (PWR) with a gross

electric output of 1000 MW. The unit

has four circulation loops, each

including a main circulation pump

and a horizontal steam generator. The

pressurizer is connected to one of the

main circulation loops. Some BNPP-1

Safety Systems are included:

1. Reactor Protection System, (RPS)

2. Turbine Stop Valves, (TSV)

3. BRU-K and BRU-A ⇔ (FASD-A and


4. Emergency Feed Water System,


5. Main Steam Isolation Valves


6. ECCS HP and LP, (TH)

7. Accumulators, (YT)

Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February

8. Pressurizer Safety Valves (PSV),


Also some BNPP-1 Safety Support

Systems are included:

1. Nuclear Component Cooling

System, (TF)

2. Secured Closed Cooling Water

System, (VJ)

3. Service Water System, (VE)

4. Emergency Diesel Generators,


5. Heating, Ventilation and Air Conditioning;

The LOOP event represents

approximately more than 26 % of the

Core Damage (CD) in Bushehr-1

VVER-1000 reactor. In order to

increase the reliability of the auxiliary

power supply system and the

emergency supply system, transmission

lines with different voltages

of grid are commonly used. Two grids

of 400 kilovolt (kV) (main grid) and

230 kV (auxiliary grid) and also 10 kV

buses of the normal power supply

system are used in Bushehr Power


Loss of offsite power is an event

linked with the loss of the power

supply of 10 kV buses from the on-site

normal operation sources and out-site

sources (400 and 230 kV of grid)

being external relative to the NPP. A

dependent failure of the system of the

normal heat removal through the

turbine condensers is a result of LOOP.

After the voltage in 10 kV (BA, BB,

BC, BD) buses has been lost, the safety

system buses (BU, BV, BW, BX) are

disconnected from them, the system

diesel generator (DG) are switched

on, stepped start-up automatic equipment

comes to actuate and safety

system services are connected.

LOOP results in [3]:

p Reactor coolant pump (RCP) shutdown

and reduction of the coolant

flow rate through the reactor;

p Actuation of the reactor pro tection

system and closing of the turbine

stop valves;

p Opening of BRU-A (Fast-acting

steam dump valves with discharge

to atmosphere (FASD-A)).

After the above-mentioned functions

have been performed, the operator

realizes the reactor plant cool down

through the secondary circuit using

BRU-A and brings the reactor plant

into the cold shutdown state. When

the LOOP event occurs, the reactor

must be scrammed, main and emergency

feed water supply have

to provide for steam generators (SG)

and discharge the steam to the

atmosphere, The cooling circuit

pressure must be adjusted through the

opening and closing of the discharge

and safety valves.

For achieve cold shutdown, the

following safety functions must be

performed [3]:

p Actuation of emergency pro tection

(EP) and reactor power reduction

down to the residual heat release

level (function A);

p Provision of main steam collector

(MSC) tightness (function T);

p Restriction of the pressure increase

in the secondary circuit (function


p Provision of the SG steam line

tightness after the actuation of the

steam generator steam releasing

valves (SRD) (function C4);

p Bringing of reactor plant into the

cold shutdown state (function CS).

There are safety systems for safety

functions that are required for

achieving safe mode. The safety

functions and safety systems and

their characteristics are presented in

Table 1.


Safety Functions Safety Systems Success Criteria

Description Code Code Description

Bringing reactor to subcritical state

and keeping it in this condition in the

entire range of operating parameters

A RPS Emergency protection system Insertion into the core of required number of CPS CRs

(control and protection system control rods)

Ensuring MSC leak-tightness T TSV



Secondary circuit pressure increase

limitation (SGs are not isolated

from MSC)

Bringing reactor plant to cold shutdown

condition (SG are not isolated

from MSC)

Heat removal from core via secondary

circuit within 24 hours over opened

circuit (SGs are not isolated from MSC)

Ensuring steam lines tightness in

section that non isolated from SG after

actuating of the SRD



Turbine stop valves (TSVs)

Turbine control valves (TCV)

Main steam valves (MSVs)

Fast-acting valves for steam dump

to atmosphere (FASD-A)

SG safety valves (SGSVs)

CS CDSS Fast-acting valves for steam dump

to atmosphere (FASD-A)

Emergency feed water system


Pressurizer safety valves (SVP)

Additional boron injection system


Low pressure emergency core

cooling system (TH10...40)

Planned cooldown line (PCL)

HO’’ HRSO Fast-acting valves for steam dump

to atmosphere (FASD-A)

Steam generators safety valves


Auxiliary feed water pumps (AFWP)

Emergency feed pumps (EFWP)

Makeup system for deaerators and

tanks of EFWP (UD)






MSC relief valves to atmosphere

for steam discharge into the

atmosphere (FASD-A)

Cut-off gate valves upstream FASD-A

Safety valves of SG (SG SV)

Closure of TSV or TCV or MSV in each

of four live steam lines

Opening of FASD-A or one SGSV in one SG

Operation of one FASD-A in cooldown mode and

water supply to one SG from one EFWP, when the

connection lines between RS tanks are opened


reducing of primary circuit pressure to 2 MPa

by opening of one SVP or conducts injection

into pressurizer from one channel of TW system


activation of one channel TH10...40

for operation along planned cooldown line

Operation of one FASD-A

in the mode P 2 =const or one SG SV


water supply to one SG from or AFWP and deaerator

makeup from the makeup system (UD) or

water supply to one SG from one EFWP and

makeup of tank in the operating train of EFWP

from the makeup system (UD), or

water supply to two SGs from one out of two EFWPs in

each subsystem of emergency feed water system, when

the connection lines between RS tanks are opened

Closing of FASD-A or cut-off gate valve,

closing SG SV (in case of FASD-A has failed to open)

in 4 (C4), 3 (C3), 2 (C2), 1 (C1) SGs

| Tab. 1.

Characteristics of safety functions and relative safety systems [3].

Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February


| Fig. 1.

Event tree for LOOP event.

Event tree could be constructed

according the safety functions and

safety systems, Figure 1. There are ten

states for accident sequences.

p Sequences 1 occurs under the

actuation of the reactor emergency

protection, provision of MSC

tightness, restriction of pressure

increase in the secondary circuit,

provision of SG steam line tightness

after the actuation of FASD-A

or SGSV and after the reactor plant

is brought into the cold shutdown

state (realization of functions A,

T, O’, C4, CS). The final state of

reactor plant is cold state.

p Sequences 2 occurs when the

reactor plant fails to be brought

into the cold shutdown state

( failure to perform CS function). In

this case heat removal from the

core through the secondary circuit

is performed through FASD-A or

SGSV during 24 hours with the

water being supplied to SG from

FWP or AFWP or EFWP (realization

of HO” function). The final

reactor plant state is hot state.

p Sequences 7 occurs in case of

non-closing (after opening) of

steam dump devices at all

4 SGs, which leads to core damage

due to full loss of heat removal via

secondary circuit.

p Sequences 8 occurs in case of

opening failure of all steam dump

devices FASD-A and SG SV, which

leads to full loss of heat removal via

secondary circuit.

p Sequences 10 occurs in case of

failure of reactor emergency protection

system, which is conservatively

considered as core


p Sequences 3 occurs in case of nonperformance

(by the operator) of

function of putting reactor plant

into cold state and failure of systems

for heat removal via secondary

circuit through open cycle.

p Sequences 4, 5, 6 occur in case of

closing failure of steam dump

Top events







| Tab. 2.

Fault tree analysis for top events occurrence probability.

(discharge) devices (SDD) at 1, 2,

or 3 SGs and a failure of water

supply to SGs from AFWP and

EFWP. Without working of heat

removal system, these sequences

lead to core damage state.

p Sequences 9 occurs at non-closing

of TSV, TCV and MSV, which leads

to steam lines leak in part isolated

from SG.

It should be noted that according to

the cut-off criteria (1,0E-8 1/year)

mentioned in FSAR, the development

of sequences 5, 6 and 9 have been

withdrawn. Also for sequences 3 and

4, it is assumed that the heat removal

is performed only through the secondary

circuit and the heat removal

through the primary circuit by bleed &

feed system is not considered.

Failure probability







Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February

Basic Event Code Description SAPHIRE Analysis FSAR Analysis

Probability RRR RIR

CCF-VE-2-ALL CCF VE11-41D001PMR 8.7E-01 1.23E+00 2.09E+03 1.29E-01

CCF-VJ-2-ALL CCF VJ11-41D001PMR 8.7E-01 1.23E+00 2.09E+03 1.29E-01

CCF-UF-8-ALL CCF UF40-70D002PMR 8.90E-02 1.20E+00 2.89E+03 3.90E-02

CCF-TL08-1-ALL CCF TL08D015 016,019,020 FAST 1.25E-02 1.10E+00 4.90E+02 3.43E-02

CCF-PS-02-ALL CCF of switches 11-14BU,V,W,X02A 1.7E-02 1.06E+00 3.6E+02 3.37E-02

HUM-BRU Actuation by operator of BRU coolibg down mode 1.02E-02 1.3E+00 8.58E+00 3.3E-02

12BV-BASIC Switchgear failure 4.36E-02 1.18E+00 6.96E+00 2.87E-02

11BU-BASIC Switchgear failure 4.36E-02 1.42E+00 6.78E+00 2.81E-02

DEP-UD HUM-UD-RS* HUM-UD-DEAR 1.02E-02 1.07E+00 2.15E+01 2.34E-02

CCF-UF-3-ALL CCF UF40-70D002PMS 1.25E-02 1.04E+00 2.66E+02 1.80E-02

CCF-LP-02-ALL CCF TH10-40D001PMR 1.83E-02 1.03E+00 5.10E+02 1.80E-02

CCF-DGS-ALL CCF GY10,11-40,41 DGS 3.32E-02 1.21E+00 2.65E+02 1.50E-02

CCF-UF-2-ALL CCF UF40-70D001COS 1.00E-02 1.11E+00 2.65E+02 1.44E-02

RA40S004VMC BZOK fails to close 3.97E-02 1.18E+00 5.41E+00 1.42E-02

MAINT-TF2 Unavailability due to maintenance TF20 1.85E-02 1.14E+00 2.60E+00 1.38E-02

MAINT-TL08-20 Unavailability due to maintenance TL08-20 1.85E-02 1.17E+00 3.02E+00 1.38E-02


MAINT-TL08D016 Unavailability due to maintenance TL08D016 1.85E-02 1.17E+00 3.01E+00 1.38E-02

MAINT-VJ2 Unavailability due to maintenance VJ21 1.85E-02 1.30E+00 2.36E+00 1.31E-02

MAINT-VE2 Unavailability due to maintenance 1.85E-02 1.30E+00 2.36E+00 1.31E-02

MAINT-UF50 Unavailability due to maintenance 1.85E-02 1.30E+00 3.27E+00 1.31E-02

CCF-UF-7-ALL CCF UF40-70D001COR 2.9E-02 1.36E+00 3.10E+03 1.28E-02

MAINT-VE1 Unavailability due to maintenance 1.85E-02 1.30E+00 2.36E+00 1.20E-02

MAINT-VJ1 Unavailability due to maintenance VJ11 1.85E-02 1.30E+00 2.36E+00 1.20E-02

MAINT-UF40 Unavailability due to maintenance 1.85E-02 1.30E+00 3.27E+00 1.20E-02

CCF-UF-1-ALL CCF UF42-72S002VMR 9.72E-02 1.07E+00 2.80E+03 1.15E-02

CCF-UF-4-ALL CCF UF42-72S001VMR 9.72E-02 1.07E+00 2.80E+03 1.15E-02

CCF-UF-5-ALL CCF UF42-72S003VMR 9.72E-02 1.07E+00 2.80E+03 1.15E-02

CCF-EHRS-01-ALL CCF of SG SV to open 1.01E-02 1.03E+00 1.06E+02 1.12E-02

CCF-NHRS-19-ALL CCF of RL62-92S001 VMO 6.52E-02 1.22E+00 4.18E+01 1.08E-02

MAINT-TL08-10 Unavailability due to maintenance TL08-10 1.85E-02 1.17E+00 3.01E+00 1.02E-02

MAINT-TL08D015 Unavailability due to maintenance TL08D015 1.85E-02 1.17E+00 3.01E+00 1.02E-02

MAINT-TF1 Unavailability due to maintenance TF10 1.85E-02 1.14E+00 2.59E+00 1.02E-02

CCF-EHRS-03-ALL CCF RA10-40 S003 to open 7.48E-03 1.05E+00 3.21E+01 9.90E-03

RA40S006VMC MOV fails to close 6.02E-03 1.08E+00 2.01E+01 9.70E-03

TH10D001PMR Pump fails to run 1.8E-03 1.20E+00 4.41E+00 9.69E-03

CCF-PS-01-ALL CCF 11-14EA 15-45 2.56E-03 1.71E+00 3.19E+02 8.05E-03

CCF-TL08-4-ALL CCF TL08D015 016,019,020 FAR 9.2E-03 1.11E+00 6.01E+02 7.72E-03

CCF-VE-3-ALL CCF VB96-99N001 6.48E-03 1.11E+00 3.15E+03 7.58E-03

CCF-LP-01-ALL CCF TH10-40D001PMS 7.9E-03 1.12E+00 5.01E+02 6.87E-03

CCF-VJ-1-ALL CCF VJ11-41D001PMS 1.25E-03 1.01E+00 1.91E+02 6.75E-03

TL08D016FAST Failure to start (1/d) 1.25E-03 1.13E+00 5.64E+00 6.75E-03

TL08D015FAST Failure to start (1/d) 1.25E-03 1.09E+00 5.64E+00 6.28E-03

14BX-BASIC Switchgear failure 4.36E-03 1.02E+00 1.50E+00 5.97E-03

CCF-VE-1-ALL CCF VE11-41D001PMS 1.25E-03 1.11E+00 1.45E+02 5.44E-03

RA40S003VRO FSDV-A fails to open 7.48E-03 1.44E+00 2.21E+00 5.24E-03

TL08D016FAR Failure to run 9.2E-03 1.11E+00 7.48E+00 5.09E-03

UF50D002PMR Pump fails to run 8.78E-03 1.21E+00 7.61E+00 4.92E-03

UF40D002PMR Pump fails to run 8.78E-03 1.11E+00 7.50E+00 4.81E-03

TL08D015FAR Failure to run 9.2E-03 1.41E+00 5.81E+00 4.71E-03

Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February

Basic Event Code Description SAPHIRE Analysis FSAR Analysis


| Tab. 3.

Importance analysis for basic events.

Probability RRR RIR

VJ21D001PMR Pump fails to run 8.78E-03 1.17E+00 6.11E+00 4.68E-03

VE21D001PMR Pump fails to run 8.78E-03 1.17E+00 6.11E+00 4.68E-03

VJ11D001PMR Pump fails to run 8.78E-03 1.11E+00 7.01E+00 4.58E-03

VE11D001PMR Pump fails to run 8.78E-03 1.11E+00 7.01E+00 4.58E-03

CCF group code Description SAPHIRE Analysis


Probability RRR RIR

FSAR Analysis

CCF-VE-2 CCF group VE11-41D001PMR 5.26E-01 1.10E+00 1.41E+04 1.47E-01

CCF-VJ-2 CCF group VJ11-41D001PMR 5.26E-01 1.10E+00 1.41E+04 1.47E-01

CCF-UF-8 CCF group UF40-70D002PMR 1.59E-02 1.01E+00 1.45E+04 5.90E-02

CCF-TL08-1 CCF group TL08D015 016,019,020 FAST 6.32E-02 1.01E+00 2.99E+03 5.35E-02

CCF-PS-02 CCF group of switches 11-14BU,V,W,X02A 1.52E-02 1.18E+00 8.28E+02 3.70E-02

CCF-UF-3 CCF group UF40-70D002PMS 8.16E-02 1.51E+00 7.11E+02 2.14E-02

CCF-LP-02 CCF group TH10-40D001PMR 3.31E-02 1.01E+00 3.87E+03 4.97E-02

CCF-TL08-4 CCF group TL08D015 016,019,020 FAR 1.45E-02 1.15E+00 2.10E+3 2.24E-02

CCF-EHRS-03 CCF group RA10-40 S003 (open) 4.27E-02 1.14E+00 1.21E+02 2.11E-02

CCF-UF-7 CCF group UF40-70D001COR 5.26E-02 1.71E+00 4.16E+03 1.87E-02

CCF-DGS CCF group GY10,11-40,41 DGS 3.1E-02 1.25E+00 1.59E+03 1.86E-02

CCF-UF-2 CCF group UF40-70D001COS 6.53E-02 1.22E+00 2.47E+03 1.70E-02

CCF-PS-01 CCF group 11-14EA 15-45 3.37E-02 1.61E+00 3.11E+03 1.57E-02

CCF-EHRS-05 CCF group RA10-40 S004 (BZOK closure) 4.17E-02 1.33E+00 2.99E+01 1.55E-02

CCF-EHRS-01 CCF group of SG SV (opening) 5.77E-02 1.61E+00 6.15E+02 1.18E-02

CCF-VJ-1 CCF group VJ11-41D001PMS 3.06E-02 1.21E+00 3.41E+03 1.18E-02

CCF-NHRS-19 CCF group RL62-92S001 VMO 3.73E-02 1.09E+00 3.54E+01 1.08E-02

CCF-EHRS-18 CCF group RA10-40S006 (valve closure) 1.38E-02 1.91E+00 2.40E+02 1.07E-02

CCF-VE-1 CCF group VE11-41D001PMS 1.38E-02 1.19E+00 2.19E+03 1.01E-02

CCF-LP-01 CCF group TH10-40D001PMS 5.15E-03 1.22E+00 3.10E+03 9.72E-03

CCF-VE-3 CCF group VB96-99N001 1.58E-03 1.71E+00 4.01E+03 5.37E-03

CCF-LP-08 CCF group TH10-40S007VMO 7.77E-03 1.31E+00 2.11E+03 6.50E-03

CCF-LP-09 CCF group TH10-40S013VMO 3.42E-03 1.55E+00 1.02E+03 6.50E-03

CCF-TF-11 CCF group TF10-40D001PMR 1.59E-03 1.06E+00 8.64E+01 5.39E-03

CCF-TF-09 CCF group TF60S001-004VMC 1.38E-03 1.17E+00 2.02E+03 5.06E-03

CCF-EHRS-02 CCF group RA10-40S001,S002 (closure) 2.63E-3 1.12E+00 1.41E+02 4.62E-3

CCF-PS-05 CCF group of switches 11-14BU,V,W,X03A 1.52E-03 1.10E+00 2.98E+01 4.39E-03

CCF-TF-01 CCF group TF10-40D001PMS 1.44E-03 1.08E+00 7.98E+01 3.82E-03

CCF-NHRS-08 CCF group RR12-22D001 PMR 2.45E-03 1.44E+00 3.40E+01 3.24E-03

CCF-UV31-1 CCF group UV31-34D009FAS 2.04E-03 1.72E+00 1.97E+01 1.66E-03

CCF-EHRS-07 CCF group RS12-42D001PMR 3.32E-03 1.18E+00 2.11E+01 1.64E-03

CCF-EHRS-09 CCF group RS17-47D001PMR 3.35E-03 1.48E+00 4.12E+01 1.64E-03

CCF-EHRS-15 CCF group RS12-42S005VCO 4.63E-03 1.01E+00 3.15E+01 1.46E-03

CCF-LP-21 CCF group TH90S005,006VMO 5.09E-03 1.32E+00 2.14E+00 1.45E-03

Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February

CCF group code Description SAPHIRE Analysis


Probability RRR RIR

FSAR Analysis

CCF-LP-20 CCF group TH90S001,002VMO 5.09E-03 1.34E+00 2.19E+00 1.45E-03

CCF-TL08-2 CCF group TL08D014 023,021,026 FAS 2.04E-03 1.29E+00 2.94E+01 1.36E-03

CCF-NHRS-12 CCF group RR12-22S004VMO 3.92E-03 1.08E+00 1.95E+01 1.34E-03

CCF-NHRS-07 CCF group RR12-22D001 PMS 1.44E-04 1.00E+00 1.29E+01 9.14E-04

CCF-LP-16 CCF group TH11,12-41,42S02VCO 4.63E-04 1.12E+00 2.14E+01 8.58E-04

CCF-LP-17 CCF group TH11,12-41,42S03VCO 4.63E-04 1.12E+00 2.14E+01 8.58E-04

CCF-YP-2 CCF group YP21-23S007VSO(VMO) 2.14E-04 1.01E+00 1.88E+01 7.87E-04

CCF-NHRS-09 CCF group RR12-22S001 VCO 5.79E-04 1.44E+00 2.84E+01 7.83E-04

CCF-TF-05 CCF group TF10-40S011VMO 3.13E-04 1.00E+00 3.14E+01 7.74E-04

CCF-YP-1 CCF group YP21-23S006VSO(VMO) 1.92E-04 1.02E+00 1.26E+01 7.12E-04

CCF-UV31-2 CCF group UV31-34D009FAR 1.45E-04 1.81E+00 3.21E+01 6.89E-04

CCF-TJ-2 CCF group TH10-40S005VCO 3.89E-04 1.48E+00 2.07E+03 6.82E-04

CCF-NHRS-11 CCF group RR12-22S003VMO 3.92E-04 1.85E+00 2.75E+01 5.24E-04


CCF-EHRS-06 CCF group RS12-42D001PMS 8.57E-04 1.11E+00 3.21E+01 4.52E-04

CCF-EHRS-08 CCF group RS17-47D001PMS 8.57E-04 1.11E+00 3.21E+01 4.52E-04

CCF-TL08-3 CCF group TL08D017 024,018,025 FAST 6.32E-04 1.05E+00 2.31E+01 3.95E-04

CCF-TL08-5 CCF group TL08D014 023,021,026 FAR 1.45E-04 1.69E+00 2.82E+01 3.74E-04

CCF-NHRS-14 CCF group RR13-23S001VMC 2.42E-04 1.16E+00 2.38E+01 3.40E-04

CCF-EHRS-19 CCF group RA10-40S003(throttle) 1.87E-04 1.21E+00 3.54E+01 3.18E-04

CCF-NHRS-13 CCF group RR13-23S001VMO 3.92E-04 1.00E+00 1.15E+01 2.48E-04

CCF-NHRS-10 CCF group RR12-22S002VMO 3.92E-04 1.00E+00 1.15E+01 2.48E-04

CCF-UF-6 CCF group UF43-73S010VMO 3.13E-04 1.21E+00 3.09E+01 1.56E-04

CCF-EHRS-14 CCF group RS12-42S002VMO 3.13E-04 1.04E+00 3.47E+01 1.56E-04

CCF-LP-15 CCF group TH11,12-41,42S01VMO 3.42E-04 1.17E+00 8.21E+00 1.38E-04

CCF-EHRS-13 CCF group RS12-42S003VMO 2.65E-04 1.41E+00 2.79E+01 1.33E-04

CCF-MSV CCF group MSV 8.25E-04 1.14E+00 2.13E+00 1.22E-04

CCF-TL08-6 CCF group TL08D017 024,018,025 FAR 1.45E-04 1.00E+00 1.30E+01 1.14E-04

CCF-LP-32 CCF group TH10-40S010VMO 3.42E-05 1.00E+00 4.11E+00 8.57E-05

CCF-YP-5 CCF group YP21-23S001VFO 3.03E-05 1.00E+00 6.21E+01 8.14E-05

CCF-PS-07 CCF group of switches 11-14BU,V,W,X04A 1.52E-5 1.02E+00 1.59E+00 7.31E-5

CCF-TF-06 CCF group TF10-40S012VCO 3.89E-05 1.58E+00 7.54E+00 6.72E-05

CCF-TSV-2 CCF group of TSV,TCV 6.41E-05 1.11E+00 7.98E+00 4.69E-05

CCF-TSV-4 CCF group of TSV,TCV 6.41E-05 1.11E+00 7.98E+00 4.69E-05

CCF-TSV-1 CCF group of TSV,TCV 6.41E-05 1.11E+00 7.98E+00 4.69E-05

CCF-TSV-3 CCF group of TSV,TCV 6.41E-05 1.11E+00 7.98E+00 4.69E-05

CCF-UV21-07 CCF group UV22,23D002,UV21,24D010FAS 2.04E-05 1.09E+00 1.41E+00 3.07E-05

CCF-TF-12 CCF group TF21,31D001PMR 8.1E-05 1.22E+00 1.97E+00 2.05E-05

CCF-EHRS-11 CCF group RS12-42S001VCO 3.88E-05 1.84E+00 2.01E+01 1.66E-05

CCF-UV21-11 CCF group UV21-24D002FAR 1.39E-05 1.00E+00 1.03E+00 1.39E-05

CCF-LP-33 CCF group TH10-40S037VCO 3.89E-06 1.07E+00 2.05E+00 7.63E-06

CCF-UV21-08 CCF group UV21-24D001FAR 1.45E-07 1.34E+00 1.77E+00 8.44E-07

| Tab. 4.

Importance analysis for CCFs.

Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February

Safety function 10 kv

failure probability







| Tab. 5.

Failure probability of 10 kv Safety functions.

Core Damage States









Total CD





Frequency per year










[1] Cepin, M. (2014). Assessment of loss of offsite power

initiating event frequency, Proceedings of the 23 rd

international conference nuclear energy for new Europe,

Portorož, Slovenia.

[2] Faghihi, F., Ramezani, E., Yousefpour, F., Mirvakili, S.M.

(2008). The Level-1 probability safety assessment of the

Iranian heavy water reactor using SAPHIRE software.

Reliability Engineering and System Safety, 93, 1377–1409.

[3] FSAR of BNPP-1. (2003). Final Safety Analysis Report of

Bushehr Nuclear Power Plant, Ministry of Russian Federation

of Atomic Energy (Atomenergoproekt), Moscow.

[4] Henneaux, P., Labeau, P. E., Obama, J. M. (2016). Reliability

of offsite power of nuclear power plants in evolving power

systems, Conference: Congrès Lambda Mu 20 de Maîtrise

des Risques et de Sûreté de Fonctionnement, Saint Malo,

France, DOI: 10.4267/2042/61785.

[5] Jiao, F., Ding, S., Li, J., Zheng, Z., Zhang, Q., Xiao, Z., Zhou, J.

(2018). Analysis of Loss of Offsite Power Events at China’s

Nuclear Power Plants, Sustainability, 10, 2680.

[6] Jung, W.S., Yang, J.E., Ha, J. (2004). An Approach to Estimate

SBO Risks in Multi-unit Nuclear Power Plants with a Shared

Alternate AC Power Source. In: Spitzer C., Schmocker U., Dang

V.N. (EDS) Probabilistic Safety Assessment and Management,

Springer, London.

[7] Kvarfordt, K.J., Wood, S.T., Smith, C.L. (2006). Systems

Analysis Programs for Hands-On Integrated Reliability

Evaluations (SAPHIRE 7.25) Code Reference Manual: user

guide and input requirements.

[8] Volkanovski, A., Avila, A. B., Veira, M. P. (2016). Statistical

Analysis of Loss of Offsite Power Events, Science and

Technology of Nuclear Installations, Volume 2016, Article ID

7692659, 9 pages.

| Tab. 6.

Frequency of CDSs.

Results and discussion

Allocated event tree should be

constructed for achieving the final

CDF. Event tree could be developed

due to the safety functions and safety

systems, Figure 1. Evaluating the

frequency of occurrence of initiation

event and top events in event tree

calculated by appropriate fault trees.

The failure probability of top events

are evaluated by appropriate fault

trees, Table 2. The failure probability

of each top event must be evaluated by

using a logical combination of basic

events through logic gates. For this

purpose, the fault trees of all safety

systems are considered. The information

of basic events for fault tree

analysis entered in code. Also common

cause failures (CCFs) evaluated

by using alpha factor model. Importance

analysis for some basic events

and several CCFs are presented in

Table 3, 4 respectively (compared

with FSAR results).

Because of the 10 kV buses play an

important role in the LOOP accident

analysis, the failure probability of

their safety systems buses (BU, BV,

BW, BX) are also given in the Table 5.

Final CDSs and their cor responding

frequencies are presented in Table 6.

There are ten end states for sequences.

Two of end states lead to core successful

state and eight of end states lead

to core damage state. The highest

frequency of CDSs related to sequences

number 3. Total core damage

frequency considered by frequencies

of eight CDSs. Total CDF is 3.40E-06

per year. According FSAR calculation,

total CDF is 3.84E-06 per year. The

full event tree diagram is shown in

Figure 1.


LOOP plays a major role in BNPP core

damage (about 26 %), all safety

aspects of the reactor must be used

to prevent the occurrence of the

accident. In this paper, level-1 PSA

considered for LOOP event in BNPP.

As reviewed in this paper,

sequences number 3, 7, 8, 9 and 10

are very important. The highest

frequency of CDSs related to sequence

number 3. Total CDF for initiating

event LOOP is calculated 3.40E-06 per

year. The difference between the

value calculated in the FSAR and

the value obtained in this study is

because of the development of

sequences 5, 6 and 9 have been withdrawn.

Also for sequences 3 and 4,

it is assumed that the heat removal

is performed only through the

secondary circuit and the heat

removal through the primary circuit

by bleed & feed system is not considered.

Also CCF has a significant

effect on CDF. Neglecting the CCFs

would lead to misleading results.

In general, the results obtained in

this paper are well-matched with the

results of the FSAR. This study shows

that the probabilistic analysis of

beyond design basis accidents is



Mohsen Esfandiari

Gholamreza Jahanfarnia

Department of Nuclear


Science and Research Branch

Islamic Azad University

Tehran, Iran

Kamran Sepanloo

Ehsan Zarifi

Reactor and Nuclear Safety

Research School

Nuclear Science and Technology

Research Institute (NSTRI)

Tehran, Iran

Environment and Safety

Probabilistic Analysis of Loss of Offsite Power (LOOP) Accident in Bushehr VVER-1000/V446 Nuclear Power Plant ı Mohsen Esfandiari, Kamran Sepanloo, Gholamreza Jahanfarnia and Ehsan Zarifi

atw Vol. 65 (2020) | Issue 2 ı February

Experimental Study of Thermal Neutron

Reflection Coefficient for two-layered


Khurram Mehboob

In this research project, the thermal neutron reflection coefficient has been measured (albedo αth) for different

combinations of reflector materials for several thicknesses by using 3.0 Curie Americium Beryllium ( 241 Am- 9 Be) neutron

source and a BF3 detector. The maximum value of neutron reflection from paraffin has been measured as 0.734 ± 0.020

appropriate to the value 0.83 mentioned in the literature. The reflection of neutrons has been measured for two-layered

medium i.e. copper-aluminum, copper-wood, wood-paraffin, and paraffin-iron of various thickness in a horizontal

arrangement. MATLAB has been used for the analytical simulation by devolving pseudocode that solves the diffusion

equations in two different mediums. It has been observed that the reflection coefficient increases exponentially

by introducing a 2 nd layer, only if the 2 nd medium has less diffusion length and higher diffusion coefficient.

The experimental results have been found in concord with analytical results. Poisson distribution has been used

for uncertainties analysis.


Commonly neutron reflection is used

for the bulk analysis of chemical

sampling, neutron dosimetry, detection

of mines and underground

explosive, boron neutron capture

therapy, detection of moisture in

hydrogenous materials and enhancement

of multiplication factor in a

nuclear reactor. The neutron reflection

is (albedo αth) a quotative

measure of the effectiveness of the

nuclear reactor core. The neutron

reflection coefficient of different

materials has been used to reduce the

critical core size and fuel mass in

nuclear reactors [1]. The reflector is

characterized by its reflection coefficient.

The neutron reflection

coefficient is defined as the ratio of

back-scattered neutrons to the total

incident neutron fluence in a diffusing

medium [2]. A good reflector is

characterized by its high scattering

cross section and low absorption cross

section having high slowing-down

power with small atomic weight [3].

Reflection of neutrons depends on the

reflector composition and geometrical

configurations [4].

In recent years, the studies have

been carried out for the measurement

of neutron reflection from different

types of reflectors. S. Dawahra et al.

[1] have used beryllium, heavy water,

graphite and light water as to measure

the efficacy of these reflectors in a

10 MW reactor using MCNP4C code.

Whereas the reflection coefficients of

the neutron from single voided

reflectors and multilayered reflectors

have been measured experimentally

by Mirza et al. [5] and Mehboob et

al. [6] respectively. Both pieces of

research have reported the increase in

thermal neutron reflection with

increasing in reflector thickness.

However, recently Rubina et al. [7]

have experimentally and theoretically

studied the response of BF 3 detector

using three reflector materials i.e.

aluminum, wood, and Perspex. The

Monte Carlo base theoretical studies

have been carried out by developing a

computation code in MATLAB.

However, only a few studies have been

carried out to measure the reflection



| Fig. 1.

Block diagram of experimental and Detection Setup.

Research and Innovation

Experimental Study of Thermal Neutron Reflection Coefficient for two-layered Reflectors ı Khurram Mehboob

atw Vol. 65 (2020) | Issue 2 ı February


| Fig. 2.

Single layer thermal neutron reflection coefficient (albedo) (a) experimental (b) analytical Simulation.

coefficient in two layers reflectors. In

this work, the experimental and

theoretic study of reflection of the

neutron from the different combination

of reflector medium has been

studied using 241 Am- 9 Be neutron

source with 7.2 × 10 6 neutrons/

second neutron emission rate, neutron

ab sorber cadmium sheet, paraffin wax

as a neutron moderator and a BF 3

detector. During experimentation

first, each selected material has been

set to its saturation thickness then a

second material is added a second

layer to measure its effect on reflection

coefficient (albedo). Analytical simulation

has been carried out and compared

with the experimental results.

1 Materials and method

The BF 3 detector is cylindrical in

shape with the cylindrical outer

cathode and small diametral tungsten

wire. The cylindrical case is usually

made up of aluminum due to its less

neutron interaction correction. The

operating voltage of proportional BF 3

detector for gas multiplication is the

order of 100 V to 500 V. These type of

BF 3 detectors are limited to the

temperature up to 100 °C as pulse

height resolution decreases beyond

the room temperatures.

2.1 Experimental Setup and


A three Curie cylindrical 241 Am- 9 Be

neutron source with the neutron

emission rate of 7.22 × 10 6 neutrons/

second was placed in the 64 × 6 × 64

wooden container homogenously

filled with paraffin wax. The neutron

source was enclosed in a cylinder

placed in a container at the depth of

11 cm from the top level. In order to

approximate the thermal flux, a thick

layer of 7 cm of paraffin wax was

placed on the top of the container.

The detector was placed over the slap

within a supporting groove. The

reflectors were placed horizontally

at the top of the slab as shown in

Figure 1. A semi-cylindrical cadmium

sheet was placed on the half side of

the detector to make the detector

sensitive to thermal neutrons from its

other half side. The transmitted

reflected flux could be measured by

rotating the detector at the angle of

180°. The interaction of neutrons in

BF 3 detector is depicted in Equation 1.


The measuring electronics was set up

according to the NIM standard

as shown in Figure 1. Since a large

electric field is required for the gas

multiplication, therefore, the detector

was operated at about 1500 V using

external high-tension supply. The

electronic pulse from BF3 detector

passes through the preamplifier

which shapes the pulse and fed to

the amplifier to achieve a user-defined

gain. The unipolar pulse is then fed

to timing single-channel analyzer

(TSCA) where a logical signal is

received as an output. The discriminator

level was fixed In TSCA to

reduce the noise and false pulses.

Logical signals were recorded in the

counter/timing unit. A cathoderay

oscilloscope (CRO) and a personal

computer rebased multichannel analyzer

(MCA) were adjusted to be such

that the bipolar pulses were received

by the gateway to SCA.

Four different material e.g. wood,

copper, aluminum, and paraffin

wax of different thickness and different

combinations were used. These

materials have been selected due to

the typical materials used in neutron

shielding and for neutron reflection

[8]. The experimental setup was

arranged as shown in Figure 1 the

counts for all the reflectors for various

thicknesses in different combinations

were recorded through TSCA and

the corresponding spectrum was

col lected on the MCA such that in the

first set of observations, the cadmium

cover faced the neutron source, and in

the second set, it was reversed. The

reflection coefficient (Albedo) was

measured for various thicknesses

of reflectors. The uncertainties

in the experimental measurements

have been carried out by Poisson distribution.

The uncertainty in albedo is

given by Equation 2.


2 Result and discussions

First, the thermal neutron reflection

coefficient (albedo) paraffin, wood,

aluminum, and copper were measured

to its saturation value (Figure 2). The

saturation value of albedo for paraffin

wax, wood, copper and aluminum has

been found 0.734 ± 0.020, 0.699 ±

0.002, 0.12 ± 0.001 and 0.27 ± 0.001

respectively. A good com parison has

been seen in perinatal and analytical

simulated results. A heard wood slabs

have been used for in this experiment

whose composition is a mixture of

carbohydrates, cellulose, minerals,

and water. For analytical simulation,

Research and Innovation

Experimental Study of Thermal Neutron Reflection Coefficient for two-layered Reflectors ı Khurram Mehboob

atw Vol. 65 (2020) | Issue 2 ı February

| Fig. 3.

Effect of wood as a 2 nd reflector to copper on thermal neutron albedo.

| Fig. 4.

Effect of aluminum as a 2 nd reflector to copper on thermal neutron albedo.


| Fig. 5.

Effect of paraffin as a 2 nd reflector to wood on thermal neutron albedo.

| Fig. 6.

Effect of iron as a 2 nd reflector to paraffin on thermal neutron albedo.

the combination of the hardwood is

chosen as 50.2 % carbon, 6.2 %

hydrogen, 43.5 % oxygen, and 0.1 %

nitrogen [9]. The paraffin reflection

coefficient has been measured

0.73 ± 0.01 that is comparable to the

value listed in the literature (0.83)


The reflection coefficient (albedo)

for wood has been found 4.8 % less

than the paraffin. The albedo for

different reflectors first increased

exponentially then reached to the

saturation value. The maximum

reflection coefficient (albedo) for

monolithic wood has been measured

0.699 ± 0.003, which is comparable

able to analytical simulated value


The situation value for neutron

reflection coefficient (albedo) for

copper has been measured 0.12 ±

0.001, which is comparable to value

0.11 reported by Doty, D. R. [11].

Whereas the saturation value of the

reflection coefficient for aluminum

has been measured to 0.27 ± 0.001.

Since the copper has a higher cross

section (Σs = 0.6709 cm-1) compare

to aluminum (Σs = 0.08976 cm-1)

therefore the copper albedo curve

is little steeper as compared to

aluminum. The experimental and

analytical simulated results for

paraffin, wood, aluminum, and

copper are depicted in Figure 2.

In order to see the effect of the 2 nd

layer in neutron reflection coefficient

combinations of different reflectors

has been used. The 2 nd reflector has

been introduced after the saturation

thickness of the first reflector. The

effect of wood as a 2 nd reflector to

copper is depicted in Figure 3. As

predicted the reflection of neutron

increased abruptly as the wood is

added as a 2 nd reflector. The saturation

from copper has been received

at 5 cm of thickness. Addition of

wood as a 2 nd reflector at this point

showed an exponential increase in

reflection coefficient. this is because

of the less scattering correction of

copper and higher scattering crosssection

of wood. Similar behavior

has been seen in analytical simulated


Theoretically, the 2 nd layer with

higher reflection and diffusion

coefficient contributes in increasing

the reflection coefficient. Glasstone

and Edlund [12] derived the

thermal neutron albedo as a function

of 2D/L.

Similarly, aluminum as a 2 nd reflector

plays the same role when added

after cooper saturation thickness. An

increment with a slope of 1.0 × 10 -4

has been observed with aluminum

as a 2 nd reflector. A similar slope has

also been reported by Doty, D. R. [11]

in his experimental study with the

increase in aluminum thickness. The

effect of aluminum as a 2 nd reflector to

copper is depicted in Figure 4. The

experiment results have been found

inconsistent with the analytical

simulated results.

If the 2 nd reflector has nearly the

same reflection coefficient as that for

1 st reflector then no significant effect

has been seen to total reflection coefficient.

This effect has been observed by

introducing the paraffin as the 2 nd

reflector to wood. Since the wood

and paraffin nearly have the same

saturation reflection coefficients and

for both reflectors, the 2D/L value is

almost similar. Therefore, no significant

effect has been observed for

paraffin as the 2 nd reflector to wood.

The effect of paraffin as a 2 nd reflector

to wood is depicted in Figure 5.

Research and Innovation

Experimental Study of Thermal Neutron Reflection Coefficient for two-layered Reflectors ı Khurram Mehboob

atw Vol. 65 (2020) | Issue 2 ı February


Some perturbations have been

observed as an effect of iron as a 2 nd

reflector to paraffin. Since the ratio of

two times of diffusion coefficient to

diffusion length (2D/L) for iron is

greater but the reflection coefficient

is small compared to paraffin. Therefore,

the reflection from the ion is

small compared to the reflection

from paraffin. The experimental

and analytical simulated results

are depicted in Figure 6. Doty, D. R.

[11] has reported the saturated

albedo for iron is 0.4. whereas we

have found the saturated reflected

value of iron is 0.304 comparable

to 0.4.

3 Conclusion

In this work, the experimental and

theoretic study of reflection of the

neutron from the different combination

of reflector medium has been

studied using 241 Am- 9 Be neutron

source with 7.2 × 106 neutrons/

second neutron emission rate, neutron

absorber cadmium sheet, paraffin

wax as a neutron moderator and a BF 3

detector. Pseudocode has been

developed in MATLAB for analytical

simulation. For analytical simulation

cross-sectional and diffusion, lengths

have been taken from appendix II Table

II.3 of [13] and the National Physical

Laboratory [14].

First, the thermal neutron albedo

reflection coefficient for aluminum,

paraffin wax, copper has been

measured and compared with analytical

simulated results. Then the

effect of the 2 nd layer to the 1 st reflector

has been studied. The results

indicate that if the 2 nd reflector has a

higher reflection coefficient than the

first type of reflector then the reflection

of neutrons increased abruptly.

This has been seen by introducing

wood and aluminum as the 2 nd reflector

to copper (Figure 3, 4).

Similarly, when the 2 nd reflector

has the same or smaller reflection

coefficient compare to 1 st reflector

then no significant effect has been

seen. This effect has been observed

by introducing paraffin and iron as a

2 nd reflector to wood and paraffin

respectively (Figure 5, 6). A higher

amount of fluctuation and perturbation

has been observed when the

2 nd reflector was introduced. Poisson

distribution has been used for uncertainty


The results indicate that the 2 nd

reflector has a significant effect on the

total thermal neutron reflection

coefficient. The effect of 2 nd reflector

could be used to enhance shielding

configurations and improve the

compact shielding for reactors.


This project was funded by the

Deanship of Scientific Research

(DSR), King Abdulaziz University,

Jeddah, under grant No. (D-211-135-

1440). The authors, therefore, gratefully

acknowledge the DSR technical

and financial support.


[1] Dawahra, S., Khattab, Saba, G., Study the effects of different

reflector types on the neutronic parameters of the 10MW

MTR reactor using the MCNP4C code. Ann. Nucl. Energy,

2015; 85: 1115–1118.

[2] Stacey.M.W. Nuclear reactor Physics, 2nd Edition Willey-VGC

Veller GmbH & Co. KgaA, 2001: ISBN 978-3-527-40679-1.

[3] Albarhoum, M. Graphite reflecting characteristics and

shielding factors for Miniature Neutron Source Reflectors.

Ann. Nucl. Energy. 2011: 38, 14–20

[4] Csikai, J., Buczko, C.M.. The concept of the reflection crosssection

of thermal neutrons. Appl. Radiat. Isot. 1999; 50:


[5] Mirza, S.M., Tufail, M., Liaqat, M.R. Thermal neutron albedo

and diffusion parameter measurements for monolithic and

geometric voided reflectors. Radiat. Meas. 2006; 41: 89–94.

[6] Mehboob, K., Ahmed, R., Ali, M., Tabassum, U. Thermal

neutron albedo measurements for multilithic reflectors. Ann.

Nucl. Energy. 2013: 62, 1–7.

[7] Rubina, N. et al. Experimental and theoretical study of BF 3

detector response for thermal neutrons in reflecting materials,

Nucl. Eng. Tech. 2018; 50: 439-445.

[8] Neeley, G.W., Newell, D.L, Larson, S.L., et al., Reactivity Effects

of Moderator and Reflector Materials on a Finite Plutonium

System, SAIC, 2004, Rev. I, US NRC-Public Documents.

[9] Ragland, K. W., Aerts, D. J. Properties of wood for combustion

analysis. Bioresource. Technol. 1991; 37: 161-168.

[10] Kogan, A.M., et al. The reflection of neutrons of various energies

by paraffin and by water. Atomnaya Energ. 1959.

[11] Doty, D. R. An absolute measurement of thermal neutron albedo

for several materials. U.S. Naval civil engineering laboratory,

1965: Y-F008-08-05-201, DASA-11.026.

[12] Glasstone, S., Edlund, M.C , The Elements of Nuclear Reactor

Theory. D, Van Nostrand Corporation, New York. 1952.

[13] Lamarsh, J. R. Introduction to Nuclear Engineering. 3rd Edition,

Prentice Hall. 2001, ISBN: 0-201-82498-1,

[14] NPL, National Physical Laboratory. 4.7.3 Attenuation of fast

neutrons: neutron moderation and diffusion, 2012: URL:


Khurram Mehboob (Ph.D.)

Associate Professor

Department of Nuclear

Engineering, Faculty of


King AbdulAziz University (KAU),

P. O. Box 80204

Jeddah 21589 Saudi Arabia

Research and Innovation

Experimental Study of Thermal Neutron Reflection Coefficient for two-layered Reflectors ı Khurram Mehboob

atw Vol. 65 (2020) | Issue 2 ı February

Workshop on the “Safety of Extended

Dry Storage of Spent Nuclear Fuel” –

SEDS 2019

Florian Rowold, Klemens Hummelsheim and Maik Stuke

For the third time now, the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH held its workshop “Safety of

Extended Dry Storage of Spent Nuclear Fuel (SEDS)”. The workshop met its initial objectives and expectations from

2017 and now it provides an annual platform for national and international experts to exchange information and

discuss recent technical and scientific progress and developments. Taking place from the 5 th to 7 th of June 2019 in

Garching, the workshop was attended by nearly 50 experts from 24 institutes of 7 countries. For Germany, the broad

range of experts was represented by universities and research organizations, technical support organizations, fuel

vendors, and the Federal Ministry for Economic Affairs and Energy. With 17 oral contributions the science-focused

agenda of the workshop reflected the broad diversity in current research projects. The topics comprised material

behavior of claddings and sealings, simulation approaches for thermal cask evaluations and thermo-mechanical fuel

rod performance. Furthermore, specific aspects were addressed such as non-destructive testing of casks or aging

management issues and regulatory aspects. On a positive note, it could be seen that the number of research projects

with an experimental focus has increased since the last year.



Peter Kaufholz from GRS in Garching

opened the workshop with a pre -

sentation entitled “Dry storage of

spent fuel and high-level waste in

Germany: Situation and Technical

Safety Aspects”, where he talked

about the scientific issues connected

to the condition of long-term stored

spent fuel with high burn-up. Technical

challenges have been identified

with rising interests on the extended

dry storage in the recent past. Matter

of interest are e.g. the drying conditions,

the hydrogen terminal solid

solubility, fission gas release, pin

pressure and cladding strain. The

proof of cladding integrity is not only

important for the storage itself but

especially for transport and conditioning

afterwards. Science and

engineering need to focus on the

reduction of the existing uncertainties

in the prediction of degradation

phenomena in extended dry storage

of spent nuclear fuel.

Karim Ben Ouaghrem from the

French technical support organization

Institut de Radioprotection et de Sûreté

Nucléaire (IRSN) presented a “Dry

storage overview and IRSN studies”.

Upon request from the French government,

IRSN published a report on

existing concepts of spent fuel storage

in France and worldwide. Considering

the characteristics of different fuel

types and storage concepts (wet or dry,

on-site or centralized), the assets and

limiting factors of dry storage were addressed.

Currently, IRSN is conducting

a study on safety issues raised by the

assessments of the package design

safety report of the dual-purpose casks

(DPC) and by the preparations of the

DPC for transport. To guarantee the

safety of transport after storage, the

topics that need to be evaluated are

the impact of material aging (e.g. cladding,

neutron resin), characterization

of monitored parameters during

storage (e.g. lid interspace pressure,

temperature) and the controls performed

before shipment (e.g. corrosion,

screw tightening check).

Timur Kandemir from the new

operator of the storage facilities for

spent fuel and high-level waste in

Germany, the Bundesgesellschaft für

Zwischenlagerung (BGZ), gave an overview

on the “Aging Management at

German Interim Storage Facilities

for Spent Fuel and High-Level

Waste”. The guidelines for a periodic

safety review (PSR) and an aging management

for spent fuel storage facilities

were introduced in 2014 after a twoyear

pilot phase. The periodic safety

review is an integral verification of the

facility safety status at regular intervals

of ten years, whereas the aging

management includes continuous control

of aging effects during storage operation.

The outcomes and findings

from the aging management are being

incorporated into the PSR, whereas

the aging management measures are

reviewed in the PSR and adapted if

necessary. The aging management

measures are limited to accessible cask

areas, safety relevant systems, components

and buildings. A graduated

approach in accordance to the protection

goal relevance of the systems and

com ponents is applied within the

aging management.

Andreja Peršič from the Slovenian

Nuclear Safety Administration (SNSA)

reported about the “Regulatory

Aspects Regarding New Dry Storage

of Spent Nuclear Fuel at the Krško

NPP”. To prevent severe accidents and

mitigate their consequences, the Krško

nuclear power plant (NPP) assessed

the options to reduce the risks associated

with spent fuel which is currently

stored in the spent fuel pool. The new

dry storage facility at the Krško NPP

will have a capacity of 2.600 spent fuel

assemblies in 70 casks of the Holtec HI-

STORM MPC design. It is designed for

a minimum operation of 60 years and

the construction shall begin in 2021.

The licensing process is challenging for

the operator as well as for the regulator.

Even though the storage technology is

proven, site specific conditions and

regulatory requirements make the

process unique. Aging management

already had to be considered in the

design phase of the licensing process

and the aging management program is

one of the important preconditions for

operation license issuing. A surveillance

program is required as well as an

environmental and seismic qualification

of systems, structures and components.

Furthermore, a systematic

approach to evaluate operating experience

is mandatory.

Aaron W. Colldeweih from the Paul

Scherrer Institute (PSI) in Switzerland

presented some details from a running

PhD in his talk “Impact of hydrogen

on fuel cladding properties and

example of Delayed Hydride

Cracking”. He started with a brief

overview of the research work on the

behavior of nuclear fuel claddings

under the influence of hydrogen. In

this respect, thermo-mechanical testing

is performed on hydrogen diffusion,

precipitation and hydride reorientation,

creep and fracture toughness.

Post-test examinations comprise

classical analytical methods like metallography,

but also Focused Ion Beam

(FIB), Scanning Electron Microscopy

(SEM) including Back Scattered Electron

detection (BSE) and Electron


Workshop on the “Safety of Extended Dry Storage of Spent Nuclear Fuel” – SEDS 2019 ı Florian Rowold, Klemens Hummelsheim and Maik Stuke

atw Vol. 65 (2020) | Issue 2 ı February



Backscatter Diffraction (EBSD) as well

as neutron radio graphy. Finite-

Element- Modelling is used to simulate

new geometries and test conditions.

Regarding the delayed hydride cracking

(DHC) investigations, differently

shaped Zircaloy-2 cladding tubes with

and without initial axial and radial

cracks are prepared and undergo the

described testing methods. The goal of

the work is to clarify the role of cladding

toughness for the DHC behavior.

Elmar W. Schweitzer from

Framatome GmbH, Germany, gave a

lecture on the “End-of-Reactor-Life

State of Spent Nuclear Fuel as Major

Input for Long Term Dry Storage

Fuel Integrity Assessment” from a

vendor’s point of view. Framatome as a

manufacturer of nuclear power plants

and has been delivering fuel assemblies

for operation of the plants. The behavior

of nuclear fuel under irradiation

up to end-of-life (EOL) is a prerequisite

for evaluating the additional damage

permissible during the dry storage

period. Limitation of temperature and

hoop stress by the present design criteria

is the best way to circumvent any

issues arising from long-term storage

of used fuel. Nevertheless, an exact

knowledge of the EOL state of the fuel

rods is necessary in order to assess

effects related to hoop stress and cladding

strain. Also, parts from the fuel

assembly structure, e.g. guide tubes,

spacer grids, water channels, fuel

channels etc. start to raise interest,

since these structures are important for

a safe repacking of the spent fuel from

the storage and transport cask into a

disposal cask. Mechanical properties of

irradiated cladding and fuel assembly

components (fast neutron fluence, corrosion

state) are necessary for transport

evaluation of the spent fuel.

The presentation of Dimitri

Papaioannou from the European

Commission Joint Research Centre

(JRC) in Germany was titled “Experimental

Studies on the Mechanical

Stability of Spent Nuclear Fuel

Rods”. He presented recent experimental

results from the spent fuel studies

at the JRC in Karlsruhe on safety

issues associated to handling and

transportation of nuclear fuel rods. In

the experiments, a pressurized rod

segment has been subject to dynamic

impact and quasi-static three-pointbending

tests. The devices are installed

in a hot cell. The rod segment stemmed

from a PWR fuel rod with burn up

67 GWd/tHM. A high-speed camera

was used to record the impact test and

thereby to determine the deflection

and absorbed energy. In the threepoint-bending

tests, the load, pressure

and displacement were recorded and

plotted. Post-test examinations were

carried out to characterize the released

mass upon rupturing in both experiments.

The final goal of these investigations

is to determine criteria and

conditions governing the response of

spent fuel rods to an external mechanical

load in accident scenarios.

Uwe Zencker from the Bundesanstalt

für Materialforschung und -prüfung

(BAM), Germany, gave a talk about

“Brittle failure of spent fuel claddings

during long-term dry interim

storage”. The current research project

BRUZL, which translates to fracturemechanical

analysis of spent fuel claddings

during long-term dry interim

storage, has the general aim to develop

risk assessment methods for potential

brittle failure under mechanical loads

after extended dry storage. The project

foresees ring compression tests with

unirradiated cladding samples with

representative hydride distribution.

Additional finite-element-analysis of

the ring compression tests will include

fracture-mechanical calcu lations, allowing

failure analysis and the identification

of failure criteria dependent on

hydride distribution (density, orientation,

and size), properties of cladding

material, mechanical load, and temperature.

The project is funded by the

Federal Ministry for Economic Affairs

and Energy (BMWi).

Another new research project was

introduced by Benedict Bongartz from

the University of Hannover, Germany.

He gave a presentation on the

“ Investigation of the temporal

rearrangement behavior of zirconium

hydride precipitates in interim

and final storage”. Within this work,

the specific experimental equipment

and the required process technology is

set up to load cladding tubes with hydrogen

contents of up to 500 wppm.

After the cladding tubes have been

loaded with hydrogen, a combination

of cooling and mecha nical stress application

is planned in order to recreate

and investigate the reorientation of the

hydrides in a laboratory environment.

The hydride precipitation in the zirconium

cladding will be investigated

with classical materials science investigations

such as metallography,

scanning and transmission electron

microscopy and X-ray diffraction. Additionally,

new investigation methods

such as X-ray microscopy are envisaged

to obtain new three- dimensional geometric

data about the precipitates.

In his talk, Marc Péridis from GRS

gave an update on his work about

“Temperature fields in a loaded

spent fuel cask”. The temperature is a

key parameter during dry storage since

it governs most of the claddings aging

mechanisms. As both, high and low

temperatures are relevant for different

effects, conservative models or a limited

consideration only on the hottest

fuel zone are insufficient for safety

studies. Considerably more, it is necessary

to carry out best-estimate calculations.

A generic detailed cask model,

inspired by the GNS CASTOR® V/19,

was set up and used to calculate the

temperature propagation from the inventories

to the cask body with

COBRA-SFS. The comparison of the

results with similar models in


good agreement. Within the recent

work, ParaView was introduced as a

graphic interface to visualize the

COBRA-SFS results. In the future, the

COBRA-SFS model is intended to be

used for transient calculations. This

will enable the user to describe the

temperature evolution during the

drying process, which has an important

impact on the material properties.

In the second contribution about

thermal modeling, Marta Galbán

Barahona from ENUSA, Spain, re ported

about her work progress with the

presentation entitled “Analysis in

Spent Nuclear Fuel Cask Using

COBRA-SFS”. In comparison to the

GRS work, ENUSA used the COBRA-

SFS code to simulate a storage cask of

the TN-24P type. The results obtained

for the helium filled TN-24P cask were

compared to measured temperature

data. There was a particularly good

agreement in the center of the fuel

assembly, where the maximum temperature

is located. In the peripheral

assemblies, the maximum differences

in temperature values were approximately

15 °C. Recently implemented

post-process scripts allowed a simpler

evaluation of the data with graphics

and colored maps. As a result of the

scripts, parameters such as helium flux

could be analyzed, where an unusual

flux distribution was found. Sensitivity

studies have been performed to analyze

the impact on the tem peratures. It

was found, that the impact of the specific

flux distribution was negligible.

Francisco Feria from CIEMAT, Spain,

gave a talk about the “Progress on the

modeling of in-clad hydrogen behavior

within FRAPCON-xt”. FRAPCONxt

in its base version is a fuel performance

code, which has been extended

to simulate fuel rods under dry storage

conditions. The code has been further

developed to model the inclad hydride

radial reorientation as a continuation

of the modelling derived on hydrogen

migration/precipitation. Moreover, an

uncertainty quantification method

has been adapted to predict the

best estimate plus the corresponding


Workshop on the “Safety of Extended Dry Storage of Spent Nuclear Fuel” – SEDS 2019 ı Florian Rowold, Klemens Hummelsheim and Maik Stuke

atw Vol. 65 (2020) | Issue 2 ı February

uncertainty. Recent results allowed the

conclusion that realistic scenarios prevent

the formation of radial hydrides,

whereas giving credit to limiting conditions

would not allow ruling out this

degrading mechanism. Depending on

experimental data made available,

further work is foreseen to validate the

modelling with more representative

data based on irradiated material and

to derive a technological limit concerning

the cladding embrittlement due to

radial hydrides formed.

To further address and support

thermo-mechanical activities, Felix

Bold from GRS held a presentation

called “Proposal of a Benchmark

Describing the Thermo-Mechanical

behavior During Dry Storage”,

wherein he invited all interested

parties, who are willing to improve

their modeling experience and to

share their knowledge about the

extended storage of spent nuclear fuel.

The goal of this benchmark is the

prediction and the code-to-code

comparison of the thermo-mechanical

parameters such as cladding temperature,

hoop stress and strain as well

as the hydrogen and hydride behavior

during the storage period. For the

starting conditions it is planned to use

fuel rod data or output of a fuel performance

code capable of simulating the

fuel rod state at the end of operation.

This will provide rod and pellet geometry,

corrosion, internal gas state

and the initial hydrogen load. The

transient conditions will include the

change of environment from water

cooling in the spent fuel pool to helium

atmosphere in the cask as well as the

temperature changes during reactor

shut down and the drying process. The

results of the benchmark will be published

and presented in 2020 on the

4th GRS workshop.

With his talk about the “Long-term

evaluation of sealing systems for

radioactive waste packages”,

Matthias Jaunich provided a round-up

of another important research area addressed

by the BAM. The work

performed for many years now, aims at

understanding the long-term behavior

of the sealing systems during possible

extended storage and sub sequent transportation

scenarios. It comprises accelerated

aging tests on metallic and elastomeric

seals and covers experimental

investigations to get a database on the

component/ material behavior. Based

on the results, appropriate analytical

descriptions and models were developed.

For the metal seals, a linear

logarithmic correlation and an extrapolation

of the remaining seal force and

useable resilience for up to 100 years

seems possible, but the question about

the confidence range has yet to be

answered. The elastomer seals exhibited

hardness increase and sealing

force decrease during the aging test.

Deriving from the tests, the researchers

were able to present an approach for a

lifetime prediction of the seals.

With his presentation about

“Radio nuclides present at inner

PWR fuel rod segment Zircaloy cladding

surfaces in the context of safety

of extended dry storage of spent

nuclear fuel”, Michel Herm from the

Karlsruhe Institute of Technology (KIT)

shifted the focus to another interesting

issue. Beside the often-discussed hydrogen

effects, the fuel rod cladding

could also be affected by various other

processes during reactor operation and

beyond. Precipitates of fission or activation

products, e.g. Cs, I, Te, and Cl,

present at the fuel-cladding interface,

possibly exhibit corrosive properties

and thus affect the integrity of the

cladding. Therefore, irradiated pressurized

water reactor UO 2 and MOX

fuel rod segments were prepared and

examined. The composition of agglomerates

found on the inner surfaces of

the plenum area and in fuel-cladding

interaction layers were analyzed by

means of SEM-EDS/WDS, XPS, and

synchrotron radiation- based techniques.

In addition, the present radionuclide

inventory was compared

to calculated values using a MCNP/

CINDER approach.

In a second contribution from the

KIT, Mirco Große reported on the

“ Experimental Simulation of Long-

Term Dry Storage in the QUENCH

Facility of KIT – Availabilities and

Plans”. The QUENCH facility at KIT is

dedicated for tests simulating design

basis and beyond design basis accidents

in light water reactors on a fuel

rod bundle scale. However, this test

bundle can also be used for long-term

cooling experiments simulating dry

storage conditions. A description of the

facility and reports about the planned

experiments within the framework of

the collaborative research project

SPIZWURZ was given. The project

investigates the behavior of cladding

materials under typical long-term

storage conditions. The experimental

bundle consists of 21 to 31 fuel rod

simulators with Zircaloy-4, ZIRLO and

Dx/D4 Duplex claddings. The rods are

electrically heated and can be pressurized

separately. The inner pressure

will be up to 5 MPa. The test will start

with a maximum cladding temperature

of 500 °C at the hottest bundle position

and the temperature will be reduced by

1 K/day during a period of 8 month.

The axial and radial hydrogen distribution

will be measured post-test by

neutron imaging methods. Metallographic

investigations will be used to

determine the change in the hydride


Michael Wagner from the Technical

University of Dresden, Germany, closed

the workshop with the last talk about

the “Investigations on potential

methods for the long-term monitoring

of the state of fuel elements in

dry storage casks: recent results”.

Four non-invasive measuring methods

were assessed regarding their suitability

for the condition monitoring of

the cask inventory by means of

simulations and experiments. For this

purpose, damage scenarios of the cask

inventory were assumed in a CASTOR

V/19. The identified scenarios based

on investigations on damage mechanisms.

The simulations and experiments

showed that the measurement

of neutron and gamma radiation fields

and muon imaging have the greatest

potential as monitoring methods.

These will also be further investigated

in a follow-up project. In principle,

the acoustic methods have a high

informative value, but the transfer of

experimental results to real con ditions

is difficult. Thermography showed a

low practicality as a monitoring method

due to its limited expressiveness.

The 2019 SEDS workshop showed

that the topic of extended storage of

spent nuclear fuel with all its different

aspects is continuing to draw a large

interest in the scientific landscape. In

fact, the efforts in terms of research

projects and collaborations increased

in the recent past. Especially for Germany

and European countries, where

very high burnup and mixed oxide

fuels were used and dry storage in

casks is a preferred option for the

spent fuel, this is a positive sign, since

not all knowledge gaps are answered

yet and require further work. The annual

workshop “Safety of Extended

Dry Storage of Spent Nuclear Fuel”

established itself as a place to address

those knowledge gaps and exchange

information in a broad community on

a very scientific level. This year the

4th workshop will be held again as a

three-day event at GRS in Garching

during the first week of June 2020.


Florian Rowold

Klemens Hummelsheim

Maik Stuke

Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH

Bereich Stilllegung und Entsorgung

Abteilung Stilllegung und



Boltzmannstr. 14,

85748 Garching b. München




Workshop on the “Safety of Extended Dry Storage of Spent Nuclear Fuel” – SEDS 2019 ı Florian Rowold, Klemens Hummelsheim and Maik Stuke

atw Vol. 65 (2020) | Issue 2 ı February



Einladung zum Vortrag


KTG Inside


für den Inhalt:

Die Autoren.


Natalija Cobanov,


Gesellschaft e. V.


Robert-Koch-Platz 4

10115 Berlin

T: +49 30 498555-50

F: +49 30 498555-51



Stand der weltweiten Entwicklung der Kernenergie

von Dr. Ludger Mohrbach

am Donnerstag, den 19. März 2020 um 17:30 Uhr,

PreussenElektra GmbH, Tresckowstraße 5,


Die einzige heute verfügbare Option zur Lösung des

weltweiten Energieversorgungsproblems zu bisher

gewohnten Kosten, bei vergleichsweise geringen CO 2 -

Emissionen und einer gesicherten Energieversorgung ist

neben der nur regional weiter ausbaubaren Großwasserkraft

die Kernspaltungsenergie, die technologisch derzeit

weltweit von 31 Ländern genutzt wird, bei fünf weiteren

Newcomern durch Neubau erschlossen wird und in vier

weiteren in der Planung ist.

Die Kernbrennstoffe Uran und Thorium sind für viele

Jahrhunderte ausreichend vorhanden und bei Nutzung in

fortgeschrittenen Reaktoren für viele Tausend Jahre. Die

Entsorgung in tiefen geologischen Erdschichten war und

ist aufgrund der kleinen Rückstandsmassen technisch und

wirtschaftlich realisierbar.

Historisch und ganzheitlich betrachtet ist die Kernenergie

ein sehr sicherer Energieträger. Bezogen auf die

MWh erzeugte Energie gibt es keine Stromerzeugungsart,

bei der weniger Menschen zu Schaden kommen.

Gleichwohl ist das weltweit einzige Land, das heute

einen echten Ausstieg betreibt, Deutschland. So hat z. B.

Frankreich, von der deutschen Öffentlichkeit kaum

reflektiert, kürzlich eine Laufzeitverlängerung um

( zunächst) zehn Jahre beschlossen.

Nukleare Sektorkopplung über die Erzeugung von

synthetischen Brennstoffen, den Wärmemarkt und

insbesondere auch Meerwasserentsalzung kann das

Klimaproblem zu heutigen Bereitstellungskosten lösen.

Der Anteil der Kernenergie von derzeit ca. 11 % an der

weltweiten Strom- und damit von ca. 6 % an der Primärenergieerzeugung

sollte somit so schnell wie möglich

gesteigert werden. Welche technischen Optionen hierfür

zur Verfügung stehen, insbesondere auch in Bezug auf

weiterentwickelte Kernreaktoren der Generation IV, wird

im Vortrag vorgestellt.

Im Anschluss an den etwa einstündigen Vortrag wird es

ausreichend Gelegenheit für weitere Diskussionen geben.

Interessierte KTG-Mitglieder sowie Freunde und

Bekannte sind herzlich eingeladen.

Mit freundlichen Grüßen

Dr.-Ing. Hans-Georg Willschütz

Sprecher KTG-Sektion NORD

Thomas Fröhmel

Stellv. Sprecher der KTG-Sektion NORD

PS: Wir bitten um eine namentliche Anmeldung

der Teilnehmer bis zum 3. März 2020 an

Dr.-Ing. Ludger Mohrbach studierte Maschinenbau mit der Vertiefungsrichtung

Reaktortechnik an der Ruhr- Universität Bochum und promovierte

dort 1989 zur Thermohydraulik des Schnellen Brüters. Bis 2019 war

er als persönlicher Referent der Geschäftsführung, Referent und Leiter der

Abteilung „Kerntechnik“ beim inter nationalen Technischen Verband der

Kraftwerksbetreiber VGB in Essen tätig.

Herzlichen Glückwunsch!

Wenn Sie künftig eine

Erwähnung Ihres

Geburtstages in der

atw wünschen, teilen

Sie dies bitte der KTG-

Geschäftsstelle mit.

Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag

und wünscht ihnen weiterhin alles Gute!

März 2020

50 Jahre | 1970

10. Dr. Stefan Nießen, Erlangen

13. Dipl.-Ìng. (FH) Michael Remshardt,


55 Jahre | 1965

22. Karsten Müller Kleinmachnow

60 Jahre | 1960

23. Peter Reimann, Lingen

65 Jahre | 1955

06. Prof. Dr. Peter-Wilhelm Phlippen,


70 Jahre | 1950

23. Hans-Dieter Schmidt, Dortmund

75 Jahre | 1945

04. Dr. Bernd Hofmann,


11. Dr. Ulrich Krugmann, Erlangen

11. Joachim Lange, Burgdorf

15. Bernhard Brand, Forchheim

20. Dipl.-Ing. mult. Herbert Niederhausen,


76 Jahre | 1944

02. Dr. Peter Schnur, Hannover

10. Prof. Dr. Reinhard Odoj, Hürtgenwald

11. Hamid Mehrfar, Dormitz

77 Jahre | 1943

16. Dipl.-Ing. Jochen Heinecke, Kürten

20. Dipl.-Ing. Jörg Brauns, Hanau

80 Jahre | 1940

01. Dipl.-Ing. Wolfgang Stumpf, Moers

03. Dipl.-Ing. Eberhard Schomer, Erlangen

18. Dipl.-Ing. Friedhelm Hülsmann,


81 Jahre |1939

01. Prof. Dr. Günter Höhlein,


82 Jahre | 1938

14. Dr. Peter Paetz, Bergisch Gladbach

KTG Inside

atw Vol. 65 (2020) | Issue 2 ı February

84 Jahre | 1936

19. Dr. Hermann Hinsch, Hannover

85 Jahre | 1935

02. Dipl.-Ing. Joachim Hospe, München

20. Dr. Jürgen Ahlf, Neustadt in Holstein

87 Jahre | 1933

30. Dipl.-Phys. Dieter Pleuger, Kiedrich

89 Jahre | 1931

17. Dipl.-Ing. Hans Waldmann, Schwabach

90 Jahre | 1930

25. Dr. Hans-Ulrich Borgstedt, Karlsruhe




First small modular reactors

open a new world

of applications

(wna) The two barge-mounted reactors

onboard Akademik Lomonosov

have started providing electricity to

the coastal town of Pevek in Russia.

This marks the official start of operations

for the world’s first small modular

reactors and makes today a historic

one for the global nuclear industry.

World Nuclear Association Director

General Agneta Rising warmly welcomed

the news, “It is fantastic to see

this innovative new floating nuclear

power plant begin operating just in

time for the winter celebrations.

It will provide much needed clean

electricity and heat to this remote

arctic community.”

In celebration of the accomplishment

and in preparation for the New

Year, Christmas tree lights were

switched on using electricity from the

reactors. The plant will be linked up to

the local district heating network

sometime in 2020. While the two

reactors with a combined output of

64 megawatts represent only a small

addition to global nuclear generating

capacity, they mark an important

evolution in nuclear technology.

Large reactors and SMRs are not

so much competing technologies as

complementary partners. Large reactors

produce huge amounts of reliable,

low-cost, low-carbon electricity

while SMRs expand the range of

useful nuclear applications.

Rising continued, “There are

around 50 advanced nuclear technologies

under development at the

moment with many countries pursuing

novel designs and seeking to use

nuclear technology for new and

exciting applications. This may be the

world’s first SMR, but many more will

soon follow. These smaller reactors

are well-suited for supplying electricity

to hard-to-reach regions as well as

serving smaller grids and industrial

centres. We are at the dawn of a new

era in nuclear technology.”

| (20211012)

ROSATOM’s first of a kind

floating power unit connects

to isolated electricity grid

in Pevek, Russia’s Far East

(rosatom) The floating power unit

(FPU) Akademik Lomonosov has been

connected to the grid, generating electricity

for the first time in the isolated

Chaun-Bilibino network in Pevek,

Chukotka, Russia’s Far East. This

happened after the Russian regulator

Rostekhnadzor issued an operations

permit, as well as permission to connect

to the northern electricity grid

maintained by Chukotenergo JSC.

Pevek residents marked this

symbolic day by turning on the fairy

lights on the town’s Christmas tree.

Rosatom’s Director General Alexey

Likhachev said: “After its connection

to the grid, Akademik Lomonosov

becomes the world’s first nuclear

power plant based on SMR-class technology

to generate electricity. This is a

remarkable milestone for both the

Russian and the world’s nuclear

energy industry. This is also a major

step in establishing Pevek as the new

energy capital of the region”.

The project has been welcomed by

scientists, nuclear energy experts and

environmentalists across the world.

Kirsty Gogan, Head of Energy for

Humanity, an NGO (London), said:

“For hard-to-reach regions, with a

climate that is simultaneously too

harsh to support the use of renewable

energies and too fragile to continue its

heavy dependence on fossil fuels,

small nuclear, including floating

plants, is the only answer. Akademik

Lomonosov is the first step towards

demonstrating its potential for decarbonisation

of the Arctic and beyond”.

Connecting the FPU generators to

the network was carried out after

parameter synchronisation with the

coastal network. This happened after

the completed construction of the onshore

facilities, ensuring the transfer of

electricity from the FPUs to Chukotka’s

high voltage networks. A vast amount

of work was also carried out on constructing

the heat supply networks.

Connecting the FNPP to Pevek’s heat

networks will be completed in 2020.

| ROSATOM’s first of a kind floating power unit connects to isolated electricity

grid in Pevek, Russia’s Far East, Credit: Rosatom

Once the FNPP will begin commercial

operations, it will make it Russia’s

11th nuclear power plant. It will also

mark the first time in Russia’s nuclear

energy history that two nuclear power

plants (the Akademik Lomonosov

FNPP and the Bilibino NPP) operate in

the same region.

Notes to the editor:

The nuclear FPU Akademik Lomonosov

is equipped with two KLT-40C reactor

systems (each with a capacity of 35

MW) similar to those used on icebreakers.

It is designed by Rosatom to work as

a part of the Floating Nuclear Thermal

Power Plant (FNPP). The vessel is

144 metres long and 30 metres wide,

and has a displacement of 21,000

tonnes. Akademik Lomonosov – the

first ship of this kind – was named for

18th century Russian scientist Mikhail

Lomonosov. Aka demik Lomo nosov is a

pilot project and a ‘working prototype’

for a future fleet of floating nuclear

power plants and on-shore installations

based on Russian-made SMRs. The

small power units will be available for

deployment to hard- to-reach areas of

Russia’s North and Far-East, as well as

for export.

SMR-based nuclear power plants

(featuring reactors of less than

300 MWe each), floating and on-shore,

are designed to made it possible to supply

electricity to hard-to-reach areas,

smaller grids and off-grid installations.

These small nuclear reactors can operate

non-stop without the need for refuelling

for three to five years, thereby

considerably reducing the cost of electricity

generation. Whilst variable


atw Vol. 65 (2020) | Issue 2 ı February



renewable energy installations such as

wind and solar for such areas require an

expensive a polluting diesel back-up or

an expensive energy storage, small nuclear

power plants ensure uninterrupted

electricity supply even for energy intensive

users. The reactors have the potential

to work particularly well in regions

with extended coastlines, power

supply shortages, and limited access to

electrical grids. The plant can be delivered

to any point along a coast and connected

to existing electrical grids.

| (20211018);


U.S.: Policymakers and energy

companies plan to reduce

carbon and know they’ll need


(nei) In mid-January 2020, the

leaders of the House Energy and Commerce

Committee released an overview

of the Climate Leadership and

Environmental Action for our Nation’s

(CLEAN) Future Act, a forthcoming

bill to put the United States on a path

to reduce carbon emissions.

At the heart of the bill is a requirement

for electricity providers to increase

the portion of their power that

comes from clean sources, including

nuclear energy, and to reach 100 percent

clean by 2050. The bill builds

upon the commitments that states

and companies have been making

to significantly reduce carbon emissions.

Prior to 2017, 28 states enacted

some form of legally binding requirement

to deploy clean electricity. A typical

state target would require 20 percent

of the state’s electricity to come

from clean sources. Only two of those

28 states adopted technology-inclusive

policies that would allow carbon-free

nuclear energy to meet the goal.

Now, states have clearly changed

their perspective. In the last three

years, 13 states have created or updated

their standards and they tend to be

much more ambitious and inclusive.

The majority of these call for 100 percent

clean electricity and the majority

are technology-neutral, which will

allow nuclear to be part of the

generating portfolio to meet these


Analysts at the think tank Third

Way created a tool that tracks who is

making commitments to reduce emissions

in the U.S. They have an online

calculator that allows you to see the

targets set by states, cities and companies

to reduce emissions. It paints an

interesting picture of how this landscape

has changed in recent years: it’s

not just state governments that are


The map that Third Way shows

makes it clear that utilities are charting

a path to carbon reductions. Twenty-eight

electric sector companies

have publicly put forward targets for

their generating portfolios. This is a

meaningful segment of the power sector

with companies including American

Electric Power Co., Duke Energy

Corp., DTE Energy, Xcel Energy Inc.,

Southern Co. and NRG Energy Inc.,

among others.

Their targets are ambitious. Almost

all call for something like 80 percent

carbon reductions or even 100 percent

clean electricity. With this ambition

comes a recognition that nuclear

needs to be among the tools available

to meet these goals. Of the 14 commitments

made in 2019, 12 were technology-inclusive.

This trend is very important. Carbon

reduction policies must be defined

to include nuclear energy as part

of the available solutions. Nuclear energy

makes up more than 55 percent

of carbon-free energy in the U.S.,

making it a key component of any plan

to reduce carbon emissions. Including

nuclear will also help to reduce costs

and maintain reliability as emissions

are reduced.

In 2020, we can expect to see a

great deal of attention on policy proposals

to reduce carbon emissions.

States and utilities have already begun

to map out where we need to go and

including nuclear as part of the solution

will help to get us there.



NEA: Nuclear and social

science nexus: challenges and

opportunities for speaking

across the disciplinary divide

(oecd-nea) The NEA organised a workshop

on the "Nuclear and Social

Science Nexus: Challenges and

Oppor tunities for Speaking Across the

Disciplinary Divide" on 12‐13 December

2019. The first‐of‐its‐kind event

brought together over 100 participants,

including social science and

humanities researchers, academic

nuclear engineers, practitioners and

policy makers. The participants

examined the current scope of

research in the social sciences with a

focus on nuclear energy, and identified

ways of transforming research

findings into recommendations for

practice. The two‐day workshop

aimed to build intellectual bridges

across the nuclear and social sciences,

as well as the academic and practitioner

divides. Selected papers from

the workshop will be published in a

forthcoming special issue of the

nuclear engineering journal Nuclear

Technology. Workshop participants

expressed a keen interest in developing

inter- and transdisciplinary

research collaborations and continuing

their dialogue beyond the workshop.

The NEA will work to identify

opportunities for such collaborations

in the coming months.

| (20211108);


Foratom: Just Transition

Mechanism must support all

low-carbon options

(foratom) FORATOM welcomes the

EU’s goal of providing financial support

to coal-dependent regions in

order to assist them in their decarbonisation

efforts. Indeed, the transition

to a low-carbon economy should not

come at the detriment to society.

Therefore, we fully support EU funds

being earmarked to help people transition

from jobs in carbon-intensive

sectors into low-carbon industries.

That being said, FORATOM regrets

the European Commission’s proposal

to exclude such funds being used for

nuclear power plants. Several reports

published over the last 18 months

(IPCC, IEA and even the Commission

itself) highlight that low-carbon

nuclear is an essential component of a

low-carbon economy. Actually, at the

end of last year, several Member

States made it clear that in order to

commit to the 2050 decarbonisation

targets then they must be allowed to

invest in nuclear power.

“The benefits of transitioning

workers from the coal into the nuclear

industry have already been demonstrated

in both France and the UK”,

states FORATOM Director General

Yves Desbazeille. “We therefore find it

hard to justify such a proposal by the

Commission. At the end of the day, the

EU should be focusing on helping

people in these regions to transition

into low-carbon industries. Limiting

the low-carbon sectors which will be

eligible for such funds will make

achieving our low-carbon targets

without leaving anyone behind a lot

more difficult – if not impossible”.


atw Vol. 65 (2020) | Issue 2 ı February

Operating Results October 2019

Plant name Country Nominal











Energy generated, gross


Month Year Since


Time availability


Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 745 687 447 6 453 520 268 108 728 100.00 96.98 99.88 96.13 100.30 96.14

OL2 Olkiluoto BWR FI 910 880 745 687 012 6 113 476 258 010 019 100.00 91.81 99.99 91.33 100.24 91.08

KCB Borssele PWR NL 512 484 745 374 368 5 512 786 167 234 474 99.12 85.37 99.12 85.30 98.11 82.20

KKB 1 Beznau 7) PWR CH 380 365 745 283 936 2 411 259 129 745 369 100.00 87.69 100.00 87.52 100.33 86.87

KKB 2 Beznau 7) PWR CH 380 365 745 283 005 2 386 203 136 736 610 100.00 86.71 100.00 86.53 100.05 85.95

KKG Gösgen 7) PWR CH 1060 1010 745 785 806 6 680 686 320 556 214 100.00 87.32 99.99 86.80 99.51 86.38

KKM Mühleberg BWR CH 390 373 745 284 160 2 778 890 130 183 205 100.00 100.00 99.93 99.76 97.80 97.66

CNT-I Trillo PWR ES 1066 1003 745 786 501 6 935 768 254 227 436 100.00 90.24 100.00 89.87 98.41 88.58

Dukovany B1 1) PWR CZ 500 473 548 261 151 2 923 685 115 153 179 73.56 82.33 70.72 81.81 70.11 80.14

Dukovany B2 PWR CZ 500 473 745 367 280 2 083 213 110 317 384 100.00 58.76 99.69 58.15 98.60 57.11

Dukovany B3 PWR CZ 500 473 745 356 858 3 029 376 109 527 417 100.00 85.54 100.00 85.18 95.80 83.04

Dukovany B4 PWR CZ 500 473 745 372 929 3 617 748 110 061 017 100.00 99.85 100.00 99.70 100.12 99.17

Temelin B1 PWR CZ 1080 1030 745 805 805 6 302 005 120 663 047 100.00 80.67 99.97 80.43 99.96 79.83

Temelin B2 PWR CZ 1080 1030 745 811 582 6 607 392 115 879 909 100.00 83.47 100.00 83.24 100.68 83.70

Doel 1 2) PWR BE 454 433 92 40 658 2 291 598 137 736 060 12.39 68.11 11.98 67.77 11.69 67.67

Doel 2 2) PWR BE 454 433 0 0 2 533 531 136 335 470 0 77.50 0 76.20 0 76.14

Doel 3 PWR BE 1056 1006 745 798 464 6 397 257 261 529 742 100.00 82.90 100.00 82.30 101.11 82.54

Doel 4 PWR BE 1084 1033 745 811 772 7 662 270 268 035 680 100.00 100.00 100.00 96.60 98.91 95.32

Tihange 1 PWR BE 1009 962 745 740 727 7 293 792 306 124 650 100.00 100.00 99.94 99.98 98.56 99.18

Tihange 2 3) PWR BE 1055 1008 128 131 910 2 286 338 256 938 268 17.18 31.32 17.09 30.70 16.94 29.92

Tihange 3 PWR BE 1089 1038 745 799 779 7 746 449 278 973 722 100.00 99.97 100.00 99.31 99.15 97.99



Plant name












Energy generated, gross


Time availability


Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Since Month Year Month Year Month Year


KBR Brokdorf DWR 1480 1410 745 986 722 8 378 370 358 946 180 100.00 86.83 93.96 81.51 89.20 77.28

KKE Emsland DWR 1406 1335 745 1 028 413 8 746 219 355 565 188 100.00 87.04 100.00 86.94 98.20 85.25

KWG Grohnde DWR 1430 1360 745 1 009 147 8 685 609 386 259 822 100.00 88.06 99.92 87.78 94.09 82.70

KRB C Gundremmingen SWR 1344 1288 745 999 296 8 419 961 339 361 715 100.00 86.98 100.00 86.41 99.34 85.42

KKI-2 Isar DWR 1485 1410 745 1 070 070 9 904 910 363 630 723 100.00 95.13 99.98 94.81 96.36 91.05

GKN-II Neckarwestheim DWR 1400 1310 745 1 028 300 8 376 010 338 202 844 100.00 92.83 100.00 84.61 98.85 82.12

KKP-2 Philippsburg 4) DWR 1468 1402 745 1 005 849 8 843 211 375 004 366 100.00 87.61 99.95 87.38 90.50 81.28

The European nuclear industry

currently sustains more than 1.1 million

jobs in the EU and generates more

than half a trillion euros in GDP according

to a study by Deloitte. Looking

ahead to 2050, the authors believe

that, on average, the industry would

support more than 1.3 million jobs annually

and generate €576 billion per

year in GDP. This shows that nuclear

offers benefits both in terms of decarbonising

the power sector and providing

European citizens with much

needed jobs.

| (20211028);


Russian NPPs set a new

record in terms of electric

power output

(rosatom) In 2019, the Russian nuclear

power plants (affiliate companies of

the Rosenergoatom Joint-Stock

Company) set a new electric power

output record – over 208.784 billion

kilowatt-hours, which means they

have grown their joint production and

exceeded their previous record of

2018 (204.275 billion kWh) by over

4.5 billion kWh.

The FAS assignment for 2019 has

been delivered at the rate of 103 %

with the planned production of

202.7 billion kWh.

The biggest contributions into the

new Company’s record were from the

Rostov (over 33.8 billion kWh), the

Kalinin (over 31 billion kWh), and the

Balakovo NPPs (over 30 billion kWh).

Thus, a share of nuclear power

plants in Russia’s energy mix has increased

up to 19.04% in 2019 (in

2018 this indicator was 18.7%). In

the United Energy Grid (UEG) of

Russia, without considering electricity

generation by Bilibino NPP which

operates in the isolated power system,

a generation share of nuclear power

plants has increased up to 19.3 %

(19.1 % in 2018).

| (20211041);

ASN issues a position statement

on the orientations of

the generic phase of the

fourth periodic safety reviews

of the 1300 MWe reactors

(asn) On 11 December 2019, ASN

issued a position statement on the

orientations of the generic phase of

the fourth periodic safety review of

EDF’s 1300 MWe nuclear reactors.

ASN considers that the general

objectives set by EDF for this review

are acceptable in principle. However,

it asks EDF to modify or supplement

these general objectives for this safety

review, to consider certain baseline

requirements for reassessment of the

safety of its facilities and to add study

topics to its review programme. The

requests made by ASN are to a large


Net-based values

(Czech and Swiss

nuclear power

plants gross-based)












Hereof traction supply


Incl. steam supply


New nominal

capacity since

January 2016


Data for the Leibstadt

(CH) NPP will

be published in a

further issue of atw

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB


atw Vol. 65 (2020) | Issue 2 ı February



extent based on those made in 2016

for the fourth periodic safety review of

the 900 MWe reactors.

In France, the operating lifetime of

a nuclear reactor is not defined in advance.

However, pursuant to Article L.

593-18 of the Environment Code, the

licensee of a basic nuclear installation

must conduct a periodic safety review

of its facility every ten years. The periodic

safety review must be able to verify

the facility’s compliance with the

rules that apply to it and to update the

assessment of the risks and drawbacks

it constitutes for public health and

safety and the protection of the environment,

while notably taking account

of the condition of the facility,

experience acquired during operation,

changing knowledge and the

rules applicable to similar facilities.

The review thus leads the licensee to

improve the safety level of the facility.

Following this review, ASN issues a

position statement on the conditions

for the continued operation of the


In 2017, EDF initiated the fourth

periodic safety review of its twenty

1300 MWe nuclear power reactors. As

with the previous periodic safety

reviews and in order to take advantage

of the standardised nature of its

reactors, EDF intends to carry out this

periodic safety review in two stages:

p a “generic” periodic review phase,

concerning subjects common to all

the 1300 MWe reactors. This

generic approach is a means of

pooling and sharing studies of

facility ageing control, obsolescence

and compliance, as well

as the safety reassessment and

design studies for any modifications

to the facilities;

p a “specific” periodic safety review

phase, concerning each individual

reactor and which is scheduled to

run from 2027 to 2035. This phase

addresses the particular characteristics

of the facility and its environment,

for example the level of natural

hazards to be considered and

the condition of the facility.

The “generic” periodic safety review

phase begins with a definition of the

objectives assigned to this periodic

safety review. In this respect, EDF

transmitted a “periodic safety review

guidance file” which specifies its objectives.

Following the generic studies

phase, ASN will also issue a position

statement on the adequacy of the

modifications planned by EDF.

For the particular purpose of the

1300 MWe reactors fourth periodic

safety review, which is aiming for

continued operation beyond 40 years,

ASN wished to promote broader participation

by the stakeholders as of the

generic phase objectives definition

stage. Thus ASN’s position was the

subject of a discussion meeting with

the stakeholders (members of the

HCTISN, the ANCCLI and CLIs, plus

qualified personalities) at the ASN

headquarters on 16 October 2019 and

a public consultation on the website

from 17 October to 17 November

2019. The comments collected more

specifically led ASN to ask EDF to produce

a summary at the end of the

generic periodic safety review phase,

presenting the safety differences that

will persist between the 1300 MWe

reactors and the Flamanville EPR

reactor, and to reformulate the request

concerning organisational and

human factors.

| (20211106);

Company News

Taiwan opts for GNS containers

(gns) During an international tender

procedure, GNS has been awarded a

contract by Taiwan Power Company

(TPC) for the development of containers

for the transport and interim

storage of intermediate and low-level

radioactive waste. Within the scope

of the upcoming national decommissioning

projects, this is the first

contract awarded internationally by

TPC after the decommissioning of

Chinshan Nuclear Power Plant had

been announced in 2019. The containers

are dedicated for metallic

waste from the dismantling of the

reactors and primary peripherals from

all Taiwanese nuclear power plants.

The order comprises the development

of a total of five different

| GNS design “SBoX®” (type B(U)) container (20210919).




container types (1x type B(U), 4x type

IP-2). The containers are based on

the proven GNS designs “SBoX®” (type

B(U)) and steel sheet containers (type


The scope of supply also includes

the complete handling and loading

equipment as well as the preliminary

plan for cutting the reactor and primary

peripherals. Additionally, the

order also comprises five prototypes,

which will be manufactured by domestic

partner companies in Taiwan,

training courses and the cold handling

at Chinshan NPP.

Edward H.C. Chang, Director of

Nuclear Backend Management Department

at TPC: “During the open tender

process GNS convinced Taipower with

their experienced packaging solutions

and their proven technology, which

are believed as reliable and efficient.

We expect that through this bilateral

cooperation, Taipower will achieve

the localization of container‘s mass

production in the future.”

Dr. Linus Bettermann, Head of

Sales Department Casks at GNS: “The

order from Taiwan proves the international

competitiveness of our container

systems. The decision of TPC

underlines the leading role of GNS as

a supplier of packaging for nuclear

waste, which occurs in large quantities

especially during nuclear power

plant decommissioning.”

| (20210919);

Framatome signs a cooperation

agreement with Japan

on the development of fast

neutron reactors

(framatome) Framatome has signed a

cooperation agreement in Tokyo with

the CEA and Japanese organizations

JAEA, MHI and MFBR on the development

of fast neutron reactors. This

agreement follows the agreement

established in 2014 for the ASTRID

program, through which a great many


The GNS SBoX ® is a container for interim

storage and final disposal of all kinds of

radioactive waste from nuclear facilities. It

consists of welded heavy-walled steel sheets.

With an empty weight of 16,500 kg, the

maximum payload is normally 8,500 kg. The

outer dimensions are 2,000 * 1,600 * 1,700 mm

(l * w * h).

The GNS SBoX ® is available with round or

rectangular lid systems.There are

connections for drying and filling facilities

integrated in the lid [3], which come with

separate closure lids [2]. For protection

against mechanical damages and ingress of

dust the lid of the GNS SBoX ® is additionally

covered with a protection plate [1].

The GNS SBoX ® can be delivered with an

integrated heating system, which enables

short drying times and a low surface

temperature during und after drying. This

reduces the overall drying cycle time



atw Vol. 65 (2020) | Issue 2 ı February

technical outcomes have been jointly

achieved and which has enabled close

collaborative ties to be established

between the parties.

The new agreement aims to further

research on high-stakes topics for this

reactor technology. Subjects of interest

include severe accidents, thermalhydraulics

and fuel behavior, justification

of material performance and

durability, under-sodium inspection

and instrumentation. This agreement

will contribute to maintain and to

develop the Framatome' skills and

expertise in the field of fast reactors.

| (20210920);


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











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

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


) 1 23.00





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






) 1




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


) 1

| Uranium spot market prices from 1980 to 2019 and from 2008 to 2019.

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












Jan. 2008

Jan. 2009

Jan. 2010

) 1

Jan. 2011

Jan. 2012

Jan. 2013

Jan. 2014

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020



Westinghouse wins ovation

i&c modernization project at

Kozloduy units 5&6

(westinghouse) Westinghouse Electric

Company announced that it has

signed a contract with Kozloduy

Nuclear Power Plant (NPP) in Bulgaria

to migrate the current Ovation

platform- based information and control

(I&C) systems at units 5&6 to its

latest standard, bringing even more

competitiveness and efficiency in the

way these plants are operating.

Kozloduy will migrate to the latest

Ovation platform, which will include

the integration of a Safety Parameter

Display System, Emergency Operator

Procedures (EOP) and partial modernization

of the Full-Scope Simulator.

“The digitalization and modernization

of the operating nuclear fleet

is a key part of our client’s long-term

operations and a strategic priority for

Westinghouse,” said Tarik Choho,

president of Westinghouse’s Europe,

Middle East and Africa (EMEA) Operating

Plant Services Business Unit.

“We are pleased to support Kozloduy

5&6 in their efforts to utilize the best

available technology and supply

cost-competitive and clean energy to

Bulgaria for decades to come.”

The Ovation platform is one of the

most advanced I&C platforms for the

energy sector and is widely used at

both operating and new nuclear

plants. As the supplier of the Ovation

platform to the nuclear industry,

Westinghouse has implemented

Ovation at the Kozloduy NPP for

more than 15 years and the platform

has proven to be safe, reliable and

very cost-efficient. Westinghouse is

committed to support Kozloduy units

5&6 in maintaining the Ovation

platform for at least another 30 years,

supporting Kozloduy’s plans to

operate units 5&6 at least until 2049.

| (20210921);







Jan. 2008

Jan. 2009

Jan. 2010

Jan. 2011

Jan. 2012

Market data

(All information is supplied without


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”. 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)].

Jan. 2013


Jan. 2014

Jan. 2015

Jan. 2016


p Uranium: 19.25–26.50

p Conversion: 4.50–6.75

p Separative work: 39.00–50.00


p Uranium: 21.75–29.20

p Conversion: 6.00–14.50

p Separative work: 34.00–42.00


January 2019

p Uranium: 28.70–29.10

p Conversion: 13.50–14.50

p Separative work: 41.00–44.00

February 2019

p Uranium: 27.50–29.25

p Conversion: 13.50–14.50

p Separative work: 42.00–45.00

March 2019

p Uranium: 24.85–28.25

p Conversion: 13.50–14.50

p Separative work: 43.00–46.00

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

| Separative work and conversion market price ranges from 2008 to 2019. The price range is shown.


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 2020









Jan. 2008

Jan. 2009

Jan. 2010

Jan. 2011

Jan. 2012

April 2019

p Uranium: 25.50–25.88

p Conversion: 15.00–17.00

p Separative work: 44.00–46.00

May 2019

p Uranium: 23.90–25.25

p Conversion: 17.00–18.00

p Separative work: 46.00–48.00

June 2019

p Uranium: 24.30–25.00

p Conversion: 17.00–18.00

p Separative work: 47.00–49.00

July 2019

p Uranium: 24.50–25.60

p Conversion: 18.00–19.00

p Separative work: 47.00–49.00

August 2019

p Uranium: 24.90–25.60

p Conversion: 19.00–20.00

p Separative work: 47.00–49.00

September 2019

p Uranium: 24.80–26.00

p Conversion: 20.00–21.00

p Separative work: 47.00–50.00

October 2019

p Uranium: 23.75–25.50

p Conversion: 21.00–22.00

p Separative work: 47.00–50.00

November 2019

p Uranium: 23.95–26.25

p Conversion: 22.00–23.00

p Separative work: 48.00–50.00

| Source: Energy Intelligence

Jan. 2013


Jan. 2014

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020


atw Vol. 65 (2020) | Issue 2 ı February



John Shepherd is a

freelance journalist

and communications



NBER working paper


IAEA director-general


Brookhaven National

Lab project


IEE analysis


Climate of Opinion Frowns on Germany

as Nuclear Exit Continues

Germany’s sad shuffle towards a nuclear exit has continued with the closure of another clean energy power station.

Unit 2 of Germany’s Philippsburg nuclear power plant was disconnected from the grid on 31 December, marking the

end of 35 years of operation.

Although planned, the closure came as economists

released a model of Germany’s electrical system to see

what would have happened if it had kept shuttered nuclear

plants running. According to economists at the US National

Bureau of Economic Research (NBER), keeping nuclear

plants online would have saved the lives of 1,100 people a

year who succumb to air pollution released by coal- burning

power plants.

The NBER working paper said lost nuclear electricity

production due to the phase-out was replaced primarily

by coal-fired production and net electricity imports. “The

social cost of this shift from nuclear to coal is approximately

$12 billion dollars per year.” More than 70 % of this cost

came from “increased mortality risk associated with exposure

to the local air pollution emitted when burning fossil

fuels”, the NBER paper said. Even the largest estimates of

the reduction in the costs associated with nuclear accident

risk and waste disposal due to the phase-out are far smaller

than 12 billion dollars.

If further evidence of Germany’s ill-judged nuclear exit

were needed, look no further than a separate report from

the Institute of Energy Economics (IEE) at the University of

Cologne, which concluded the country could “significantly”

miss its target of covering 65 % of gross electricity

con sumption with renewables by 2030. Analysis by an IEE

team calculated that gross electricity consumption could

rise to 748 terawatt hours (TWh) by 2030. At the same time,

electricity generation from renewables would rise to

345 TWh. “The share of renewable energies would thus be

only 46 %, instead of the targeted 65 %.

When will the anti-nuclear brigade face up to climate

reality? Thankfully, the International Atomic Energy

Agency (IAEA) has long since shaken off its reticence to say

anything that might be deemed as ‘promoting’ nuclear

power. The agency’s new director-general, Rafael Mariano

Grossi, used one of his first major speeches since taking office

to hammer home the fact that nuclear power is already

reducing carbon dioxide emissions by about two gigatonnes

annually. He said that was the equivalent of taking more

than 400 million cars off the world’s roads every year.

The IAEA chief, who was speaking at a side event during

the COP 25 UN Climate Change Conference in Madrid,

warned that while 30 countries currently use nuclear

power, if any major users were to halt nuclear energy

programmes overnight “this would have very serious

consequences for CO 2 emissions”.

And Grossi rightly pointed out that nuclear energy

should not been seen as being in competition with renewables.

“In order to achieve climate change goals and ensure

sufficient energy for the future, we need to make use of all

available sources of clean energy,” he said.

In contrast to Germany, the US nuclear industry has a

spring in its step for the new year – thanks to a pre-

Christmas vote by Congress that included $1.5 billion for

nuclear energy programmes in appropriations for the

2020 fiscal year. The nuclear cash boost represented a

12.5 % increase over the previous year.

In addition, Congress supported a seven-year reauthorisation

of the Export-Import Bank (the US’ official export

credit agency), which the country’s Nuclear Energy

Institute (NEI) said would help to level the playing field for

American companies competing against foreign stateowned


The reauthorisation marked what the NEI said was a

“welcome departure” from a series of short-term authorisations

since 2012 – which had made US nuclear suppliers

pursuing long-term projects particularly vulnerable to

perceptions that the Bank’s future was in doubt.

According to the NEI, more than 95 % of the world’s

nuclear construction projects are being built outside of the

US and, to compete, US suppliers must be able to offer

competitive financing to potential customers. “In international

nuclear energy markets, a competitive export

credit agency is a requirement to bid on virtually every

project,” the NEI said.

Indeed, the Trump administration can be credited with

offering increasingly positive signals to the benefit of the

domestic nuclear industry.

The US Department of Energy has selected Brookhaven

National Laboratory in New York State as the site for a

planned new research facility that will benefit the global

nuclear physics community. The Electron Ion Collider

(EIC), which will be designed and built over 10 years at an

estimated cost between $1.6 and $2.6 billion, will smash

electrons into protons and heavier atomic nuclei “in an

effort to penetrate the mysteries of the ‘strong force’ that

binds the atomic nucleus together.

In the UK, which will formally leave the European

Union on 31 January 2020, the future of investment in the

fading but much-needed nuclear park faces another

tumultuous year.

EDF has appointed Rothschild as financial advisers to

the Sizewell C project and the French energy giant said it is

“working on sales documents to be issued once we have

clear government policy on the detail of the funding

model”. EDF wants to start building the Sizewell C plant,

comprising two UK EPR nuclear reactor units, in 2022.

Meanwhile, a parliamentary report in uranium-rich Australia

said the federal government should consider a partial

lifting of the current moratorium on nuclear energy to allow

the deployment of new and emerging technologies.

This year is also expected to see a milestone development

in the United Arab Emirates, where the first of four

nuclear reactor units at the Barakah nuclear power plant is

said to be aiming to start up within months. The first of

Barakah’s units had been due to come online in late 2017,

but faced regulatory and related delays.

The start of electricity generation at Barakah will make

the UAE the first country in the region to deliver commercial

nuclear power – and others in the oil-producing region,

including Saudi Arabia, are keen to follow.

As the world’s petro giants gear up for a nuclearpowered

future, one can only hope nations still addicted to

fossil fuels take note.

Nuclear Today

Climate of Opinion Frowns on Germany as Nuclear Exit Continues ı John Shepherd

Kommunikation und

Training für Kerntechnik

International sicher agieren


English for the Nuclear Industry

Im internationalen Dialog ist Englisch die universelle Sprache. Dies gilt für Geschäfts beziehungen

im Allgemeinen ebenso wie für die Branche der Kerntechnik im Speziellen. In Deutschland gewinnen

der internationale Austausch und damit die englische Sprache eine noch größere Bedeutung,

insbesondere bedingt durch die auf das Jahr 2022 begrenzte Stromerzeugung aus Kernenergie.






Introductions and informal conversations

Preparing a meeting

Intercultural communication

PR in the nuclear industry

Den Teilnehmerinnen und Teilnehmern wird über eine praxisorientierte Didaktik und unter der

Verwendung „kerntechnischen Vokabulars“ das notwendige Know-how für den beruflichen Alltag

vermittelt. Dabei gilt es sprachlich bedingte Kommunikationsbarrieren mit internationalem Kollegium

und Kunden zu überwinden.


Diese 2-tägige Schulung richtet sich an Fach- und Führungskräfte, Projektverantwortliche

sowie Mitarbeiterinnen und Mitarbeiter aus allen Fachbereichen, bei denen Englisch für die

organisationsinterne und/oder externe Kommunikation von Bedeutung ist.

Maximale Teilnehmerzahl: 12 Personen


Angela Lloyd

Language and intercultural trainer (English native speaker)

Wir freuen uns auf Ihre Teilnahme!


2 Tage

1. bis 2. April 2020

Tag 1: 10:30 bis 17:30 Uhr

Tag 2: 09:00 bis 16:30 Uhr


Kaiserin Friedrich-Haus


Robert-Koch Platz 7

10115 Berlin


898,– € ı zzgl. 19 % USt.

Im Preis inbegriffen sind:

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inkl. Mittagessen



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5. – 6. Mai 2020

Estrel Convention Center


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