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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
option.
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
and
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 –
63
EDITORIAL
Editorial
EU “Green Deal” – Just Not with Nuclear Energy?

atw Vol. 65 (2020) | Issue 2 ı February
EU-„Green Deal“ – nur nicht mit der Kernenergie?
EDITORIAL 64
„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
werden.
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
und
p ein jährlich erwirtschaftetes BIP von mehr als 500 Milliarden
Euro.
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 –
Editorial
EU “Green Deal” – Just Not with Nuclear Energy?

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atw Vol. 65 (2020) | Issue 2 ı February
66
Issue 2 | 2020
February
CONTENTS
Contents
Editorial
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
Report
Workshop on the “Safety of Extended Dry Storage
of Spent Nuclear Fuel” – SEDS 2019 109
KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .112
News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .113
Cover:
Vogtle Unit 4 Containment Vessel
©2019 Georgia Power Company
Contents:
Unit 3 Low Pressure Turbine
©2019 Georgia Power Company
Nuclear Today
Climate of Opinion Frowns on Germany
as Nuclear Exit Continues 118
Imprint 92
G
E/G
= German
= English/German
Insert: AiNT – Aus- und Fortbildungsprogramm 2020
Contents

atw Vol. 65 (2020) | Issue 2 ı February
Feature
Major Trends in Energy Policy
and Nuclear Power
67
CONTENTS
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
Contents

atw Vol. 65 (2020) | Issue 2 ı February
68
INSIDE NUCLEAR WITH NUCNET
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
operations.
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
costly.
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.
Author
NucNet
The Independent Global Nuclear News Agency Editor
responsible for this story: David Dalton
Avenue des Arts 56 2/C
1000 Bruxelles
www.nucnet.org
DID YOU EDITORIAL KNOW...?
69
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
KernD
Robert-Koch-Platz 4
10115 Berlin
Germany
E-mail: presse@
KernD.de
www.KernD.de
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
Source:
“The Private and
External Costs of
Germany’s Nuclear
Phase-out”,
Stephen Jarvis,
Olivier Deschenes,
Akshaya Jha,
NBER Working Paper
No. 26598
Did you know...?

atw Vol. 65 (2020) | Issue 2 ı February
Calendar
70
2020
CALENDAR
19.02. – 21.02.2020
International Power Summit 2020. Hamburg,
Germany, Arena International,
www.arena-international.com
26.02.2020
TotalDECOM – International Conference. London,
UK, TotalDECOM, www.totaldecom.com
02.03. – 03.03.2020
Forum Kerntechnik. Berlin, Germany, VdTÜV & GRS,
www.tuev-nord.de
02.03. – 06.03.2020
International Workshop on Developing a
National Framework for Managing the Response
to Nuclear Security Events. Madrid, Spain, IAEA,
www.iaea.org
08.03. – 12.03.2020
WM Symposia – WM2019. Phoenix, AZ, USA,
www.wmsym.org
08.03. – 13.03.2020
IYNC2020 – The International Youth Nuclear
Congress. Sydney, Australia, IYNC, www.iync2020.org
15.03. – 19.03.2020
ICAPP2020 – International Congress on Advances
in Nuclear Power Plants. Abu-Dhabi, UAE, Khalifa
University, www.icapp2020.org
18.03. – 20.03.2020
12. Expertentreffen Strahlenschutz. Bayreuth,
Germany, TÜV SÜD, www.tuev-sued.de
22.03. – 26.03.2020
RRFM – European Research Reactor Conference.
Helsinki, Finland, European Nuclear Society,
www.euronuclear.org
25.03. – 27.03.2020
H2020 McSAFE Training Course. Eggenstein-
Leopoldshafen, Germany, Karlsruhe Institute of
Technology (KIT), www.mcsafe-h2020.eu
29.03. – 02.04.2020
PHYSOR2020 — International Conference on
Physics of Reactors 2020. Cambridge, United
Kingdom, Nuclear Energy Group,
www.physor2020.com
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,
NUGENIA, www.nugenia.org
30.03. – 01.04.2020
INDEX International Nuclear Digital Experience.
Paris, France, SFEN Société Française d’Energie
Nucléaire, www.sfen-index2020.org
31.03. – 03.04.2020
ATH'2020 – International Topical Meeting on
Advances in Thermal Hydraulics. Paris, France,
Société Francaise d’Energie Nucléaire (SFEN),
www.sfen-ath2020.org
08.04. – 09.04.2020
International SMR & Advanced Reactor Summit
2020. Atlanta, GA, USA, Nuclear Energy Insider,
www.nuclearenergyinsider.com
19.04. – 24.04.2020
International Conference on Individual
Monitoring. Budapest, Hungary, EUROSAFE,
www.eurosafe-forum.org
20.04. – 22.04.2020
World Nuclear Fuel Cycle 2020. Stockholm,
Sweden, WNA World Nuclear Association,
www.world-nuclear.org
05.05. – 06.05.2020
KERNTECHNIK 2020.
Berlin, Germany, KernD and KTG,
www.kerntechnik.com
10.05. – 15.05.2020
ICG-EAC Annual Meeting 2020. Helsinki, Finland,
ICG-EAC, www.icg-eac.org
11.05. – 15.05.2020
International Conference on Operational Safety
of Nuclear Power Plants. Beijing, China, IAEA,
www.iaea.org
12.05. – 13.05.2020
INSC — International Nuclear Supply Chain
Symposium. Munich, Germany, TÜV SÜD,
www.tuev-sued.de
12.05. – 14.05.2020
KELI – Conference for Electrical Engineering, I&C
and IT in generation plants. Bremen, Germany,
VGB PowerTech, www.vgb.org
14.05.2020
Nuclear Solutions Exhibition. Warrington, UK,
Industrial Exhibition, www.nuclear-solutions.co.uk
17.05. – 22.05.2020
BEPU2020– Best Estimate Plus Uncertainty International
Conference, Giardini Naxos. Sicily, Italy,
NINE, www.nineeng.com
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, www.snamc2020.jpn.org
20.05. – 22.05.2020
Nuclear Energy Assembly. Washington, D.C., USA,
NEI, www.nei.org
31.05. – 03.06.2020
13 th International Conference of the Croatian
Nuclear Society. Zadar, Croatia, Croatian Nuclear
Society, www.nuclear-option.org
06.06. – 12.06.2020
ATALANTE 2020. Montpellier, France, CEA,
www.atalante2020.org
07.06. – 12.06.2020
Plutonium Futures. Montpellier, France, CEA,
www.pufutures2020.org
08.06. – 10.06.2020
8 th Asia Nuclear Business Platform. Yogyakarta,
Indonesia, Nuclear Business Platform,
www.nuclearbusiness-platform.com
08.06. – 12.06.2020
20 th WCNDT – World Conference on
Non-Destructive Testing. Seoul, Korea, EPRI,
www.wcndt2020.com
15.06. – 19.06.2020
International Conference on Nuclear Knowledge
Management and Human Resources Development:
Challenges and Opportunities. Moscow,
Russian Federation, IAEA, www.iaea.org
15.06. – 20.07.2020
WNU Summer Institute 2020. Japan, World Nuclear
University, www.world-nuclear-university.org
02.08. – 06.08.2020
ICONE 28 – 28 th International Conference on
Nuclear Engineering. Disneyland Hotel, Anaheim,
CA, ASME, www.event.asme.org
01.09. – 04.09.2020
IGORR – Standard Cooperation Event in the International
Group on Research Reactors Conference.
Kazan, Russian Federation, IAEA, www.iaea.org
09.09. – 10.09.2020
VGB Congress 2020 – 100 Years VGB. Essen,
Germany, VGB PowerTech e.V., www.vgb.org
09.09. – 11.09.2020
World Nuclear Association Symposium 2020.
London, United Kingdom, WNA World Nuclear
Association, www.world-nuclear.org
16.09. – 18.09.2020
3 rd International Conference on Concrete
Sustainability. Prague, Czech Republic, fib,
www.fibiccs.org
16.09. – 18.09.2020
International Nuclear Reactor Materials
Reliability Conference and Exhibition.
New Orleans, Louisiana, USA, EPRI, www.snetp.eu
28.09. – 01.10.2020
NPC 2020 International Conference on Nuclear
Plant Chemistry. Antibes, France, SFEN Société
Française d’Energie Nucléaire,
www.sfen-npc2020.org
28.09. – 02.10.2020
Jahrestagung 2020 – Fachverband Strahlenschutz
und Entsorgung. Aachen, Germany, Fachverband
für Strahlenschutz, www.fs-ev.org
07.10. – 08.10.2020
3 rd India Nuclear Business Platform. Mumbai,
India, Nuclear Business Platform,
www.nuclearbusiness-platform.com
12.10. – 17.10.2020
FEC 2020 – 28 th IAEA Fusion Energy Conference.
Nice, France, IAEA, www.iaea.org
21.10. – 23.10.2020
2 nd Africa Nuclear Business Platform.
Accra, Ghana, Nuclear Business Platform,
www.nuclearbusiness-platform.com
26.10. – 30.10.2020
NuMat 2020 – 6 th Nuclear Materials Conference.
Gent, Belgium, IAEA, www.iaea.org
09.11. – 13.11.2020
International Conference on Radiation Safety:
Improving Radiation Protection in Practice.
Vienna, Austria, IAEA, www.iaea.org
24.11. – 26.11.2020
ICOND 2020 – 9 th International Conference on
Nuclear Decommissioning. Aachen, Germany,
AiNT, www.icond.de
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,
NEA, www.sammi-2020.org
17.12. – 18.12.2020
ICNESPP 2020 – 14. International Conference on
Nuclear Engineering Systems and Power Plants.
Kuala Lumpur, Malaysia, WASET, www.waset.org
This is not a full list and may be subject to change.
Calendar

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
plant.
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
factor.
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.
TWh
Source: World Nuclear Association and IAEA Power Reactor Information Service (PRIS)
GWe
3000
2500
2000
1500
1000
500
0
West & Central Europe
South America
North America
East Europe & Russia
Asia
Africa
1970
1972
1974
1976
1978
Source: World Nuclear Association, IAEA PRIS
%
1980
1982
1984
1986
1988
| Fig. 1.
Nuclear electricity production 1970 to 2018.
450
400
350
300
250
200
150
100
50
0
Not operating
1971
1973
1975
Operating
1977
1979
1981
1983
1985
1987
1989
| Fig. 2.
Nuclear generation capacity operable (net) 1971 to 2018.
90
80
70
60
50
0
1970
1974
1978
1982
1986
Source: World Nuclear Association, IAEA PRIS
1990
1990
1992
1994
1996
1998
2000
2002
2004
1991
1993
1995
1997
1999
2001
2003
2005
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
Switzerland.
The continued good operation of reactors is an
indication of the potential for longer operations. In the US
1994
1998
| Fig. 3.
Global average capacity factor 1970 to 2018.
2002
2006
2010
2014
2018
2006
2008
2010
2012
2014
2016
2018
2007
2009
2011
2013
2015
2017
71
FEATURE | MAJOR TRENDS IN ENERGY POLICY AND NUCLEAR POWER
Feature
Highlights of the World Nuclear Performance Report 2019 ı Jonathan Cobb

atw Vol. 65 (2020) | Issue 2 ı February
FEATURE | MAJOR TRENDS IN ENERGY POLICY AND NUCLEAR POWER 72
Number of reactors
18
16
14
12
10
8
6
4
2
0
%
100
90
80
70
60
50
40
30
20
10
0
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)
>90%
80-90%
70-80%
60-70%
50-60%
40-50%
0-40%
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
1983
35,000
30,000
25,000
20,000
15,000
10,000
5000
0
1984
1985
1986
Source: World Nuclear Association, IAEA PRIS
1987
1988
1989
1990
1991
1992
1993
1958
1960
1962
1964
1966
1968
1970
1994
1996
1972
1974
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
Asia
Africa
1954
1956
| Fig. 7.
Capacity of first grid connection 1954 to 2018.
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
1976
1978
1980
1982
1984
1986
1988
1990
1992
1994
1996
1998
2000
2002
2004
2006
2008
2010
2012
2015
2016
2017
2018
2014
2016
2018
%
100
80
60
40
20
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
capacity.
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
Feature
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
2050.
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.
Author
Dr Jonathan Cobb
Senior Communication Manager
World Nuclear Association
Tower House, 10 Southampton Street
London WC2E 7HA, UK
FEATURE | MAJOR TRENDS IN ENERGY POLICY AND NUCLEAR POWER 73
Feature
Highlights of the World Nuclear Performance Report 2019 ı Jonathan Cobb

atw Vol. 65 (2020) | Issue 2 ı February
74
Neue Wege der Öffentlichkeitsbeteiligung in atomrechtlichen
Verfahren
Tobias Leidinger
SPOTLIGHT ON NUCLEAR LAW
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.
I
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
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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
würden.
SPOTLIGHT ON NUCLEAR LAW 75
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.
Autor
Prof. Dr. Tobias Leidinger
Rechtsanwalt und Fachanwalt für Verwaltungsrecht
Luther Rechtsanwaltsgesellschaft
Graf-Adolf-Platz 15
40213 Düsseldorf
tobias.leidinger@luther-lawfirm.com
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76
ENERGY POLICY, ECONOMY AND LAW
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
profiles.
Background
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
curves.
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
Introduction
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.
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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
besides.
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.
Methodology
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
intervals.
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
importance.
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
disagree).
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.
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ENERGY POLICY, ECONOMY AND LAW 78
| 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
selected.
This workshop also defined the
information requirements that would
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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
compared.
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
ENERGY POLICY, ECONOMY AND LAW 79
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
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ENERGY POLICY, ECONOMY AND LAW 80
| 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]
[14].
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.
Acknowledgements
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
CompariCube®.
References
[1] “Hierarchy of Controls”. U.S. National Institute for
Occupational Safety and Health. Retrieved 2017-01-31.,”
[Online].
[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: ssrn.com/abstract=2696000.
[Accessed 16 March 2018].
[3] R. Schapire, “COS511: Theoretical Machine Learning,” 1 May
2014. [Online]. Available: https://www.cs.princeton.edu/
courses/archive/spring14/cos511/scribe.../0501.pdf.
[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://
cran.r-project.org/web/packages/bbmle/bbmle.pdf. [Accessed
16 March 2018].
[7] A. B. Collier, “Fitting a model by maximum likelihood,”
18 August 2013. [Online]. Available:
http://www.exegetic.biz/blog/2013/08/fitting-a-model-bymaximum-likelihood/.
[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:
https://www.tvo.fi/Materiality%20analysis%20and%
20responsibility%20aspects#.
[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.
ac.uk/thebeam/2018/07/17/beyond-consultation-hybridforums-for-the-development-of-nuclear-energy/.
[Accessed 30 September 2019].
[14] Times & Star, “Volunteers are needed for nuclear think-tank,”
18 September 2019. [Online]. Available:
https://www.timesandstar.co.uk/news/
17908571.volunteers-needed-nuclear-think-tank/.
[Accessed 30 September 2019].
Authors
Howard Chapman
Maria Cormack
Caroline Pyke
John-Patrick Richardson
Reuben Holmes
National Nuclear Laboratory
Limited
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
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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
terrorism.
2 Related work
For nuclear terrorism countering, the
continuous monitoring system like as
Argonne’s ARG-US RFID, ARG-US
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
Transcom
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
operation
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
81
ENVIRONMENT AND SAFETY
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ENVIRONMENT AND SAFETY 82
| 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
rays.
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;
respectively.
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
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| 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/
hour.
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
GSM/GPRS.
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
application.
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
minutes).
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
Calculation
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.
ENVIRONMENT AND SAFETY 83
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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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ENVIRONMENT AND SAFETY 84
(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
system.
4.3.1 Tracking through Google
maps
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:
//maps.google.com/?q=30.053795,
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
intervention.
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
ENVIRONMENT AND SAFETY 86
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
below.
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.
Acknowledgment
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.
References
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[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
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[6] INVAP, “ORNL/ORIGEN Version 2.1”, 2004.
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[10] Markku Anttila, “Radioactive characteristics of the spent fuel
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[14] C. Renaudie, M. Markgraf, O. Montenbruck, and M. Garcia,
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Authors
M. I. Youssef
Faculty of Engineering
Al Azhar University
Cairo, Egypt
M. Elzorkany
National Telecommunication
Institute
Cairo, Egypt
G. F. Sultan
Morsi
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.
ENVIRONMENT AND SAFETY 87
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
improvement
management of NPPs
2.1 Nuclear safety culture
dynamic improvement
model
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
culture.
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
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ENVIRONMENT AND SAFETY 88
| 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
model.
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
developed.
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
CNNP
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
methods
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
of NSCEMS
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.
WANO
Ten Traits
INPO
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.
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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
level.
4 Design and application
of the comprehensive
nuclear safety culture
quantitative evaluation
model
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
judgment.
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
analysis
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
evaluation.
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
analysis
Main negative
attributes
PI.2- Evaluation LA.5- Provide resources LA.5- Provide resources
LA.7- Change
Management
LA.5- Provide resources
NS.3- Risk control
throughout the whole
work process
NS.5- High quality
procedures
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
Assessment
PI.2- Evaluation
LO.5- Training
WE.1- Respect is Evident
NS.5- High quality
procedures
ENVIRONMENT AND SAFETY 89
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ENVIRONMENT AND SAFETY 90
NSC
level
Assignment
range
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
Instinctive
reaction
Passive
management
Active
management
Employee
participation
Team mutual
assistance
Continuous
improvement
Highly
self-discipline
| 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
network
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.
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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.
(1)
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
levels.
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
processing.
(2)
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.
(3)
| 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 91
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ENVIRONMENT AND SAFETY 92
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)
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
determined.
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)
C=b
C=a=b
C=a
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
feasible.
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
NPPs.
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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
NPPs.
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.
References
4 positive observations
5 general observations
6 negative observations
[1] Park, Kyung S.; Lee, Jae in. A new method for estimating
human error probabilities: AHP-SLIM. Reliability Engineering
and System Safety. 4(2008), 578-587.
[2] T. Lee *, K. Harrison. Assessing safety culture in nuclear
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[3] Wang Li, Nuclear Safety Cultural Conflicts and Its Countermeasures:
Revelation from the Fukushima Nuclear Accident.
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(Social Sciences Edition), 2013, Vol. 26.24-29.
[4] Fernandez-Muniz, Beatriz, etal.Safety culture: Analysis of the
causal relationships between its key dimensions. JOURNAL
OF SAFETY RESEARCH.2007,Vol.38,No. 6, 627-641.
[5] Björn Wahlström. Systemic thinking in support of safety
management in nuclear power plants. Safety Science.
109 (2018), 201–208.
[6] Schein, E.H., 2010. Organizational Culture and Leadership,
fourth ed. Jossey-Bass,San Francisco.
[7] International Atomic Energy Agency (IAEA) Safety series
No.75-INSAG-4. Safety Culture.1991.
[8] World Nuclear Operators Association (WANO) Guideline
2006-02. Principles for a Strong Nuclear Safety Culture.2006.
[9] Ziedelis Stanislovas, etal. Human based roots of failures
in nuclear events investigations. ATW-INTERNATIONAL
JOURNAL FOR NUCLEAR POWER. 2012, Vol. 57, No. 10,
596-601.
[10] Hazmimi Kasim, etal.The relationship of safety climate
factors, decision making attitude, risk control, and risk
estimate in Malaysian radiation facilities. Safety Science.
113 (2019), 180–191.
[11] Institute of Nuclear Power Operations(INPO) 12-012.Healthy
Nuclear Safety Culture Traits.2013.
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[12] Sang Min Han a, Seung Min Lee, etal. Development of
Nuclear Safety Culture evaluation method for an operation
team based on the probabilistic approach. Annals of Nuclear
Energy. 111 (2018), 317–328.
[13] Young Gab Kim, etal.Approach for safety culture evaluation
under accident situation at NPPs ; an exploratory study using
case studies. Annals of Nuclear Energy.121 (2018), 305–315.
[14] Han, Kiyoon, etal. Development of a New Methodology for
Quantifying, Nuclear Safety Culture. ATW-INTERNATIONAL
JOURNAL FOR NUCLEAR POWER. 2017,Vol. 62, No.1, 30–35.
[15] Stammsen,S ,Gloeckle,W. Capturing safety culture in plant
inspections - KOMFORT, an oversight tool of Baden-
Wurttemberg’s nuclear regulatory authority. ATW-
INTERNATIONAL JOURNAL FOR NUCLEAR POWER. 2007,
Vol. 52, No.11, 731–735.
[16] Markus Schöbel, etal. Digging deeper! Insights from a
multi-method assessment of safety culture in nuclear power
plants based on Schein’s culture model. Safety Science.
95 (2017), 38–49.
[17] Joanna Martyka,Kazimierz Lebecki.Safety Culture in High-
Risk Industries. International Journal of Occupational Safety
and Ergonomics (JOSE) ,2014, Vol. 20, No. 4, 561–572.
[18] Liu, Meiyu, Shi, Jing. A cellular automata traffic flow model
combined with a BP neural network based microscopic lane
changing decision model. JOURNAL OF INTELLIGENT
TRANSPORTATION SYSTEMS.2019, Vol.23,No. 4, 309-318.
[19] Esra Ilbahar, Ali Karaşan, etal. A novel approach to risk
assessment for occupational health and safety using
Pythagorean fuzzy AHP & fuzzy inference system. Safety
Science. 103 (2018), 124–136.
[20] Qian Wang, Rong Han, etal. Research on energy conservation
and emissions reduction based on AHP-fuzzy synthetic
evaluation model: A case study of tobacco enterprises.
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Authors
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
Operation
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 93
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
ENVIRONMENT AND SAFETY 94
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-
Schmehausen.
2 Sicherheitsanforderungen
an zukünftige
(V)HTR-KKW.
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
Ereignissen;
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
Störungen;
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
AVR-Versuchs-KKW
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
Graphiteinbauten;
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
wurde;
p Die unerwartet geringe Menge an
Graphitstaub;
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.
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4 Erfahrungen mit
dem 300 MW el -THTR-
Demonstrations-KKW
in Hamm-Uentrop/
Schmehausen
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
richtig.
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
Rohrleitungen;
p Kühlgasströmung von oben nach
unten;
p Abschalt- und Regelstäbe in den
Graphiteinbauten.
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
Kraftwerk;
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
Endlager;
Diese umfassenden Erfahrungen
ermöglichen den Bau neuer HTR-
Groß-Kernkraftwerke [8; 14 – 19].
5 Die Sicherheit der
Kugel-Brennelemente
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 95
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ENVIRONMENT AND SAFETY 96
Totalemission in die Umgebung nicht
zu einer zu hohen Belastung geführt
hätte.
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
Treaty
(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
Graphit-Moderatorelementen
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
Anforderungen.
Ausgenommen hiervon sind alle
sekundären Anlagen, wie Stromerzeugung,
Trinkwasserproduktion
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-
Prinzip“.
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
Spannbetonbehälter;
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
Bunkerbereich.
7 Die Konstruktion
sicherheitsrelevanter
Komponenten
Sicherheitsrelevante Komponenten
sind:
p Der Spannbetonbehälter mit
Linerkühlsystem, Isolierung und
Stützkonstruktion:
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
kann.
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
ist.
Letztlich kann der Druck im SBB
durch Absenken des Druckes des
Primär-Heliums und Abpumpen in
das Heliumlager druckentlastet
werden.
p Die Primärgasgebläse
Eingebaut werden mehrere Gebläse,
bspw. sechs. Sie haben die
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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
lagern.
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
Notsteuerstellen:
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
Heliumlager.
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
Bedeutung.
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
erfüllt.
Ein derartig hoher Sicherheitsstandard
kann von keinem der derzeit
in Betrieb oder Planung befindlichen
KKW erreicht werden.
Alle Konstruktionsprinzipien sind
erprobt.
Es gilt:
„Der sicherste Reaktor ist auch der
wirtschaftlichste Reaktor“.
Literatur:
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
6/2011.
18. Urban Cleve: „The Technology of High Temperature Reactors
and Production of Nuclear Heat“. University of Cracow,
NUTECH-2011.
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”.
Author
Dr.-Ing. Urban Cleve
Ex. CTO/HA-Leiter Technik
of BBC/Krupp Reaktorbau GmbH,
Mannheim
Hohenfriedbergerstr. 4
44141 Dortmund, Germany
ENVIRONMENT AND SAFETY 97
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
ENVIRONMENT AND SAFETY 98
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.
Introduction
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
features.
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
FASD-C)
4. Emergency Feed Water System,
(RS)
5. Main Steam Isolation Valves
(MSIV) ⇔ BZOK
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),
(YP)
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,
(GY)
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
Plant.
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
O’);
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.
ENVIRONMENT AND SAFETY 99
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
TCV
MSV
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
O’ SRD
Open
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
(EFWP)
Pressurizer safety valves (SVP)
Additional boron injection system
(TW)
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
(SGSV)
Auxiliary feed water pumps (AFWP)
Emergency feed pumps (EFWP)
Makeup system for deaerators and
tanks of EFWP (UD)
C4
C3
C2
C1
SRD
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
AND
reducing of primary circuit pressure to 2 MPa
by opening of one SVP or conducts injection
into pressurizer from one channel of TW system
AND
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
AND
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
ENVIRONMENT AND SAFETY 100
| 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
damage.
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
A_
T_
Cs_
Ho”_
O’_
C4_
| 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
2.6E-07
2.88E-08
5.32E-03
1.71E-03
1.38E-08
1.98E-06
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
ENVIRONMENT AND SAFETY 101
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
ENVIRONMENT AND SAFETY 102
| 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
RIR
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
RIR
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
ENVIRONMENT AND SAFETY 103
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
Reference
ENVIRONMENT AND SAFETY 104
11BU
12BV
13BW
14BX
| Tab. 5.
Failure probability of 10 kv Safety functions.
Core Damage States
CD3
CD4
CD5
CD6
CD7
CD8
CD9
CD10
Total CD
2.78E-02
2.79E-02
2.76E-02
2.7E-02
Frequency per year
2.71E-06
5.90E-07
3.18E-09
2.85E-10
1.94E-09
3.87E-09
8.58E-09
7.74E-08
3.40E-06
[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.
Conclusion
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
necessary.
Authors
Mohsen Esfandiari
Gholamreza Jahanfarnia
Department of Nuclear
Engineering
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
Reflectors
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.
Introduction
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
105
RESEARCH AND INNOVATION
| 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
RESEARCH AND INNOVATION 106
| 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
measurements
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.
(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)
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,
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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.
RESEARCH AND INNOVATION 107
| 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)
[10].
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
0.71.
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
results.
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.
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atw Vol. 65 (2020) | Issue 2 ı February
RESEARCH AND INNOVATION 108
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
analysis
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.
Acknowledgements
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.
References
[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:
487–490.
[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:
http://www.kayelaby.npl.co.uk/atomic_and_nuclear_physics/4_7/4_7_3.html
Author
Khurram Mehboob (Ph.D.)
Associate Professor
Department of Nuclear
Engineering, Faculty of
Engineering,
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.
109
REPORT
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
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110
REPORT
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
COCOSYS and ANSYS CFX showed
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
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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
orientation.
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.
Authors
Florian Rowold
Klemens Hummelsheim
Maik Stuke
Gesellschaft für Anlagen- und
Reaktorsicherheit (GRS) gGmbH
Bereich Stilllegung und Entsorgung
Abteilung Stilllegung und
Zwischenlagerung
Forschungszentrum,
Boltzmannstr. 14,
85748 Garching b. München
111
REPORT
Report
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
112
Inside
Einladung zum Vortrag
KTG INSIDE
KTG Inside
Verantwortlich
für den Inhalt:
Die Autoren.
Lektorat:
Natalija Cobanov,
Kerntechnische
Gesellschaft e. V.
(KTG)
Robert-Koch-Platz 4
10115 Berlin
T: +49 30 498555-50
F: +49 30 498555-51
E-Mail:
natalija.cobanov@
ktg.org
www.ktg.org
Stand der weltweiten Entwicklung der Kernenergie
von Dr. Ludger Mohrbach
am Donnerstag, den 19. März 2020 um 17:30 Uhr,
PreussenElektra GmbH, Tresckowstraße 5,
Hannover
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
thomas.froehmel@preussenelektra.de
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,
Leingarten
55 Jahre | 1965
22. Karsten Müller Kleinmachnow
60 Jahre | 1960
23. Peter Reimann, Lingen
65 Jahre | 1955
06. Prof. Dr. Peter-Wilhelm Phlippen,
Geilenkirchen
70 Jahre | 1950
23. Hans-Dieter Schmidt, Dortmund
75 Jahre | 1945
04. Dr. Bernd Hofmann,
Eggenstein-Leopoldshafen
11. Dr. Ulrich Krugmann, Erlangen
11. Joachim Lange, Burgdorf
15. Bernhard Brand, Forchheim
20. Dipl.-Ing. mult. Herbert Niederhausen,
Gebhardshain
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,
Garbsen
81 Jahre |1939
01. Prof. Dr. Günter Höhlein,
Unterhaching
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
113
NEWS
Top
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
News

atw Vol. 65 (2020) | Issue 2 ı February
114
NEWS
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); www.rosatom.ru
www.fnpp.info
World
U.S.: Policymakers and energy
companies plan to reduce
carbon and know they’ll need
nuclear
(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
goals.
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
acting.
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.
(20211007)
| www.nei.org
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); www.oecd-nea.org
Europe
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”.
News

atw Vol. 65 (2020) | Issue 2 ı February
Operating Results October 2019
Plant name Country Nominal
capacity
Type
gross
[MW]
net
[MW]
Operating
time
generator
[h]
Energy generated, gross
[MWh]
Month Year Since
commissioning
Time availability
[%]
Energy availability
[%] *) Energy utilisation
[%] *)
Month Year Month Year Month Year
OL1 Olkiluoto BWR FI 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
115
NEWS
Plant name
Type
Nominal
capacity
gross
[MW]
net
[MW]
Operating
time
generator
[h]
Energy generated, gross
[MWh]
Time availability
[%]
Energy availability
[%] *) Energy utilisation
[%] *)
Month Year Since Month Year Month Year Month Year
commissioning
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); www.foratom.org
Reactors
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); www.rosatom.ru
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)
1)
Refueling
2)
Inspection
3)
Repair
4)
Stretch-out-operation
5)
Stretch-in-operation
6)
Hereof traction supply
7)
Incl. steam supply
8)
New nominal
capacity since
January 2016
9)
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
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atw Vol. 65 (2020) | Issue 2 ı February
116
NEWS
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
facility.
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);
www.french-nuclear-safety.fr
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).
1
2
3
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
IP-2).
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); www.gns.de
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
DESCRIPTION
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
significantly.
News

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); www.framatome.com
Uranium
Prize range: Spot market [USD*/lb(US) U 3O 8]
140.00
120.00
100.00
80.00
60.00
40.00
20.00
0.00
Year
Year
Separative work: Spot market price range [USD*/kg UTA]
Conversion: Spot conversion price range [USD*/kgU]
180.00
) 1 23.00
160.00
140.00
120.00
1980
Yearly average prices in real USD, base: US prices (1982 to1984) *
1985
1990
1995
2000
2005
) 1
2010
2015
2019
Uranium prize range: Spot market [USD*/lb(US) U 3O 8]
140.00
) 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.
120.00
100.00
80.00
60.00
40.00
20.00
0.00
22.00
20.00
18.00
16.00
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
117
NEWS
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);
www.westinghouse.com
100.00
80.00
60.00
40.00
20.00
0.00
Jan. 2008
Jan. 2009
Jan. 2010
Jan. 2011
Jan. 2012
Market data
(All information is supplied without
guarantee.)
Nuclear Fuel Supply
Market Data
Information in current (nominal)
U.S.-$. No inflation adjustment of
prices on a base year. Separative work
data for the formerly “secondary
market”. Uranium prices [US-$/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
Year
Jan. 2014
Jan. 2015
Jan. 2016
2017
p Uranium: 19.25–26.50
p Conversion: 4.50–6.75
p Separative work: 39.00–50.00
2018
p Uranium: 21.75–29.20
p Conversion: 6.00–14.50
p Separative work: 34.00–42.00
2019
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.
)1
In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.
* Actual nominal USD prices, not real prices referring to a base year
Sources: Energy Intelligence, Nukem; Bilder/Figures: atw 2020
14.00
12.00
10.00
8.00
6.00
4.00
2.00
0.00
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
www.energyintel.com
Jan. 2013
Year
Jan. 2014
Jan. 2015
Jan. 2016
Jan. 2017
Jan. 2018
Jan. 2019
Jan. 2020
News

atw Vol. 65 (2020) | Issue 2 ı February
118
NUCLEAR TODAY
John Shepherd is a
freelance journalist
and communications
consultant.
Sources:
NBER working paper
https://bit.ly/
35QIQLd
IAEA director-general
remarks
https://bit.ly/30iJShT
Brookhaven National
Lab project
https://bit.ly/
2FMq7pp
IEE analysis
https://bit.ly/
35SYQMy
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
competitors.
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

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