atw 2017-06

nucmag.com
2017
6
ISSN · 1431-5254
16.– €
378
AMNT 2017:
Opening Address
384 ı AMNT 2017: Best Paper
Emplacement Radiation Exposure Calculations
for Generic Deep Geological Repositories
392 ı Environment and Safety
Retrofitting a Spent Fuel Pool Spray System
396 ı Operation and New Build
Cyber Security in Nuclear Power Plants
402 ı Decommissioning and Waste Management
Validation of Spent Nuclear Fuel Nuclide Composition Data

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atw Vol. 62 (2017) | Issue 6 ı June
India Goes Nuclear
Dear reader, India is a country of diversity and with its 1.3 billion citizens India is after China not only second most
populous country in the world but also the most populous democracy in the world. This reflects the great responsibility
for the countries’ politics, to create social conditions, which further maintain and strengthen the democracy. Economic
growth is thereby unquestionably one of the most important components in order to expand social services and to create
quality of life. Also due to this India’s economy grew in the past 10 years on average by 11 % per year, even in times of a
global financial crisis as of 2007.
With a view to the energy supply India is today, after China
and the USA the third largest energy consumer in the
world. Despite of extensive available energy resources,
India developed to an important energy importer of fossil
fuels. With a share of over 75 % of the energy generation
the importance of coal for the energy and electricity supply
is very dominant. In the year 2013 around 692 million tons
of coal were used of which 159 million tons were imported.
The 22 nuclear power plants with a gross capacity of
6,780 MW have a share of around 2.2 % of the country’s
total power generation capacity of 303,071 MWe and of
the generation of around 3.5 % through the production of
35 gigawatt hours in 2016. Due to the combination of a
today comparatively low per capita rate of electricity
consumption in the amount of 1,000 kWh per inhabitants
and aspired growth, as well as the need to provide
electricity to the approximately 240 million persons in
India which do not have any access to electricity today, it
will certainly further increase. Until the 2020s a doubling
is expected. It should be noted for India, that the agriculture
proportion on the energy consumption – especially for the
irrigation of fields- up to one third, clarifies that a secure
supply of energy is not a question of comfort put also of a
primary care.
This poses the country and its decision makers to great
challenges. In order to manage this situation no options
are excluded. Thus, growth for all energy carriers in India
is expected and aimed for in the upcoming years until the
middle of the 2020s with strong differing degrees of the
single energy carrier. Ambitioned is the extension of
renewables, with a target setting especially for wind of
+60,000 MW and photovoltaic of +100,000 MW, which
corresponds in total to a fourfold increase. But also the
coal-fired generation will further increase.
....and nuclear energy?
Research and development of nuclear energy in India have
a long national tradition. Today’s Bhaba Atomic Research
Centre near Mumbai was established in the 1950s. A first
light water reactor or rather heavy water moderated
pressurised water reactor of the Canadian type CANDU
was put in operation in 1969 or rather 1972 at the sites
Tarapur and Rajasthan. The advancement of nuclear
energy within the international network was then
inhibited, as India, being a nuclear power, had not signed
the Nuclear Non-Proliferation Treaty.
India’s nuclear economy depended thus on its own
development or rather further development of the
expansion. Through standardising the heavy water reactor
technology the possibility was given to establish an own
productive reactor type and to commission until today
18 plants. The long term perspective of nuclear energy will
be underlined with the prototype establishment of a fast
sodium-cooled 500-MW-reactor at the site Kalpakkam,
whose commissioning is planned for autumn 2017.
Additionally the Indian Department for Atomic Energy just
recently communicated, that two further 600-MW-breeder
reactor shall follow in Kalpakkam
With the end of the East-West conflict the relationships
of many countries with India changed in the matter of
nuclear technology. Through cooperation with Russia two
WWER-reactors with each 1,000 MW of (output) were
established and put in operation at the site Kudankulam
as of the year 2002. An agreement with the Nuclear
Suppliers Group in 2008 opened up the path for a couple
bilateral agreements for the expansion of nuclear energy.
Miscellaneous new-build projects are mentioned and
negotiated repeatedly since this year in order to achieve
the expansion target of +10 % capacity per year until the
year 2025.
India is getting serious also in the matter of nuclear
energy. The Indian Prime Minister Narendra Mori
announced in May 2017 as a first step for the government
the initiation of a national nuclear energy expansion
program. This programme shall bring a strong push
to entire Indian economy: 10 nuclear power plant projects
on the basis of the Indian heavy water reactor technology
with an overall performance of 6,700 MW* – the
same amount as the currently operated ones – shall be
accommodated within the next 5 years.
The full investment is mentioned with 11 b. $. Just in
the country’s nuclear industry 33,400 new, qualified jobs
shall be generated in this manner. With the experience
made from the establishment and commissioning of
today’s operating heavy water reactors and through the
standardisation of 10 new plants, the „fleet construction
program“ shall generate synergies through an „Economy
by Number“ and thus electrically enable the aspired and
comparatively low investment costs of 1,650 $ per installed
kilowatt.
India’s paths of a future energy supply are diverse and
include the path of using nuclear energy; step by step and
under the target set of further 80,000 MW until the end of
the 2020’s.
Christopher Weßelmann
– Editor in Chief –
* At the power
generation 1 MW of
installed power
output corresponds
due to a higher
availability
approximately to
4 MW of installed
wind power and
8 MW of installed
photovoltaic
capacity – the often
underestimated
difference between
labour and
performance
367
EDITORIAL
Editorial
India Goes Nuclear

atw Vol. 62 (2017) | Issue 6 ı June
EDITORIAL 368
*Bei der Stromerzeugung
entspricht
1 MW installierter
Kernenergieleistung
aufgrund höherer
Arbeitsverfügbarkeit
etwa 4 MW installierter
Windkraft und
8 MW installierter
Fotovoltaikleistung –
der häufig unterschätzte
Unterschied
von Leistung und
Arbeit.
Indiens Weg mit der Kernenergie
Liebe Leserin, lieber Leser, Indien ist ein Land der Vielfalt und mit 1,3 Milliarden Menschen nicht nur zweitbevölkerungsreichstes
Land der Welt nach China, sondern auch bevölkerungsreichste Demokratie der Welt. Dies
spiegelt die große Verantwortung für die Politik des Landes wieder, gesellschaftliche und soziale Rahmenbedingungen
zu schaffen, die diese Demokratie weiter zusammenhalten und festigen.
Wirtschaftliches Wachstum ist dabei unzweifelhaft eine
der wichtigsten Komponenten, um das Sozialwesen
auszubauen und Lebensqualität zu schaffen. In den vergangenen
10 Jahren ist Indiens Wirtschaft auch deshalb
um durchschnittlich 11 % pro Jahr gewachsen und dies
selbst in den Zeiten der globalen Finanzkrise ab 2007.
Mit Blick auf die Energieversorgung ist Indien heute
nach China und den USA drittgrößter Energieverbraucher
der Welt. Trotz umfangreicher vorhandener Energie reserven
hat sich Indien zum bedeutenden Energie importeur
fossiler Energieträger entwickelt. Die Bedeutung von
Kohle für die Energie- und Stromversorgung ist sehr dominant
mit einem Anteil von über 75 % der Stromerzeugung.
Hier wurden im Jahr 2013 rund 692 Millionen Tonnen
Steinkohle eingesetzt, von denen 159 Millionen Tonnen
importiert wurden. Die 22 Kernkraftwerke haben mit
6.780 MW Bruttoleistung einen Anteil von rund 2,2 % an
der Gesamtstromerzeugungsleistung des Landes von
303.071 MWe und an der Erzeugung von rund 3,5 % durch
die Produktion von 35 Gigawattstunden in 2016.
Aufgrund der Kombination von heute vergleichsweise
niedrigem pro Kopf Stromverbrauch in Höhe von 1.000
kWh pro Einwohner und angestrebtem Wachstum sowie
der Notwendigkeit den etwa 240 Millionen Menschen in
Indien, die heute keinen Zugang zu Elektrizität haben,
einen solchen zu verschaffen, wird sich dieser weiter
erhöhen. Bis in die 2020er-Jahre wird eine Verdoppelung
erwartet. Dabei ist für Indien zudem zu bemerken, dass
der Anteil der Landwirtschaft am Stromverbrauch –
wesent lich zur Bewässerung von Feldern – von bis zu
einem Drittel verdeutlicht, dass eine gesicherte Stromversorgung
keine Frage des Komforts, sondern auch der
Grundversorgung ist.
Dies stellt das Land und seine Entscheider vor enorme
Herausforderungen. Um sie zu bewältigen, wird keine
Option ausgeschlossen. Daher wird für alle Energieträger
in Indien in den kommenden Jahren bis Mitte der 2020er
ein Wachstum erwartet und angestrebt, mit unterschiedlich
starker Ausprägung der einzelnen Energieträger.
Ambitioniert ist der Ausbau der Erneuerbaren, mit einer
Zielvorgabe vor allem beim Wind von +60.000 MW und
der Fotovoltaik von +100.000 MW, was insgesamt einer
Vervierfachung entspricht. Aber auch die Kohleverstromung
wird sich weiter erhöhen.
... und die Kernenergie?
Forschung und Entwicklung der Kernenergie in Indien
haben eine lange nationale Tradition. In den 1950er- Jahren
wurde das heutige Bhaba Atomic Research Center nahe
Mumbai errichtet. An den Standorten Tarapur und
Rajasthan wurden 1969 bzw. 1972 ein erster Leichtwasserreaktor
bzw. ein schwerwassermoderierter Druckwasserreaktor
vom kanadischen CANDU-Typ in Betrieb
genommen. Die Weiterentwicklung der Kernenergie im
internationalen Verbund war dann aber gehemmt, da
Indien als Atomwaffenmacht nicht den Atomwaffensperrvertrag
unterzeichnet hatte. Indiens Nuklearwirtschaft
war somit auf eigene Entwicklungen bzw.
Weiterentwicklungen bei Ausbau angewiesen. Durch
Standardisierung der Schwerwasser reaktortechnologie
gelang es, eine eigene leistungsfähige Reaktorlinie zu
etablieren und bis heute 18 solcher Anlagen in Betrieb zu
nehmen. Die Langfristperspektive der Kern energie wird
mit der prototypischen Errichtung eines schnellen natriumgekühlten
500-MW-Reaktors am Standort Kalpakkam
unterstrichen, dessen Inbetriebnahme für Herbst 2017
angekündigt ist. Zudem wurde jüngst vom indischen
Department for Atomic Energy mitgeteilt, dass weitere zwei
600-MW-Brutreaktoren in Kalpakkam folgen sollen.
Mit dem Ende des Ost-West-Konfliktes änderten
sich auch die Beziehungen vieler Länder in Sachen
Kernenergietechnologie mit Indien. Im Rahmen von
Kooperationen mit Russland wurden so ab 2002 zwei
WWER-Reaktoren mit jeweils 1.000 MW Leistung am
Standort Kudankulam errichtet und in Betrieb genommen.
Eine Vereinbarung mit der Nuclear Suppliers Group im
Jahr 2008 eröffnete zudem den Weg für eine Reihe von
bilateralen Vereinbarungen zum Ausbau der Kernenergie.
Verschiedenste Neubauvorhaben werden seit diesem Jahr
immer wieder genannt und verhandelt, um das Ausbauziel
von +10 % Kapazität pro Jahr bis 2025 zu erreichen.
So macht Indien jetzt ernst, auch in Sachen Kernenergie.
Der indische Premierminister Narendra Mori gab
im Mai 2017 für die Regierung als den ersten Schritt
die Initiierung eines nationalen Kernenergieausbauprogramms
bekannt. Dieser soll der indischen Wirtschaft
insgesamt einen großen Schub verschaffen: Innerhalb der
kommenden 5 Jahre sollen 10 Kernkraftwerksprojekte auf
Basis der indischen Schwerwasserreaktortechnologie
mit einer Gesamtleistung von 6.700 MW* – also in
gleicher Höhe, wie die derzeit betriebenen – aufgenommen
werden. Der Gesamtinvestitionsumfang wird
mit 11 Mrd. $ angegeben. Allein in der kerntechnischen
Industrie des Landes sollen so 33.400 neue, qualifizierte
Arbeitsplätze geschaffen werden. Mit den Erfahrungen
aus Errichtung und Inbetriebnahme der heute laufenden
Schwerwasserreaktoren und durch die Standardisierung
der 10 Neuanlagen soll das „Flottenbauprogramm“
Synergien durch eine „Economy by Number“ erzeugen und
so die angestrebten vergleichsweise niedrigen Investitionskosten
von 1.650 $ pro installiertem Kilowatt elektrisch
ermöglichen.
Indiens Wege der zukünftigen Energieversorgung sind
vielfältig und schließen den Weg der Kernenergienutzung
mit ein; Schritt für Schritt und einer Zielvorgabe von
weiteren 80.000 MW bis Ende der 2020er-Jahre.
Christopher Weßelmann
– Chefredakteur –
Editorial
India Goes Nuclear

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atw Vol. 62 (2017) | Issue 6 ı June
370
Issue 6
June
CONTENTS
378
AMNT 2017:
Opening Address
| | Research, an inevitable part of know-how preservation and development. View of the FRM II upper head reactor components during
commissioning. (Photo: FRM II)
Editorial
India Goes Nuclear. . . . . . . . . . . . . . . . . . . . 367
Indiens Weg mit der Kernenergie . . . . . . . . . . 368
Abstracts | English . . . . . . . . . . . . . . . . . . . 372
Abstracts | German . . . . . . . . . . . . . . . . . . . 373
AMNT 2017
48 th Annual Meeting on Nuclear Technology
(AMNT 2017): Opening Address . . . . . . . . . . . 378
Ralf Güldner
48 th Annual Meeting on Nuclear Technology
(AMNT 2017): Impressions . . . . . . . . . . . . . . . 382
Inside Nuclear with NucNet
Q&A: Poland’s Progress on the Road
to New Nuclear . . . . . . . . . . . . . . . . . . . . . . 374
NucNet
Calendar . . . . . . . . . . . . . . . . . . . . . . . 376
DAtF Notes. . . . . . . . . . . . . . . . . . . . . .377
384
| | Different angles of the phantom with respect to POLLUX.
AMNT 2017: Best Paper
Monte-Carlo Based Comparison of the
Personal Dose for Emplacement Scenarios
of Spent Nuclear Fuel Casks in Generic
Deep Geological Repositories . . . . . . . . . . . . . 384
378
Héctor Saurí Suárez, Bo Pang, Frank Becker and Volker Metz
| | Dr. Ralf Güldner delivering his Opening Address.
Contents

atw Vol. 62 (2017) | Issue 6 ı June
371
Spotlight on Nuclear Law
The 15 th German Atomic Energy Act Amendment
to the Implementation of the EURATOM Nuclear
Safety Directive . . . . . . . . . . . . . . . . . . . . . . 391
Die 15. AtG-Novelle zur Umsetzung
der EURATOM-Sicherheits-Richtlinie. . . . . . . . . 391
Christian Müller-Dehn
Environment and Safety
Retrofitting a Spent Fuel Pool Spray System
for Alternative Cooling as a Strategy for Beyond
Design Basis Events . . . . . . . . . . . . . . . . . . . 392
Christoph Hartmann and Zoran Vujic
392
| | AP1000® Plant Spent Fuel Pool Spray System.
Operation and New Build
Cyber Security in Nuclear Power Plants and its
Portability to Other Industrial Infrastructures . . . 396
Sébastien Champigny, Deeksha Gupta, Venesa Watson
and Karl Waedt
|408
418
| | Nodalization for the primary system of AP1000.
Research and Innovation
Reliability Analysis on Passive Residual Heat
Removal of AP1000 Based on Grey Model . . . . . 408
Qi Shi, Zhou Tao, Muhammad Ali Shahzad, Li Yu and Jiang Guangming
Experimental Investigation of a Two-Phase
Closed Thermosyphon Assembly for Passive
Containment Cooling System . . . . . . . . . . . . . 413
Kyung Ho Nam and Sang Nyung Kim
Displacement of Cryomodule
in CADS Injector II . . . . . . . . . . . . . . . . . . . . 418
Yuan Jiandong, Zhang Bin, Wang Fengfeng, Wan Yuqin, Sun Guozhen,
Yao Junjie, Zhang Juihui and He Yuan
| Model of the cryomudule.
CONTENTS
KTG Inside . . . . . . . . . . . . . . . . . . . . . . 422
News . . . . . . . . . . . . . . . . . . . . . . . . . 424
396
| | Overview of cyber security portfolio.
Decommissioning and Waste Management
Validation of Spent Nuclear Fuel Nuclide
Composition Data Using Percentage Differences
and Detailed Analysis . . . . . . . . . . . . . . . . . . 402
Man Cheol Kim
Nuclear Today
Clean Energy Proposals are Chance for Nuclear
to have Rightful Place at Policy Table . . . . . . . . 430
John Shepherd
Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 375
AMNT 2018: Call for Papers . . . . . . . . . . . . . Insert
DAtF: Kernenergie in Zahlen 2017 . . . . . . . . . Insert
Contents

atw Vol. 62 (2017) | Issue 6 ı June
372
ABSTRACTS | ENGLISH
Q&A: Poland’s Progress on the Road to New
Nuclear
NucNet | Page 374
Conflicting reports have emerged from Poland
about plans for its first reactors. Professor Grzegorz
Wrochna of the Polish National Centre for Nuclear
Research says the programme is on track and a
business model is expected soon. The previous
government’s programme calls for 6 GW in two
locations. The government recently published a
strategy for responsible development which calls
for the nuclear programme to be sped up. However,
no capacity figures were included. The industry
should not be bound by a rigid number. In time,
maybe we will speak of 4 GW or 12 GW, but it will
depend on market needs and financial possibilities.
The biggest risks do not come from cancellation or
public opinion. They come from delays.
48 th Annual Meeting on Nuclear Technology
(AMNT 2017): Opening Address
Ralf Güldner | Page 378
The past twelve months in German nuclear energy
policy have been characterised mainly by legislative
clearing up work which has been pending since
the decision for an accelerated phase-out of nuclear
power in 2011. This applies particularly to the
reorganisation of financing in nuclear waste
management. The other major political work
package was the amendment to the Site Selection
Act. Our real challenge though is nuclear expertise.
This is important for research, for industry but
above all for the state itself. The question of
expertise covers the whole range of scientific and
technical knowledge relating to nuclear technology:
basic nuclear research, reactor safety research,
radiochemistry, radiological protection, nuclear
applications in medicine, industry and agriculture,
to mention but a few examples.
Monte-Carlo Based Comparison of the
Personal Dose for Emplacement Scenarios
of Spent Nuclear Fuel Casks in Generic Deep
Geological Repositories
Héctor Saurí Suárez, Bo Pang,
Frank Becker and Volker Metz | Page 384
In the operational phase of a deep geological
disposal facility for high-level nuclear waste, the
radiation field in the vicinity of a waste cask is
influenced by the backscattered radiation of the
surrounding walls of the emplacement drift. For a
comparison of disposal of spent nuclear fuel in
various host rocks, it is of interest to investigate the
influence of the surrounding materials on the
radiation field and the personal radiation exposure.
In this generic study individual dosimetry of
personnel involved in emplacement of casks with
spent nuclear fuel in drifts in rock salt and in a clay
formation was modelled.
The 15th German Atomic Energy Act
Amendment to the Implementation of the
EURATOM Nuclear Safety Directive
Christian Müller-Dehn | Page 391
The 15th German Atomic Energy Act Amendment
has now passed the parliamentary legislative
procedure with the decision of the Bundestag in the
third reading of 30 March 2017. The publication in
the Federal Law Gazette (Bundesgesetzblatt) is still
pending. The background of the amendment is the
addition to the Euratom safeguards directive
adopted by the European Council in July 2014. This
directive has to be implemented in the national
regulations of the EURATOM Member States.
However, since most of these supplements were
already standard in German atomic law, the
regulatory requirements for Germany were low.
This is also explicitly stated in the statement to the
act.
Retrofitting a Spent Fuel Pool Spray System
for Alternative Cooling as a Strategy for
Beyond Design Basis Events
Christoph Hartmann and Zoran Vujic | Page 392
Due to requirements for nuclear power plants to
withstand beyond design basis accidents, including
events such as happened in 2011 in the Fukushima
Daiichi Nuclear Power Plant in Japan, alternative
cooling of spent fuel is needed. Alternative spent
fuel cooling can be provided by a retrofitted spent
fuel pool spray system based on the AP1000 plant
design. As part of Krško Nuclear Power Plant’s
Safety Upgrade Program, Krško Nuclear Power
Plant decided on, and Westinghouse successfully
designed a retrofit of the AP1000® plant spent fuel
pool spray system to provide alternative spent fuel
cooling.
Cyber Security in Nuclear Power Plants
and Its Portability to Other Industrial
Infrastructures
Sébastien Champigny, Deeksha Gupta,
Venesa Watson and Karl Waedt | Page 396
Power generation increasingly relies on decentralised
and interconnected computerised systems.
Concepts like “Industrial Internet of Things” of the
Industrial Internet Consortium (IIC), and “Industry
4.0” find their way in this strategic industry. Risk
of targeted exploits of errors and vulnerabilities
increases with complexity, interconnectivity
and decentralization. Inherently stringent security
requirements and features make nuclear
computerised applications and systems a benchmark
for industrial counterparts seeking to hedge
against those risks. Consequently, this contribution
presents usual cyber security regulations and
practices for nuclear power plants. It shows how
nuclear cyber security can be ported and used in an
industrial context to protect critical infrastructures
against cyber-attacks and industrial espionage.
Validation of Spent Nuclear Fuel Nuclide
Composition Data Using Percentage
Differences and Detailed Analysis
Man Cheol Kim | Page 402
Nuclide composition data of spent nuclear fuels
are important in many nuclear engineering
applications. In reactor physics, nuclear reactor
design requires the nuclide composition and the
corresponding cross sections. In analyzing the
radiological health effects of a severe accident on
the public and the environment, the nuclide
composition in the reactor inventory is among the
important input data. Nuclide composition data
need to be provided to analyze the possible
environmental effects of a spent nuclear fuel
repository. They will also be the basis for identifying
the origin of unidentified spent nuclear fuels or
radioactive materials.
Reliability Analysis on Passive Residual Heat
Removal of AP1000 Based on Grey Model
Qi Shi, Zhou Tao, Muhammad Ali Shahzad,
Li Yu and Jiang Guangming | Page 408
It is common to base the design of passive systems
on the natural laws of physics, such as gravity, heat
conduction, inertia. For AP1000, a generation-III
reactor, such systems have an inherent safety
associated with them due to the simplicity of their
structures. However, there is a fairly large amount
of uncertainty in the operating conditions of these
passive safety systems. In some cases, a small
deviation in the design or operating conditions can
affect the function of the system. The reliability of
the passive residual heat removal is analysed.
Experimental Investigation of a Two-Phase
Closed Thermosyphon Assembly for Passive
Containment Cooling System
Kyung Ho Nam and Sang Nyung Kim | Page 413
After the Fukushima accident, increasing interest
has been raised in passive safety systems that
maintain the integrity of the containment building.
To improve the reliability and safety of nuclear
power plants, long-term passive cooling concepts
have been developed for advanced reactors.
In a previous study, the proposed design was
based on an ordinary cylindrical Two-Phase
Closed Thermosyphon (TPCT). The exact assembly
size and number of TPCTs should be elaborated
upon through accurate calculations based on
experiments. While the ultimate goal is to propose
an effective MPHP design for the PCCS and experimentally
verify its performance, a TPCT assembly
that was manufactured based on the conceptual
design in this paper was tested.
Displacement of Cryomodule in CADS
Injector II
Yuan Jiandong, Zhang Bin, Wang Fengfeng,
Wan Yuqin, Sun Guozhen, Yao Junjie,
Zhang Juihui and He Yuan | Page 418
As Cryomodule can easily reduce higher power
consumption and length of an accelerator and the
accelerator can be operated more continuously.
The Chinese academy of sciences institute of
modern physics is developing an accelerator driven
subcritical system (CADS) Injector II. Cryomodules
are extremely complex systems, and their design
optimization is strongly dependent on the
accelerator application for which they are intended.
Clean Energy Proposals are Chance
for Nuclear to Have Rightful Place at Policy
Table
John Shepherd | Page 430
Foratom, the Brussels based trade association
for the nuclear industry in Europe, published a
position paper on the European Commission’s
‘Clean Energy for All Europeans’ package of EU
legislative proposals. The proposals seek to improve
the functioning of the energy market and ensure all
energy technologies can compete on a level-playing
field without jeopardising climate and energy
targets. If Europe seeks to have a coherent and
inclusive energy policy, which encompasses all
lowcarbon contributors, nuclear must be allowed a
place at the policy table.
Abstracts | English

atw Vol. 62 (2017) | Issue 6 ı June
Fortschritte in Polen auf dem Weg
zur Nutzung der Kernenergie
NucNet | Seite 374
Zu den Bauplänen Polens für erste Kernkraftwerke
gibt es widersprüchliche Meldungen. Professor
Grzegorz Wrochna vom Polnischen Nationalen
Zentrum für Kernenergieforschung sieht das Programm
auf dem richtigen Weg und erwartet zeitnah
die Vorlage eines geeigneten Geschäftsmodells. Das
Programm der vorherigen polnischen Regierung
sah 6 GW Leistung an zwei Standorten vor. Kürzlich
stellte die jetzige Regierung eine Strategie für die
Entwicklung vor und betonte, dass das Nuklearprogramm
beschleunigt werden muss. Konkrete
Ausbauzahlen, an die die Industrie gebunden wird,
wurden nicht genannt. Der voraussichtliche Bedarf
an Kapazitäten liegt zwischen 4 und 12 GW. Markt
und Finanzierungsmöglichkeiten sollen diesen
bestimmen. Die größten Risiken liegen nicht in
einer Abkehr vom Programm oder der öffentlichen
Meinung, sondern von Verzögerungen.
48. Annual Meeting on Nuclear Technology
(AMNT 2017): Eröffnungsansprache
Ralf Güldner | Seite 378
In der deutschen Kernenergiepolitik waren die
vergangenen zwölf Monate vor allem geprägt
von gesetzgeberischen Arbeiten, die nach dem
beschleunigten Ausstieg aus der Kernenergie 2011
anstanden. Das gilt insbesondere für die
Neu ordnung der Finanzierung in der nuklearen
Ent sorgung. Das andere große politische Arbeitspaket
war die Novelle des Standortauswahl gesetzes.
Unsere eigentliche Herausforderung ist aber die
kerntechnische Kompetenz. Dies gilt für die
Forschung, die Industrie, aber vor allem auch für
den Staat selbst. Die Frage der Kompetenz betrifft
das gesamte Spektrum des wissenschaftlichen und
technischen Wissens um die Kerntechnik: die
grundlegende Kernforschung, die Reaktorsicherheitsforschung,
die Radiochemie, den Strahlenschutz,
kerntechnische Anwendungen in Medizin,
Industrie und Landwirtschaft etc.
Monte-Carlo basierter Vergleich der
Personendosis in Szenarien zur Einlagerung
von Behältern mit bestrahltem Kernbrennstoff
in generischen Tiefenlagern
Héctor Saurí Suárez, Bo Pang,
Frank Becker and Volker Metz | Seite 384
In der Betriebsphase eines Tiefenlagers für hochradioaktive
Abfälle wird das Strahlenfeld um einen
Lagerbehälter durch die Rückstreustrahlung von den
Wänden der Einlagerungsstrecken verändert. Daher
ist für einen Vergleich der Einlagerung von
abgebranntem Kernbrennstoff in verschiedenen
Wirtsgesteinen von Interesse, den Einfluss der unterschiedlichen
Wandungsmaterialien auf die Strahlenexposition
der dort Beschäftigten zu ermitteln. In
dieser generischen Studie wurde die individuelle
Dosimetrie von Beschäftigten bei Einlagerung von
Behältern mit abgebranntem Kernbrennstoff in
Steinsalz und einer Tonformation untersucht.
Die 15. AtG-Novelle zur Umsetzung
der EURATOM-Sicherheits-Richtlinie
Christian Müller-Dehn | Seite 391
Die 15. AtG-Novelle (AtG: Atomgesetz) hat das
parlamentarische Gesetzgebungsverfahren mit dem
Beschluss des Bundestages 30.3.2017 nunmehr
durchlaufen, harrt aber noch der Veröffentlichung
im Bundesgesetzblatt. Hintergrund aller Regelungen
sind die Ergänzungen der EURATOM- Sicherheits-
Richtlinie, die der Europäische Rat im Juli 2014
beschlossen hat und die bis spätestens August 2017
in den nationalen Regelungen der EURATOM-
Mitgliedsstaaten zu verankern sind. Da die meisten
dieser Ergänzungen jedoch bereits geltender
Standard im deutschen Atomrecht waren, waren die
für Deutschland umsetzungsbedürftigen Regelinhalte
gering. Dies wird ausdrücklich auch in der
Gesetzesbegründung festgehalten.
Nachrüstung eines Pool-Spraysystems eines
Brenn elementlagerbeckens als alternative
Strategie der Kühlung für auslegungsüberschreitende
Ereignisse
Christoph Hartmann und Zoran Vujic | Seite 392
Aufgrund von Anforderungen an Kernkraftwerke,
die über Auslegungsstörfälle hinausgehen, einschließlich
derer, wie sie im Jahr 2011 im Kernkraftwerk
Fukushima Daiichi in Japan auftraten, ist eine
alternative Kühlungsmethode von Brennelementlagerbecken
erforderlich. Diese alternative Kühlung
kann durch ein nachgerüstetes abgebranntes
Lagerbecken-Sprühsystem nach dem AP1000-
Anlagendesign bereitgestellt werden. Im Rahmen
des Sicherheits-Upgrade-Programms des Kernkraftwerks
Krško entschied sich der Betreiber für die
Nachrüstung mit dem Westinghouse-System des
AP1000®.
Cybersecurity in Kernkraftwerken und
ihre Anwendung in weiteren industriellen
Infrastrukturen
Sébastien Champigny, Deeksha Gupta,
Venesa Watson and Karl Waedt | Page 396
Stromerzeugung ist verstärkt auf dezentralisierte
und vernetzte Rechensysteme angewiesen. Begriffe
wie „Industrial Internet of Things“ des Industrial
Internet Consortium (IIC) und „Industrie 4.0“
bahnen sich heute ihren Weg auch in diese
bedeutende Industriebranche. Die Risiken einer
gezielten Ausnutzung von Fehlern und Schwachstellen
nehmen mit der Komplexität, mit dem
Vernetzungsgrad und mit der Dezentralisierung zu.
Die inhärent strengen Sicherheitsanforderungen
der Kernenergiebranche und die langjährige
Berücksichtigung von Anforderungen im Bereich
Cybersecurity in der Entwicklung von Produkten
und in projektbegleitenden Maßnahmen machen
sie zum Gold-Standard der Risikovorbeugung.
Die gewonnenen Erkenntnisse können für die
Ableitung angepasster Sicherheitsvorkehrungen
anderer Branchen dienen. Aus diesem Blickwinkel
heraus wird das Thema Cybersecurity betrachtet.
Der Artikel zeigt gängige Regularien und die Vorgehensweisen
zum Schutz vor Cyberangriffen in
Kernkraftwerken, sowie auch deren zahlreichen
Übertragungsmöglichkeiten auf andere kritische
Infrastrukturen, um sie gegen Cyberangriffe und
Industriespionage zu wappnen.
Validierung der Nuklid-Zusammensetzung
von abgebranntem Kernbrennstoff unter
Verwendung von Prozentsatzdifferenzen
und einer detaillierten Analyse
Mann Cheol Kim | Seite 402
Die Informationen zur Nuklid-Zusammensetzung
von abgebranntem Kernbrennstoff sind in vielen
Anwendungen wichtig. In der Reaktorphysik
erfordert die Konstruktion des Kernreaktors Informationen
zur Nuklidzusammensetzung. Für die
Analyse der radiologischen Folgen eines schweren
Unfalls ist die Nuklidzusammensetzung eine
wichtige Eingangsgröße. Ebenso ist sie eine
Einganggröße und Grundlage für die Auslegung
eines geologischen Endlagers. Ein weiteres Feld
ist die Identifizierung der Herkunft von nicht
identifizierten verbrauchten Kernbrennstoffen oder
radioaktiven Stoffen.
Zuverlässigkeitsanalyse für die passive
Restwärmeabfuhr eines AP1000-Reaktors
basierend auf einem „Grey model“
Qi Shi, Zhou Tao, Muhammad Ali Shahzad,
Li Yu und Jiang Guangming | Seite 408
Passive Systeme stützen sich auf Naturgesetze
der Physik, wie z.B. Schwerkraft, Wärmeleitung
oder Trägheit. Beim Generation-III-Reaktor
AP1000, einem Generation-III-Reaktor, bieten
solche Systeme, verbunden mit einfachen Designstrukturen,
inhärente Sicherheitsfunktionen. Unter
Betriebsbedingungen werden für passive Sicherheitssysteme
Unsicherheiten angegeben. In einigen
wenigen Fällen kann eine geringe Abweichung von
Konstruktions- oder Betriebsbedingungen die
Funktion des Systems beeinträchtigen. Die Zuverlässigkeit
eines solchen System wird analysiert.
Experimentelle Untersuchung einer
zweiphasigen geschlossenen Thermosyphon-
Baugruppe für ein passives
Containment-Kühlsystem
Kyung Ho Nam und Sang Nyung Kim | Seite 413
Nach dem Unfall von Fukushima stieg da Interesse
an passiven Sicherheitssystemen, die die Integrität
des Containments aufrechterhalten. Zur weiteren
Erhöhung von Zuverlässigkeit und Sicherheit von
Kernkraftwerken, wurden passive Kühlkonzepte für
die Langfristkühlung fortgeschrittener Reaktoren
entwickelt. In einer früheren Studie basierte ein
vorgeschlagenes Design auf einem einfachen
zylindrischen zweiphasigen geschlossenen Thermosyphon
(TPCT). Baugröße und Anzahl der erforderlichen.
Eine neue TPCT-Baugruppe wurde getestet,
die auf der Grundlage des hier vorgestellten Designs
entwickelt wurde.
Auslegung eines Kryomoduls für den CADS
Injektor II
Yuan Jiandong, Zhang Bin, Wang Fengfeng,
Wan Yuqin, Sun Guozhen, Yao Junjie,
Zhang Juihui und He Yuan | Seite 418
Durch den Einsatz von Cryomodule können die
Leistungsaufnahme von Beschleunigern und ihre
Länge reduziert sowie ein kontinuierlicher Betrieb
unterstützt werden. An der Chinesischen Akademie
der Wissenschaften, Institut für moderne Physik
wird ein beschleunigergetriebenes subkritisches
System (CADS) Injektor II entwickelt, bei dem
Cryomodule zum Einsatz kommen. Cryomodule
sind äußerst komplexe Systeme und ihre Designoptimierung
hängt stark von der Beschleunigeranwendung
ab, für die sie bestimmt sind.
Vorschläge für „Saubere Energie“
sind eine Chance für die Kernenergie
John Shepherd | Seite 430
Foratom, der Brüsseler Verband der kerntechnischen
Industrie in Europa, hat ein Positionspapier
zur EU-Gesetzgebung „Clean Energy for All
Europea ns“ veröffentlicht. Die Vorschläge zielen
darauf ab, das Funktionieren des Energiemarktes zu
verbessern und sicherzustellen, dass alle Energietechnologien
gleichermaßen verwendet werden
können, ohne dass die Klima- und Energieziele
gefährdet werden. Wenn Europa eine kohärente
und integrative Energiepolitik anstrebt, die alle
kohlenstoffarmen Optionen umfasst, muss die
Kernenergie mit Berücksichtigung finden.
373
ABSTRACTS | GERMAN
Abstracts | German

atw Vol. 62 (2017) | Issue 6 ı June
374
INSIDE NUCLEAR WITH NUCNET
Q&A: Poland’s Progress on the Road
to New Nuclear
NucNet
Conflicting reports have emerged from Poland about plans for its first reactors, but Professor Grzegorz
Wrochna of the Polish National Centre for Nuclear Research says the programme is on track and a business
model is expected soon.
NucNet: There have been various media reports
in Poland about the country’s nuclear new-build
project, with some saying it has been postponed.
Could you tell us more about the current project status?
Grzegorz Wrochna: If you depend on the media you will
get a confused picture. The situation is rather straightforward.
There is a delay. PGE EJ1, the company responsible
for building the first nuclear plant in Poland, announced
that a tender would be started in December 2015, but this
has not happened. Based on this delay, some media has
speculated that the programme is on hold, but that is not
true. It is just the tendering procedure which has been
suspended while all the other work – site surveys, preparation
of the nuclear regulatory body, changing the nuclear
law – is all going ahead and going well.
The programme prepared by the previous government [in
office from November 2011 until October 2015] is still valid.
The cabinet accepted this programme, but asked the ministry
of energy to present a new schedule and business model
by spring 2017. So I hope soon we will have a plan ready to
be shown to the government by the minister of energy.
NucNet: Have any of the conditions for the nuclear
programme changed?
Grzegorz Wrochna: The global economic situation has
changed. When the previous government prepared the
nuclear programme, it was difficult to get financing for this
kind of investment. Therefore, the condition was that the
tender should concern all elements, including reactor
design, construction, the first few years of operation, fuel
and, finally, financing, which was the most important part.
The organisations pitching for the contract would be asked
to present everything, including the financing.
Now conditions are different. The cost of borrowing
money has decreased and it is easier to find loans at low
interest rates. The new government decided to split the
tender into a technical part and a financial part, each to be
considered separately. The detailed model has not been
decided, but this, most probably, will be the new direction.
NucNet: What about the schedule?
Grzegorz Wrochna: The original plan assumed the project
would take 10 years from the investment decision to the actual
operation of the first reactor. This was based on International
Atomic Energy Agency (IAEA) documents, which were
in turn based on the experience of other nuclear countries.
Many countries have managed to build nuclear power
units in 10 years. In Poland, it turned out to be impossible
under existing Polish laws, which did not allow many of
the regulatory processes to run in parallel to each other. In
other words, to get each consecutive decision, we first
needed to get feedback from authorities on previous
decisions. When PGE EJ1 and the NCBJ (Polish National
Centre for Nuclear Research) recalculated the schedule, it
turned out that Poland would need 16 years to have its
nuclear programme operational, six years longer than
originally anticipated. The media took this as a delay, but
rather it was just a ‘procedural discovery’.
We can now aim for commercial operation some time
around 2028, but have to wait for the ministry of energy to
officially present its schedule.
NucNet: What is the government’s vision of the country’s
energy mix? What roles do coal and nuclear play?
Grzegorz Wrochna: In the not so distant past Poland was
100 % independent concerning its sources of electric supply,
but this was based almost entirely on coal. More than 90 %
of electricity was produced from coal. This has changed a
little, with the increased but still limited participation of
renewables and gas. Coal will continue to dominate the
energy landscape for many years, because it is our domestic
resource, essential for our security of supply.
But this is not enough, because we hope the Polish
economy will grow along with the demand for energy. And
we will not have any means other than nuclear and energy
imports to meet this growing demand. If we want to
maintain our energy independence, nuclear remains the
only viable option. This does not mean there is competition
between nuclear and coal. We do not have to choose
between the two. We still need to build new coal-fired
plants to replace old, inefficient ones. But the investment
timeframe for coal-fired plants is a few years, while for
nuclear it will be more than 10 years. Even if the government
decides overnight to go for 100v% nuclear, nothing
will change for coal for a few decades.
NucNet: Is the introduction of nuclear energy a politicised
issue in Poland or is there a sense of consensus among
different parties and stakeholders?
Grzegorz Wrochna: There is no consensus between the
political parties. But there is a consensual understanding
that our energy mix is too dependent on domestic resources.
We do not have much wind or solar potential in Poland,
hydro is being used but cannot be expanded much further,
and we have some domestic gas, but it is far from sufficient
to meet our needs. The only sufficient domestic resources
are coal and then nuclear. We have no other choice.
NucNet: How does the Polish public see the new-build
programme? Are there concerns about safety?
Grzegorz Wrochna: Public opinion is reasonably positive
about nuclear. The most recent polls showed more than
60 % in favour of nuclear in Poland and, surprisingly even
for us, about 48 % said they would have a reactor close to
their homes. People see this as an opportunity for economic
prosperity.
I think we, the scientists, have done a good job telling
the public about nuclear energy. The way we communicated
what happened in Fukushima was very important.
People are now well aware that we do not have tsunamis or
earthquakes of these magnitudes in Poland.
NucNet: Poland has pledged to build four to five units with
combined output of 6 GW, by the mid-2030s. Is this realistic?
Grzegorz Wrochna: The previous government’s programme
calls for 6 GW in two locations. The number of reactors
per site would have depended on the technology choice.
The government recently published a strategy for
responsible development which calls for the nuclear
programme to be sped up. However, no capacity figures
were included. The industry should not be bound by a rigid
number. In time, maybe we will speak of 4 GW or 12 GW,
but it will depend on market needs and financial possibilities.
The first reactor will be the most challenging. I believe
Inside Nuclear with NucNet
Q&A: Poland’s Progress on the Road to New Nuclear ı NucNet

atw Vol. 62 (2017) | Issue 6 ı June
it is possible to complete this first unit by 2027-2029 and
then we could go for a total of 6 GW by the early-2030s.
NucNet: The Polish Nuclear Roadmap includes a plan to
deploy a high-temperature gas-cooled reactor (HTR). Can
you elaborate on these plans.
Grzegorz Wrochna: The Polish nuclear programme is in
nature a light-water reactor (LWR) investment project.
The Polish industry will be part of the supply chain, but not
much will be gained in terms of intellectual property and
technological know-how. Fundamentally, we will order
existing reactor designs, pay for them and build them.
But once we have spent so much money on building a
nuclear plant it might be better to spend a little bit extra
and make even greater gains for the economy. We could
invest in R&D, which would have lasting benefits for us.
Poland has an extensive chemical industry, which
consumes a lot of heat, produced from coal or imported
natural gas. If we want to become more independent, we
need an alternative source of heat for industry. And it is
here that high temperature reactor HTR nuclear technology
could play a big part.
HTRs produce high-temperature steam at about 550 °C.
We could safely and easily replace an old gas or coal-fired
boiler at a chemical plant with an HTR which would produce
the same amount of heat. We are talking about 6 GW, but
this time in heat rather than in electricity, distributed among
10 or more sites. This is a parallel programme, but there are
obvious synergies between the two – supply chain, regulation,
and the scientific part. We would really like to see the
HTR programme as a spin-off from the main LWR programme.
What we plan is to build about 10 to 20 HTRs in
Poland by 2050. We have the capacity for this. The first
should be in operation by 2031-2032. The need of Europe
we estimate about 100-200 of such reactors.
The HTR programme is also mentioned in government
policy. Last year the ministry of energy established a
committee for HTR deployment. That committee is
preparing an intermediate report and this year we are
planning to establish a company to start designing an HTR,
based on international experience. Preparation for the first
demonstrator will be supported by the Gemini+ initiative,
which is being funded by Euratom. Within the framework
of the € 4 million project, NCBJ scientists will be coordinating
international preliminary works aimed at
implementing HTRs. This could eventually help the first
European HTR become a reality in Poland.
NucNet: Finally, what are the challenges and risks for the
new-build programme?
Grzegorz Wrochna: The biggest risks do not come from
cancellation or public opinion. They come from delays. In
Europe, all major investments, power stations and other
infrastructure, experience cost overruns and take longer
than expected. In the past, the designs were several
thousand pages long and the investment agreements a few
pages. Today it is the opposite – designs are general and
often standardised, while investment agreements have
become long and cumbersome. Nuclear is no exception.
This is a malaise that has affected all major investments
in Europe. I hope the time spent preparing the nuclear
programme in Poland will help avoid delays.
Author
NucNet
The Independent Global Nuclear News Agency
Editor responsible for this story: Kamen Kraev
Avenue des Arts 56
1000 Brussels, Belgium
www.nucnet.org
INSIDE NUCLEAR WITH NUCNET 375
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Q&A: Poland’s Progress on the Road to New Nuclear ı NucNet

atw Vol. 62 (2017) | Issue 6 ı June
376
CALENDAR
Calendar
2017
04.06.-07.06.2017
37 th Annual Canadian Nuclear Society Conference.
Niagara Falls, ON, Canada, www.cns-snc.ca
06.06.-09.06.2017
International Conference on Topical Issues in
Nuclear Installation Safety: Safety Demonstration
of Advanced Water Cooled Nuclear Power Plants.
Vienna, Austria. International Atomic Energy Agency
(IAEA), www.iaea.org
11.06.-17.06.2017
ENYGF 2017 – European Nuclear Young
Generation Forum. Manchester, United Kingdom,
ENS YGN, www.enygf.org
11.06.-15.06.2017
ANS Annual Meeting. 10 th International Topical
Meeting on Nuclear Plant Instrumentation,
Control and Human Machine Interface Technology
(embedded topical meeting). San Francisco, CA,
USA, American Nuclear Society (ANS), www.ans.org
13.06.-14.06.2017
Journees thematiques fusion — Journees
thematiques fusion AFF CCS. Cadarche, France,
Commission Cryogenie et Supraconductive (AFF),
affccs.grenoble.fr
19.06.-21.06.2017
ATOMEXPO 2017. Moscow, Russia, 2017.atomexpo.ru
19.06.-20.06.2017
EURELECTRIC Annual Convention &
Conference 2017. Lisbon, Portugal, Eurelectric,
www.eurelectric.org
26.06.-30.06.2017
Third PETRUS-ANNETTE PhD and Early-Stage
Researchers Conference 2017 – Radioactive
Waste Management and Disposal. Lisboa,
Portugal, Petrus and Annette (Euratom),
petrus-annette-2017.strikingly.com/
27.06.-28.06.2017
New Nuclear Build – NNB 2017. London, UK,
Nuclear Industry Association, www.niauk.org
27.06.-29.06.2017
Power-Gen Europe 2017. Cologne, Germany,
PennWell, www.powergeneurope.com
26.06.-30.06.2017
International Conference on Fast Reactors and
Related Fuel Cycles. Yekaterinburg, Russia, International
Atomic Energy Agency (IAEA), www.iaea.org
27.06.-04.08.2017
World Nuclear University Summer Institute.
Uppsala, Sweden, World Nuclear Association,
www.world-nuclear.org
31.07.-04.08.2017
AccApp'17 – 13 th International Topical Meeting
on Nuclear Applications of Accelerators. Quebec
City, Quebec, Canada, American Nuclear Society
(ANS), www.ans.org, ccapp17.org
06.08.-09.08.2017
Utility Working Conference and Vendor
Technology Expo – The Nuclear Option – Clean,
Safe, Reliable & Affordable. Amelias Island, FL,
USA, American Nuclear Society (ANS), uwc.ans.org
20.08.-25.08.2017
24 th International Conference on Structural
Mechanics in Reactor Technology. Busan, Korea,
SMIRT Organisation Committee, www.smirt24.org
23.08.-01.09.2017
Frédéric Joliot/Otto Hahn (FJOH) Summer School
FJOH-2017 – Uncertainties in nuclear reactor
systems analysis: Improving understanding,
confidence and quantification. Karlsruhe, Germany,
Nuclear Energy Division of Commissariat à l’énergie
atomique et aux énergies alternatives (CEA) and Karlsruher
Institut für Technologie (KIT), www.fjohss.eu
27.08.-02.09.2017
INCC – 5 th International Nuclear Chemistry
Congress. Gothenburg, Sweden. Chalmers
University of Technology Division of Nuclear
Chemistry (Organisation), www.chalmers.se
03.09.-08.09.2017
NURETH 17 – 17 th International Topical Meeting
on Nuclear Reactor Thermal Hydraulics. Xi’an,
China, nureth17.com
03.09.-06.09.2017
15 th IAEE European Conference Heading Towards
Sustainability Energy Systems: by Evolution
or Revolution? Vienna, Austria, AAEE/IAEE,
www.iaee.org
10.09.-14.09.2017
2017 Water Reactor Fuel Performance Meeting.
Jeju Island, Korea, Korean Nuclear Society, the
Atomic Energy Society of Japan, the Chinese Nuclear
Society, the American Nuclear Society and the
European Nuclear Society, wrfpm2017.org
10.09.-15.09.2017
2017 Nuclear Criticality Safety Division Topical.
Carlsbad, New Mexico, USA. American Nuclear
Society (ANS), www.ans.org
11.09.-14.09.2017
Nuclear Energy in New Europe – NENE 2017.
Bled, Slovenia, Nuclear Society of Slovenia,
www.nss.si/nene2017
13.09.-14.09.2017
VGB CONGRESS 2017 – Generation in Competition.
Essen, Germany, VGB PowerTech e.V., www.vgb.org
13.09.-15.09.2017
World Nuclear Association Symposium 2017.
London, United Kingdom, World Nuclear Association
(WNA), www.world-nuclear.org
17.09.-20.09.2017
2 nd International CNS Conference on Fire Safety
and Emergency Preparedness in the Nuclear
Industry. Toronto, ON, Canada, Canadian Nuclear
Society (CNS), www.cns-snc.ca
18.09.-22.09.2017
61 st IAEA General Conference. Vienna, Austria,
Inter national Atomic Energy Agency (IAEA),
www.iaea.org
24.09.-28.09.2017
PSA 2017 – 2017 International Topical Meeting
on Probabilistic Safety Assessment and Analysis.
Pittsburgh, Pennsylvania, USA, American Nuclear
Society (ANS), www.ans.org, psa.ans.org
01.10.-04.10.2017
11 th International Conference on CANDU
Maintenance and Nuclear Component. Toronto,
ON, Canada, Canadian Nuclear Society (CNS),
www.cns-snc.ca
01.10.-04.10.2017
SIEN 2017 – International Symposium for Nuclear
Energy. Bucharest, Romania, www.sien.ro
04.10.-05.10.2017
Fire Safety in Nuclear Power Plants. Bruges,
Belgium, Bel V, Gesellschaft für Anlagen- und
Reaktor sicherheit (GRS) gGmbH, Bundesamt
für kerntechnische Entsorgungssicherheit (BfE),
www.belv.be, www.grs.de
10.10.-12.10.2017
4 th International Symposium on the System of
Radiological Protection. Paris, France, IRSN,
icrp-erpw2017.com
17.10.-20.10.2017
27 th Atomic Energy Research (AER) Symposium.
Munich, Germany, Contact: Gesellschaft für
Anlagen- und Reaktorsicherheit (GRS) gGmbH,
www.grs.de
17.10.-18.10.2017
49. Kraftwerkstechnisches Kolloquium. Dresden,
Germany, Technische Universität Dresden,
www.kraftwerkskolloqium.de
21.10.-28.10.2017
IEEE Nuclear Science Symposium and Medical
Imaging Conference. Atlanta, Georgia, USA, IEEE,
www.nss-mic.org
23.10.-28.10.2017
Fourth International Conference on Nuclear Power
Plant Life Management. Lyon, France, International
Atomic Energy Agency (IAEA), www.iaea.org
07.11.-09.11.2017
10 th International Symposium Release of
Radioactive Materials Requirements for
Exemption and Clearance. Berlin, Germany, TÜV
Nord Akademie, www.tuev-nord.de/tk-rrm
25.10.-26.10.2017
Chemistry in Power Plants. Koblenz, Germany,
VGB PowerTech e.V., www.vgb.org
29.10.-02.11.2017
2017 ANS Winter Meeting and Nuclear
Technology Expo. Washington, DC, USA, American
Nuclear Society (ANS), www.ans.org
26.11.-30.11.2017
International Symposium on Future I&C for Nuclear
Power Plants. Gyeongiu, Korea, www.isofic.org
27.11.-30.11.2017
ICOND 2017 – International Conference on
Nuclear Decommissioning. Aachen, Germany,
Aachen Institute for Nuclear Training GmbH,
www.icond.de
01.12.-02.12.2017
ThermAc 2016 – Aquatic Actinide Chemistry and
Thermodynamics at elevated Temperatures.
Dresden, Germany, HZDR, www.hzdr.de
05.12.-07.12.2017
POWER-GEN International. Las Vegas, NV, USA.
PennWell, www.power-gen.com
2018
26.02.-01.03.2018
Nuclear and Emerging Technologies for Space
2018. Las Vages, NV, USA. American Nuclear
Society (ANS), www.ans.org
08.04.-11.04.2018
International Congress on Advances in Nuclear
Power Plants – ICAPP 18. Charlotte, NC, USA,
American Nuclear Society (ANS), www.ans.org
22.04.-26.04.2018
Reactor Physics Paving the Way Towords More
Efficient Systems – PHYSOR 2018. Cancun, Mexico,
www.physor2018.mx
29.05.-30.05.2018
49 th Annual Meeting on Nuclear Technology
AMNT 2018 | 49. Jahrestagung Kerntechnik.
Berlin, Germany, DAtF and KTG,
www.nucleartech-meeting.com – Save the Date
17.09.-20.09.2017
FONTEVRAUD 9. Avignon, France, Société
Française d’Energie Nucléaire (SFEN),
www.sfen-fontevraud9.org
30.09.-05.10.2018
Pacific Nuclear Basin Conferences – PBNC 2018.
San Francisco, CA, USA, American Nuclear
Society (ANS), www.ans.org
14.10.-18.10.2018
12 th International Topical Meeting on Nuclear
Reactor Thermal-Hydraulics, Operation and
Safety – NUTHOS-12. Qingdao, China
Calendar

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Notes
Press Release 16 May 2017
AMNT: President Warns Against
Loss of Nuclear Expertise
In his speech at the 48 th Annual Meeting on Nuclear
Technology (AMNT 2017), the President of the DAtF
( German Atomic Forum), Dr. Ralf Güldner, warned against
the loss of nuclear expertise and of nuclear research and
industry in Germany. Güldner said that the challenge for
nuclear technology in Germany lay in the long-term
provision of expertise. He said this applied to research,
industry and the state itself and that it was premised on
using this expertise, for example, in industrial projects for
upgrading plants or in development. He continued that
the international demand for German safety expertise,
which enjoys an excellent reputation, contributed
significantly to maintaining it. He warned that the decision
to phase out nuclear energy must not constitute a risk of
losing this expertise.
In his speech, Güldner said, “Nuclear safety research
forms the basis for expertise in safety issues in which
Germany intends to play a long-term role and exert its
influence. If we want to continue participating in the international
discussion of safety standards, then continuity in
safety research is absolutely essential.” He complained
that, especially in the case of innovative topics, reactor
safety research was now being regarded as superfluous
and that many federal state governments no longer wanted
anything to do with it. He said that university chairs were
not being refilled and universities and research institutes
were shaped as to withdraw from areas that were not
assigned to waste management or dismantling.
Güldner therefore suggested a new beginning for safety
research, “The solution might lie in a new Centre of
Expertise for Nuclear Safety where current issues could be
dealt with without the burden of past conflicts. Here, it
may be possible to pool capacities, to network research,
state and industry and to create an attractive hub for our
international collaboration.”
He drew attention to the fact that nuclear energy would
continue to contribute to the security of the power supply
in Germany. He indicated that the political consensus on
the transformation of the German energy sector would
also be implemented by operating the plants until 2022.
And with regard to this he stated, “There must therefore be
no factually unfounded complications to operation of the
nuclear power plants in the last few years.” Güldner
pointed out that the facilities for uranium enrichment and
fuel assembly production were explicitly excluded from
the phase out of nuclear energy use and he rejected any
efforts to expand the phase out.
DATF EDITORIAL NOTES
377
New Brochure
The DAtF has published the new edition of
its nuclear power statistics flyer with status April 2017:
“Kernenergie in Zahlen 2017”
3 The flyer can be downloaded and ordered
at kernenergie.de under the headings
Downloads and Shop.
For further details
please contact:
Nicolas Wendler
DAtF
Robert-Koch-Platz 4
10115 Berlin
Germany
E-mail: presse@
kernenergie.de
www.kernenergie.de
DAtF Notes

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AMNT 2017
48 th Annual Meeting
on Nuclear Technology (AMNT 2017):
Opening Address
16 to 17 May 2017, Berlin
Ralf Güldner
Ladies and Gentlemen, Welcome to our 48 th Annual Meeting on Nuclear Technology in Berlin on behalf of the
DAtF and the German Nuclear Society. It is my pleasure to see you again in Berlin. As in other years, we offer a
comprehensive program, providing insights into many aspects of nuclear technology and contributing to the
international exchange of knowledge and experience in industry, research, politics and administration.
Ladies and Gentlemen,
The content of our Meeting is already reflected in the
Plenary Session with its fixed topics relating to politics,
industry, expertise, communication and waste management.
In the section on politics, Steffen Kanitz, rapporteur
for nuclear energy of the CDU/CSU parliamentary group in
the Bundestag, will provide us with an overview of a turbulent
year for German nuclear energy policy. Guido Knott,
Chairman of the Board of Management of PreussenElektra
GmbH, will give us an understanding of the challenges
involved in operating nuclear power plants cost-effectively
in Germany. We are looking forward to the panel discussion
on dismantling and I would also like to draw your attention
to our workshop on the preservation of skills after the lunch
break today and tomorrow morning.
Special thanks are due to our partners in the exhibition
and for the sponsoring without which our meeting would
not even be possible. The exhibition provides you with the
opportunity to make personal contact with a large number
of companies and organisations in our industry with the
chance for a direct exchange of ideas and information.
We have further increased the number of international
partners involved. I want to draw your attention to the
Czech pavilion and also the UK’s pavilion. Our British
colleagues are facing historic decisions in their own
country and in respect of the future relationship with
Europe. We hope that solutions will be found during the
Brexit negotiations that enable constructive cooperation in
nuclear technology to continue in the future. This applies
not least to the new British construction projects.
Upheavals, new beginnings and the travails of
everyday business
The past twelve months in German nuclear energy policy
have been characterised mainly by, what I would call, late
legislative clearing up work which has been pending since
the decision for an accelerated phase-out of nuclear power
in 2011. This applies particularly to the reorganisation of
financing in nuclear waste management where a whole
legislative package has been used to implement a change
of system in many areas. This process is not yet quite
complete. The laws themselves have not yet entered into
force due to being examined for conformity with EU law,
and the contractual arrangement sought between the
nuclear power plant operators and the government has not
yet been signed.
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Of course, there is criticism of the arrangements.
• On the one hand there are fundamental reservations
regarding limitation of liability.
• For the operators, however, the high risk premium on
the waste management costs represents an additional
burden that is unexpected and hard to bear and
which is now likely to increase yet gain in the wake of
recalculations.
Overall, however, the reorganisation in waste management
will satisfy the conditions of the phase-out. A
situation with permanent separation between responsibility
for action and financing, potentially unlimited
secondary liability and a ban on using nuclear energy
could not have existed in the long run.
selection step, the localisation of subareas on the white
map, actually be completed by 2021 as is currently the
aim? The division and clear definition of tasks between BfE
(Federal Office for the Safety of Nuclear Waste Management)
and BGE (Federal Company for Final Disposal) is
another issue. It applies particularly to final repository
research which now has many new tasks. It has not yet
been specified who will be responsible for final disposal
research in the future.
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AMNT 2017
The other major political work package – and Mr Kanitz
will report on this in detail very shortly – was the
amendment to the Site Selection Act (StandAG). Although
political agreement was reached in spring 2013 on the
search for a new site for a final repository for high active
waste, in many details the law was still poorly conceived
and left the Final Repository Commission with a number of
unanswered questions along the way. Dr. Bernhard Fischer
and Professor Gerd Jäger called on our industry’s expertise
while working on this constructively and with tremendous
dedication. As part of the practical implementation, transfer
of the DBE (German Company for the Construction and
Operation of Waste Repositories) to the government was
completed here in Berlin yesterday. As an industry we have
contributed to describing the path for a solution in the
search for a new final repository. Now it is the politicians’
task to implement the set framework consistently.
Despite everything that has been achieved – here
we should also mention reorganisation of the regulatory
and institutional structure for final disposal, the 15 th
amendment to the Atomic Energy Act and the first
consolidated Radiological Protection Act in German legal
history – there is still work to be done.
Reorganisation in waste management also includes the
transfer of responsibility for interim storage to the state.
This is an even bigger change to the system than that for
final disposal but it is not so much in the public eye. A
whole series of operational challenges will arise when the
operational responsibility changes. This change was set in
motion when the Federal Company for Interim Storage
was set up and the aim, for the central interim storage
facilities, is for it to be completed during the course of this
year, for the site-based high-level interim storage facilities
early in 2019 and for the LLW/ILW storage facilities a year
later. The first steps have been taken and the choice of
Essen as the company’s headquarters will have a positive
effect, particularly on preserving the necessary skills, as a
result of GNS employees transferring over. Together, the
nuclear power plant operators and the GNS are handing
over a well-functioning system in which high safety
requirements are applicable. They are thus making an
important contribution to the reorganisation of responsibility
in nuclear waste management.
Now we come to the practical test of implementing the
Site Selection Act. When will the Federal Company for
Final Disposal, as the project developer, be capable of
working operationally? What time frame must we
realistically assume for the whole process? Will the first
Due to the concentration on high active waste, another
waste management issue has faded somewhat into the
background: What exactly is happening with the Konrad
facility? Are the plans for completing it by 2022 still valid?
When will regular operation actually start? How is the outflow
from the interim storage facilities to be prioritised?
These questions are important for any region throughout
Germany that has an interim storage facility, a state
collecting facility or a dismantling project. By bundling the
interim storage facilities in a federally-owned company,
I see opportunities for bringing more common sense to the
discussions and accelerating the processes.
In the comments of the Federal Court of Auditors for
2016, there is criticism that the Federal Government has
not adequately exercised supervision of the Konrad project
over the years. It recommends using the reorganisation of
tasks, which is welcomed by the Federal Court of Auditors,
to document the current situation, to make contractual
agreements with the BGE and to implement closer
monitoring. These considerations sound reasonable and
the ongoing restructuring provides an excellent opportunity
to get such project management off the ground; it
could be the culmination, so to speak, of the many reforms
in this legislative period.
At the same time, of course, we have the operation of
the nuclear power plants which we will safely continue
with and which we would also like to continue costeffectively.
Last January showed yet again that, particularly
during the so-called “dark doldrums”, grid operators and
reserve capacities are gradually reaching their limits. The
reserve capacity requirement of 10,400 MW now specified
by the Federal Network Agency for the coming winter
speaks for itself.
There is political consensus on the phased exit from
nuclear energy which we will implement by 2022 with our
expertise and also by safe operation. So there must be no
factually unfounded complications to the operation of the
nuclear power plants in the last few years.
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AMNT 2017
Ladies and Gentlemen,
The major issues of the future for nuclear technology in
Germany are dismantling on the one hand and nuclear
expertise on the other hand. These questions affect us all
and are long-term issues.
Dismantling by consensus
Dismantling is on the right track. The first decommissioning
and dismantling licences within the scope of
phasing out nuclear energy have been issued – for Isar 1,
Neckarwestheim 1, Biblis and Philippsburg 1. Important
preliminary work has been carried out at all the sites; they
are free of fuel or work is ongoing to ensure this. It is
important here that the flask and storage licences still
outstanding are issued on time. However, it is only in the
coming years that the considerable breadth of the projects
will become apparent.
In recent years, when requesting factual information
about dismantling and when consulting with citizens,
cooperation with the authorities has been good and the
support of politicians has been helpful. It is important for
speedy dismantling to maintain the consensus which now
exists between state and operators and to push the projects
forward efficiently on this basis. I have little sympathy here
with the traditional adversaries of nuclear energy who for
ideological reasons are now fighting the dismantling
process as well. In Germany dismantling is being carried
out in compliance with the highest safety standards and,
just like the construction and operation of a nuclear power
plant, it is subject to constant inspection by the authorities
and their experts.
Preserving and developing nuclear expertise
Our real challenge though is nuclear expertise. This is
important for research, for industry but above all for the
state itself. Many people may simply not be aware of this.
The topic of preserving and building up nuclear
expertise by shifting operational responsibility to federallyowned
companies will gain relevance particularly in the
waste management sector. Taking into account all the
authorities and public companies that operate in the waste
management sector, we could soon be talking about up to
4,000 employees. Together with the civil servants and
government employees in other areas of nuclear technology,
in expert appraisal and in research, it may be assumed
that in the future at least a sixth of the more than 30,000
employees in the industry will be assigned to the public
sector. In the long term, this will require appropriate
training of skilled staff and targeted human resources
planning. It can only be successful if there are positive
prospects for young people who employers would like to
win over for the important work ahead. It also needs to
include appropriate public discussion of the subject.
The question of expertise covers the whole range of
scientific and technical knowledge relating to nuclear
technology: basic nuclear research, reactor safety research,
radiochemistry, radiological protection, nuclear applications
in medicine, industry and agriculture, to mention but
a few examples.
Let’s take reactor safety research which is closely linked
to the operation of nuclear power plants. Reactor development
in particular is now subject to the accusation of being
redundant; sometimes it is regarded as outmoded or even
illegitimate. Nuclear safety research, however, forms the
basis for expertise in safety issues in which Germany has
stated its intention to play a long-term role and exert
its influence. If we want to continue participating in
the international discussion of safety standards, then
continuity in safety research is absolutely essential.
Our nuclear expertise, however, can only develop in
collaboration with scientifically attractive partners in
other countries. To win them over for this purpose requires
appropriate facilities and experts who are able to offer
added scientific value. This applies to all topics, especially
innovations and new design concepts. After all, we need to
be able to knowledgeably have a say too. Consider, for
example, a development in fuel assemblies, such as that
which Seth Grae, CEO of the Lightbridge Corporation from
the USA, will be presenting later. In the long run, scepticism
about research or even a ban on research has never done
any industrialised country any good.
In practice, however, we see that teaching and research
are being thinned out, that university chairs are not being
refilled and, under political pressure or for image reasons
and in a spirit of anticipatory obedience, whole institutes
are withdrawing from those areas that are not assigned to
waste management or dismantling.
Centre of Expertise for Nuclear Safety?
The question here is: what can we do? On the one hand,
the Federal Government wants and needs to access the
appropriate expertise and it also has the funds for this. On
the other hand, many federal state governments want
nothing more to do with the subject and are thus shaping
the orientation of universities and research institutes. The
solution might lie in a new Centre of Expertise for Nuclear
Safety where current issues could be dealt with without
the burden of past conflicts. Here, it may be possible to
pool capacities, to network research, state and industry
and to create an attractive hub for our international
collaboration. A new start such as this might provide
young people who want to become involved in nuclear
technology with credibly fascinating tasks, good prospects,
respect and appreciation. Perhaps such a project would not
require the very broad general consensus but rather a
viable coalition of people with insight.
Nuclear energy – long-term reality in Europe
Insight also includes the realisation that other countries
are not following our path. Now, after many years of delay,
the new construction projects of Olkiluoto and Flamanville
have reached the preparations for commissioning and are
no longer merely a mirage. The Hinkley Point C project has
received its first partial permit. By the way, all four reactors
will be constructed using instrumentation and control
equipment made in Germany. In the United Kingdom, in
addition to the EPR by Areva, the AP 1000 by Westinghouse
has also received confirmation in the Generic Design
Assessment and the ABWR by Hitachi will follow by the
end of the year.
Things are also happening east of Germany: a few
months ago unit 1 of the Novovoronezh II nuclear power
plant went online – with German instrumentation and
control equipment and a planned operating period up to
2077. Unit 1 of the Leningrad II nuclear power plant, which
is set to replace the old Chernobyl-type plants, is in start-up
commissioning. Construction of the first nuclear power
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plant in Belarus is scheduled and the projects in Paks and
Hanhikivi are also being pushed forward consistently. Our
Czech partners also have expansion plans, not least with a
view to preventing CO2. There will be no shortage of
interested parties as no less than six suppliers have already
expressed an interest. In Poland, the site selection process
for the first nuclear power plant has entered the concrete
phase within the defined area. If safety is also going to be a
concern for us in the coming decades then it must be
Germany’s goal to count permanently as a partner in safety
with recognised expertise. However, the repetition of
demands for phase out is not sufficient, what is needed in
fact is a constructive attitude.
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AMNT 2017
Nuclear technology – part of the location for
industry and science
And let’s not forget that Germany will also benefit from
nuclear technology in many respects and in the long term.
The research reactors in Munich, Berlin and Mainz are not
only used for basic research, they also do a great deal for
applied research and industrial development. They are
also indispensable for direct applications in industry and
medicine. Nuclear technology is also found elsewhere:
such as in non-destructive material testing, plant breeding,
in medical diagnosis and therapy. Nuclear technology is
directly linked to our status as a country of science and
technology.
And let’s not forget economic value creation. Many
internationally recognised nuclear technology companies
are both important employers and taxpayers. This
industrial value chain made up of manufacturers, suppliers
and service providers also requires nuclear expertise,
especially in safety engineering. Germany has a good
reputation in this field and German products and services
related to nuclear safety are in great demand. Obstructing
export will not increase nuclear safety for Germany, for our
neighbours or for the world. And vital expertise can only
develop while it’s in use, e.g. in industry, and therefore in
the medium term largely in exports.
This also applies to companies involved in the fuel cycle
in Germany which are now frequently becoming the target
of political debate. These facilities are explicitly excluded
from the phase out of nuclear energy use and we reject any
efforts to expand the phase out. The Federal Government
may well profess uranium enrichment and fuel assembly
manufacturing in Germany as centres of expertise. When it
comes to using the expertise of these companies for
operational and waste management safety, for the
subject of non-proliferation and for security-policy risk
assessments, then it is not so distant. In this field too, Germany
would like to have its own knowledge and it’s the
same here as with reactor safety. Those who want to
perfect the phase out will also perfect the loss of expertise.
This cannot and must not be our aim.
of our meeting. I would like to thank you all very much for
your contribution to the AMNT, which in 2017 has once
again become our industry’s most important platform for
exchanging knowledge and experience in Germany.
I would also like to thank all those taking part who make
our AMNT so diverse and enriching.
The German Atomic Forum’s traditional reception,
which you are cordially invited to attend, will take place
this evening from 7 pm. It will flow seamlessly into the
usual social evening which we are all looking forward to.
As in previous years, our exhibitors hope you will accept
their invitation to join them.
Ladies and Gentlemen,
I wish everyone a successful meeting with lively discussions
and valuable insights. And please don’t forget to enjoy your
participation here and your stay in the vibrant city of
Berlin.
Author
Dr. Ralf Güldner
President of the DAtF
(German Atomic Forum)
Robert-Koch-Platz 4
10115 Berlin, Germany
Successful AMNT
Ladies and Gentlemen,
Maintaining and developing expertise in addition to
national and international networking are ultimately the
key tasks of the AMNT. In this case, the commitment and
expertise of those who participate in designing the
programme, who are responsible for the sessions and give
presentations in their specialist fields, form the backbone
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48th Annual Meeting on Nuclear
Technology (AMNT 2017): Impressions
AMNT 2017
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AMNT 2017
Monte-Carlo Based Comparison of
the Personal Dose for Emplacement
Scenarios of Spent Nuclear Fuel Casks in
Generic Deep Geological Repositories
Héctor Saurí Suárez, Bo Pang, Frank Becker and Volker Metz
The paper “Monte­
Carlo Based Comparison
of the Personal
Dose for Emplacement
Scenarios of
Spent Nuclear Fuel
Casks in Generic Deep
Geological Repositories”
by Héctor Saurí
Suárez, Bo Pang,
Frank Becker and
Volker Metz has been
awarded as Best
Paper of the 48 th
Annual Meeting on
Nuclear Technology
(AMNT 2017), Berlin,
16 and 17 May 2017.
1 Introduction When a high-level nuclear waste cask is transported to its final position in a deep geological
disposal facility, the radiation exposure received by the workers in such a facility is expected to be significantly influenced
by the materials of the surrounding layers. Moreover, the question arises if there is an enhanced directional dependent
influence on the personal radiation exposure in such facilities since certain amount of backscattered radiation comes
from the back and lateral sides. Hence, it is of interest to study the influence of the worker’s position and its orientation
on the personal dose assessment.
In the current study, the generalpurpose
Monte-Carlo N-Particle code
MCNP6 [Pelowitz et al., 2013] was
employed to calculate the radiation
field around POLLUX® type shielding
casks [Janberg and Spilker, 1998;
Filbert et al., 2011] loaded with spent
nuclear fuel (SNF), which were
emplaced in horizontal drifts of deep
geological repositories. Furthermore,
a simplified mathematical phantom
was used to represent a worker inside
the facility, in order to calculate
the personal radiation exposure for
working scenarios with MCNP6.
Emplacement in two different geological
disposal facilities was considered,
i.e. a horizontal drift in rock salt (from
now on in short as “rock salt drift” or
RSD) and a horizontal drift in a clay or
shale formation (from now on in short
as “clay drift” or CLD). In contrast to a
repository in rock salt, drifts and
access galleries of a repository in soft
rock, such as clay and shale, have
to be reinforced with concrete lining
with a thickness of several decimetres
[e.g. Chen et al., 2014; Leon Vargas
et al., 2017]. The radiation field was
calculated in terms of ambient dose
equivalent for both drifts at different
positions to the shielding cask, which
is disposed on the ground of the drift.
In order to study the dependence of
the worker’s orientation towards the
cask on the personal exposure,
simulations with different angles
between phantom and POLLUX® cask
were performed in RSD. Finally, a
comparison between the calculated
personal dose rate during a working
scenario in RSD and in CLD was
conducted.
2 Methodology
2.1 Waste inventories considered
for POLLUX® type casks
Based on the average inventory of
used fuel elements discharged from
pressurized water reactors (PWR)
in Germany [Peiffer et al. 2011], a
representative waste loading of 90 %
uranium oxide (UOX) fuel and 10 %
mixed-oxide (MOX) fuel with a burnup
of 55 GWd/t(HM) was considered.
The POLLUX® self-shielding cask
[ Janberg and Spilker, 1998; Filbert et
al., 2011], designed for deep geological
disposal in RSD, was employed for
both RSD and CLD. In our model for
disposal in RSD, a POLLUX® cask,
loaded with fuel rods of ten PWR fuel
assemblies was numerically simulated.
This corresponds to a waste load of
about 5.45 metric tons heavy metal
(tHM). The cask with fuel rods of one
MOX and nine UOX fuel assemblies is
herein after referred to as POLLUX-10.
Due to temperature restrictions
regard ing emplacement of casks with
heat-generating waste in clay and
shale formations [Leon Vargas et al,
2017], for the emplacement in a CLD
the maximum amount of fuel assemblies
per cask was set to three. Therefore,
one POLLUX® type cask with an
homogeneous mixture of two thirds
PWR-UOX and one third PWR-MOX
fuel corresponding to one MOX and
two UOX fuel assemblies, comprising
1.64 tHM (herein after referred to as
POLLUX-3M), and two POLLUX® casks
with an homogeneous mixture of
PWR-UOX corresponding to three
UOX fuel assemblies, comprising
1.64 tHM (herein after referred to as
POLLUX-3U), were employed for the
emplacement in a CLD. In general, the
MOX fuel rods are supposed to be
placed in the center of the cask
surrounded by the UOX fuel rods. This
arrangement provides an additional
shielding for neutrons coming from
MOX fuel rods. Hence, a homogeneous
mixture will give conservative results
since the MOX fuel is homogeneously
distributed and supposed to be less
shielded. Two zones were defined in a
fuel rod, i.e. an “active zone” which
contains the fuel pellets and an
“ inactive zone” which corresponds to
the top and bottom of the fuel rod and
it is mainly composed of Zircaloy
cladding [Janberg and Spilker, 1998].
The effective density in these zones
was calculated according to the equation:
where m zone is the mass of the corresponding
zone and V canisterzone is the
total volume available in the POLLUX®
type cask for that zone.
An average burnup of 55 gigawattdays
per metric ton of heavy metal
(GWd/tHM) was assumed for both
UOX and MOX SNF. Before emplacement
a cooling time of the SNF was
assumed to be 50 years after discharge
from the reactor core. This duration
corresponds to an assumed interim
storage time before disposal of SNF in
a deep geological disposal facility to
be built in 2050 according to BMUB
[2015]. Isotope mass of the SNF in dependence
of the cooling time was taken
from [Peiffer et al., 2011]. The SNF
inventory is composed of hundreds of
different isotopes, but many of them
have negligibly small activities. As
investigated in a previous study [Pang
et al., 2016], for the waste inventory
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| | Fig. 1.
MCNP6 model of the emplacement drift with a POLLUX® cask loaded with irradiated UOX and MOX fuel. Black dots represent the position of the F5 tallies, where the letter indicates the axis
direction and the number the distance in meters to the cask surface.
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considered in this study, neutrons
dominate the radiation field and
exposure outside the shielding cask.
Hence, only those isotopes that
contrib ute significantly to neutron
activity were considered when
defining the radiation sources for
simulations with MCNP6.
For the considered fuel inventory
the main contributor to neutron emissions
is the spontaneous fission of
244 Cm (90 % of the total emission) and
246 Cm (5 % of the total emission),
while the contribution due to (α,n)
reactions, mainly stemming from
interactions with 18 O, is less than 5 %.
The total neutron source strength
for the POLLUX-10 inventory is
1.66 · 10 +9 neutrons/sec (n/s),
while those for the POLLUX-3M
and POLLUX-3U inven tory are
9.02 · 10 +8 n/s and 3.24 · 10 +8 n/s,
respectively.
2.2 Calculation of the ambient
dose equivalent rate Ḣ*(10)
In the generic model for a repository
for heat generating waste of Stahlmann
et al. [2015], an emplacement
drift has a length of 57 m (RSD) and
63 m (CLD), respectively. The drift is
surrounded by a host rock layer of at
least 100 m thickness and several
decimetres of concrete lining in the
case of CLD. To simplify the calculations,
the thickness of the drift walls,
i.e. rock salt for POLLUX-10 and
concrete lining for POLLUX-3M and
POLLUX-3U, was set to 1 m in the
MCNP6 model, which is sufficient to
account for possible interactions of
the radiation with the drift wall
materials. With respect to interactions
of neutrons and photons with clay and
concrete, both materials are characterized
by similar densities and elemental
/ oxidic compositions, dominated
by SiO 2 , CaO, Al 2 O 3 and H 2 O.
Figure 1 shows the modelling of the
deep geological disposal facility and
the POLLUX® type cask with MCNP6.
As a simplification, only one cask
(cylindrical form with a length of
5.5 m and an outer diameter of
1.56 m) was placed on the ground of
the drift with its bottom surface at
2.63 m distance to the drift end side.
Detailed geometrical information of
POLLUX® type cask and the generic
emplacement drifts can be found in
[Janberg and Spilker, 1998; Filbert et
al., 2011] and [Stahlmann et al.,
2015], while the respective detailed
MCNP6 models were already
described in [Pang et al., 2016], hence
they are not shown here. Since the
radiation scattered by the drift layers
might have an important impact on
the radiation field, a third drift was
also modelled. This one has the same
geometry as the ones described above
but the surrounding wall layers were
replaced by air, representing a cask
free in air (from now on in short as
FIA).
As denoted by black dots in Fig. 1,
twelve MCNP6 point detector F5 tallies
[Pelowitz et al., 2013] were employed
to calculate the neutron fluence rate
and the ambient dose equivalent rate
Ḣ*(10) at different positions inside the
drift. Tallies X1, Y1 and Z1 (see Fig. 1)
were defined to compare Ḣ*(10) at 1 m
distance to the cask surface in the
respective directions. To study the
change of Ḣ*(10) with the distance to
the cask, the tallies X1 to X10 (see
Fig. 1) were also employed. The neutron
fluence- to-ambient-dose-equivalent
conversion coefficients given by
ICRP [1996] were employed to convert
the F5 tally results into Ḣ*(10). To
assess the precision of the result,
MCNP6 produces a wealth of information
about a simulation, which is
represented by ten statistical checks
[see Pelowitz et al. (2013)]. To pass the
ten statistical checks, 2 · 10 +7 particles
were required per simulation.
2.3 Calculation of the personal
dose equivalent rate Ḣ p (d)
To obtain the personal dose equivalent
rate Ḣ p (d), a worker inside the
drift was represented in this study
with a simplified anthropomorphic
phantom. This phantom is a virtual
representation of the BOMAB (BOttle
MAnnikin ABsorber) phantom, which
models the head, neck, chest,
abdomen, thighs, calves, and arms
with cylinders or elliptical cylinders.
A detailed description of its components
can be found in [U.S.
Department of Energy, 2016]. Figure 2
shows the MCNP6 model of the
phantom used in the current study.
| | Fig. 2.
MCNP6 representation of the BOMAB phantom with a cylindrical detector
at the front side (chest dosimeter) and at the back side (back dosimeter).
As recommended by ICRP, [2007]
the personal dose equivalent rate
Ḣ p (d) at a depth d=10 mm gives a
conservative assessment of the effective
dose rate under most irradiation
conditions. However, this requires the
personal dosimeter to be worn at a
position on the body which is representative
with respect to the exposure.
In general it is recommended to wear
a dosimeter in front of the chest,
where Ḣ p (d) is supposed to give a conservative
estimation of the effective
dose even in cases of lateral or isotropic
radiation incidence on the body
[ICRP, 2007]. However, in cases of
exposure from the back, the question
arises if a dosimeter worn at the front
still appropriately assesses the effective
dose. For a worker inside an
emplacement drift, as investigated in
the current study, a certain amount of
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radiation incidents on the backside
due to the backscattered radiation by
the surrounding drift layers. Therefore,
in order to study the influence
of the backscattered radiation, two
cylindrical detectors (2 cm radius and
0.2 cm length) were modelled in the
phantom (see Fig. 2): one on the front
side (representing a dosimeter worn
in front of the chest) and another one
on the back side (representing a
dosimeter worn at the back side), both
at 10 mm depth to calculate Ḣ p (d).
The personal dose equivalent rate
Ḣ p (d) is calculated as:
Where Ḋ n (E n ) is the neutron absorbed
dose rate, and Ḋ γ (E γ ) is the gamma
absorbed dose rate. The quality factor
for photons (Q γ ) is equal to 1; while
for neutrons (Q n ) it is dependent on
the neutron energy (E n ) and the linear
energy transfer (L) according to:
The MCNP6 energy deposition tally
F6 [Pelowitz et al., 2013] was used to
calculate the absorbed dose rate (D).
Tabulated values for Q n (E n ) were
taken from [Siebert and Schuh macher,
1995] to convert the F6 tally results to
dose equivalents.
| | Fig. 3.
Different angles of the phantom with respect to POLLUX.
To study the effect the orientation
of the worker with respect to the
shielding cask, five simulations with
the phantom at angles of 0°, 15°,
45°, 60°, and 90° with respect to
POLLUX-10 symmetrical axis (see
Figure 3) were performed in RSD. For
each simulation, the Ḣ p (10) obtained
with the front dosimeter and the sum
of the Ḣ p (10) obtained with the front
and back dosimeter were compared to
check if the use of only one dosimeter
may underestimate the received dose
rate. To reduce the calculation time,
the MCNP6 variance reduction technique
“geometry splitting” [Pelowitz et
al., 2013] was applied. Using geometry
splitting, a weighting is assigned in
the following way: regions near the
tallies (cylindrical detectors in the
phantom) are assigned with a greater
importance than regions farther away.
When a particle leaves a region it is
split/killed according to the importance
ratio adjusting the weight of the
remaining particles to leave the tally
unbiased. A total of 1 · 10 +8 particles
were required per simulation to pass
the ten MCNP6 statistical checks.
2.4 Comparison of Ḣ p (10) in
the rock salt and clay
formation drifts during a
typical working scenario
The above explained methodology,
i.e. the use of two dosimeters for the
estimation of Ḣ p (10) was applied to
the working scenario of POLLUX®
| | Fig. 4.
MCNP6 model of the four steps for a POLLUX® disposal scenario (for details see text).
disposal in the emplacement drift
based on the proposal of DBE TECH-
NOLOGY GmbH [Filbert et al., 1995].
Figure 4 shows the MCNP6 models of
four main working steps in the cask
disposal procedure as well as the main
components. A description of their
geometry can be found in [Bollingerfehr
et al., 2011]. The four steps are:
first the cask is transported on a
carriage through the drift with an
electric locomotive with a driver
sitting inside the cabin (see Fig. 4a).
Once it arrives at the disposal position,
as shown in Fig. 4b, the cask is slowly
positioned under a storage equipment
which elevates the cask from the
carriage to allow locomotive and
carriage to drive back. Once the locomotive
is driven back, the storage
equipment places the cask on the
ground (see Fig. 4c). Finally, as shown
in Fig. 4d, the locomotive moves
the storage equipment to the next
disposal position.
Since the cask geometry of
POLLUX-3M and POLLUX-3U are
equal to that of POLLUX-10, the same
steps as described in Fig. 4 were simulated
for the disposal of POLLUX-10,
POLLUX-3M and POLLUX-3U cask. To
compare the radiation exposure, the
same or a similar amount of SNF
should be disposed in both emplacement
drifts, i.e. one POLLUX-10 cask
in RSD containing fuel rods of one
MOX and nine UOX fuel assemblies
(5.45 tHM) and three casks in CLD
(one POLLUX-3M and two POLLUX-
3U, in total 4.92 tHM). As the working
steps are the same for the disposal
of POLLUX-10, POLLUX-3M and
POLLUX- 3U, Ḣ p (10) was employed to
compare the radiation exposure in the
different working steps as described
above.
The driver sitting inside the cabin
was represented in this study by
the phantom (see Fig. 2). Since the
worker stays all the time inside the
cabin and faces the shielding cask,
the angle between phantom and cask
is always 0°. However, the amount
of backscattered radiation may be
further increased due to the reflection
by the cabin walls and backscattered
neutrons from the drift walls. To
perform the MCNP6 simulations,
geometry splitting was employed in
the drift and inside the locomotive
cabin to reduce the number of transported
particles. To pass the ten
MCNP6 statistical checks, 4 · 10 +8
particles were required for simulation
of the transport and location under
the storage equipment. For the placement
and retreat of the storage
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a) Inside the gallery b) Free in air
| | Fig. 5.
Spectral neutron fluence rate calculated with MCNP6 (for details see text).
equipment 1.5 · 10 +9 particles were
required per simulation, since the
distance between the cask and the
phantom is larger.
3 Results and discussion
3.1 The ambient dose
equivalent rate Ḣ*(10) in
the emplacement drift
Figure 5a and Figure 5b show the
spectral fluence rate calculated with
MCNP6 at 1 m distance to the cask
surface in X direction inside the RSD
and CLD as well as FIA. The relative
error of the fluence rates in each
energy bin is in general less than 4 %,
except for some bins with fluence
rates lower than 0.005 cm −2 s −1 . The
effect of the backscattered radiation
can be observed for the RSD with the
local minimum of the spectral fluence
rate between 2 · 10 -3 to 3 · 10 -3 MeV
(Fig. 5a), which is caused by elastic
neutron scattering on 23 Na (one of the
main isotopes of the surrounding rock
salt), which has a peak in the crosssection
at 2.8 · 10 -3 MeV. In the CLD,
the maximum between 1 · 10 -8 and
1 · 10 -6 MeV (Fig. 5a) shows the presence
of moderated neutrons mainly
due to interactions with 16 O content of
the concrete lining (mainly composed
of CaO, SiO 2 , Al 2 O 3 , H 2 O).
Figure 6 shows Ḣ*(10) at 1 meter
distance to cask surface in the
different drifts as well as FIA. Since
the cask shielding in the X direction is
the thickest, Ḣ*(10) outside the cask
in the X direction is lower than that
in Y and Z directions. For the FIA
scenarios, Ḣ*(10) in the X direction
outside the POLLUX-10 cask is 24 %
lower than that outside the POLLUX-
3M, while in the Y and Z directions
Ḣ*(10) outside the POLLUX-10 cask is
24 % and 20 % higher than outside
the POLLUX-3M, respectively. This
can be explained due to the influence
of the inactive zone at the top and
bottom of the fuel rods (X direction).
Figure 7 shows the neutron spectra
before and after the Zircaloy layer in
the inactive zone for POLLUX-3M and
POLLUX-10. The total neutron fluence
rate before the inactive zone is
higher for POLLUX-10 (1538 cm −2 s −1 )
than for POLLUX-3M (861 cm −2 s −1 ).
According to the calculated density of
the SNF (see Equation 1), which is for
POLLUX-10 (active and inactive zone)
3 times larger than for POLLUX-3M,
neutrons emitted in the X direction
are stronger shielded by the Zircaloy
layer in POLLUX-10 (total neutron
fluence after the inactive zone
331 cm −2 s −1 ) than in POLLUX-3M
( total neutron fluence after the
inactive zone 340 cm −2 s −1 ).
| | Fig. 6.
Ḣ*(10) at 1 meter distance from the cask calculated with MCNP6 (for details see text).
| | Fig. 7.
Spectral neutron fluence rate for the SNF calculated with MCNP6 (for details see text).
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| | Fig. 8.
Ḣ*(10) at different distances in the X direction for POLLUX-10 in rock salt, POLLUX-3M and POLLUX-3U
in clay formation and POLLUX-3U in free air.
Figure 8 shows Ḣ*(10) at different
distances in the X direction
for POLLUX- 10, POLLUX-3M and
POLLUX- 3U as well as also for their
respective FIA cases. Ḣ*(10) for
POLLUX- 10 in RSD is between 30 %
(at the cask surface) and 80 % (at
10 m distance) higher than for
POLLUX- 10 FIA. For POLLUX-3M and
POLLUX-3U, Ḣ*(10) in CLD is between
15 and 75 % higher than FIA. This
reveals the important role of the
backscattered radiation on the radiation
field in a geological disposal
facility. Figure 8 shows further that up
to 1 m distance, Ḣ*(10) is higher (up
to 9 %) for POLLUX-3M in CLD than
for POLLUX-10 in RSD. From this
point on, the higher moderation of the
concrete layers and the larger reflection
of the salt layers lead to a higher
Ḣ*(10) in the RSD (between 10 % and
40 %). Since only spent UOX was
loaded in a POLLUX-3U cask, Ḣ*(10)
for the case of POLLUX-3U is in
general 63 % lower than that for
POLLUX-3M.
Angle
[degrees]
Ḣ p (10) Total
[µSv/h]
Ḣ p (10) Chest
[µSv/h]
3.2 Influence of the angle
between phantom and
disposal cask on Ḣ p (10)
Table 1 shows Ḣ p (10) obtained with
the detector at the chest Ḣ p (10) Chest
and at the back Ḣ p (10) Back of the
phantom (see Fig. 2). Also included in
the table are the sum of both detectors
Ḣ p (10) Total and the contribution of
each detector to Ḣ p (10) Total . When the
frontal body part of the phantom is
facing the cask, corresponding to an
angle of 0°, the main contribution to
Ḣ p (10) Total comes from the detector at
the chest. However, as the angle between
phantom and cask increases,
the contribution of the detector at the
back increases. This phenomenon
arrives its maximum when the phantom
has an angle 90° with the cask. In
this case, the dose rate obtained with
each detector represents approximately
50 % of Ḣ p (10) Total . Hence, the
addition of Ḣ p (10) Chest and Ḣ p (10) is a
simple way to account for the angular
dependence.
% Chest
[%]
Ḣ p (10) Back
[µSv/h]
% Back
[%]
0 1.4 1.2 86 0.19 14
15 1.5 1.3 87 0.20 13
45 1.2 0.97 82 0.22 18
60 1 0.75 75 0.26 25
90 0.89 0.45 50 0.44 50
| | Tab. 1.
Total Ḣ p (10) values and the contribution of the chest and back detectors with different angles between
phantom and disposal cask. Calculations were performed in a RSD at 5 meter distance to POLLUX-10
surface.
3.3 Comparison of Ḣ p (10) in
the rock salt and clay
formation drifts during
a working scenario
Table 2 shows the calculated dose
rate Ḣ p (10) Chest and Ḣ p (10) Back for the
working steps of the disposal scenario
shown in Fig. 4. Ḣ p (10) Total in the table
refers to the sum of Ḣ p (10) Chest and
Ḣ p (10) Back while % Chest and % Back are
their percentage contribution to
Ḣ p (10) Total , respectively. In the table,
POLLUX-10 refers to the calculated
Ḣ p (10) for each working step of the
disposal in a RSD, while POLLUX-3
refers to the sum of the calculated
Ḣ p (10) for two POLLUX-3U and one
POLLUX-3M. The calculated dose rate
for the disposal of only one POLLUX-
3M and one POLLUX-3U is also
included in Tab. 2.
For the simulated working steps,
the angle between phantom and
source is always 0°. Hence, the main
contribution to Ḣ p (10) Total is coming
from the chest detector. However,
comparing with the results at 0° given
in Tab. 1, the contribution of the back
dosimeter to Ḣ p (10) Total for the worker
inside the cabin is higher than that for
the worker standing alone in the drift.
This effect is attributed to additional
backscattered radiation due to the
cabin walls and locomotive elements.
The calculated dose rate for each
working step is similar for POLLUX-3M
and for POLLUX-10, while that for
POLLUX-3U is 60 % lower that for
POLLUX-3M, since no spent MOX fuel
was stored in POLLUX-3U. However,
to dispose the same amount of waste
as in a POLLUX-10, one POLLUX-3M
and two POLLUX-3U have to be employed.
Therefore, Ḣ p (10) Total is 30 %
higher for the transport and location
in a CLD that in a RSD. For the
placement and retreat in CLD the
Ḣ p (10) Total is more than a 40 % higher
than that in the RSD. This reveals that
the selection of the host rock can play
an important role in the radiation
exposure of the workers in such facilities.
The developed methodology can
be applied to assess the exposure
during the different steps of nuclear
waste disposal. In this work the same
geometrical parameter were considered
for both emplacement drifts.
However, due to the lower loading
capacity of the cask in CLD, a larger
disposal space is required resulting in
a larger repository compared to a
repository in rock salt [e.g. DBE-Tec,
2016]. This leads to a longer transport
distance and also longer exposure
duration. Since the transport of the
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Step Cask Ḣ p (10) Total Ḣ p (10) Chest % Chest Ḣ p (10) Back % Back
Transport POLLUX-3M 2.8 2.4 87 0.35 13
POLLUX-3U 0.72 0.63 88 0.09 12
POLLUX-3 4.2 3.7 88 0.52 12
POLLUX-10 2.8 2.3 83 0.49 17
Location POLLUX-3M 2.6 2.2 85 0.38 15
POLLUX-3U 0.72 0.63 87 0.09 13
POLLUX-3 4.1 3.5 86 0.57 14
POLLUX-10 2.6 2.1 80 0.50 20
Placement POLLUX-3M 0.18 0.15 86 0.02 14
POLLUX-3U 0.04 0.03 85 0.01 15
POLLUX-3 0.26 0.22 86 0.04 14
POLLUX-10 0.19 0.16 82 0.03 18
Retreat POLLUX-3M 0.07 0.05 73 0.02 27
POLLUX-3U 0.02 0.01 85 0.002 15
POLLUX-3 0.10 0.08 77 0.02 23
POLLUX-10 0.06 0.05 88 0.01 12
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| | Tab. 2.
Total Ḣ p (10) values and the contribution of the chest and back detectors for the different disposal steps in RSD and CLD.
cask is one of the steps with the
highest personal dose rate, it is
plausible to assume that the total
personal exposure for disposal in clay
formation is higher than for disposal
in rock salt.
Since a precise description of the
duration of each working step is
still unknown, only a dose rate
comparison is conducted in this study.
The following example will try to
demonstrate the importance of this
description. Assuming that five hours
are required to dispose a POLLUX-10,
where four hours are for transport (to
simplify, only transport in a drift is
considered) and the fifth hour is
equally divided amongst the other
three steps (20 min/step). For the
disposal in CLD the transport of
each cask will take longer since the
needed space within the drift is larger
(assuming 7 h). This will lead to a
dose of 12.26 μSv and 30.79 μSv for
RSD and CLD, respectively. However
as more and more casks are disposed
in the drift, the transport time will
reduce. Assuming that only 1 h is
required for the transport when the
drift is almost full, and since the time
for the other three steps will be the
same, it will lead to a dose of 3.77 μSv
and 5.65 μSv for RSD and CLD,
respectively.
As illustrated in the example
above, the duration of the working
steps (especially the transport) plays a
decisive role in the personal dose.
Therefore, a precise description of
the different steps is necessary for
a proper comparison between the
different disposal options and to
provide recommendations for minimizing
the occupational radiation
exposure.
4 Summary and
conclusions
In the current study, the ambient
dose equivalent rate Ḣ*(10) and the
personal dose equivalent rate Ḣ p (10)
were calculated for emplacement of
casks with spent UOX / MOX fuel
within two generic deep geological
disposal facilities. In the rock salt drift
a POLLUX-10 was placed, while for
the clay drift with concrete lining a
POLLUX-3M and two POLLUX-3U
were disposed. In addition, casks free
in air were also investigated. Results
show that the backscattered radiation
of the host rock layers or the concrete
lining increases Ḣ*(10) in the disposal
drift in comparison with a cask FIA.
Ḣ*(10) for POLLUX-10 in RSD is
between 30 % (at the cask surface)
and 80 % (at 10 m distance) higher
than for POLLUX-10 free in air. For
POLLUX-3M and POLLUX-3U, Ḣ*(10)
in CLD is between 15 and 75 % higher
than FIA. The higher increase for
POLLUX-10 is caused by the neutron
reflection of the rock salt layers, while
in the clay drift the presence of oxygen
in the concrete lining moderates the
neutrons resulting in a lower increase.
For the calculation of Ḣ p (10) a
mathematical phantom was modelled
with two detectors, one at the front
side of the chest and another one
at the back side. Calculations with
different angles between the phantom
and the cask show that there is an
angular dependence of the registered
dose rate values. This effect is
enhanced if the dose rate is obtained
with only one dosimeter. Therefore, it
was proposed to sum up the dose rate
obtained with both dosimeters. This
methodology was applied to the
working scenario for the disposal of a
POLLUX® type cask in an emplacement
drift. The results of the investigated
scenario, where the worker is
sitting inside the locomotive cabin
and always facing the cask, show that
the main contribution to Ḣ p (10) comes
from the front detector. However, due
to the additional neutron scatterings
at the cabin, the contribution of the
back detector to Ḣ p (10) is up to 10 %
higher that with the worker just
standing alone in the drift. Therefore,
a study of the effective dose under
this irradiation conditions should be
performed to verify if Ḣ p (d) is still
a conservative assessment.
The calculated personal dose rate
for each working step is similar for
POLLUX-3M and for POLLUX-10 but is
40 % higher than that for POLLUX-3U.
However, to dispose the same amount
of waste as in the RSD, three casks
have to be placed in the CLD. Therefore
each disposal step has to be
carried out three times (one POLLUX-
3M and two POLLUX-3U, in Tab. 2
summarized as POLLUX-3), which
leads to a higher dose rate (between
35 % and 40 % depending of the
working step) for the disposal in CLD.
In this study the same geometrical
parameter where considered for both
galleries. However, due to the lower
loading capacity of the cask in CLD, a
larger emplacement drift is required.
AMNT 2017
Monte-Carlo Based Comparison of the Personal Dose for Emplacement Scenarios of Spent Nuclear Fuel Casks in Generic Deep Geological Repositories ı Héctor Saurí Suárez, Bo Pang, Frank Becker and Volker Metz

atw Vol. 62 (2017) | Issue 6 ı June
390
AMNT 2017
This leads to a longer transport
distance and also longer exposure
durations. Since the transport of the
canister is one of the steps with the
highest personal dose rate, it is
plausible to assume that the total
personal exposure for disposal in a
clay formation drift is higher than for
disposal in a rock salt drift.
Acknowledgements
The authors would like to thank our
colleagues of DBE TECHNOLOGY
GmbH for fruitful discussions regarding
the emplacement of POLLUX®
casks and for providing a movie showing
the working scenario in a drift in
rock salt. This study was financially
supported by the German Federal
Ministry of Education and Research
(BMBF; grant number 15S9082E)
as part of the joint research project
ENTRIA – Disposal Options for Radioactive
Residues: Interdisciplinary
Analyses and Development of Evaluation
Principles.
| | M. Sc. Héctor Saurí Suárez, winner of the “AMNT 2017 Best Paper Award”
during his presentation of the paper “Monte-Carlo Based Comparison of the
Personal Dose for Emplacement Scenarios of Spent Nuclear Fuel Casks in
Generic Deep Geological Repositories” at the 48 th AMNT in Berlin, Germany.
| | “AMNT 2017 Best Paper Award” ceremony: Dr. Alexander Zulauf, NUKEM
Technologies Engineering Services GmbH; Dr. Ralf Güldner, President
of DAtF; M. Sc. Héctor Saurí Suárez; Frank Apel, Chairperson of KTG;
Dr. Ron Dagan, Karlsruhe Institute of Technology (KIT) (f.l.t.r.)
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& Neydek, J. (1995): Direkte Endlagerung
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| | Filbert, W., Tholen, M., Engelmann, H.J.,
Graf, R., & Brammer, K.-J. (2011).:
Disposal of Spent Fuel from German
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WM2011 Conference, February 27 -
March 3, 2011, Phoenix, USA
| | International Commission on Radiological
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| | Janberg, K. & Spilker, H. (1998): Status of
the development of final disposal casks
and prospects in Germany. Nuclear
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| | Leon Vargas, R. Stahlmann, J. &
Mintzlaff, V. (2017): Thermal impact in
the geo metrical settings in deep
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retrievability and monitoring.
Proceedings of the International
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Conference, IHLRWMC2017,
April 9 -13, 2017, Charlotte, USA,
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| | Pang, B., Saurí Suárez, H., & Becker, F.
(2016): Individual dosimetry in disposal
repository of heat-generating nuclear
waste. Radiation Protection Dosimetry,
first published online May 5. 2016
| | Peiffer, F., McStocker, B., Gründler, D.,
Ewig, F., Thomauske, B., Havenith, A., &
Kettler, J. (2011): Abfallspezifikation
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Gesellschaft für Anlagen und Reaktorsicherheit
(GRS) mbH. GRS-278.
| | Pelowitz, D.B., Goorley, J.T., James, M.R.,
Booth, T.E., Brown, F.B, Bull, J.S., ...
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| | Siebert, B.R.L., & Schuhmacher, H.
(1995): Quality factors, ambient and
personal dose equivalent for neutrons,
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| | Stahlmann, J., Mintzlaff, V. & Leon
Vargas, R. (2015): Generische Tiefenlagermodelle
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der radioaktiven Reststoffe:
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Aspekte für die Auslegung. ENTRIA-
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Authors
M. Sc. Héctor Saurí Suárez a
Dr. Bo Pang a,b
Dr. Frank Becker a
Dr. Volker Metz a
(a) Institute for Nuclear Waste
Disposal (INE)
Karlsruhe Institute of Technology
(KIT)
Hermann-von-Helmholtz-Platz 1
76344, Eggenstein-Leopoldshafen,
Germany
(b) College of Physics and Energy
Shenzhen University
Nanhai Avenue 3688
518060, Nanshan District,
Shenzhen, China
AMNT 2017
Monte-Carlo Based Comparison of the Personal Dose for Emplacement Scenarios of Spent Nuclear Fuel Casks in Generic Deep Geological Repositories ı Héctor Saurí Suárez, Bo Pang, Frank Becker and Volker Metz

atw Vol. 62 (2017) | Issue 6 ı June
Die 15. AtG-Novelle zur Umsetzung der
EURATOM-Sicherheits- Richtlinie
391
Christian Müller-Dehn
Die 15. AtG-Novelle (AtG: Atomgesetz) hat das parlamentarische Gesetzgebungsverfahren mit dem Beschluss des
Bundestages in der dritten Lesung vom 30.3.2017 und der Befassung im Bundesrat vom 12.5.2017 nunmehr vollständig
durchlaufen, harrt aber noch der Veröffentlichung im Bundesgesetzblatt. Hintergrund aller Regelungen sind die
Ergänzungen der EURATOM-Sicherheits-Richtlinie, die der Europäische Rat im Juli 2014 beschlossen hat und die
bis spätestens August 2017 in den nationalen Regelungen der EURATOM-Mitgliedsstaaten zu verankern sind. Da
die meisten dieser Ergänzungen jedoch bereits geltender Standard im deutschen Atomrecht waren, waren die für
Deutschland umsetzungsbedürftigen Regelinhalte gering. Dies wird ausdrücklich auch in der Gesetzesbegründung
festgehalten.
Die in Deutschland danach noch umzusetzenden Regelungen
lassen sich drei Regelungskreisen zuordnen: Der
Einführung eines periodischen Topical Peer Reviews
für die kerntechnischen Anlagen, der Erweiterung der
Betreiber pflichten sowie der Etablierung von Informations-
und sonstigen Pflichten für die atomrechtlich
zuständigen Behörden.
Unstreitig neu und regelungsbedürftig ist das Topical
Peer Review, das jetzt ausführlich in § 24b Abs. 2 AtG
geregelt wird. Danach soll, beginnend im Jahr 2017, für in
Betracht kommende und sich im Geltungsbereich dieses
Gesetzes befindliche kerntechnische Anlagen mindestens
alle 6 Jahre eine Selbstbewertung hinsichtlich ausgewählter
technischer Themen vorgenommen werden. Das
bereits begonnene Topical Peer Review für 2017, für
das nun auch die gesetzliche Legitimation geschaffen
wird, hat gemäß der europaweiten Vorgabe technische
Fragen zum Alterungsmanagement zum Gegenstand.
Wenn und soweit ein EU-weit abgestimmtes Thema zum
nächsten oder einem späteren Topical Peer Review nur
Anlagen im Leistungsbetrieb betreffen sollte, wäre das
Topical Peer Review in Deutschland dann entbehrlich.
§ 7 c AtG, also die Norm, die die Pflichten des Genehmigungsinhaber
regelt, wächst und wächst. Die Norm
wird in dreierlei Hinsicht ergänzt:
So hat der Genehmigungsinhaber sicher zu stellen, dass
auch Auftragnehmer und Unterauftragnehmer über die
zur Erfüllung der atomrechtlichen Pflichten erforderlichen
personellen Mittel verfügen. Freilich wird hierzu in der
Gesetzesbegründung festgehalten, dass dies nur der Klarstellung
dient und bereits zuvor materiell galt.
Außerdem wird in § 7c Abs. 2 AtG eine neue Nummer 4
eingefügt, die den Betreiber im Rahmen seiner Kommunikationspolitik
zur Information der Öffentlichkeit,
insbesondere der lokalen Bevölkerung und von Interessenträgern
verpflichtet. Rechtspolitisch mag eine
solche detaillierte Regelung zur Öffentlichkeitsarbeit eines
Industrieunternehmens befremden, aufgrund der zahlreichen
nationalen Regelungen, insbesondere in der AtVfV
(Verordnung über das Verfahren bei der Genehmigung
von Anlagen nach § 7 des Atomgesetzes (Atomrechtliche
Verfahrensverordnung)), der AtSMV (Verordnung über
den kerntechnischen Sicherheitsbeauftragten und über
die Meldung von Störfällen und sonstigen Ereignissen
(Atomrechtliche Sicherheitsbeauftragten- und Meldeverordnung
– AtSMV)) und den Sicherheitsanforderungen
an Kernkraftwerken, besteht freilich insoweit bereits eine
so große Regelungsdichte, dass sich hier nur noch
nachdrücklich die Frage aufdrängt, ob überhaupt noch
ein Umsetzungsbedarf bestand.
Drittens wird in § 7c AtG ein neuer Absatz 3 eingefügt,
der den Genehmigungsinhaber verpflichtet, angemessene
Verfahren und Vorkehrungen für den anlageninternen
Notfallschutz vorzusehen. Aufgrund der bestehenden
gesetzlichen Verpflichtungen gemäß §§ 7d, 19a Abs. 4 AtG
und §§ 51, 53 StrlSchV sowie weiteren Konkretisierungen
im untergesetzlichen Regelwerk, nämlich den Sicherheitsanforderungen
an Kernkraftwerke, den Leitfäden zur
periodischen Sicherheitsüberprüfung und einschlägigen
RSK- und SSK-Empfehlungen (RSK: Reaktor-Sicherheitskommission,
SSK: Strahlenschutzkommission) bestand
insoweit allerdings keine ausfüllungsbedürfte Rechtslücke.
Die entsprechende Regelung dient somit im
Ergebnis lediglich der Transparenz gegenüber europäischen
Organen. Der sehr hohe Standard hinsichtlich der
mit der 15. AtG geregelten Betreiberpflichten spiegelt sich
in der Stellungnahme des nationalen Normen-Kontrollrates
wieder, der hierfür nur einen sehr geringen Aufwand
pro kerntechnischer Anlage wiedergibt.
Abgerundet werden die Neuregelungen durch Verpflichtungen,
die sich an Behörden richten. Dies ist
zum einen die Verpflichtung der zuständigen Behörden
nach § 24 a Abs. 1 AtG, die Öffentlichkeit über den bestimmungsgemäßen
Betrieb kerntechnischer Anlagen
sowie über meldepflichtige Ereignisse und Unfälle zu
informieren. Weiterhin wird das für die kerntechnische
Sicherheit und den Strahlenschutz zuständige Bundesministerium
verpflichtet, unverzüglich zu einer internationalen
Überprüfung einzuladen, falls es zu einem
Unfall in einer kerntechnischen Anlagen käme, der Maßnahmen
des externen Notfallschutzes erforderte. Beide
Pflichten sind neu und daher auch umsetzungsbedürftig.
Die Änderungen der EURATOM-Sicherheits-Richtlinie
waren darauf gerichtet, die kerntechnische Sicherheit
in Europa weiter zu erhöhen. Vor diesem Hintergrund
belegt der hier aufgezeigte sehr geringe Regelungsbedarf
zur Umsetzung der EURATOM-Sicherheits-Richtlinie
nochmals und sehr nachdrücklich das hohe Niveau der
Anforderungen an die kerntechnischen Anlagen in
Deutschland, das bereits zuvor bestanden hatte und von
diesen erfüllt wird.
Author
Dr. Christian Müller-Dehn
Senior Vice President Nuclear Regulation and Policy
PreussenElektra GmbH
Tresckowstraße 5
30457 Hannover, Deutschland
SPOTLIGHT ON NUCLEAR LAW
Spotlight on Nuclear Law
The 15 th German Atomic Energy Act Amendment to the Implementation of the EURATOM Nuclear Safety Directive ı Christian Müller-Dehn

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392
ENVIRONMENT AND SAFETY
Retrofitting a Spent Fuel Pool Spray
System for Alternative Cooling as a
Strategy for Beyond Design Basis Events
Christoph Hartmann and Zoran Vujic
Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as
happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed.
Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant
design. As part of Krško Nuclear Power Plant’s Safety Upgrade Program, Krško Nuclear Power Plant decided on, and
Westinghouse successfully designed a retrofit of the AP1000® plant spent fuel pool spray system to provide alternative
spent fuel cooling.
1 Introduction
Following the tsunami and resulting
events in 2011 at the Fukushima
Daiichi Nuclear Power Plant in
Japan, the Western European Nuclear
Regulators Association (WENRA) updated
the safety reference levels in
its report “WENRA Reactor Safety
Reference Levels,” [1] to incorporate
lessons learned from the event.
The update includes establishing an
independent heat removal system
for the spent fuel pool to maintain
the integrity of used fuel assemblies
being temporarily stored there in the
unlikely event of a beyond design
basis accident. The AP1000® nuclear
power plant design foresees provisions
for beyond design basis events.
This includes failure of the spent fuel
pool walls or floor, which would result
in the spent fuel pool draining and
fuel assemblies being uncovered. This
design is also in agreement with the
Nuclear Energy Institute (NEI) issue
of NEI 06-12, Revision 2, “B.5.b Phase
2 & 3 Submittal Guideline” [2], where
an external spent fuel pool makeup
and spray strategy is recommended.
For events with extended loss of AC
power, that is, station blackout, and/
or loss of heat sink due to the spent fuel
pool draining or partially draining,
spent fuel cooling can be provided by
a spent fuel pool spray system. A spent
fuel pool spray system based on the
AP1000® plant design can be retrofitted
for existing nuclear power
plants. In the case of an uncontrolled
spent fuel pool water level drop to
such an extent that the spent fuel pool
would be completely dried out, an
emergency spray system is the best
practical solution that can be applied
for sufficient cooling of the spent fuel
assemblies.
2 AP1000 Plant Spent Fuel
Pool Cooling
The AP1000® plant design features
multiple, diverse lines of defense to
ensure spent fuel cooling can be
maintained for design basis and
beyond design basis events.
During normal and abnormal
conditions, defense-in-depth and duty
systems provide highly reliable spent
fuel pool cooling. These systems are
driven by offsite AC power or the
onsite standby diesel generators.
For unlikely events with extended
loss of AC power, that is, station
blackout, and/or loss of heat sink,
passive systems provide spent fuel
pool cooling. These passive systems
require minimal or no operator action
and are sufficient for at least 72 hours
under all possible loading conditions.
After 72 hours, several different
means are provided to continue spent
fuel pool cooling using installed plant
equipment, as well as off-site equipment.
Even for beyond design basis
events with postulated spent fuel pool
damage and multiple failures in the
passive safety-related systems and
active defense-in-depth systems, the
AP1000® plant spent fuel pool spray
system provides an additional line of
defense to prevent spent fuel damage.
The spent fuel pool is located in a
hardened section of the Auxiliary
Building and contains used fuel that
has been removed from the nuclear
reactor core. Typically, 64 fuel assemblies
are removed from the reactor
| | Fig. 1.
AP1000® Plant Spent Fuel Pool Spray System. Spray headers and nozzles (left) and section view of spray pattern from nozzle (right).
Environment and Safety
Retrofitting a Spent Fuel Pool Spray System for Alternative Cooling as a Strategy for Beyond Design Basis Events ı Christoph Hartmann and Zoran Vujic

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| | Fig. 2.
Pipe routing of spent fuel pool alternate cooling with mobile heat exchanger (MHX) and spray system.
during refueling every 18 months and
stored in the spent fuel pool. The
AP1000® plant’s spent fuel pool has
the capacity to cool up to 889 spent or
used fuel assemblies, which are
continuously submerged beneath
approximately 7.6 m of water.
The spent fuel assemblies continue
to generate decay heat naturally even
when they are removed from the
reactor and are placed in the spent
fuel pool. This decay heat will
decrease significantly over time so
that older spent fuel produces less
heat than spent fuel that has recently
been removed from the reactor.
The spent fuel in the spent fuel
pool is cooled by transferring the
decay heat from the used fuel to the
water in the spent fuel pool. The spent
fuel pool water is, in turn, pumped
through a loop with a heat exchanger
where it is cooled and decay heat is
transferred to a second water cooling
system. The cooled water is then
returned from the second water cooling
system to the spent fuel pool and
the decay heat is transferred to the
environment. There are two identical
spent fuel pool cooling trains, though
only one pump and heat exchanger in
one of the two trains are in operation
in most circumstances.
3 AP1000 plant spent fuel
pool spray system
The AP1000® spent fuel pool spray
system is designed to cool the spent
fuel during a beyond design basis
event in accordance with the B.5.b
guideline [2].
The AP1000® plant spent fuel pool
spray system has two redundant spray
headers located on either side of the
spent fuel pool. There are 16 spray
nozzles on each header (Figure 1,
left). One header receives water
through either gravity-fed draining of
the passive containment cooling water
storage tank, which is located on top
the Shield Building, or from a flanged
connection located in the truck bay,
which is used with an onsite portable
pump. The other header receives
water from the fire protection water
tanks and the diesel-driven or electric
motor-powered fire protection system
water pumps. Spray nozzles distribute
water spray in the form of a hollow
spray cone over the fuel assemblies.
Only one spray header is required
to assure sufficient cooling of the
exposed spent fuel due to sensible
heat and latent heat from water spray
vaporization (Fig. 1, right).
The spray system used to cool the
spent fuel pool during a postulated
loss-of-large-area event is sized to
provide an adequate amount of
spray to the hottest fuel assembly that
will enter the spent fuel pool. The
analytical basis for determining the
minimum amount of spray needed to
cool a fuel assembly is adapted from
the calculation used in Section 3.3
of the Sandia report, “Mitigation of
Spent Fuel Pool Loss-of-Coolant
Inventory Accidents And Extension of
Reference Plant Analyses to Other
Spent Fuel Pools” [3]. Further, to
prevent pressurization inside the
Fuel Handling Building, the system
includes a relief panel to release steam
that is produced during the cooling
process.
4 Krško nuclear power
plant safety upgrade
program
Krško Nuclear Power Plant was
already in the process of making
significant upgrades as a result of
applying for a license extension in
2009 to operate beyond 2023. The
Krško Safety Upgrade Program
was designed in response to the
Slovenian Nuclear Safety Administration’s
re gulations and interpretation
of reference safety levels
from the report, “ WENRA Reactor
Safety Reference Levels” [1], concerning
reasonable measures to prevent
and mitigate severe accidents in
preparation for the possibility of
extending original plant operating
licenses. The reference safety levels
within the report were updated in
2014 to incorporate lessons learned
from the event at the Fukushima site.
The measures defined in the frame of
the Krško Safety Upgrade Program
are in agreement with the nuclear
industry’s response to the Fukushima
accident and the resulting update of
the safety reference levels proposed
by WENRA. This includes plant upgrades
and design changes to address
design extension conditions defined
in the report and beyond design basis
accidents.
Krško’s Safety Upgrade Program is
divided into various projects being
carried out during three phases. The
Spent Fuel Pool Alternative Cooling
Project is in the scope of Phase 2. The
project is scheduled to be completed
by the end of 2017.
The Spent Fuel Pool Alternate
Cooling Project shall assure alternate
cooling of used fuel assemblies
by using a mobile heat exchanger
or spray system (see Figure 2).
Furthermore, it shall assure depressurization
of the Fuel Handling
Building by using relief panels to
release steam produced during the
cooling process.
The systems of the Spent Fuel Pool
Alternate Cooling Project are designed
to assure that heat is removed from
the spent fuel during Design Extension
Conditions A and B and to mitigate
spent fuel damage. The operational
conditions for the systems of the Spent
Fuel Pool Alternate Cooling Project
are classified according to the plant’s
severe accident scenarios, following
the WENRA guidance document “ Issue
F: Design Extension of Existing
Reactors” [1].
ENVIRONMENT AND SAFETY 393
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ENVIRONMENT AND SAFETY 394
| | Fig. 3.
Spent fuel pool spray system simplified process flow diagram.
5 Spent fuel pool spray
system function and
design
The spent fuel pool spray system will
be used in in the event of scenarios
described in Design Extension Conditions
A and B, which postulate a
highly unlikely drainage of the spent
fuel pool. In such scenarios, cooling
water can be provided by a spray
header with fixed spray nozzles
installed along the spent fuel pool
walls. Westinghouse conducted
experi mental testing to determine the
exact number of nozzles required and
their optimal positions. Water can be
supplied by two diverse sources: the
fire protection system and the Sava
River.
The fire protection system can be
flexibly connected using fire hoses.
The Sava River water can be pumped
using a mobile pump unit and fire hoses.
The mobile pump unit is powered
directly by a diesel motor. Either water
source is connected to the spent fuel
pool spray system with hose connections
that are installed inside and
outside of the Fuel Handling Building.
A simplified process flow diagram of
the spent fuel pool spray system is
shown in Figure 3.
The Krško Nuclear Power Plant’s
spent fuel pool spray system is
designed to perform its designated
safety function under Design Extension
Conditions. This includes being
designed to meet the seismic performance
requirements for operation
and mitigation during and after a
design extension condition earthquake,
which is equal to twice the
design requirements for a safe shutdown
earthquake for the existing
systems, structures and components
of the Krško Nuclear Power Plant.
The spent fuel pool spray system’s
permanently installed equipment will
be protected against flood events with
additional margin, even in the case
of a Sava River bank failure. The
system’s permanently installed equipment
is also designed to withstand
extreme winds and tornados. In order
to keep the design of the system as
simple as possible only local indicators
are used and there are no electrically
driven components.
As in the AP1000® spent fuel pool
spray system, the Krško Nuclear Power
Plant’s spent fuel pool spray system:
• Will provide a sufficient amount of
cooling water to maintain the
spent fuel cladding temperature at
lower than 400 °C for a long term
during the loss of ultimate heat
sink,
• Is sized to provide an adequate
amount of spray to the hottest fuel
assembly that will enter the spent
fuel pool during a postulated lossof-large-area
event,
• Has an analytical basis for determining
the minimum amount
of spray needed to cool a fuel
assembly adapted from the calculation
used in Section 3.3 of the
Sandia Letter Report “Mitigation of
Spent Fuel Pool Loss-of-Coolant
Inventory Accidents And Extension
of Reference Plant Analyses to
Other Spent Fuel Pools” [3].
An example of the water spray coverage
and distribution through the
nozzles throughout the spent fuel
pool is shown in Figure 4. The red
circles show the area inside of the
spent fuel pool covered by water
through the spray nozzles, which are
installed at the side walls of the spent
fuel pool. Darker red areas show
possible overlap, whereas water that
does not fit the spent fuel pool geometry
hits the spent fuel pool walls.
Westinghouse conducted experi mental
testing to optimize the spray configuration
around the spent fuel pool edge
and to verify the spray water distribution
and overlap.
Inside the spent fuel pool is a two
region rack design. Recently offloaded
fuel, which is comprised of the hottest
fuel assemblies, is placed in the
Region 1 racks. The Region 2 racks
provide high density storage for
| | Fig. 4.
Example of the spent fuel pool spray system’s nozzle coverage and distribution of water spray.
Environment and Safety
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cooled, irradiated fuel assemblies. Per
the initial cooling time defined in
plant procedures (typically 4.5 years
or three fuel cycles), the most recently
cooled, irradiated fuel assemblies are
transferred from the Region 1 racks to
the Region 2 racks.
6 Experimental determination
of spray nozzle
coverage and distribution
A specific flow density and a uniform
coverage of cooling water over a large
rectangular area are required to cool
fuel assemblies in the unlikely event
that they become uncovered during a
beyond design basis accident. The
experimental testing used to determine
the optimal coverage area, spray
flow density and distribution of the
spent fuel pool spray nozzles for cooling
uncovered fuel assemblies in case
of Design Condition Extensions A and
B include testing different types of
nozzles with different volume flow
rates as the decay heat decreases with
cooling time (see Region 1 and Region
2 in Fig. 4). The measurement setup
of the experimental testing at the
Lechler GmbH Technology Center
Metzingen is shown in Figure 5.
• Volume Flow:
variable adjustable, depending on
spray nozzle type and pressure
The coverage area measurements
were performed by using adequate
collecting canisters with the same
dimensions as the rack cell canisters
at specified positions (see Fig. 5).
Pictures and video recording documented
the testing process and resulting
water distribution. An example of
the measurement setup is shown in
Figure 6.
| | Fig. 6.
Example of water distribution from a spray
nozzle [4].
7 Managed challenges
Retrofitting the AP1000® spent fuel
pool spray system to Krško Nuclear
Power Plant’s existing systems and
structures posed a few challenges.
These included:
• Planning the pipe routing and
meeting the space requirements
for pipe supports (see Figure 7),
• Assuring pipe routing had minimal
impact to equipment already existing
inside of the Fuel Handling
Building,
• Planning the installation process
for the system to meet the space
requirements in the Fuel Handling
Building,
• Assuring the number of fuel assembly
racks that cannot be used for
spent fuel storage due to nozzles
and/or supports which protrude
into the pool are as low as possible,
• Designing the retrofit system with
consideration of the fuel assembly
loading pattern at Krško Nuclear
Power Plant to assure that the
spray nozzles provide the required
water spray amount, coverage and
distribution to cool the spent
fuel pool during Design Extension
Conditions A and B.
Each of these challenges was successfully
resolved.
8 Summary
Due to requirements for nuclear
power plants to withstand beyond
design basis accidents, including
events such as happened in 2011 in
the Fukushima Daiichi Nuclear Power
Plant in Japan, alternative cooling of
spent fuel is needed. Alternative spent
fuel cooling can be provided by a
retrofitted spent fuel pool spray
system based on the AP1000® plant
design. As part of Krško Nuclear Power
Plant’s Safety Upgrade Program,
Krško Nuclear Power Plant decided
on, and Westinghouse successfully
designed a retrofit of the AP1000®
plant spent fuel pool spray system to
provide alternative spent fuel cooling.
The spent fuel pool spray system
will be installed inside and outside the
Fuel Handling Building. For diverse
water supply, sources such as the fire
protection system and river water
were considered and chosen by Krško
Nuclear Power Plant. The system has a
robust design that employs local
measurements and indicators and
ENVIRONMENT AND SAFETY 395
| | Fig. 5.
Exemplary measurement setup for experimental
testing of spray nozzle coverage and
distribution [4].
The following conditions were
applied to determine the flow density
and the coverage area:
• Spray height:
variable adjustable, depending on
local conditions around the spent
fuel pool
• Setting angle horizontal:
variable adjustable, depending on
local conditions around spent fuel
pool and spray nozzle type
• Pressure:
variable adjustable, depending
on spray nozzle type and hydraulic
design of spent fuel pool spray
system
| | Fig. 7.
Depiction of the local conditions around the Krško nuclear power plant spent fuel pool.
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396
OPERATION AND NEW BUILD
does not require electrically driven
components. Westinghouse determined
the required amount of water
spray needed to cool the spent fuel
pool during Design Extension Conditions
A and B analytically; designed
the system’s hydraulics to provide
sufficient flow rates for the volume;
defined the corresponding pipe
diameters; and determined the pipe
routing, based on the on the local
space restrictions around Krško
Nuclear Power Plant’s spent fuel pool.
Westinghouse also conducted experimental
testing at the Lechler GmbH
Technology Center Metzingen to determine
the coverage and distribution of
the spent fuel pool spray nozzles and
to confirm the spray height, setting
angle horizontal orientation, pressure
and volume of water flow.
The spent fuel pool spray system is
planned to be installed by the end of
2017.
References
[1] Western European Nuclear Regulators
Association (WENRA): WENRA Safety
Reference Levels for Existing Reactors,
September 2014.
[2] Nuclear Energy Institute (NEI): B.5.b
Phase 2 & 3 Submittal Guideline,
Revision 2, NEI 06-12, December 2006.
[3] Sandia National Laboratories:
Mitigation of Spent Fuel Pool Loss-of
Coolant Inventory Accidents And
Extension of Reference Plant Analyses
to Other Spent Fuel Pools, Sandia Letter
Report, Rev. 2, November 2006.
[4] Lechler GmbH, Technology Center
Metzingen.
Authors
Dipl.-Ing. Christoph Hartmann
Project Engineer Safety
Engineering
Dr.-Ing. Zoran Vujic
Marketing Manager Business
Development
Westinghouse Electric Germany
GmbH
Dudenstraße 6
68167 Mannheim, Germany
Cyber Security in Nuclear Power Plants
and its Portability to Other Industrial
Infrastructures
Sébastien Champigny, Deeksha Gupta, Venesa Watson and Karl Waedt
Introduction This technical contribution provides a snapshot of the current cyber security efforts in different
industry domains. We argue that stringent security controls (countermeasures) that are already in place for nuclear
power plants (NPP) can be ported to other industry domains. A reason for this is that the nuclear domain is more
formally regulated, thus graded security requirements were already mandated long before the critical infrastructure
debates started and before gradual enforcement of the European and national legislation.
Note: Generally, in the nuclear and
industrial automation domain, the
term “control” is used mainly to
denote Instrumentation and Control
(I&C), Industrial Automation and
Control Systems (IACS) or SCADA
(Supervisory Control and Data Acquisition)
referring to control theory
tasks. However, in the security context,
the term “Security Control” is
ubiquitous, and means any countermeasure
that can reduce the systems
risk due to security threats. Countermeasures
are not limited to add-on
provisions at the components or systems
level. For example, they also include
provisions at the software
source code level.
In Section 1, we will provide an
overview of current international and
national cyber security guidance, and
how this guidance evolved for International
Atomic Energy Agency (IAEA),
Nuclear IEC and selected countries.
Section 2 summarises the increasing
cyber security efforts for Industrial
Automation and Industry 4.0 as well
as its Chinese “Manufactured in China
2025” and US “Industrial Internet of
Things” counterparts. Section 3
provides reasons for the portability
of Security Controls from Nuclear
to other industrial infrastructure.
Summary provides an outlook on the
newest cyber security-related activities
in the different domains, and
concludes with a summary of the
main steps that are necessary for
achieving and maintaining a target
security level.
1 Cyber security and safety
requirements for NPPs
In the nuclear domain, for Safety,
Human Factors Engineering, Physical
Security, Radiation Protection and
Cyber Security, the international
top-level guidance is provided by the
IAEA. The IAEA guidance is regularly
updated based on priorities set by
yearly or bi-yearly meetings of representatives
of all IAEA member states.
The overall IAEA Cyber Security
guidance is refined, e.g. for Instrumentation
& Control (I&C) and
Electrical Systems (ES), by the
Nuclear IEC subcommittees. However,
each country may supersede the
international guidance by providing a
mandatory higher priority regulation,
as will be addressed in section 1.4 for
selected countries.
1.1 Stringent and graded
security requirements for
I&C already since 1986
Safety and security grading are
essential when addressing critical
industrial infrastructures. Grading by
Safety Categories in IEC 61226 and
Safety Classes in IEC 61513, were
already in place since the first editions
of these standards. The softwarespecific
requirements for software
implementing Category A or Category
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B and C I&C functions, are also graded
by the respective standards IEC
60880:1986 and IEC 62138. The first
edition of IEC 60880:1986 already
contained explicit requirements on
security during software development
and security during software deployment,
two essential phases in the software
development lifecycle.
1.2 Overall IAEA Cyber security
Guidance
The IAEA Cyber security Guidance is
published in the IAEA Nuclear Security
Series (NSS). Currently the top-level
guidance is IAEA NSS 17 from 2011.
Developing this guidance took several
years with considerable input by
member states provided since 2006,
and essential agreements being
achieved during the first major IAEA
cyber security conference in summer
2011. IAEA NSS 17 introduces a graded
security approach with 5 security
levels and recommendations on
security zones.
IAEA NSS 17 is complemented by
IAEA NSS 8 on preventive and protective
measures against insider
threats, and further IAEA NSS guidance,
including IAEA NSS 12, on a
comprehensive educational program
in nuclear security.
1.3 Nuclear IEC Cyber security
Standards
Subsequently, the three major Nuclear
IEC cyber security standards will be
introduced.
1.3.1 The Top-level Nuclear IEC
Cyber security Standard
After initial attempts to structure
the top-level nuclear IEC standard
according to nuclear safety and other
criteria, finally, a core-team devised
the alignment with the most popular
information security standard ISO/
IEC 27001:2005 then in place. This
structuring was proposed mainly in
order to reduce the initial training
needs of security staff already familiar
with the mainstream standards, and
in order to avoid annexes with cumbersome
mappings.
While ISA99 experts were involved
in the development of the first toplevel
nuclear IEC 62645:2013 cyber
security standard, an alignment
with ISA99 industrial cyber security
standards or the corresponding IEC
62443-x-x was ultimately not attempted,
as several planned parts of the IEC
62443-x-x series were not yet available
and because the Security grading
follows a different approach, as will be
addressed in a subsequent section.
| | Fig. 1.
Safety functions, process functions and I&C functions.
1.3.2 Coordinating safety and
cyber security by IEC 62859
Whether safety and cyber security
should be considered jointly or subsequently,
is a part of ongoing debates
in different industry domains. For
nuclear, the security grading is
directly related to the potential impact
of a security attack on nuclear safety.
Figure 1 shows the hierarchical
refinement from Safety Objectives
(level 1) down to I&C Functions (level
3 and 4). Main safety objectives are
control of reactivity, residual heat
removal and confinement of radioactive
material.
Cyber security is applied at the
level of I&C and IT equipment while
considering the potential impact of
manipulations on Safety Functions
and Safety Objectives.
IEC 62859:2016 [1] specifies the
main requirements for coordinating
safety and cyber security. In other
industries, work on this important
topic was just started, e.g. by the new
working group WG20 of IEC TC65.
1.3.3 Detailed security controls
for nuclear by IEC 63096
Similar to the alignment of IEC 62645
with ISO/IEC 27001, the new working
draft IEC 63096 is being aligned
with the ISO/IEC JTC1/SC27 WG1
standard ISO/IEC 27002:2013. This
nuclear IEC standard extends the
generic security controls of ISO/IEC
27002 by recommendations for each
security level: BR (Baseline Requirements),
S3, S2 and S1 (highest
security level). It also provides
guidance for the main I&C and ES
(Electrical Systems) lifecycle phases:
Product & Platform Development,
Engineering and Operation & Maintenance.
Additionally, it provides
security control specific guidance for
legacy I&C and ES systems.
As a sector-specific standard,
similar to ISO/IEC 27009 [3], for
non-nuclear utilities, IEC 63096
provides guidance that is structured
and formatted in principle in line with
ISO/IEC 27009 which provides
common guidance on the elaboration
of sector-specific security controls and
Information Security Management
Systems (ISMS) standards.
1.4 International and national
nuclear cyber security
regulations
Table 1 lists the international standards
discussed above, along with the
national standards for Germany, USA
and the UK. In Germany, SEWD
(Schutz gegen Störmaßnahmen oder
sonstige Einwirkungen Dritter/Protection
against Disruptive Acts or Other
Intervention of Third Parties) is a requirement
found in §6 para. 2 no. 4
Atomic Energy Act. released in 1959.
[7]. Licenses for the storage of nuclear
fuels are only granted once risks and
threats, as a result of SEWD, can be
considered as negligible. Created
by Congress in 1974, the USA’s NRC
regulates commercial nuclear power
plants and other uses of nuclear
materials. NRC RG 5.71 [4] provides
guidelines for the protection of digital
computer and communication systems
and networks from cyberattacks,
against which licensees should provide
assurance. The Nuclear Energy Institute
(NEI) 08-09 “Cyber Security
Plan for Nuclear Power Reactors” provides
a generic template for a cyber
security plan, which must be used
by licensees to develop their cyber
security plans to be submitted to the
NRC [8]. The HMG IA (Information
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International and national nuclear laws and regulations
| | Tab. 1.
Examples of international and national nuclear cyber security regulations.
Assurance) Standard is intended for us
by IA practitioners, working especially
with UK Government ICT systems, as
the foundation for their Information
Risk Management Policy. This standard
provides a methodology by which
these practitioners can “identify, assess
and determine the level of risk to an ICT
system and a framework for the selection
of appropriate risk treatments.”
Requirements from these international
nuclear Cyber Security
standards are applicable for the whole
nuclear power plant. Figure 2 shows
the scope of applicability of these
requirements using the example of a
typical nuclear I&C architecture.
In Figure 3, the relationships
between safety standards (in purple)
and security standards (in orange)
from different industries are indicated.
All the individual fields have
their own specific standards for safety
and security. For example, IEC 60601
and IEC 62304 are the safety standards
referred in medical field.
| | Fig. 2.
An example of a nuclear I&C architecture (© AREVA).
| | Fig. 3.
Safety and Security Interface at the Standards Level (© IEC TC65).
2 Gradual consideration
of information security
in Industry 4.0 and IoT
Industry 4.0 and “Manufactured in
China 2025” are governed by a “Reference
Architecture Model Industry 4.0”
(RAMI) or similar which are typically
represented by cubes subdivided as
6x6x6 or 5x5x5. The 3 axis of the cube
are “Layers”, “Hierarchy Levels” and
“Value Streams”. None of the 6 Layers
(Business, Functional, Information,
Communication, Integration and
Asset) explicitly contains cyber
security. Similarly along the other two
axes, cyber security is not explicitly
included. This is due to the fact that
security and interoperability are
considered as integral components in
multiple of the 3D elements that built
up the complete cube, see Figure 4.
2.1 Generic information
security
One purpose of generic security standards
is to be applicable by any size of
an organization, e.g. a one- employee
service provider or a multinational organization.
The ISO/IEC 27000 series
takes credit on meeting this criterion.
Still, beyond these generic information
security standards in the 27000 to
27021 range, additional standards in
the 27031 to 27050 and other ranges
provide more in-depth guidance.
2.2 IT security for power
generating plants
VGB-S-175 addresses generic security
requirements, Defense-in-Depth
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| | Fig. 4.
Reference Architectural Model Industry 4.0 by ZVEI (© Plattform Industrie 4.0).
principles, redundancy and diversity,
risk management, risk analysis and
security countermeasures for both,
new built and power plant modernization
projects.
Furthermore, VGB provides guidance
on intrusion detection and
prevention (addressed in more detail
by ISO/IEC 27039), patch management
(addressed in more detail by
IEC 62443-2-3), security gateways
(addressed in more detail in ISO/
IEC 27033-4), wireless (ISO/IEC
27033-6), documentation of security
incidents (ISO/IEC 27035-3) and
additional countermeasures.
2.3 Emerging industrial
automation security
Cyber security for Industrial Automation
mainly builds on the ISA99
specific standards which are published
as IEC 62443-x-x. The 13 parts
of this series are not yet complete. The
security grading is based on the risk
an attacker imposes and on its
strength. This regularly leads to controversy,
as the strength of an attacker
can change over time, e.g. today’s
“script kiddies” have other malicious
tools as compared to 10 years earlier.
2.4 Initial Industry 4.0 and IoT
proposals
Despite its current incompleteness,
IEC 62443-x-x builds a solid basis for
cyber security in the Industry 4.0
RAMI framework. Interoperability is a
key component of Industry 4.0. The
multipart IEC 62541 defines the Open
Connectivity Unified Architecture
(OPC UA) not just as a communications
protocol, but as a communication
architecture that supports
among other services, interoperability
between digital technologies from
different vendors. The services, as
provided by the layers of the platformindependent
OPC UA, include the
semantics of an information model,
address spaces, discovery services,
alarm functions, etc.
AREVA NP implements Embedded
OPC UA, for example, in its SIPLUG®
family of monitoring sensors, as
shown on Figure 5. Hence, it can
directly be connected to reporting and
trend surveillance systems. This
feature drastically reduces the costs
for interconnecting the respective
sensor devices with equipment from
different vendors, as deployed worldwide
at NPP sites [6].
Part 2 of IEC 62541 provides the
security framework for OPC UA, the
main aim of which is to provide
security for the data exchanges facilitated
by this architecture.
While there seems to be general
acceptance on OPC UA as a part of
Industry 4.0 and IoT, the final hard
real-time communication protocols
and the respective security solutions
are still to emerge.
| | Fig. 5.
SIPLUG® OPC UA based example.
3 Portability of cyber
security knowledge and
features from nuclear to
other industrial infrastructures
The subsequent sections exemplify
some domains where solutions from
the nuclear domain can be adapted
and applied to other domains.
3.1 Joint functional safety and
cyber security consideration
One benefit of IEC 62859:2016, as
compared to generic safety & security
related solution, is its well delimited
context of the applicability for NPPs.
The grading is well defined based on
the maximum impact in the nuclear
context. The transition between the
safety states is also well understood
due to comprehensive deterministic
and probabilistic safety analyses.
These results from the functional
safety experts can directly be leveraged
by the security staff. This
approach can be transferred and
adjusted for other business domains.
The security grading has to be
adjusted to the possible impact levels
in the respective business domain.
Similarly, an analysis is needed and
feasible on which security events can
lead to a similar impact as the respective
safety events (like equipment
faults, failures of supporting assets,
spurious actuations). Based on this
mapping, a risk management process
can be modified in order to adjust
and justify the criticality assignment
(assignment of security degrees to
systems) and to apply complementary
security controls.
3.2 Security grading
The generic information security
standards like ISO/IEC 2700x define
no security grading- also called
security levels or levels of trust. Unfortunately,
in some industries the
grading may be defined based on
criteria that may change over time.
Thus, the strength of an attacker may
change while the impact will not
change or only in well-justified (and
easily identifiable) circumstances, e.g.
after power up-rating of an NPP.
As for nuclear, in implementing a
long-term stable impact-based grading
approach, the overall risk management
and security control adjustment
requirements could be considerably
reduced.
3.3 Security awareness
training
Safety Culture and Security Culture
have a long tradition in nuclear, see
e.g. IAEA NSS 7 “Nuclear Security
Culture” from 2008. With humans as
the strongest and also as the weakest
link in the security chain, specific
security training is essential. Such
training can be adapted for other
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business domains and for different
staff roles, like operators, service
engineers, physical security staff,
cyber security staff and management.
3.4 Strong preventive security
controls
Often mimicking the activities of their
counterparts in the office IT world,
cyber security safety staff deploy
network or host monitoring systems,
like Network and Host Intrusion
Detection Systems (IDS). These detective
security controls may be the only
option in an office IT environment,
where the exact content and frequency
and destination of messages
sent via communication networks
cannot be predicted. However, for
nuclear and for many other industries,
like process automation and discrete
manufacturing, the data exchange is
of a periodic nature, e.g. with fixed
communication cycle times.
This allows the implementation of
strong Preventive Security controls
beyond baseline firewall filtering. In
many cases, the network architecture
may be adjusted to include Data
Diodes as (preferably optical) Physically
Unidirectional Security Gateways.
Applying these network architecture
level improvements ensures
reaching and maintaining the required
target security degree. An
example of preventive security control
is provided in Figure 6. On the left
half of the figure, an automation
system is shown in its standard configuration.
On the right half of the
figure, the automation system is protected
by patented software called
OPANASec. OPANASec is both a
preventive and a detective measure
against cyber-attacks on the automation
system. It protects the system’s
integrity by detecting any read or
write access to the automation system
and announces it to the operator in
the main control room, by means of a
red traffic light for example. It also
prevents information retrieval and
any modifications of the automation
system by locking read and write
access.
3.5 Forensic readiness
Reports on system intrusions and
manipulations without a trace to the
identity and location of hackers or
threat agents are, in general, frequently
reported in technical magazines,
but also more and more by
commercial media. Typically, the
reason for this is that no forensic
readiness specific security controls are
in place. Also, the implementation of
the forensic readiness security controls
(e.g. log files related) may not be
adequate for the target security level.
As for nuclear, this can be improved
by systematically performing attack
tree analyses and assigning appropriate
forensic readiness security
controls in line with the security
grading.
3.6 Incident response
While incident response on Safety
related incidents has a long tradition
with nuclear, cyber security incident
management is currently in the focus
of the first IAEA financed cyber
security R&D with 14 international
partners.
As one of the major partners in the
IAEA Coordinated Research Proposal
(CRP) J02008, AREVA NP, together
with one of its German partner Universities,
can leverage the results for
other business domains.
3.7 Security testing
The appropriate assignment of
security controls based on a continuous
risk management, is essential
for achieving a high security posture.
However, the implementation or configuration
of some security controls
may be flawed. Even more important,
the implementation and configuration
of the software and FPGA-based
systems may include vulnerabilities,
some of which may be security
relevant.
This mandates a selective, prioritised,
in-depth penetration and
fuzz-testing. We are currently working
on an extensive R&D together with
multiple German partner universities
and several Cyber security PhD candidates,
as part of the partially BMWi
Ministry funded the SMARTEST R&D
project on “smart” (model based)
cyber security testing. The respective
results can be leveraged, as most
of the six (6) Industrial Automation
platforms deployed in NPPs and
analyzed by the project, are also
deployed in other industries.
3.8 Security modelling
The I&C and ES Architecture of NPPs
comprises multiple distributed I&C
systems that are built-up from several
subsystems and components. Modelling
these systems together with
models of the physical process (including
pumps, valves …) is common
practices for several decades. Typically,
this includes simulators which
run in real-time or faster than realtime.
There are modelling approaches
which include the security control
definitions into existing 3D models
(for physical security related security
controls) and 2D models, e.g. for
network architectures. These models
support the systematic generation and
analysis of attack trees, far beyond
any paper-based manual analysis.
This approach can be leveraged by
using the same modelling framework
(e.g. AutomationML from the Industry
4.0 context) for other business
domains. The initial investment in
defining the models is compensated
not only by the more comprehensive
analysis, but also by the opportunities
that the models provide for training of
different staff and even for advertising
security features of the customer
products.
| | Fig. 6.
Security control using patented software OPANASec.
3.9 Security asset management
Implicit asset identification is unavoidable
in order to purchase and install
the equipment. However, an asset
management in line with ISO 55000
and ISO/IEC 19770-x (4 layers of
maturity) is needed in order to leverage
the relevant knowledge about
assets. This is a precondition for
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correct patch management. It can be
well applied in many industries.
3.10 Secure human-machine
interaction
Main control rooms and I&C maintenance
rooms equipped with HMI
equipment are common for power
plants and stringently regulated for
NPPs, e.g. with regard to the explicit
documentation for plant operators.
Different solutions exist for secure
human-machine interaction. An example
of it is a Qualified Display
System (QDS), which limits functionalities
accessible to the operator. The
respective security provision may be
transferred or adapted for other HMI
related user activities.
3.11 Domain specific application
security controls
The semi-formal approach of the upcoming
ISO/IEC 27034-x is applied to
the nuclear context. A key concept is
the Application Security Controls
(ASCs). An ASC provides a semiformal
definition of a security control.
It also includes the indication of the
security grade that the ASC can meet,
the status of the ASC implementation
(e.g. whether verification and validation
were completed), the role assignment
according to RACI (Responsible,
Accountable, Consulted, Informed)
and the specification of links to other
ASCs. AREVA NP even considers advanced
features, like ASC inheritance,
not yet included in the current ISO/
IEC 27034-x standard versions.
As an example of the adaptation of
the ASCs concept, the default grading
of 10 levels of trust has to be adjusted
to the domain specific grading, or a
grading has to be introduced for
the target domain. Additionally, the
accompanying concepts of an Organization
Normative Framework and an
Application Normative Framework
can be adapted.
This is in line with the key concepts
of ASCs, which promote the development
and delivery of high-quality
specialised ASCs by standardsconforming
sub-suppliers.
3.12 Advanced persistent
threats
Targeted Advanced Persistent Threats
(APT), like Stuxnet, are the most
feared attack scenarios in any business
domain. The combination of several
of the aforementioned approaches,
including a comprehensive asset
management, semi-formal modelling
of the assets and supporting assets,
semi-formal description of the
| | Fig. 7.
Overview of cyber security portfolio.
Application Security Controls, targeted
security testing, Forensic
Readiness Security Controls and
further security controls related to the
secure software development will
support in systematically increasing
the security posture and thus, the
effort needed to be spent by an APT
agent.
Similar APT analysis can be performed
for other business domain,
provided the above listed preparations,
like asset management and
semi-formal modelling are already in
place or are implemented.
The knowledge areas described
above should be organised in different
products and services offered to
selected critical industries for efficient
application. An example of how to
implement this is shown in Figure 7.
Summary
Monitoring agencies like the “US
Industrial Control Systems Cyber
Emergency Response Team (ISC-
CERT)”, the “French National Agency
for Information systems’ security” and
the “German Federal Office for Information
Security” (BSI) all record steep
increases in cyberattacks on companies
and institutions in general, and
on critical infrastructures in particular.
For example, the BSI reported an
increase of 20% in the number of
known malicious program versions,
from 2015 to 2016, up to 560 million a
year. Hence, overall public awareness
of cyber security threats, as well as of
legislators, of power plant operators
and of their owners, is also on a steep
rise.
Preemptive cyber security measures
not only avoid loss of revenues,
costs of crisis management, costs of
reimbursements and higher insurance
premiums. They also avoid upcoming
legal penalties for infringement of
an increasingly intransigent legislation.
AREVA NP’s long-standing expertise
in nuclear cyber security relies on
in-depth knowledge of industrial and
legislative requirements and of the
corresponding companies’ protection
needs. As shown above, it applies to a
great extent to any industrial infrastructure
using control systems. Not
only the energy sector, but also the
manufacturing sector, the water and
wastewater systems sector and the
defense industrial base sector benefit
from such an expertise.
References
[1] IEC 62859:2016, Nuclear Power Plants –
I&C Systems – Requirements for Coordinating
Safety and Cyber security.
[2] IEC 62443-3-3:2013, Industrial communication
networks – Network and system
security – Part 3-3: System security
requirements and security levels.
[3] ISO/IEC 27009:2016, Information
technology – Security techniques –
Sector- specific application of ISO/IEC
27001 – Requirements.
[4] U.S. Nuclear Regulatory Commission
(2010). Regulatory Guide 5.71 Cyber
Security Programs for Nuclear Facilities.
Available at: https://www.nrc.gov/
docs/ML0903/ML090340159.pdf.
[5] Th. Poussier, S. Gomes-Augusto, K.
Waedt: Cyber security Aspects of a
Safety Display System. IAEA Inter national
Conference on Computer
Security in a Nuclear World: Expert
Discussion and Exchange, Vienna,
2015-06.
[6] OPC Foundation (2016), Unified Architecture:
Interoperability for Industrie 4.0
and the Internet of Things. Available at:
https://opcfoundation.org/wpcontent/uploads/2016/05/OPC-UA-
Interoperability-For-Industrie4-and-IoT-
EN-v5.pdf.
[7] (BMUB) Federal Ministry for the
Environ ment, Nature Conservation,
Building and Nuclear Safety (2015):
Constitution and Laws. Available at:
http://www.bmub.bund.de/en/topics/
nuclear-safety-radiological-protection/
nuclear-safety/legal-provisionstechnical-rules/constitution-and-laws/.
OPERATION AND NEW BUILD 401
Operation and New Build
Cyber Security in Nuclear Power Plants and its Portability to Other Industrial Infrastructures ı Sébastien Champigny, Deeksha Gupta, Venesa Watson and Karl Waedt

atw Vol. 62 (2017) | Issue 6 ı June
402
DECOMMISSIONING AND WASTE MANAGEMENT
[8] Department of Homeland Security
(2015). Cyber security Framework
Implementation Guidance for U.S.
Nuclear Power Reactors. Available at:
https://www.us-cert.gov/sites/default/
files/c3vp/framework_guidance/
nuclear-framework-implementationguide-2015-508.pdf.
Validation of Spent Nuclear Fuel Nuclide
Composition Data Using Percentage
Differences and Detailed Analysis
Man Cheol Kim
1 Introduction Nuclide composition data of spent nuclear fuels are important in many nuclear engineering
applications. In reactor physics, nuclear reactor design requires the nuclide composition and the corresponding cross
sections. In analyzing the radiological health effects of a severe accident on the public and the environment, the nuclide
composition in the reactor inventory is among the important input data. Nuclide composition data need to be provided
to analyze the possible environmental effects of a spent nuclear fuel repository. They will also be the basis for identifying
the origin of unidentified spent nuclear fuels or radioactive materials.
The Spent Fuel Isotopic Composition
(SFCOMPO) database [1–3], which
was originally developed by the Japan
Atomic Energy Research Institute and
is now managed by the Organization
for Economic Co-operation and Development/Nuclear
Energy Agency (OECD/
NEA), provides measured nuclide composition
data of spent nuclear fuels.
The SFCOMPO database has been
widely used to validate computer codes
and nuclear data libraries for spent
fuel and fuel cycle applications. For
example, Lee [4] validated TR4PEP, a
depletion code combining the continuous-energy
Monte Carlo transport
code TRIPOLI-4.3 and the point
depletion code PEPIN-2, using the
Takahama-3 post-irradiation examination
results provided in the SFCOMPO
database. Fast et al. [5] compared the
code calculation results obtained using
SCALE 6.0 and 6.1 with measurement
data from Obrigheim nuclear power
plant (NPP) to investigate the validity
of the correlations for burnup calculations
involving key nuclides such as
Cs-134, Cs-137, and Eu-154. Nicolaou
[6] tested the potential of isotopic
fingerprinting for nuclear forensics
purposes using the measurement data
provided in the SFCOMPO database.
With the recognition of the importance
of extending the SFCOMPO
database, the Expert Group on
Assay Data of Spent Nuclear Fuel
Authors
Sébastien Champigny
MBA, Dipl.-Phys., M.Eng.
Product manager cyber security for
critical infrastructures
Deeksha Gupta
M.Sc. in Nuclear Sci. & Tech.
Cyber security PhD Candidate
(EGADSNF), which is in charge of
main taining the OECD/NEA SFCOMPO
database, is trying to obtain new assay
data that are not open to the public or
are open but not widely available.
Suyama et al. [7] mentioned that the
use of the OECD/NEA framework was
intended to facilitate the collection of
new data from member countries.
Possible candidates for newly added
data are summarized in Gauld and
Rugama [8] and the state-of-the-art
report by EGADSNF [9]. For this
purpose, Suyama et al. [10] provided
additional measurement data from
Ohi-1 and Ohi-2 with detailed information
and specifications so that they can
be added to the SFCOMPO database.
Raap et al. [11] reported on the expansion
of the SFCOMPO database by
the addition of measured data from
CANDU reactors, MAGNOX reactors,
VVERs, and RBMKs for use in developing
isotopic signatures for nuclear
forensics purposes.
As the EGADSNF admitted in the
state-of-the-art report [9], measurement
data were added to the SFCOMPO
database as reported by laboratories,
without peer review. Validation of
the data to assess the quality of the
measurements is con sidered a priority
task for improvement of the SFCOMPO
database. The measurement data in
the SFCOMPO database have been
validated in several recent studies.
Venesa Watson
Master in Computer Forensics
Cyber security PhD Candidate
Dr. Karl Waedt
Senior expert Cyber Security
Concepts & Architecture
AREVA GmbH
Paul-Gossen-Straße 100
91052 Erlangen, Germany
Gauld et al. [12] described the recent
experience of Oak Ridge National
Laboratory in validating the measured
isotopic composition data of spent
nuclear fuel. Among the 118 PWR
experimental assay data, 87 (73.7 %)
were from the SFCOMPO database.
Gauld et al. [12] reported problems
such as highly erratic Am-241
measurement data from Takahama-3
due to possible errors in the adjustment
of the time of discharge and
physically impossible measurement
results due to possible typographical
errors. However, the details on how the
SFCOMPO database should be revised
are not clearly described. Okumura et
al. [13] described how the measurements
of Se-79, Tc-99, Sn-126, and
Cs-135 for the Cooper, Calvert Cliffs-1,
and H. B. Robinson-2 reactors in the
SFCOMPO database should be revised
by applying the latest nuclear data,
especially the half-lives of the four
nuclides. For example, the calculatedto-experimental
value for Se-79
changed from 5.5 to 0.92 after the
application of the latest half-life of
Se-79 provided by Bienvenu et al. [14].
This paper proposes a simple
method for analysis and validation of
nuclide composition data of spent
nuclear fuels such as those found in
the SFCOMPO database. The proposed
method consists of a simplified
code calculation, the assumption of a
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| | Fig. 1.
Distribution of U-235 initial enrichment and burnup of 246 samples in OECD/NEA SFCOMPO database.
constant power history, comparison of
the measured data and code calculation
results, and detailed analysis of
those data that deviate significantly
from other data to identify the causes
of the deviations. During such a crosscheck
process, many, if not all, of the
errors in either the measured data
or the code calculations could be
identified. The proposed method is
described in Section 2. The application
of the proposed method to
nuclide composition data of spent
nuclear fuels from Obrigheim NPP is
described in Section 3, and the identified
data and associated findings are
analyzed in detail. Section 4 presents
the conclusion of this paper.
2 Approach to validation
of nuclide composition
data
2.1 Overview of the data in
the SFCOMPO database
The OECD/NEA SFCOMPO database
provides 10,282 measurement data
from 246 samples of nuclear fuels
irradiated in 14 reactors. Figure 1
shows the distribution of the U-235
initial enrichment and burnup of 246
samples in the SFCOMPO database.
The initial enrichment ranges from
1.45 wt % for the samples from
Fukushima-Daiichi-3 and Monticello
to 4.11 wt % for those from
Takahama-3, and is concentrated
in the range from 2.5 to 3.5 wt %.
The burnup ranges from 2.21 to
71.84 GWd/MTU. Note, however, that
the samples from Monticello show
exceptionally high burnup (exceeding
40 MWd/MTU) despite the low initial
U-235 enrichment. Owing to this
abnormal overburn of low-enriched
fuels, Hermann et al. [15] rated
the data from Monticello as ‘not
recommended.’ If the samples from
Monticello are excluded, the burnup
ranges from 2.21 to 47.3 GWd/MTU.
2.2 Approach to validation
Nuclide composition data of spent
nuclear fuels such as those in the
OECD/NEA SFCOMPO database can
be validated using the percentage
differences between the computed
and measured compositions of each
isotope in each sample. Here, the
percentage difference is defined as the
calculated value minus the measured
value, divided by the measured
quantities or ratios, as follows:
(1)
The percentage difference has been
used in many code validation studies
such as those of Hermann et al. [16],
DeHart and Hermann [17], and Jang
et al. [18]. By reviewing the percentage
differences of nuclide compositions
or ratios, candidates for
detailed analysis can be identified.
The identified candidates can be
subjected to detailed analyses such as
consistency checks as a function of
burnup or parent–daughter pairs, as
performed by Gauld and Rugama [8],
and the original data sources can be
reviewed.
2.3 Code calculations
As mentioned in the state-of-the-art
report by EGADSNF [9], evaluation of
the measured data in the SFCOMPO
database requires significant effort
and is therefore a significant but
challenging objective of future activities
of the EGADSNF. Thus, using
simplified models in code calculations
would be more efficient for performing
a large number of code
calculations in a manageable and unified
way than using detailed models.
For this reason, code calculations
were performed using ORIGEN-ARP
[19,20]. Consequently, the code
calculations were performed mainly
based on important parameters such
as the fuel assembly type, initial
enrichment, burnup, operation
history, and cooling time of the samples.
Fast et al. [5] compared the code
calculation results of the ARP model
(simple and fast) and the NEWT
model (compli cated and precise) for a
sample from Obrigheim NPP and
found that the percentage differences
for most nuclides are within 20 % for
the APR model and within 10 % for
the NEWT model. To make code
calculations for a large number of
samples, the use of ORIGEN-ARP
provides an efficient way of calculating
nuclide compositions with sufficient
precision.
2.4 Consideration of operation
history
The degree of detail in the irradiation
history varies among the 14 reactors
in the OECD/NEA SFCOMPO database.
Very detailed information on the
cycle number, elapsed time, time
interval, core power density, bundle
power density, and so on is provided
for Cooper NPP. On the other hand, no
information is available for JPDR-I,
Tsuruga-1, and Takahama-3 NPPs in
the SFCOMPO database. The operation
history of Takahama-3 NPP is
available in NUREG/CR-6798 [21].
Research has found that code
calculations of nuclide compositions
are not significantly affected by the
detailed power history. Chabert et al.
[22] and Nakahara et al. [23] compared
the nuclide compositions of
spent fuels obtained using the accurate
power history and an assumed
constant power history, and found
that the principal uranium and
plutonium isotopes and other fission
products are not significantly affected
by the power history. Based on these
findings, a constant power history was
assumed as a reasonable approximation
instead of the accurate power
history, which was available for only a
limited number of samples.
3 Application to Obrigheim
NPP Nuclide Composition
Data
A total of 1,153 nuclide composition
or nuclide ratio data were provided
for 23 samples from Obrigheim NPP.
The samples were analyzed at two
different laboratories, Ispra and
Karlsruhe. Ispra analyzed 17 samples,
and Karlsruhe analyzed 10 samples.
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DECOMMISSIONING AND WASTE MANAGEMENT 404
| | Fig. 2.
Percentage differences between measured and computed nuclide compositions for Obrigheim NPP (measured at Ispra).
Four samples were analyzed in both
laboratories for cross-checking.
The proposed method was applied
to a total of 728 nuclide composition
or nuclide ratio data for 17 samples
measured at Ispra to identify those
data that deviate significantly from
other data and therefore became
the candidates for detailed analysis.
Various detailed analysis appropriate
to identify the root causes of the
significant deviation from other data
were applied and described below.
3.1 Burnup for GEROBRPWR-9
Figure 2 shows the percentage differences
between the calculated and
measured nuclide composition data for
the samples from Obrigheim NPP measured
at the Ispra laboratory. Relatively
high percentage differences between
the calculated and measured data were
observed for one sample, which
was found to be the GEROBRPWR-9
sample. According to Barbero et al.
[24], the burnup for the GEROBRPWR-9
sample measured at Karlsruhe using
the Nd-148 method (22,700 MWd/
MTU) was abnormally high compared
to the burnup measured at Ispra using
the Nd-148, non-destructive Cs-137,
and destructive Cs-137 methods
(17,130, 16,970, and 17,490 MWd/
MTU, respectively). Because the
burnup in the OECD/NEA SFCOMPO
database was based on the measurement
at Karlsruhe, the code calculation
was performed again using the burnup
measured at Ispra with the Nd-148
method (17,130 MWd/MTU). Figure 3
shows the percentage differences
after the burnup correction for the
GEROBRPWR-9 sample. It can be
seen that the deviations of the
GEROBRPWR-9 sample from other
samples were properly corrected.
Figure 3 also identifies those
nuclide composition or nuclide ratio
data that need to be analyzed in more
detail. The Pu-241/Pu-239 ratio
shows relatively high percentage
differences. The percentage differences
of Cs-137/U-238 ratio are
divided into two groups. A large
uncertainty and a large deviation in
percentage differences are observed
for Am-241 and Am-242, respectively.
Detailed analysis on each of the
identified nuclide composition or
nuclide ratio data are described in the
following sections.
3.2 Cooling time of plutonium
isotopes
As indicated in Fig. 3, the Pu-241/
Pu-239 ratio was found to have
higher percentage differences among
samples than other nuclide ratios.
Barbero et al. [24], the original source
of the measurement data, reported
the nuclide ratios of uranium and
plutonium with the dates of the
measurements. The actual cooling time
of the samples can be calculated from
the date of discharge (August 16, 1974)
and the date of measurement (e.g.,
April 12, 1978 for GEROBRPWR-3).
However, the cooling times of the
data were specified as zero in the
OECD/NEA SFCOMPO database,
which means that the data were
adjusted to the time of discharge. The
code calculations were performed
assuming that the cooling times of
the samples were zero. Because the
half-life of Pu-241 (14.325 years) is
comparable with the cooling time of
the samples, a non-negligible amount
of Pu-241 decayed out; therefore, the
calculated Pu-241/Pu-239 ratios were
higher than the measured ones.
| | Fig. 3.
Percentage differences between measured and computed nuclide compositions for Obrigheim NPP (measured at Ispra) after burnup
correction for GEROBRPWR-9 sample.
3.3 Possible errors during
Cs-137/U-238 ratio calculation
from measured data
As indicated in Fig. 3, the Cs-137/
U-238 ratios were found to fall into
two groups. One group consists of 5
samples with very small percentage
differences, and the other group consists
of 12 samples with percentage
differences of up to −35 %. The existence
of two distinct groups motivated
a detailed analysis considering possible
systematic errors in the data.
Figure 4 shows the amounts of
Cs-137 buildup and the remaining
U-238 in the spent nuclear fuel per
metric ton of final uranium. As a
burnup monitor, the amount of Cs-137
buildup shows a linear relationship
with the burnup. Although the
amount of U-238 remaining shows a
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| | Fig. 4.
Cs-137 buildup and remaining U-238 as functions of burnup for Obrigheim NPP
(measured at Ispra).
very slight decreasing trend as the
burnup increases, it can be considered
as constant at around 950 kg/MTU.
Therefore, the Cs-137/U-238 ratio is
expected to show a linear relationship
with the burnup.
Figure 5 shows the relationship
between the Cs-137/U-238 ratio and
the burnup for the data in the OECD/
NEA SFCOMPO database for
Obrigheim NPP measured at Ispra. It
was first confirmed that the data in
the SFCOMPO database are identical
with those in the original source,
Barbero et al. [24]. It can be seen that
two distinct relationships exist in
Fig. 5. The lower part is the 5 samples
with very small percentage differences.
The upper part is the 12
samples with percentage differences
of up to −35 %, which means that the
measured quantities are about 50 %
higher than the calculated quantities.
Because the half-life of Cs-137 (30.08
years) is much shorter than that of
U-238 (4.47E+09 years), the Cs-137/
U-238 ratio is expected to decrease
gradually as time passes; therefore,
the cooling time would not make the
measured quantities much higher
than the calculated quantities, which
assumed zero cooling time.
Table 1 compares the Cs-137/
U-238 data obtained directly from the
OECD/NEA SFCOMPO database, the
data calculated from the Cs-137 and
U-238 quantities provided in the database,
and the data calculated using the
code. From Fig. 4, the isotopic quantities
of Cs-137 and U-238 from the
database seem to be free of error;
hence, the Cs-137/U-238 ratio from
| | Fig. 5.
Cs-137/U-238 as a function of burnup for Obrigheim NPP (measured at Ispra).
the two isotopic quantities is expected
to provide correct results. In Tab. 1, the
‘Percentage difference I’ column lists
the difference between the Cs-137/
U-238 values obtained directly from
the database and those calculated from
the nuclide composition data for the
two nuclides (Cs-137, U-238) divided
by the Cs-137/U-238 values taken
directly from the database. The nuclide
composition data of the two nuclides in
Tab. 1 are given in kilograms per
metric ton of final uranium (kg/
MTU final). The 12 data points with
percentage differences of up to −35 %
also appear in the ‘Percentage difference
I’ column.
Code calculation results obtained
using ORIGEN-ARP are also provided
in Tab. 1, with the associated percentage
difference denoted as Percentage
DECOMMISSIONING AND WASTE MANAGEMENT 405
Burnup
(MWd/
MTU)
Cs-137
(SFCOMPO)
U-238
(SFCOMPO)
Cs-137/
U-238
(SFCOMPO)
Cs-137/
U-238
(Calculation)
Percentage
difference I
(Calculation)
Cs-137/
U-238
(Code)
Percentage
difference II
(Code)
Percentage
difference III
(Calc. vs. Code)
GEROBRPWR-2 27,900 1.88.E-03 1.88.E-03 0.24 %
GEROBRPWR-3 33,800 1.21.E+00 945 2.33.E-03 2.22.E-03 −4.53 % 2.29.E-03 −1.72 % 2.95 %
GEROBRPWR-4 20,200 7.16.E-01 957 1.35.E-03 1.30.E-03 −3.72 % 1.35.E-03 0.34 % 4.22 %
GEROBRPWR-6 36,300 1.29.E+00 943 2.48.E-03 2.38.E-03 −4.17 % 2.46.E-03 −0.61 % 3.72 %
GEROBRPWR-7 30,900 1.12.E+00 948 2.15.E-03 2.05.E-03 −4.54 % 2.09.E-03 −2.65 % 1.98 %
GEROBRPWR-8 22,900 8.27.E-01 953 2.36.E-03 1.51.E-03 −36.12 % 1.54.E-03 −34.85 % 1.99 %
GEROBRPWR-9 17,100 6.22.E-01 958 1.77.E-03 1.13.E-03 −36.28 % 1.15.E-03 −35.18 % 1.72 %
GEROBRPWR-10 25,800 9.11.E-01 951 2.59.E-03 1.66.E-03 −35.75 % 1.74.E-03 −32.77 % 4.63 %
GEROBRPWR-11 31,500 1.15.E+00 947 3.28.E-03 2.11.E-03 −35.68 % 2.11.E-03 −35.72 % −0.07 %
GEROBRPWR-12 27,700 1.00.E+00 948 2.86.E-03 1.83.E-03 −35.93 % 1.87.E-03 −34.57 % 2.12 %
GEROBRPWR-15 29,400 1.06.E+00 948 3.01.E-03 1.94.E-03 −35.47 % 1.98.E-03 −34.09 % 2.14 %
GEROBRPWR-17 38,100 1.38.E+00 942 3.94.E-03 2.54.E-03 −35.41 % 2.53.E-03 −35.74 % −0.51 %
GEROBRPWR-18 35,600 1.30.E+00 945 3.70.E-03 2.39.E-03 −35.41 % 2.44.E-03 −33.93 % 2.28 %
GEROBRPWR-19 30,200 1.12.E+00 951 3.20.E-03 2.05.E-03 −36.06 % 2.06.E-03 −35.68 % 0.60 %
GEROBRPWR-20 24,200 8.91.E-01 955 2.55.E-03 1.62.E-03 −36.44 % 1.64.E-03 −35.50 % 1.47 %
GEROBRPWR-21 25,500 8.34.E-01 953 2.39.E-03 1.52.E-03 −36.39 % 1.73.E-03 −27.62 % 13.79 %
GEROBRPWR-22 36,700 1.31.E+00 944 3.75.E-03 2.41.E-03 −35.71 % 2.52.E-03 −32.89 % 4.40 %
| | Tab. 1.
Comparison of Cs-137/U-238 data obtained directly from OECD/NEA SFCOMPO database, data calculated from isotope quantities provided in the database, and data calculated using the code.
Decommissioning and Waste Management
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim

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DECOMMISSIONING AND WASTE MANAGEMENT 406
| | Fig. 6.
Correlation of measured and calculated Am-241 data with burnup for BE124 and BE210 fuel assemblies
of Obrigheim NPP.
difference II. The definition of
Percentage difference II is similar to
that of Percentage difference I, except
that the code calculation results
replace the Cs-137/U-238 values calculated
from the nuclide composition
data for the two nuclides (Cs-137,
U-238). The 12 data points with percentage
differences of up to −35 %
can also be seen in the ‘Percentage
difference II’ column.
The ‘Percentage difference III’
column lists the differences between
the code calculation results and the
Cs-137/U-238 values calculated from
the nuclide composition data for the
two nuclides (Cs-137, U-238) divided
by the calculated Cs-137/U-238
values. The magnitudes of the percentage
differences in this column
are less than 5 %, except for one
case. Therefore, in the OECD/NEA
SFCOMPO database, the nuclide
composition data for these two
nuclides (Cs-137, U-238) seem to be
trustworthy, whereas the Cs-137/
U-238 data taken directly from
the database can be suspected
of including calculation errors.
Recalculation of the Cs-137/U-238
data in the OECD/NEA SFCOMPO
database using the measured data
for the two nuclides (Cs-137, U-238)
is recommended.
in the percentage differences for
Am-241 (up to 85 %) and Cm-242
(70 % for Karlsruhe and 45 % for
Ispra). Gauld et al. [12] also indicated
possible problems in back-calculating
isotopic compositions to a reference
date, using the measurement data
of Am-241 for Takahama-3 NPP as
an example. Note that most of
the Am-241 (half-life = 432.6 years)
at the time of measurement is from
the decay of Pu-241 (half-life =
14.325 years).
Figure 6 shows the correlation of
the measured and calculated Am-241
data with the burnup for the BE124
and BE210 fuel assemblies at
Obrigheim NPP. First, it was confirmed
that the data provided in the OECD/
NEA SFCOMPO database are identical
with those provided in the original
data source, Barbero et al. [24].
Although a gradual increase in
Am-241 quantities is expected as the
burnup increases, as can be seen in
the calculated data, a large variation
is observed in the measured data
as the burnup increases, especially
when the burnup is 31.5 GWd/MTU
( GEROBRPWR-11). Some measured
data, especially those within the
ellipse in Fig. 6, appear to be inconsistent
with other measured data.
The measured Am-241 data were
obtained by mass spectrometry or
alpha spectrometry at Ispra or alpha
spectrometry at Karlsruhe. All of the
measured data that were found to be
larger (sometimes significantly larger)
than the calculated data, which are
indicated by an ellipse in Fig. 6, were
obtained by alpha spectrometry at
Ispra. In addition, 8 of the 10 values
measured by alpha spectrometry at
Ispra are found in the ellipse. The
measurement data obtained by mass
spectrometry at Ispra and alpha
spectrometry at Karlsruhe are found
to be relatively close to the calculated
data. Owing to the relatively high
inaccuracy of the measured data
obtained by alpha spectrometry at
Ispra compared to the measured data
obtained by mass spectrometry
at Ispra and alpha spectrometry at
Karlsruhe, as shown in Fig. 6, a
detailed review of the measured data
obtained by alpha spectrometry at
Ispra seems to be necessary.
Figure 7 shows the correlation of
the measured and calculated Am-242
data from Obrigheim NPP with the
burnup. Note that the scale of the
measured data (left-hand Y-axis) is 10
times larger than that of the calculated
data (right-hand Y-axis). Figure 7
shows that the measured Am-242
data are about 10 times larger than
the calculated data for the seven
samples from Obrigheim NPP.
Because Cm-242 is produced by the
decay of Am-242, the ratio of the halflives
of Am-242 and Cm-242 provides
information on the atomic ratio of the
3.4 Large uncertainty in
Am-241 and large
deviation in Am-242
As indicated in Fig. 3, a large uncertainty
between the measured and
calculated data for Am-241 and a
large deviation between the measured
and calculated data for Am-242 were
observed. In a comparison of the
measured and computed data for a
sample from Obrigheim NPP, Fast et
al. [5] also observed a high deviation
| | Fig. 7.
Correlation of measured and calculated Am-242 data from Obrigheim NPP with burnup.
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Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim

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| | Fig. 8.
Ratios of measured Cm-242 data to measured and calculated Am-242 data for the seven samples from
Obrigheim NPP.
two nuclides in equilibrium. The halflives
of Am-242 and Cm-242 are
16.02 hours and 162.8 days, respectively,
and 82.70 % of Am-242 goes
through beta decay to form Cm-242.
Therefore, the measured values for
Cm-242 are expected to be about
200 times larger than the measured
values for Am-242 in equilibrium.
Figure 8 shows the ratios of the measured
Cm-242 values to the measured
and calculated Am-242 values for the
seven samples from Obrigheim NPP.
Although the ratio of the Cm-242
values to the calculated Am-242 values
is around 150, the ratio of the Cm-242
values to the measured Am-242 values
is generally less than 20. Therefore, it
is likely that the measured Am-242
values overestimate the actual quantity
of Am-242 by about 10 times. Detailed
analysis of how the measured Am-242
values were derived from the raw
experimental data to the data provided
in the OECD/NEA SFCOMPO database
seems to be necessary.
4 Conclusions
Nuclide composition data of spent
nuclear fuels such as those provided in
the OECD/NEA SFCOMPO database
are important in many fields including
reactor physics, fuel cycle applications,
radiological consequence
analysis, and nuclear forensics. To
reduce unnecessary uncertainties
associated with nuclide composition
data, the validation of such data is a
high-priority task.
As one of the first steps, a simple
method is proposed for identifying
the nuclide composition data that
may include errors and therefore
require detailed analysis or further
investigation. The proposed method
is based on the ORIGEN-ARP code
calculation, the assumption of a
constant power history, the percentage
differences of the calculated and
measured composition data, and
detailed analysis of the identified
data. The application of the proposed
method to the nuclide composition
data of spent nuclear fuels from
Obrigheim NPP demonstrated that
the method can effectively identify
various possible errors or data that
need to be further investigated. Errors
identified during detailed analysis
or possible errors that require further
investigation include:
• Errors in burnup measurement
(e.g., GEROBRPWR-9)
• Errors in properly considering the
cooling time (e.g., Pu-241/Pu-239)
• Errors in the ratio calculation from
measured data (e.g., Cs-137/
U-238)
• Possible systematic errors in
measurements of isotopic composition
(e.g., Am-241 and Am-242
measurements at Ispra)
Although the nuclide composition
data that were not identified as
needing detailed analysis or further
investigation cannot necessarily be
considered as definitively validated, it
is believed that the proposed method
can identify a significant portion of
the errors in the nuclide composition
data. Despite the simplicity of the
proposed method, it is believed to be a
very efficient method of identifying
those nuclide composition data that
require detailed analysis or further
investigation. For this reason, the proposed
method is expected to be useful
as the first step in validation of nuclide
composition data of spent nuclear
fuels such as those in the OECD/NEA
SFCOMPO database.
Acknowledgements
This work was supported by a grant
from the Nuclear Safety Research
Program of the Korea Foundation of
Nuclear Safety, with funding from the
Korean government's Nuclear Safety
and Security Commission. This work
was also supported by a grant from
the Nuclear Research & Development
Program of the National Research
Foundation of Korea, which is funded
by the Korean government's Ministry
of Science, ICT & Future Planning
(Grant Code:
NRF-2016M2B2A9A02945211).
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Hiroki Sakamoto and Toshiyuki Kaneko.
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96-036, Japan Atomic Energy Research
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[3] H. Mochizuki, K. Suyama, Y. Nomura,
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[5] Ivan Fast, Yuliya Aksyutina, Holger
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[6] G. Nicolaou. Discrimination of spent
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[7] Kenya Suyama, Ali Nouri, Hirold
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[8] Ian C. Gauld, Yolanda Rugama. Activities
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DECOMMISSIONING AND WASTE MANAGEMENT 407
Decommissioning and Waste Management
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim

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408
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[9] Expert Group on Assay Data of Spent
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Data for Isotopic Validation - State-ofthe-art
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DOC(2011)5, Organisation for
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[10] Kenya Suyama, Minoru Murazaki,
Kiyoshi Ohkubo, Yoshinori Nakahara,
Gunzo Uchiyama. Re-evaluation of
Assay Data of Spent Nuclear Fuel
obtained at Japan Atomic Energy
Research Institute for validation of
burnup calculation code systems,
Annals of Nuclear Energy, vol.38,
pp.930-941, 2011.
[11] M.C. Brady Raap, B.A. Collins, J.A. Lyons,
J.V. Livingston. FY13 Summary Report
on the Augmentation of the Spent Fuel
Composition Dataset for Nuclear
Forensics: SFCOMPO/NF, PNNL-23225,
Pacific Northwest National Laboratory,
Richland, Washington, March 2014.
[12] Ian C. Gauld, Georgeta Radulescu,
Germina Ilas. SCALE Validation
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Burnup Credit (BUC) Workshop,
Cordoba, Spain, October 2009.
[13] Keisuke Okumura, Shiho Asai, Yukiko
Hanzawa, Hideya Suzuki, Masaaki
Toshimitsu, Jun Inagawa, Tsutomu
Okamoto, Nobuo Shinohara, Satoru
Kaneko, Kensuke Suzuki. Analyses of
Assay Data of LWR Spent Nuclear Fuels
with a Continuous-Energy Monte Carlo
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135Cs, Progress in NUCLEAR SCIENCE
and TECHNOLOGY, Vol. 2, pp.369-374,
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[14] Philippe Bienvenu, Philippe Cassette,
Gilbert Andreoletti, Marie-Martine Bé,
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[15] O. W. Hermann, M. D. DeHart, and B. D.
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spent fuel composition data for use in
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Tennessee, February 1998.
[16] O. W. Hermann, S. M. Bowman, M. C.
Brady, C. V. Parks. Validation of the
SCALE System for PWR Spent Fuel
Nuclide composition Analyses, ORNL/
TM-12667, Oak Ridge National Laboratory,
Oak Ridge, TN, March 1995.
[17] M. D. DeHart, O. W. Hermann. An
Extension of the Validation of SCALE
(SAS2H) Isotopic Predictions for PWR
Spent Fuel, ORNL/TM-13317, Oak
Ridge National Laboratory, Oak Ridge,
TN, September 1996.
[18] Jeong-nam Jang, Hyung-moon Kwon,
Jung-suk Kim, Yong-bum Chun.
Validation of SCALE SAS2H Isotopic
Predictions for high burnup PWR spent
fuels, Transactions of the 2009 Korean
Nuclear Society Spring Meeting, Jeju,
Korea, May 22, 2009.
[19] M. J. Bell. ORIGEN – the ORNL isotope
generation and depletion code,
ORNL-4628, Oak Ridge National Laboratory,
Oak Ride, Tennessee, May 1973.
[20] http://scale.ornl.gov/origen-arp.shtml
[21] C. E. Sanders, L C. Gauld, R. Y. Lee.
Isotopic Analysis of High-Burnup PWR
Spent Fuel Samples From the
Takahama-3 Reactor, NUREG/CR-6798,
ORNL/TM-2001/259, United States
Nuclear Regulatory Commission,
Washington, DC, January 2003.
[22] Christine Chabert, Alain Santamarina,
Robin Dorel, Didier Biron, Christine
Poinot-Salanon. Qualification of the
APOLLO 2 assembly code using PWR-
UO2 isotopic assays – the importance of
irradiation history and thermomechanics
onfuel inventory prediction,
Proceedings of the American Nuclear
Society International Topical Meeting
on Advances in Reactor Physics, and
Mathematics and Computation Into
the Next Millennium (PHYSOR-2000),
Pittsburgh, Pennsylvania, May 7-11,
2000.
[23] Yoshinori Nakahara, Kenya Suyama,
and Takenori Suzaki. Technical
Development on Burn-up Credit for
Spent LWR Fuels, (Eds.), JAERI-Tech
2000-071, Japan Atomic Energy
Research Institute (JAERI), 2000 (in
Japanese). Translation published as
ORNL/TR-2001/01, Oak Ridge National
Laboratory, 2002.
[24] P.Barbero et.al. Post Irradiation Analysis
of The Obrigheim PWR Spent Fuel.
Nuclear Science and Technology, 1980.
Figure captions
Author
Man Cheol Kim
School of Energy Systems
Engineering
Chung-Ang University
84 Heukseok-ro
Dongjak-gu, Seoul 06974, Korea
Reliability Analysis on Passive Residual
Heat Removal of AP1000 Based on Grey
Model
Qi Shi, Zhou Tao, Muhammad Ali Shahzad, Li Yu and Jiang Guangming
1 Introduction It is common to base the design of passive systems [1, 2] on the natural laws of physics, such
as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated
with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating
conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can
affect the function of the system, and the failure to achieve its desired aim is termed as function failure [3].
In the reliability analysis of the passive
systems, the main sources of the
uncertainty [4] are the numerical
errors in the calculation program such
as RELAP5 and the reactor parameters.
However, a lot of experience is required
to analyze the error propagation in
such system codes. The difficult is
increased by the fact that AP1000 has
not been connected to the grid yet. In
this paper, more focus has been placed
on the uncertainties of design and
operation parameters of the reactor.
The analytic hierarchy process (AHP)
[5, 6] and artificial neural network
(ANN) [7] have been applied, in order
to perform a sensitivity analysis on different
parameters of the passive safety
systems. However, there are large
subjective qualitative considerations in
the AHP. On the other hand, ANN has a
large amount of randomness, thus
requiring a large amount of data for its
training. Hence, these methods have
many limitations. The grey correlation
method [8]-[9], which has been
applied in many fields, can make up for
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| | Fig. 1.
Passive Residual Heat Removal System.
the limitations of above statistical
methods. It does not require a large
amount of data and its results
are consistent with the qualitative
implications. For sensitivity analysis on
PRHRS, it is rare to find the application
of Grey correlation method in literature.
The grey derivative and differential
equations are defined in the grey
system model [10] to establish the
dynamic prediction, based on concepts
of space relevance and smooth discrete
function. It is used to forecast the
passive system parameters.
For AP1000, the loss of normal
feedwater accident has been taken as
an example in this paper, which
involves the drop of control rod, operation
of PRHRS, coolant pump outage
and so on. Grey correlation is used to
analyze the importance of influencing
factors on PRHRS, and the Grey model
plays an important role for predicting
the data. This provides a new viewpoint
for studying the PRHRS of
AP1000.
2 Research object
2.1 System description
As an important part of AP1000
passive core cooling system, PRHRS
[11] is used to remove the decay heat
of the core for ensuring the safety of
reactor during the accident operating
conditions. It consists of a ‘C’ type
heat exchanger, an in-containment
refueling water storage tank (IRWST)
and the corresponding pipes and
valves.
Figure 1 shows the layout of
PRHRS in AP1000. The PRHRS heat
exchanger is located at a higher elevation
than the reactor core. Its upper
head is connected to the hot leg of
reactor coolant system (RCS) and its
lower head is connected to the lower
head of the steam generator. There is
an electro valve normally in the open
state located on the inlet line. Two
parallel pneumatic valves normally in
the close state are located on the
outlet line. Once the accident takes
place, the pneumatic valves are
opened by pneumatic signal, and the
electro valve also receives a signal to
confirm its open state. Corresponding,
a completely natural circulation loop
is established and the decay heat is
removed by the IRWST using density
difference.
It is complex to gage the factors,
affecting the heat transfer capacity of
PRHRS, and its interactions with RCS.
Hence, the parametric uncertainties
have different effects on PRHRS under
different accident conditions. Loss of
normal feedwater accident belongs to
the second type accidents, meaning
medium frequency accident. When
the accident takes place, the decay
heat needs to be adequately removed
from the reactor core. Otherwise, the
reactor core may be damaged. In this
paper, the influence of parametric
uncertainties on PRHRS has been
studied under the loss of normal feedwater
accident.
X 1
X 2
| | Tab. 1.
Parameters and corresponding distribution.
2.2 Identification and quantification
of uncertainties
PRHRS relies on natural circulation,
which has a much weaker driving
force than the active systems. A small
deviation from the design and operating
conditions can lead to function
failure. According to the criterion of
International Atomic Energy Agency
(IAEA) [13], the PRHRS is considered
to fail in providing its safety function,
once the maximum coolant temperature
at the reactor outlet exceeds
beyond 350 °C, in order to avoid fuel
cladding damage. Only the epistemic
uncertainties have been considered in
the present analysis. In accordance
with the previous studies [14, 15],
some design and operating parameters
have been selected for analysis
together with the corresponding
distribution as shown in Table. 1.
2.3 Thermal hydraulic model
and verification
According to the current research
results [12], RELAP5 program can be
used to analyze the loss of normal
feedwater accident in AP1000. It has
been developed by Idaho National
Engineering and Enviromental Laboratory
for transient thermal hydraulic
analysis in light water reactors. In this
paper, RELAP5/MOD3.4 has been
used for the calculation. The nodalization
scheme of AP1000 is shown in
Figure 2.
As depicted in Fig. 2, the model
consists of reactor core, two steam
generators, pressurizer, reactor coolant
pump, PRHRS core makeup tank
(CMT) and so on. Some preliminary
calculations are performed to determine
the steady-state parameters of
AP1000 using RELAP5, in order to
ensure their consistency with the
reference standards. According to
the sequence events in transient
con ditions, some factors have been
Variable Description Average value Standard deviation Distribution
T
(K)
D
(mm)
Temperature
of IRWST
Diameter
of PRHR HX
X 3 Kin
Resistance
coefficient of inlet
X 4
X 5
X 6
H
(m)
P
(MPa)
Q
(MW)
Height of ascending
Pipeline
300 10 Normal
0.0162 0.002 Normal
50 17 Normal
9.0 0.33 Normal
Initial pressure 15.5 0.5 Normal
Initial power level 3415 11.6 Normal
RESEARCH AND INNOVATION 409
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RESEARCH AND INNOVATION 410
| | Fig. 2.
Nodalization for the primary system of AP1000.
compared with the results of
LOFTRAN, such as pressure, flow rate
etc. The comparison results show that
RELAP5 has the capability to predict
the system parameters [11] correctly.
Among above these parameters, the
reactor coolant temperature is the
most important parameter for the
PRHRS loop, as shown in Figure 3.
As depicted in Fig. 3, the results of
RELAP5 and LOFTRAN exhibit the
same trend. A similar value of maximum
temperature has been observed
in the two results. During the accident,
however, a difference in the sequence
of events and response of the reactor
control system can lead to a slight
difference in the temperature trend.
This has no significant influence on
the analysis.
3 Calculation methods
3.1 Latin hypercube sampling
Latin Hypercube Sampling (LHS) [16,
17] is an improvement over the traditional
Monte Carlo sampling method.
It can overcome its drawbacks, in that
most of sampled results lie near the
average value. In order to improve the
accuracy of the parameters, therefore,
the method covers the upper and
lower limits of the distributions.
Hence, this method has the ability to
determine any value, as long as the
parameters are known. The steps are
as follows.
(1) For each variable, the probability
distribution is divided into N nonoverlapping
equal probability
interval [0, 1/N], [1/N, 2/N],…,
[(N-1)/N, 1]. This ensures that
the degree of the correlation of
LHS is small.
(2) The random standard normal
sample matrix Z N×n is used to
represent the order of sample
points, and the integer matrix
R N×n is used to record information
regarding the ordering of the
above sample points. Hence,
R ij = k shows that the sequence
of the j th variable in the i th sampling
is k.
(3) According to in each interval, the
cumulative probability function
of each sample point in the Latin
hypercube can be obtained randomly,
as shown in Eq. (1).
(1)
Here, i = 1,..., N and j = 1,..., n.
The function r a n d (0,1) represents
a random number, which is
uniformly distributed in the [0,1]
interval.
(4) In the Latin hypercube, the sampling
points are obtained by the
method of equal probability
change, as shown in Eq. (2).
(2)
Here, φ –1 (.) is the inverse normal
distribution function.
3.2 Grey Relation Method
The Grey Relational Method [8] is a
quantitative technique for comparative
analysis. The basic idea is to determine
the exponent of each factor in
the correlation, according to their
degree of similarity with the geometry
of sequence curve. If the curve is close
for a particular factor, it would have a
high exponent in the correlation. X 0 is
defined as the target parameter, with
k referring to the sequence of the
parameter, denoted as {X 0 (k)}. It is
assumed that there are a total of m
control parameters, and a parameter j
in the same sequence k is called the
comparison sequence, denoted as
{X j (k)} (k = 1,…, N)(j = 1,…, m). In
this correlation, the exponent of each
factor can be obtained by comparing
the tendency of development between
the target and influence parameters.
These steps are shown as follows [18].
(1) The reference and comparison sequences
are normalized, as shown
in Eq. (3).
(3)
(2) The absolute value of difference
between the reference and the
comparison sequences is calculated
as Eq. (4) based on above normalization
results.
(4)
a) Results of RELAP5. b) Results of LOFTRAN.
| | Fig. 3.
Reactor coolant temperature.
(3) Maximum and minimum absolute
values are calculated shown as
Eq. (5)-(6).
(5)
(6)
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Where, the Δ max is the maximum
value of the absolute difference
from m control parameters in
accordance with j = 1,…, m after
finding N number of maximum
absolute differences in the j-type
control parameter. The procedure
of determining Δ min is similar
to Δ max .
(4) The correlation coefficients
between reference and comparison
sequences are calculated by using
Eq. (7).
(7)
Where, ρ is the resolution ratio in
the (0,1) interval.
(5) The degree of correlation γ 0j is
determined, as shown in Eq. (8).
(8)
The corresponding differential equation
of GM(1,h) is given as
(13)
For determining the â and YN parameters,
the general equations are given
as follows.
(14)
(15)
(16)
(17)
Hence, Eq. (12) can be solved by using
Eq. (15), (16), (17)
| | Fig. 4a.
Natural circulation mass flow.
RESEARCH AND INNOVATION 411
3.3 Grey Model Method
In the grey system theory [10],
the grey derivative and differential
equations are defined to establish
dynamic prediction model based on
the concepts of space relevance and
smooth discrete function. This model
(GM) is used to forecast the passive
system parameters. The general
equation of grey model is written as
GM(n, h), where the variable h is
expressed by n th –order differential
equation. The procedure is discussed
as follows.
The correlation sequence is calculated
as follows.
(9)
Corresponding accumulative value
sequence is calculated by
Where
(10)
The generated sequences of corresponding
means value is calculated
by
(11)
The algebraic equation of GM(1,h) is
written as
(12)
4 Results and analysis
(18)
4.1 Parameter uncertainties
For correlation analysis, Mendenhall
[19] reports that the sample sizes are
5 to 10 times larger than the variable.
In this paper, the key parameters,
which affect the coolant temperature,
have been shown in Tab. 1 sampled by
LHS with a size of 100. After identification
of 100 combinations for the key
parameters, RELAP5 model as verified
in Section 2.3 is used for the analysis.
Figue 4 shows 25 groups of parametric
uncertainties, which influence
the natural circulation flow rate and
coolant temperature at the outlet of
the reactor core.
As seen from Fig. 4a, the PRHRS
actuates at 400 s with an initial mass
flow rate of 250 kg/s. At 1400 s, the
reactor coolant pimp stops and the
mass flow decreases to 125 kg/s. To
drive the natural circulation system,
PRHRS relies on the difference of
density between cold and hot sections
to take away the decay heat from
the reactor core. After actuation of
the reactor safety systems, the core
temperature decreases and there is a
rise in the IRWST temperature. During
this time, there is a lesser density
difference which leads to a decrease in
the mass flow of PRHRS. The mass
flow is maintained until the decay
heat power is balanced with the
| | Fig. 4b.
Coolant outlet temperature.
cooling capacity. There are great
changes in the mass flow during
natural circulation due to uncertainty
in the parameters. This has an effect
on the coolant temperature.
As seen from Fig. 4b, the PRHRS
starts at 400 s with an initial temperature
of 560 K. The CMT is actuated at
1,500 s together with the corresponding
systems, limiting the outlet
temperature to 550 K. At 6,000 s, this
temperature decreases to about 530 K.
There is a 50 K variation in the outlet
temperature of the coolant, owing to
uncertainty in the parameters.
4.2 Grey correlation analysis
As mentioned before, maximum outlet
temperature of the coolant should
always be kept below 350 °C, in order
to avoid function failure of PRHRS.
Assuming X 0 as the maximum outlet
temperature of the coolant, X 1 as the
temperature of IRWST, X 2 as the
diameter of PRHR HX, X 3 as the
resistance coefficient of inlet, X 4 as
height of ascending pipe, X 5 as initial
pressure, and X 6 as Initial power level,
the correlation is created using
RELAP5. Eq. (3)-(8) are used for
the grey correlation with resolution
coefficients of 0.1, 0.2, and 0.5. The
results are shown in Figure 5.
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RESEARCH AND INNOVATION 412
| | Fig. 5.
Correlation degree in different resolution.
As seen from Fig. 5, using different
resolutions, the factors have different
effect on the maximum outlet temperature
of the coolant. There is a
little difference between the effects of
different factors, when the resolution
is 0.5. It shows that the above
parameters are not well distributed.
The individual characteristics of
the above parameters are gradually
distinguished, once the resolution is
reduced from 0.5 to 0.1. The order of
these parameters is established, based
on their degree of the importance, X 6 ,
X 1 , X 4 , X 5 , X 2 , X 3 . The initial power has
the strongest influence on the decay
heat after shutdown. This is followed
by temperature of IRWST, which is
the cooling source for the core. As
the height of the ascending pipe is
| | Fig. 6.
Error analysis.
increased, there is a greater density
difference between cold and hot
sections, further increasing the mass
flow rate of PRHRS. Other parameters
X5, X2 and X3 have a less pronounced
effect on the maximum outlet temperature
of the coolant.
4.3 GM(1,6) model
The Grey correlation has been used
to determine the influence of each
parameter on the maximum coolant
temperature at the reactor’s output.
Considering the relationship among
the parameters, the coolant fluid
temperature (X 0 ) is considered to
represent the main behavior of the
system. The temperature of IRWST
(X 1 ), the diameter of PRHR HX(X 2 ),
resistance coefficient (X 3 ), height of
ascending pipe(X 4 ), initial pressure(X
5 ) and the initial power level(X 6 )
are considered to represent the correlated
behavior factors. According to
Eq. (9)-(18), an in-house code has
been used to build GM (1,6) model.
The code randomly selects 90 groups
and the other 10 groups are used to
validate. The corresponding differential
equation is shown as Eq. (19).
(19)
Figure 6 shows the errors in the
coolant temperature as determined
by the results from GM(1,6) and
RELAP5.
As seen in Fig. 6, the results of
GM(1,6) agree well with RELAP5,
and the errors fall within 15%. The
Grey model can adequately predict
maximum coolant temperature at the
outlet of the reactor core using a small
amount of data, making up for the
deficiency of artificial neural network
(ANN), which becomes unstable with
a small amount of data. This is a
new way to replace thermal-hydraulic
model.
5 Conclusion
By taking the loss of normal feedwater
in AP1000 as an example, the behavior
of PRHRS has been analyzed with the
help of RELAP5, and the Grey system
method has been applied for calculating
the maximum coolant temperature
at the outlet of the reactor
core. The following conclusions are
drawn.
(1) The degree of Grey correlation is
used to analyze the importance of
influencing factors. Smaller the
resolution, more obvious is the
difference among these factors.
The behavior of the factors can be
distinguished very easily, when the
resolution is 0.1.
(2) The initial reactor power has the
greatest influence on the maximum
coolant temperature at the
reactor outlet. And the sequences
is followed by the temperature of
IRWST, height of ascending pipe,
initial pressure. Correspondingly,
the diameter and resistance coefficient
of PRHRS-HX have a lesser
effect.
(3) The GM(1,6) model is built to
predict maximum coolant temperature
at the reactor outlet. All
errors fall within 15 % range.
Research and Innovation
Reliability Analysis on Passive Residual Heat Removal of AP1000 Based on Grey Model ı Qi Shi, Zhou Tao, Muhammad Ali Shahzad, Li Yu and Jiang Guangming

atw Vol. 62 (2017) | Issue 6 ı June
Acknowledgments
The research has been funded by
Science and Technology on Reactor
System Design Technology Laboratory
Funds (2015BJ0151).
References
[1] Zhou Tao, Li Jingjing, Ru Xiaolong, et al.
Application and development of
passive technology in nuclear power
units [J]. Proceedings of the CSEE, 2013,
33(8):81-89.
[2] Zhou Tao. Passive concept and
technology [M]. Beijing, Tsinghua
university press, 2016.
[3] Burgazzi L. Evaluation of uncertainties
related to passive systems performance
[J]. Nuclear Engineering and Design,
2004, 230(1):93-106.
[4] Zhang Shunxiang, Liang Guoxing.
Application of status uncertainty
analysis methods for AP1000 LBLOCA
calculation [J]. Atomic Energy Science
and Technology, 2012,S1:330-334.
[5] Zio E, Cantarella M, Cammi A. The
analytic hierarchy process as a
systematic approach to the
identification of important parameters
for the reliability assessment of passive
systems [J]. Nuclear Engineering and
Design, 2003, 226(3):311-336.
[6] Ma G, Yu Y, Huang X, et al. Screening
key parameters related to passive
system performance based on Analytic
Hierarchy Process [J]. Annals of Nuclear
Energy, 2015, 85:1141-1151.
[7] Zio E, Apostolakis G E, Pedroni N.
Quantitative functional failure analysis
of a thermal–hydraulic passive system
by means of bootstrapped Artificial
Neural Networks [J]. Annals of Nuclear
Energy, 2010, 37(37):639-649.
[8] Liu Sifeng. Grey system theory and
application [M]. Beijing, Science Press,
2008
[9] Liu Ping, Zhou Tao, Zhang Ming et al.
Study on grey correlation degree of
influence factors on ONB in narrow
channel under natural circulation [J].
Nuclear Power Engineering, 2011,
32(4):29-32.
[10] Zhou Tao, Yang Ruichang, Qin Shiwei
et al. Study on grey model in ONB of
nature circulation [J]. Nuclear Power
Engineering, 2005, 26(2):121-124.
[11] Sun Hanhong. Third generation nuclear
power technology AP1000 [M]. Beijing,
China Power Press, 2010.
[12] Li Yankai, Lin Meng, Hou Dong et al.
Qualitative accident analysis on loss of
normal feedwater for AP1000 [J].
Atomic Energy Science and
Technology, 2012, S1:295-300.
[13] IAEA. Natural Circulation in Water
Cooled Nuclear Power Plants
Phenomena Models, and Methodology
for System Reliability Assessment. IAEA
(TEC-DOC-1474).
[14] Baosheng Wang, Dongqing Wang et
al. Efficient estimation of the functional
reliability of a passive system by means
of an improved Line Sampling method
[J]. Annals of Nuclear Energy, 2013,55:
9-17.
[15] Liu Qiang. Reliability analysis of AP1000
passive system based on artificial neural
networks [D]. Beijing: Tsinghua
University, 2014.
[16] Jiang Shuihua, Li Dianqing, Zhou
Chuangbing. Non-instrusive stochastic
finite element method for slope
reliability analysis based on Latin
hypercube sampling [J]. Chinese journal
of Geotechnical enginerring, 2013,
35(S2):70-76.
[17] Wu Guojun, Chen Weizhong, Tan
Xiaojun, et al. Program development of
finite element reliability method and its
application based on Latin Hypercube
sampling [J]. Rock and Soil Mechanics,
2015(2):550-554.
[18] Zhang Xiaolian, Hao Sipeng, Li Jun, et
al. Grey correlation based analysis on
impacting factors of maximum power
point tracking control of wind power
generating unit [J]. Power system
Technology, 2015, 39(2):445-449.
[19] W Mendenhall. Statistics for engineers
and the sciences [M]. Beijing, China
Machine Press, 2009.
Authors
Qi Shi
Zhou Tao
Muhammad Ali Shahzad
Li Yu
School of Nuclear science and
Engineering
North China Electric Power
University
Beijing, 102206, China
Beijing Key Laboratory of Passive
Safety Technology for Nuclear
Energy
Beijing, 102206,China
Jiang Guangming
Science and Technology on Reactor
System Design Technology
Laboratory
Nuclear Power Institute of China
Chengdu, 610041, China
RESEARCH AND INNOVATION 413
Experimental Investigation of a Two-
Phase Closed Thermosyphon Assembly
for Passive Containment Cooling System
Kyung Ho Nam and Sang Nyung Kim
1 Introduction After the Fukushima accident, increasing interest has been raised in passive safety systems that
maintain the integrity of the containment building. The conventional containment building is a thick, airtight reinforced
concrete structure the design of which is highly unfavorable for removing heat from the containment atmosphere to the
environment following an accident. Therefore, the sprays and/or fan coolers are installed to control the containment
pressure and temperature for maintaining the integrity of the containment. However, either sprays or fan coolers are
dependent on the power supply, which is unreliable if Design Basis Accidents (DBAs) are coupled with a station blackout
(SBO) or Extended Loss of AC Power (ELAP). Therefore, to improve the reliability and safety of Nuclear Power Plants
(NPPs), long-term passive cooling concepts have been developed for advanced reactors. In a previous study, The
proposed design was based on an ordinary cylindrical Two-Phase Closed Thermosyphon (TPCT).[1] The exact assembly
size and number of TPCTs should be elaborated upon through accurate calculations based on experiments. While the
ultimate goal is to propose an effective MPHP design for the PCCS and experimentally verify its performance, a TPCT
assembly that was manufactured based on the conceptual design in this paper was tested. Figure 1.
Research and Innovation
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim

atw Vol. 62 (2017) | Issue 6 ı June
RESEARCH AND INNOVATION 414
| | Fig. 1.
Schematic of the Passive Containment Cooling System using the Multi-Pod Heat Pipe.
2 Experiment procedure
2.1 Experimental apparatus
design
As illustrated in Figure 2, an experimental
facility was designed and
installed to acquire various types of
information related to the heat transfer
capacity of MPHP. The facility consists
of three major parts: a pressure vessel,
a coolant tank, and an experimental
TPCT assembly. An experimental TPCT
assembly is a key part of the experimental
apparatus used in this study.
It conducts a heat transfer from the
heater in the pressure tank to the
coolant in the coolant tank. This
assembly is made up of seven TPCTs,
which are a 1-m long boiling region
and condensation region, respectively,
and has a hexagonal array.
2.1.1 Design of TPCT assembly
The operation of TPCT is based on the
force of gravity and the temperature
differences between its parts; one
side is heated while the other side
is cooled. Heat transfer occurs in
TPCT due to these temperature
differences. The thermal resistance
(or the heat transfer coefficient) is
calculated for each region. These
are combined in a thermal resistance
circuit, as shown in Figure 3, to
calculate the total thermal resistance
between the pressure tank inside and
the coolant water inside the coolant
tank (the heat transfer coefficient) for
one TPCT. If R tot and ∆T are the total
thermal resistance and the temperature
difference between the pressure
tank inside, which heats the boiling
region, and the water cooling the
condensation region, respectively,
then it holds that:
(1)
where, ˙Q is the heat removal rate for a
TPCT.
To obtain an explicit expression for
R tot , the heat transfer coefficient of
each region was calculated first, and
then the heat removal of one TPCT
was calculated from this.
The total heat transfer coefficient
(resistance) for a given value of T h , T c
and ∆T bc was calculated by summing
up the aforementioned thermal resistances
in each region. We first assumed
that the temperature distribution was
uniform, that is, there was no temperature
difference between the air in the
assembly center and the air inside the
containment, as mentioned above. In
fact, a considerable temperature drop
is expected, and it is difficult to predict
the specific value. This needs to be
researched through additional experiments
or a review of the literature. For
convenience, we denote the overall
number of pipes in the boiling and
condensation regions as N b and N c ,
respectively, and the heat removal
rate per TPCT as ˙Q i . The following
equation holds for the temperature
drop at the inner boundary of the
boiling region (where the resultant
thermal resistance R5 can also be
determined):
(2)
| | Fig. 2.
Experimental apparatus for heat transfer performance test of MPHP.
| | Fig. 3.
The thermal resistance circuit in a TPCT.
Research and Innovation
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim

atw Vol. 62 (2017) | Issue 6 ı June
Similarly, R 3 can be determined by
dividing the thermal resistance of a
single pipe by the number of pipes in
the condensation region.
(3)
The resulting thermal resistance of
the pipe walls can be determined by
dividing the value for one TPCT by the
number of TPCTs. At the same time,
the heat transfer increases by the
same factor.
(4)
(5)
(6)
where, ˙Q i is the heat transfer rate per
pipe. If we assume that there is no
heat transfer or temperature drop in
the adiabatic region, then R 4 = 0 and
R 8 = 0.
The total thermal power extracted
from the containment by one assembly
can be calculated by multiplying the
reciprocal of the sum of the thermal
resistance values found above by
the temperature difference between
the containment atmosphere and the
cooling water on the top of the containment
dome shell and taking into
account the difference ∆T bc as well.
Here R MPHP is the total thermal resistance
of the MPHP.
(7)
(8)
where, R i is the thermal resistance
component of a single TPCT. Therefore,
the total heat transfer coefficient
of the MPHP assembly is:
(9)
Section Material Height, m Diameter, m P/D ratio Thickness, m FR
Pipe Stainlesssteel 1 0.03 2 0.0005
Adiabatic 304
0.3 0.2 − 0.013
| | Tab. 1.
Specifications of experimental thermosyphon assembly.
This equation will be applied to compare
the theoretical and experimental
results in Chapter 3 of this paper.
The heat transfer mechanisms in
the boiling region occur in various patterns,
which are the natural convection,
evaporation, nucleate boiling, and a
combination of each to the fill charge
ratio (FR), heat flux, etc. The fill charge
ratio indicates the percentage of the
boiling region volume that is filled by
the working fluids. In this study, it
is determined that a fill charge ratio
is 30 % of the boiling region based on
a previous study performed by Imura
[2, 3, 4]. Thus, an experimental TPCT
assembly is partially filled with distilled
water and then sealed. The specifications
of the experimental TPCT assembly
are presented in Table 1.
2.1.2 Pressure tank and Coolant
tank
A pressure tank simulates the inner
containment building, and steam is
generated at the bottom of the vessel
by two horizontally mounted immersion
electric heaters with a total
capacity of 30 kW. Air and makeup
water can be injected into the vessel,
and a drain line is located at the
bottom of the vessel. The maximum
rated operating pressure for the tank
is 1 MPa, which is insured by a safety
relief valve. The coolant tank is an
open type, and the height is sufficiently
high so that the condensation
region of the pipes can be submerged
in the coolant. The pressure tank
is fully insulated with fiberglass to
reduce heat loss so that sufficient
power can be delivered to the TPCT
assembly. Measurements have shown
that the vessel heat loss is less than
1 kW, which can be easily compensated
by the heaters, which have a
capacity of 30 kW. The capacity of a
heater is determined by considering
the predicted performance limitation
of the TPCT assembly. The maximum
heat transfer rate owing to entrainment
limitations can be calculated
through flooding correlations that are
expressed in terms of the Kutateladze
dimensionless groups [5].
(10)
where,
(11)
(12)
The maximum heat transfer rate is
predicted to be about 4 kW per pipe.
Therefore, an experimental TPCT
assembly that has seven pipes is
considered, in which the maximum
heat transfer rate is about 28 kW.
However, this limitation is a conservative
value, and thus this value is
only considered to determine the
capacity of an electrical heater.
The coolant tank is an open type
and the height is sufficiently high so
that the condensation region of pipes
can be submerged in coolant as shown
in figure 2. As shown in figure 3, the
heat generated in pressure tank transfers
and this coolant in coolant tank
will be external heat sink during the
operation.
2.3 Experiment procedure
The experiment was conducted in
the following manners. Two types of
instrumentation devices were used
in the experiment: thermocouples for
temperature measurement, and pressure
gauges/transducers for pressure
measurement. K-type thermocouples
were placed at each point to provide
temperature readings. Temperature
data were continuously recorded
during operation with the thermocouples.
The temperatures at the
pressure tank, pipe wall surface, inner
pipe, and coolant tank were recorded.
One pressure transducer was installed
to measure the overall pressure of the
vessel. A communication-based Data
Acquisition System (DAS) was set up
for this experiment. All thermocouple
leads and pressure transducer output
cables were wired into the measurement
channels. Additionally, a Silicon
Controlled Rectifier (SCR) was installed
to control the electrical power
of the heater so that the electrical power
was consistently fixed to the heater.
0.3
RESEARCH AND INNOVATION 415
Research and Innovation
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim

atw Vol. 62 (2017) | Issue 6 ı June
RESEARCH AND INNOVATION 416
Test case
Heat
input,
kW
Initial
absolute
pressure,
MPa
The predicted Air
weight fraction
at steady state,
w/o
10-#1 10 0.1 0.4
10-#2 0.14 0.45
10-#3 0.21 0.5
15-#1 15 0.1 0.35
15-#2 0.15 0.4
15-#3 0.22 0.45
20-#1 20 0.1 0.3
20-#2 0.16 0.35
20-#3 0.27 0.4
25-#1 25 0.1 0.25
25-#2 0.18 0.3
25-#3 0.29 0.35
30-#1 30 0.1 0.2
30-#2 0.14 0.25
30-#3 0.2 0.3
Area of
interest
| | Tab. 2.
Test matrix.
• Temperature profile according to the heat
input and initial air pressure
• Pressure profile in the pressure tank
Region Correlation Author
Inside
pressure tank
Inside
boiling region
of pipe
Inside
condensation region
of pipe
Inside
coolant tank



| | Tab. 3.
Heat transfer correlations for predicted value compared with the experiment results.
Uchida
Tagami
Kataoka
Murase
Imura
Nusselt
Rohsenow
The test conditions are listed in
Table 2. Test cases 10-# to 30-#
were performed to evaluate the heat
removal performance in the pressure
vessel. Heat input flowed from 10 kW
to 30 kW in each case. As mentioned
above, non-condensable gases greatly
affect the heat transfer inside the
containment because they depend
only on natural circulation, and no
power supply or fan operation for
forced circulation is possible. Furthermore,
as the MPHP assembly consists
of a multitude of long pipes, the
lengths and radial locations of
the pipe in the assembly are expected
to have a significant effect on the
passage of steam to pipes. Therefore,
the increase in concentration of noncondensable
gases owing to steam
condensation in the pipe array was
considered. For this reason, the
weight fraction range of air is determined
to be from 0.2 to 0.5 w/o.
3 Results and discussions
The empirical correlations compared
with the experimental results are
presented in Table 3. The four correlations
based on steam condensation
with non-condensable gas were chosen
for a comparison with the experimental
data. These correlations are only
dependent on the con centration of
non- condensable gas, and thus are selected
to compare with the data. [7, 8]
As shown in Figure 4, all correlations
compared with the experimental
data tend to under predict the
measured values. These correlations
were developed for steam condensation
with non-condensable gas on a
long vertical plate. In this study, the
geometry of the condensation area
making contact with a steam and air
mixture is of a cylinder type, and it
shows that air weight accumulated
on the pipe surface is lower than
pre dicted. Thus, steam condensates
better than the predicted models.
The correlation reported by Imura
et al. was compared with the experimental
data, as shown in Figure 5.
The Imura et al. correlation tends to
under predict the measured values
though the heat transfer coefficients
in the boiling region, and predictions
generally show reasonable agreement
with the majority of the points being
within the 35 % band.
The correlation reported by Nusselt
was compared with the experimental
data. This correlation covers all data
within ±10 %, as shown in Figure 6,
and shows very good agreement with
the measurements.
| | Fig. 4.
Predicted and experimentally determined heat transfer coefficients in the
pressure tank for steam condensation with non-condensable gas.
| | Fig. 5.
Predicted and experimentally determined heat transfer coefficients in the
boiling region for full pool boiling mode with distilled water.
Research and Innovation
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim

atw Vol. 62 (2017) | Issue 6 ı June
| | Fig. 6.
Predicted and experimentally determined heat transfer coefficients in the
condensation region for film condensation.
| | Fig. 7.
Comparison of the overall heat transfer rates between the predicted and
experimentally determined results (left), and variation of total thermal
resistance of a TPCT assembly vs. air weight fraction in a pressure tank (right).
RESEARCH AND INNOVATION 417
| | Fig. 8.
Comparison of the overall heat transfer rates between the predicted and experimentally determined results (Left), Variation in total thermal resistance of a TPCT
assembly vs air weight fraction in the pressure tank (Right).
The correlation reported by
Rohsenow was compared with the
experimental data, as shown in
Figure 7. Most heat transfer coefficients
of the experimental results are
much lower than those obtained by
the correlations. In this study, city
water was used as a coolant in the
experiment, and the scale generated
on the pipe surface effected as insulation.
If the accident sequence determines
that the designed water sources
are not available in sufficient quantity,
or at a sufficient rate, any water source
should be used without delay. It is
quite possibly that the city water may
be used during an accident when the
water sources are not available. For
conservatism, city water was used as a
coolant, and this condition causes the
experimental data to be much lower
than the predicted values.
The input heat transfer rates versus
the temperature difference between
the inner pressure tank and coolant
are plotted in Figure 8, and total
thermal resistance is obtained from
Eq. 9 and also presented in Fig. 8.
It shows that the concentration of
non-con densable gas in the containment
is key factor which affects total
thermal resistance and the performance
of the MPHP when the MPHP is
implemented in actual NPPs.
4 Conclusion
An analysis of experimental data and
comparison to existing widely used
correlations lead to the following
conclusions:
1. Measured heat transfer coefficients
in each region and the overall heat
transfer rate are higher than the
predicted values. This shows that
the theoretical results are conservative
when a MPHP is implemented
in an actual NPP. Additionally,
the key factor that affects the
total thermal resistance of a MPHP
assembly is non-condensable gas
concentration in the containment.
2. The experiment results show that
a TPCT consists of a 1-m long
boiling and condensation region,
respectively, and can transfer at
least 45 kW/m 2 of heat flux.
3. Based on the measured heat flux
and heat transfer capacity, a MPHP
assembly consists of 1-m long
boiling and condensation pipes,
respectively, and has about 2,000
pipes with an overall diameter of
about 1.75 m to provide 50 % heat
removal capacity. In the case of
100 % heat removal capacity, it
has 4,500 pipes and the overall
diameter is about 2.4 m.
4. Precise calculations using computer
code simulate the behavior
(pressure, temperature) of the
containment atmosphere when
the novel PCCS is in operation and
to account for other heat sources
than the decay power.
5. The development of average parameters
(lumped parameter method)
and performing param etric studies
to account for the effects of increasing
heat pipe length and array size
(steam access from the containment
to pipes). The air weight fraction
was con sidered to be up to 0.5 w/o
in this experiment, and thus this
effect was considered roughly in
this experiment.
6. Because of the large added mass
(cylindrical wall extension and/
or water tanks on the dome top,
cooling water, MPHP assemblies
with water, and their accessories),
an additional seismic evaluation of
the containment (concrete walls
and dome) is necessary.
Research and Innovation
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim

atw Vol. 62 (2017) | Issue 6 ı June
RESEARCH AND INNOVATION 418
Acknowledgement
This study was sponsored by the
Ministry of Trade, Industry and Energy
(MOTIE) and was supported by
Nuclear Convergence and Original
Technology Development Program
Grant funded by the Korea Institute
of Energy Technology Evaluation and
Planning (KETEP) (Grant code:
20111520100030)
Nomenclature
P pressure (Pa)
h heat transfer coefficient (W/m2°C)
W weight fraction (w/o)
T temperature (°C)
k thermal conductivity (W/m°C)
N number of pipe ( - )
L length (m)
D diameter (m)
ρ density(kg/m 3 )
c specific heat (J/kg-°C)
q heat flux (W/m 2 )
g gravity (m/s 2 )
h latent heat (J/kg)
μ viscosity (kg/m-s)
Q heat transfer rate (W)
Subscripts
nc non-condensable gas ( - )
b boiling region of pipe ( - )
c condensation region of pipe ( - )
hot outside boiling region of pipe ( - )
cold outside condensation region of pipe( - )
P/D Pitch-to-Diameter ratio ( - )
References
[1] G.H. Nam, J.S. Park, S.N. Kim.
Conceptual Design of Passive Containment
Cooling System for APR-1400
using Multi-Pod Heat Pipe, Nuclear
Technology. 189 (2015) 278–293.
[2] H. Imura. Heat Transfer in the Two-
Phase Closed Thermosiphon, Trans.
JSME, Vol. 45, pp.712-722, 1979.
[3] H. Imura. Critical Heat Flux in a Closed
Two-Phase Thermosyphon, Int. J. Heat
Mass Transfer, Vol26, No.8,
pp. 1181-1188, 1983.
[4] I. Khazaee, R. Hosseini, S.H. Noie.
Experimental investigation of effective
parameters and correlation of geyser
boiling in a two-phase closed thermosyphon,
Applied Thermal Engineering,
Vol. 30, pp. 406-412, 2010.
[5] S. Khandekar, et. al. Thermal performance
of closed two-phase thermosyphon
using nanofluids, Int. J. Thermal
Science, Vol. 47, 659-667, 2008.
[6] Y.G. Lee, et. al. An experimental study
of air-steam condensation on the
exterior surface of a vertical tube under
natural convection conditions, Int. J.
Heat and Mass Transfer, Vol. 104,
pp. 1034-1047, 2017.
[7] A. Dehbi. A generalized correlation for
steam condensation rates in the
presence of air under turbulent free
convection, Int. J. Heat and Mass
Transfer, Vol. 86, pp.1-15, 2015.
[8] J.C. de la Rosa, A, Escriva. Review on
condensation on the containment
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pp. 33-36, 2009.
Authors
Kyung Ho Nam
Korea Atomic Energy Research
Institute
111, Daedeok-daero 989beon-gil
Yuseong-gu, Daejeon, Korea
Sang Nyung Kim
Kyunghee University
1732, Deogyeong-daero
Giheung-gu, Yongin-si,
Gyeonggi-do, Korea
Displacement of Cryomodule
in CADS Injector II
Yuan Jiandong, Zhang Bin, Wang Fengfeng, Wan Yuqin, Sun Guozhen, Yao Junjie, Zhang Juihui and He Yuan
1 Introduction As Cryomodule can easily reduce higher power consumption and length of an accelerator,
make the accelerator can be run continuously, it is becoming increasingly important in the superconducting linac [1].
Due to the invisibility and coupled with ultra-low temperature characteristics (4 k), Cryomodule is the key points and
difficulties for a superconducting linear accelerator. The Chinese academy of sciences institute of modern physics is
developing an accelerator driven subcritical system (CADS) Injector II [2].CADS will accelerate protons with a beam
current of 10mA to about 1.5 GeV to produce neutrons for the transmutation of nuclear waste [3]. To avoid generating
beam orbit distortion, the magnet magnetic center must be on the beam axis, so the displacement of cold components
has extremely requirements [4]. From the theoretical point, there are generally three approaches to deal with the
displacement on the working condition [5]. One is to maintain the alignment upon the cooldown. In this approach, the
structure is designed so that the cooldown is absolutely symmetric. The other is to allow realignment once cold. In this
approach, components must be realigned after they reached their final cryogenic temperature. As we all know that both
the above two situations cannot easily be reached.
The last approach is to allow the
components to change in a predict able
and repeated way. There are four
different methods to realize this
objective currently. The European
organization for nuclear research
developed a double-sided Brandeis
CCD Angle Monitor (BCAM) [6]. The
Japanese high-energy accelerator
research organization adopted white
light interferometer (WLI) [7]. German
electron synchrotron [8], the institute
of high energy physics Chinese academy
of sciences [9] and Fermi national
accelerator laboratory [10] employed
a Wire Position Monitor (WPM) to
monitor the contraction. The France
large national heavy-ion accelerator
adopted a micro-alignment telescope
to align Cryomodule intuitively [11].
However, these above methods only
investigated the cryo-displacement,
did not concern the effect of the negative
pressure of the vacuum. Ref [12]
(D. Passarelli) have estimated the
pressure distribution inside the cavity
string used a mathematical model.
Ref [13] analyzed the displacement
induced by temperature differences,
but did not correlate the cryo-vacuum
displacement.
In this paper, we present a detailed
description of the principle of the
vacuum cryo-environments firstly;
and then we take out the simulation
of vacuum and cryo-displacement
Research and Innovation
Displacement of Cryomodule in CADS Injector II ı Yuan Jiandong, Zhang Bin, Wang Fengfeng, Wan Yuqin, Sun Guozhen, Yao Junjie, Zhang Juihui and He Yuan

atw Vol. 62 (2017) | Issue 6 ı June
respectively; in the section IV, we compared
the measured results with the
simu lated ones; At last, we correlated
the cryo-vacuum displacement. The
deep investigation will benefit for the
optimization [14] and upgrade [15-16]
of Cryomodule design for the CIADS.
2 Heat transfer principle
CADS injector II project includes 4
cells 6 cavities cryostats. Most cavities
and magnets are working in the
cryostat at liquid helium (LHE) temperature.
The vacuum jacket of each
cryomodule is rectangular box shaped
of dimension 4.3 m × 1.7 m × 2.1 m.
Their alignment will be carried out at
room temperature first, and then after
the compensation, the position error
of the cavities and magnets shall be
within ±0.1 mm. Finite element
method was used here to analysis the
thermal stress and displacement: we
used Solid Works to construct model
first; and then imported in ANSYS,
meshing with four surfaces unit;
finally simulated the thermal stress
and cold displacement.
Generally, heat transfer includes
the sum of thermal radiation, convection,
and sometimes conduction
transfer. Usually, more than one of
these processes occurs in a given
situation. Since the Cryomodules are
operated in a cryo-vacuum environment,
there is no convective heat
transfer in the static heat loads [17,
18]. There are only thermal radiation
[19] from the “hotter” environment
and direct thermal conduction
through the cold mass supports,
power coupler and the feedthrough
[20]. Thermal conduction is the
transfer of heat (internal energy) by
microscopic collisions of particles and
movement of electrons within a body
[21]. According to Fourier's law,
the heat flux resulting from thermal
conduction is proportional to the
magnitude of the temperature gradient,
the thermal conductivity and the
cross-sectional surface area. However,
it is inversely proportional to the
length of a conduction path and
opposite to it in sign. In many cases,
the analysis may be simplified by the
use of thermal conductivity integrals.
In this approach, the conduction heat
transfer in one dimension is given by
[22] (eq. 1):
(1)
Thermal radiation is an electromagnetic
radiation generated by the
thermal motion of charged particles in
matter. All matter with a temperature
greater than absolute zero emits
thermal radiation. When the temperature
of a body is greater than absolute
zero, inter-atomic collisions cause
the kinetic energy of the atoms or
molecules to change. This results in
charge-acceleration and/or dipole
oscillation which produces electromagnetic
radiation, and the wide
spectrum of radiation reflects the
wide spectrum of energies and accelerations
that occur even at a single
temperature. During the cool-down,
the thermal shield receives radiative
heat flux from both the external vacuum
vessel and from the internal cold
mass. In fact, the heat flux density due
to radiation transport between two
surfaces at different temperatures can
be written as following [22] (eq. 2):
(2)
Where [22] the vector Q is the heat
flux (in W/m 2 ) in the positive direction;
λ is known as the conductivity
constant or conduction coefficient
(in w/m k); A is the total crosssectional
area of conducting surface
(in m 2 ); L is the thickness of conducting
surface (in m). Where [22] σ =
5.67 10 is the Stefan-Boltzmann constant
and ε is the effective emissivity
of the king into account the view
factor and surface emissivities. For a
simple “feeling” of the order of
magnitudes, it is useful to refer to the
simple case of black body radiation
(unitary emissivity) intercepted by a
unitary surface at 2 K from a parallel
plate at different temperatures.
2 Simulation
A Vacuum displacement
To guarantee an effective cool-down
process for the Cryomodule, a
high-vacuum level must be achieved
[12]. The thorough stress and Strain
analysis of the vacuum chamber
under atmospheric pressure and selfgravity
must be carried out. According
to the Hooke's Law, the displacement
ΔL (eq. 3) is proportional to the above
composite forces F n , is inversely proportional
to the cross section size A
and the Young's Modulus E.
(3)
The transfer of energy to the gravitational
field results in the deformation
of vacuum. The Finite element model
of the cold mass and its support are
shown in Figure 1. The generalized
Hooke's law can be used to predict
the displacement (Formula 4) caused
in a given material by an arbitrary
combination of stresses. Where v is
the Poisson ratio, F x , F y and F z are the
combined forces of the x, y, z-direction
respectively.
(4)
Given that 1.5 tons of gravity were
exerted on the two insulating
RESEARCH AND INNOVATION 419
| | Fig. 1.
Model of the cryomudule.
| | Fig. 2.
Vacuum simulation.
Research and Innovation
Displacement of Cryomodule in CADS Injector II ı Yuan Jiandong, Zhang Bin, Wang Fengfeng, Wan Yuqin, Sun Guozhen, Yao Junjie, Zhang Juihui and He Yuan

atw Vol. 62 (2017) | Issue 6 ı June
RESEARCH AND INNOVATION 420
| | Fig. 3.
Cryo-Simulation
supports, one atmospheric pressure
(0.1Mpa) was applied to the six
surfaces of the vacuum chamber; the
four bottom supports were fixed. As
seen from Figure 2, the vacuum
displacement occurs mainly in the
central area around the horizontal
and vertical zone of support are 0.42
and 0.62 mm respectively; and the
central region is larger than the lateral
area, with pot-shaped.
B Cryo-displacement
The two cooling experiments were
cooled using Radiation cooling firstly
(300 to 210 K); then the cryomodule
is precooled with liquid nitrogen
(210 to 150 K). After that, the
cooldown of the cryomodule begins
with liquid helium (150 to 4.2 K). The
helium reservoir and cavity can be
filled with liquid helium once temperature
becomes 4.2 K. Liquid helium is
pumped to reduce the vapor pressure
of the liquid helium in the helium
reservoir and cavity in order to make
2 K. Internal stresses are also developed
in the statically indeterminate
structure if the free movement of the
joint is prevented. According to the
analysis of displacement in a statically
determinate structures induced by
temperature changes [24-25] (eq. 5),
if the temperature of the member is
decreased uniformly throughout its
length, a denotes the coefficient of
thermal expansion of the material
(mm/K); ΔT denotes the temperature
change (K); L denotes the length of
the structure (mm).
(5)
In the model, Magnets, the helium
tank, and its welding bracket adopted
316 L stainless steel materials, HWR
cavity and its welding bracket
used titanium material, cold quality
support components used titanium
material, collimation bracket and
cross hair targets used G11 materials.
The contact surface of support was
operated at 300 K. The thermal conductivity
of materials varies strongly
with a temperature between 300 and
77 K [17]. The surface heat load of
77 K (BPMs) was 1 W/m 2 [26-27].
The boundary conditions and load
[28] contain a self-gravity of the cold
mass assembly, a distributive load of
temperature, a force of the cold mass
assemblies and the top suspending
support. The force of the cold mass
assemblies acting on the Ti support
frame is decided by the gravity of each
cold mass assembly. According to the
mechanical characteristics of cold
mass [28], the simulated results of the
solenoid and HWR cavity were contracted
0.77 mm in Horizontal and
risen 2.98 mm in Vertical direction,
respectively (As shown in Figure 3).
4 Results and Analysis
A Vacuum displacement
As shown in Figure 4, the two processes
of vacuum pump started at
18:00 on November 30 and 16:00 on
December 2, 2016, respectively. The
vacuum level reached 0.1 Pa about
3 hours later and 10-3 Pa about
11 hours later. The two processes of
vacuum release started at 8:00 on
December 2 and 7:00 on December 4,
2016, respectively. The vacuum level
returned to 0:1 Pa about 8 hours later.
Figure 5 illustrates the vertical and
horizontal displacements monitored
by the Laser Tracker from the top and
left of the vacuum vessel. The Laser
Tracker system is able to compensate
for temperature and humidity effects
based on the measurement conditions.
The monitored displacements
are 0.42 mm in the vertical direction
and 0.62 mm in horizontal direction
respectively. The monitored and
simulated displacements are matching
very well: the differences are
0.02 mm in the vertical direction and
0.04 mm in the horizontal direction.
B Cryo-displacement
The optical instrument microalignment
telescope (MAT) was
adopted to monitor the displacement.
Fig. 5 shows the monitor results of
Hori zontal (Vertical) direction during
one thermal cycling: as the target was
located on the right (below) of
solenoid and HWR cavity, A plus sign
means that it is close to center(rise
up); A minus sign means that it is
off center (go down). After cooled
down 24 hours, cold mass has
contracted 0.8 mm in horizontal and
2.27 mm in vertical direction respectively
on average(with respect to the
pumped). And then cold mass has
warmed up to 290 K after warmed up
24 hours, and has expanded 0.5 mm
horizontal and 1 mm vertical respectively
on average.
| | Fig. 4.
Vacuum displacement (DX: horizontal; DY: vertical).
| | Fig. 5.
Cryo-displacement.
Research and Innovation
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atw Vol. 62 (2017) | Issue 6 ı June
| | Fig. 6.
Stress Analysis.
C Discussion
As shown in Figure 6 (a), due to the
supports were located on the outer
surface of the bottom vacuum chamber,
the direct vertical stress exerted
on the inside cold mass (cavity, solenoid
and so on) comprised the negative
atmospheric pressure (the force F
normal to the surface per area A),
gravity and thermal stress during the
process of cooling down. Therefore,
the complete displacements in the
vertical direction were the superposition
effect of the above stress. However,
the horizontal displacement
resulted only from the thermal stress.
Root mean square value of the
measurements was 0.03 mm in the
vertical direction and 0.02 mm in the
horizontal direction. As shown in Fig.
6 (b), the direct vertical stress exerted
on the inside cold mass (cavity, solenoid
and so on) comprised the negative
atmospheric pressure, gravity and
positive thermal stress during the
process of warming up. Therefore, the
complete displacements in the vertical
direction were the subtraction of the
above stress. As shown in Fig. 5, the
differences of cryo displacement
( released) with respect to nominal
zero resulting from plastic displacement
[29] and the not fully released
pressure of the vacuum. The above
results indicate that the reproducibility
of the horizontal and
vertical position is 0.3 mm and
0.6 mm respectively.
5 Conclusion
Cryomodules are extremely complex
systems, and their design optimization
is strongly dependent on the
accelerator application for which they
are intended. We have demonstrated
that the simulated vacuum and
cryo-displacement shows a good
agreement with the measured values.
The above data provides information
not only on the nature of the heat
exchange phenomena and their effect
on the structural stability of the internal
components of the Cryomodule,
but also benefit to an optimization
for future Cryomodules design. The
analysis procedure will be helpful for
the estimation of displacements in
working conditions like mechanical
and thermal loads. We will study the
on-line continuous monitoring system
in the future, which will further reveal
low-temperature displacement mechanism
of the cryostat.
Acknowledgment
This work was supported by the
National Natural Science Foundation
of China (No.11605262). This work
could not have been accomplished
without the advice and support of our
colleagues: Juihui Zhang and Bin
Zhang.
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Authors
Yuan Jiandong
Zhang Bin
Wang Fengfeng
Wan Yuqin
Sun Guozhen
Yao Junjie
Zhang Juihui
He Yuan
Institute of Modern Physics
Chinese Academy of Sciences
509#, Nan chang Road, Lanzhou,
China, 730000
Inside
KTG Sektion Süd und Fachgruppe Kernfusion
Vortragsveranstaltung
zur Fusionsforschung
| | Prof. Dr. Zohm (rechts) – u.a. 2014 mit dem
John Dawson Award der Amerikanischen
Physikalischen Gesellschaft und 2016 mit
dem Hannes­ Alfvén-Preis der Europäischen
Physikalischen Gesellschaft ausgezeichnet –
hier im Gespräch mit dem Sprecher der FG
Kernfusion Dr. Thomas Mull (links) und der
Sprecherin der Sektion Süd Yvonne Broy (Mitte).
Erstmals luden für den 3. Mai 2017 die KTG Sektion Süd
und die Fachgruppe Kernfusion zu einer gemeinsamen
Veranstaltung nach Erlangen ein.
Referent Prof. Dr. Hartmut Zohm – Leiter des Bereichs
Tokamak-Szenario-Entwicklung am Max-Planck-Institut
für Plasmaphysik – beantwortete die Frage „Wo steht die
Fusionsforschung?“ und ging dabei auf den Tokamak, ITER
und internationale Perspektiven ein.
Die Forschungen zum magnetischen Einschluss von
Wasserstoffplasmen mit Temperaturen von mehr als
100 Millionen °C zur Energiegewinnung aus Kernfusion
haben in den letzten Jahrzehnten große Fortschritte
gemacht. Dabei werden unterschiedliche
Fragen der Plasmaphysik, wie
z.B. Wärmetransport oder Stabilität,
experimentell und theoretisch untersucht.
Parallel dazu werden spezielle
Technologien, wie etwa der Bau großer
supraleitender Spulen, vorangetrieben.
Das Max-Planck- Institut für Plasmaphysik
betreibt dazu in Garching das
Groß experiment ASDEX Upgrade und
hat eine weitere Großanlage, Wendelstein
7-X, in Greifswald in Betrieb
genommen.
Im Vortrag ging Prof. Zohm ebenfalls
auf den Test reaktor ITER ein, der
zurzeit in einer weltweiten Zusammen
arbeit in Cadarache, Frankreich,
entsteht und eine Schlüsselrolle auf
dem Weg zur Nutzung von Kernfusionsenergie
spielen wird. Der ITER
wird in Cadarache (Frankreich) durch
| | 80 Teilnehmer verfolgten interessiert den hochinteressanten Vortrag,
der zunächst mit einer Einführung in die Tiefen der Kernfusion begann.
China, EU, Indien, Japan, Korea, Russland und die USA gebaut,
wobei jeder der Partner sogenannte „In-kind“ – Leistungen
erbringt, was die Projektsteuerung sehr komplex
macht.
Die EU-Roadmap zum Fusionskraftwerk sieht vor, das
spätestens 2050 ein erstes derartiges Kraftwerk in Betrieb
gehen soll. Der Weg vom ITER zu einem ersten DEMO ist
allerdings noch weit. Neben dem Nachweis zuverlässiger
Energieerzeugung mit abgeschlossenem Brennstoffkreislauf
sind auch Verbesserungen in Physik und Technologie
notwendig, um ein attraktiveres DEMO Design anbieten zu
können.
Alle ITER Partner haben starke nationale Aktivitäten:
die Roadmap für China sieht dabei bereits 2030 die
Inbetriebnahme des CFETR vor – China Fusion Engineering
Test Reactor.
Fazit des kurzweiligen Vortrages: Die Fusionsforschung
hat in den letzten Jahren große Fortschritte erzielt, die
nun im nächsten Schritt eine Realisierung des ITER
ermöglicht. Fusionskraftwerke könnten ab 2050 Baustein
der Energieversorgung sein, was immer noch rechtzeitig
wäre, um eine weltweite Energiewende zu vollziehen, aber
diese Entwicklung wird kontinuierliche Unterstützung
benötigen. Die deutsche Fusionsforschung ist dabei ebenso
von großer Bedeutung – nicht zuletzt mit W7-X wird auch
der Stellarator eine wichtige Rolle spielen.
Yvonne Broy
KTG Inside

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19. Dipl.-Ing. Nils Förtsch,
Eggenstein-Leopoldshafen
21. Thomas Röder, Biblis
Juli 2017
92 Jahre wird
8. Dr. Werner Eyrich, Karlsruhe
90 Jahre wird
24. Dipl.-Phys. Gerhard Hecht, Erlangen
89 Jahre wird
17. Dipl.-Ing. Karl Josef Sauerwald,
Höchstadt
85 Jahre werden
21. Dr. Wolfgang Häussermann,
Müllheim (Baden)
24. Dipl.-Ing. Joachim May, Burgwedel
27. Dr. Rainer Schwarzwälder, Glattbach
31. Dr. Theodor Dippel,
Eggenstein-Leopoldshafen
83 Jahre werden
6. Dr. Gerhard Schmidt, Darmstadt
14. Prof. Dr. Walter-H. Köhler, Wien/A
25. Dr. Leonhardt Wetzel, Liebschütz
82 Jahre wird
20. Dipl.-Ing. Ralf Wünsche, Hannover
81 Jahre werden
1. Dipl.-Ing. Hans-Jürgen Börner,
Weisenheim
11. Dipl.-Phys. Agmar Müller,
Bad Neuenahr
25. Dr. Gert Dressler, Lingen
80 Jahre werden
6. Dipl.-Ing. Paul Börner,
Steinau-Uerzell
16. Dr. Dieter Hennig, Berlin
24. Dipl.-Ing. (FH) Klaus Haase, Bruchsal
29. Dr. Herbert Reutler, Köln
79 Jahre werden
13. Dipl.-Ing. Ralf Harth, Aachen
30. Dr. Philipp Dünner, Odenthal
78 Jahre werden
10. Dr. Bernhard Steinmetz,
Bergisch Gladbach
23. Heinz Stahlschmidt, Erlangen
26. Dipl.-Ing. Ewald Passig, Bochum
77 Jahre werden
2. Prof. Dr. Anton Bayer, Ilmmünster
2. Dr. Manfred Hagen, Berlin
16. Dipl.-Ing. Dietrich Kuschel, Fulda
24. Dr. Manfred Ruhbaum,
Wächtersbach
31. Dr. Peter Schneider-Kühnle, Worms
76 Jahre werden
1. Norbert Semann, Bruchsal
24. Dipl.-Ing. Friedrich Wietelmann,
Erkrath
25. Dr. Heinz-Wilhelm Bock, Erlangen
70 Jahre werden
12. Dr. Karl-Wilhelm Zerreßen, Saarlouis
15. Dr. Manfred Rost Karlsruhe
24. Dr. Klaus-Arno Beckers, Wennigsen
24. Klaus Krist, Heddesheim
60 Jahre werden
9. Dipl.-Ing. Bernward Hartje,
Isernhagen
15. Dipl.-Ing. Reiner M. Praetorius,
Dettenheim
50 Jahre wird
28. Dr. Heinrich Harke, Scharnhorst
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atw Vol. 62 (2017) | Issue 6 ı June
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*)
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
Top
Nuclear energy’s role
in fighting climate change
under threat
(nei) Nuclear power is essential for
addressing climate change, but
significant changes are required in
technology development and global
governance in order to continue that
vital role, according to a report
released by the Global Nexus Initiative
(GNI), a policy group that for the first
time brought together experts from
the nuclear industry and leading
energy, climate change, and nuclear
security experts and organizations.
The GNI is a joint project of the
Partnership for Global Security, a
think tank with a long history of
offering innovative solutions to global
nuclear security challenges, and the
Nuclear Energy Institute, the nuclear
energy industry’s policy organization.
Maria Korsnick, President and CEO
of NEI, said: “America’s 99 nuclear
reactors have a vital role to play when
it comes to powering our economy,
protecting the environment and supporting
our nation’s influence around
the world. GNI’s recommendations
reflect a common interest in finding
policy solutions to help keep our
plants running, advance new designs
and promote the role our nuclear
suppliers play in generating jobs at
home while strengthening America’s
hand in global governance in the face
of challenges abroad. We look forward
to continuing our work with GNI
on these important issues as we chart
the future of our industry.”
Kenneth Luongo, President of the
Partnership for Global Security, said:
“GNI has responded to the realities of
the complex global environment
where the linkages between critical
issues including climate change,
nuclear power and international
security require new responses and
innovative partnerships. Nuclear
power has an important role to play
in tackling climate change, but there
are governance and geopolitical
challenges that need to be addressed.
The GNI report focuses attention on
the nexus of these issues and provides
an actionable agenda for progress that
will benefit the global community.”
Richard Meserve, former Chairman
of the Nuclear Regulatory Commission
and a member of the GNI working
group, said: “This report draws attention
to nuclear power’s geopolitical dimension,
which often is overlooked in
the debate. The nuclear rules are
shaped by the countries with the
largest market share, and traditional
leaders like the U.S. will soon be
overtaken by China and Russia. There
is a danger that the U.S. will lose the
capacity to influence the global norms
for safety, security and nonproliferation.
There thus are national
security issues at stake.”
Armond Cohen, Executive Director
of the Clean Air Task Force and a
member of the GNI working group,
said: “Nuclear energy has increasingly
come forward as a climate change
management tool as we realize how
deep and fast carbon cuts need to happen.
Nuclear can be part of a portfolio
approach – along with renewables,
carbon capture and sequestration
and improvements in efficiency –
that gives us multiple options to
decarbonize the electricity sector and
sustain economic growth. But developing
nuclear energy at sufficient
scale and speed will require both
technical innovation and close cooperation
among industry, international
regulatory bodies, civil society, and
public and private investors. I look
forward to the GNI’s involvement in
that process in the months ahead.”
Recommendations
The report, “Nuclear Power for the
Next Generation: Addressing Energy,
Climate and Security Challenges,”
addresses critical issues around
climate policy, nuclear technology
and global security. Its principal
recommendations are:
• Nuclear power is necessary to
address climate change.
Operating Results December 2016
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 910 880 744 677 943 7 306 048 247 231 855 100.00 92.73 99.26 91.61 100.13 91.40
OL2 Olkiluoto BWR FI 910 880 641 583 590 7 565 721 237 817 140 86.12 95.42 85.09 94.46 86.20 94.65
KCB Borssele PWR NL 512 484 744 381 876 3 960 315 154 804 440 99.96 89.40 99.96 89.10 100.25 89.27
KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0
KKB 2 Beznau 6,7) PWR CH 380 365 744 285 861 3 175 815 128 232 156 100.00 96.47 100.00 96.27 101.11 95.14
KKG Gösgen 7) PWR CH 1060 1010 744 795 765 8 668 128 296 610 635 100.00 93.72 99.98 93.33 100.90 93.10
KKM Mühleberg BWR CH 390 373 744 286 460 3 077 620 121 212 245 100.00 92.90 99.62 92.02 98.73 89.84
CNT-I Trillo PWR ES 1066 1003 744 791 255 8 552 866 230 493 717 100.00 92.38 100.00 92.24 99.41 90.85
Dukovany B1 PWR CZ 500 473 744 369 392 3 813 268 105 810 374 100.00 87.88 100.00 87.55 99.30 87.06
Dukovany B2 PWR CZ 500 473 0 0 2 521 816 101 322 628 0 59.04 0 58.69 0 57.58
Dukovany B3 PWR CZ 500 473 744 372 723 2 487 538 99 624 856 100.00 57.29 100.00 56.63 100.19 56.79
Dukovany B4 PWR CZ 500 473 744 371 674 3 131 703 100 528 151 100.00 73.08 100.00 72.35 99.91 71.50
Temelin B1 PWR CZ 1080 1030 156 155 902 6 111 759 97 628 159 20.97 66.67 20.97 66.52 19.40 64.60
Temelin B2 PWR CZ 1080 1030 744 812 954 6 037 562 93 864 322 100.00 63.25 99.94 62.74 101.17 63.82
Doel 1 PWR BE 454 433 744 338 530 3 169 852 - 100.00 80.57 99.89 79.64 100.04 79.26
Doel 2 PWR BE 454 433 744 342 566 3 207 325 - 100.00 80.33 99.99 79.90 100.70 79.85
Doel 3 PWR BE 1056 1006 580 598 227 7 689 354 - 77.87 83.03 74.76 82.40 75.72 82.46
Doel 4 PWR BE 1084 1033 744 811 914 9 270 685 - 100.00 98.92 100.00 98.03 100.23 96.78
Tihange 1 PWR BE 1009 962 0 0 3 005 326 - 0 34.45 0 33.85 0 33.98
Tihange 2 PWR BE 1055 1008 744 789 428 8 954 388 - 100.00 97.01 99.98 96.42 101.23 97.17
Tihange 3 PWR BE 1089 1038 744 809 473 8 226 519 - 100.00 86.78 99.99 86.19 99.83 85.94
News

atw Vol. 62 (2017) | Issue 6 ı June
• Nuclear governance needs significant
strengthening.
• Evolving nuclear suppliers impact
geopolitics.
• Innovative nuclear policy requires
“break the mold” partnerships.
The full report and additional information
can be found on the GNI
website: www.globalnexusinitiative.
org. A webcast of the media briefing
also is available.
| | www.nei.org, 8345
World
OECD figure show slight
decrease for nuclear share of
net electricity production
(nucnet) Net electricity production in
the 35 Organisation for Economic Cooperation
and Development (OECD)
member countries grew by 0.9 % in
2016 compared to 2015 with nuclear’s
share falling by 0.1% to 18.1 % figures
released by the International Energy
Agency show. Total OECD cumulative
production of nuclear electricity in
2016 was 1,873.6 TWh, a decrease
of 2.7 TWh. Europe was the only
region which decreased its nuclear
pro duction, by 19.6 TWh, or 2.4 %, to
790 TWh led by the continued
phaseout of nuclear electricity in
Germany as well as decreases in
the Czech Republic and France caused
by extended outages. There were
also operational outages in Slovenia
and Switzerland. There was a large
increase of 9.5 % in renewable
generation and a smaller, but still
significant, increase of 2.2 % for
hydro. Combustible fuels fell by
0.2 % and 0.1 %. Non-combustible
renew ables accounted for 22.4 %
of all generation compared to 21.6 %
in 2015.
| | www.oecd.org, 9345
Europe
Foratom: EU Energy Proposals
must take nuclear industry’s
views into account
(nucnet) Legislative proposals in the
European Commission’s ‘Clean Energy
for All Europeans’ package could
ensure a coherent and optimal approach
towards meeting energy and
climate objectives, provided they take
into account the views of the nuclear
energy industry, Foratom, the Brusselsbased
trade association for the industry
in Europe, said in a position paper.
The position paper said the goal of
the EU to decarbonise the economy by
more than 80% by 2050 cannot be
achieved without nuclear power.
The EC’s legislative proposals aim
to improve the functioning of the
energy market and make sure that all
energy technologies compete on a
level- playing field.
| | www.foratom.org, 3845
UK Nuclear Industry Study –
steps required to avoid Brexit
Euratom cliff edge
(nia) The Government needs to work
closely with industry in order to bring
about replacement arrangements for
Euratom in a timely manner to avoid a
cliff edge for the nuclear industry, is
the main message from a new position
paper, Exiting Euratom, published
today by the UK Nuclear Industry
Association (NIA).
The paper, prepared by the NIA
following detailed consultation and
discussion with its members, sets out
the priority areas for negotiations
with the European Commission as the
UK ceases to be a full member of the
Euratom community alongside the
process to leave the EU. The paper
also sets out the steps the UK Government
need to take to avoid serious
disruption to normal nuclear business
in the UK and across the European
Union.
The key steps for government include:
• Agreeing a replacement Voluntary
Offer Agreement with the IAEA for
a new UK safeguards regime
• Replacing the Nuclear Cooperation
Agreements (NCA) with
key nuclear markets; the Euratom
Community, United States,
Canada, Australia, Kazakhstan and
South Korea
• Clarifying the validation of the
UK’s current bilateral Nuclear
Co-operation Agreements with
Japan and other nuclear states
• Setting out the process for the
movement of nuclear material,
goods, people and services
• Agreeing a new funding arrangement
for the UK’s involvement in
Fusion 4 Energy and wider European
Union nuclear R&D programme
• Maintaining confidence in the
industry and securing crucial
investment
Addressing these priority areas will
enable the nuclear sector to continue
its work with other countries, both
within and outside the continuing EU,
as the UK ceases to be a member of the
European Union.
However, given the amount to be
concluded within the next 22 months,
there is a risk that new arrangements
will not be in place. The NIA is urging
the Government to begin these negotiations
by seeking an agreement with
the EU that existing arrangements
will continue to apply until the process
of agreeing new arrangements is
concluded, and avoiding the cliff edge
scenario that is not in the interests of
the industry, consumers, the UK or the
EU.
Tom Greatrex, Chief Executive of
the Nuclear Industry Association,
said:
“The UK civil nuclear industry is
ready and willing to work with the
Government as it begins the process of
putting replacement arrangements for
Euratom in place. The clock is ticking,
and this is a priority of increasing
urgency.
“This new report demonstrates
that without new arrangements in
place by the time the UK leaves the
Euratom community, there is scope
for real and considerable disruption.
The industry has not only set out the
priority areas to be addressed, but
also the steps we think the Government
needs to take to address those
issues.
“Government Ministers have stated
their desire to both work with industry
and to ensure the same high standards
will continue to apply as the UK leaves
the EU – there is no disagreement on
that principle.
“The Government now need to get
down to the work of putting such
arrangements in place, including a
prudent approach to ensuring there
are transitional arrangements in place,
to avoid a gap in regulation. That
would not be in the interests of the EU,
the UK or the industry globally.”
The NIA has called for a joint
industry and Government working
group to be created to help develop a
plan to preserve the essential benefits
of Euratom membership. This was
also a key recommendation by the
House of Lords Science and Technology
Committee in its report published
earlier this week.
| | www.niauk.org, 3856
Reactors
Kansai Electric to begin
restart process for Japan’s
Takahama-3 and -4
(nucnet) Kansai Electric Power Company
(Kepco) said it plans to begin the
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atw Vol. 62 (2017) | Issue 6 ı June
426
NEWS
restart process for the Takahama-3
and -4 nuclear reactor units in Fukui
Prefecture, western Japan.
The company said it had received
consent from local authorities to
restart the two 830-MW pressurised
water reactor units. Kepco said
it is scheduled to load fuel into
Takahama -4 later this week and plans
to connect the reactor to the grid in
late May and to start commercial
operation in mid-June.
The utility expects to load fuel into
Takahama-3 in mid-May and to connect
the unit to the grid in early June.
| | www.kepco.co.jp, 9345
Site work at Turkey’s Akkuyu
to begin in July 2017
(nucnet) Site Work at Turkey’s first
nuclear power plant at Akkuyu is
scheduled to begin this summer and
run for almost two years, Russian
state nuclear corporation Rosatom
said, according to the state-operated
domestic news agency RIA Novosti.
Earthworks will begin in July
with construction of the reactor pits
scheduled to begin in January 2019,
RIA Novosti said. Akkuyu, near Mersin
on the Turkey’s southern Mediterranean
coast, is to be built in
cooperation with Rosatom under a
contract signed in 2010. The station
will have four 1,200-MW VVER units.
On 3 March 2017, Akkuyu Nuclear, the
joint stock company in charge of the
project, applied for a construction
licence to the Turkish Atomic Energy
Authority.
| | www.rosatom.ru, 8345
Hungary: Construction of
initial facilities at Paks 2 to
begin in autumn
(nucnet) Construction of auxiliary
facilities for the planned two-unit
Paks 2 nuclear power station in
Hungary will begin in the autumn of
2017, Alexei Likhachev, head of
Russian state nuclear corporation
Rosatom, said in a statement.
Rosatom said auxiliary facilities
include a number of production,
storage and other buildings to be used
by contractors during the project’s
construction phase.
| | www.atomeromu.hu, 9345
Research
In ‘anti-nuclear’ Denmark:
How a reactor startup is helping
to change opinions
(nucnet) Seaborg Technologies of
Copenhagen is developing an
advanced thorium-based molten salt
reactor (MSR) and has received a
grant from the public funding agency
Innovation Fund Denmark, a move
that marks the first Danish investment
into nuclear fission research since a
1985 ban on nuclear energy.
The decision to fund the reactor,
known as the Seaborg CUBE-100
(short for Compact Used Fuel BurnEr),
is the beginning of the first Danish
venture into the development of novel
fission reactor concepts, Seaborg said.
NucNet editor-in-chief David
Dalton spoke to Seaborg’s co-founders
about the significance of the funding,
the next steps on the road to commercialisation,
and how attitudes towards
nuclear in traditionally anti-nuclear
Denmark are changing.
Full story for NucNet subscribers:
http://bit.ly/2oKDzCp
| | seaborg.dk, 3452
United States announces
€ 1 million pledge for modernization
of IAEA Nuclear
Applications Laboratories
(iaea) The United States announced a
pledge of €1 million to support the
modernization of the International
Atomic Energy Agency (IAEA) Nuclear
Applications Laboratories in Seibersdorf,
outside Vienna. These facilities
opened their doors in 1962 and play a
key role in the peaceful uses of nuclear
science and technology to assist
countries in areas such as human and
animal health, food security and the
protection of the environment.
The announcement was made
during the first day of the first session
| | Tentative schematic of the SWaB reactor.
(Illustration: Seaborg Technologies, Denmark).
of the Preparatory Committee for the
2020 Review Conference of the
Parties to the Treaty on the Non-
Proliferation of Nuclear Weapons
(NPT), May 2–12 in Vienna, Austria.
The contribution will go towards
the construction of a new Animal
Production and Health Laboratory,
one of eight laboratories that will
be upgraded under the Agency’s
Renovation of the Nuclear Applications
Laboratories (ReNuAL) and
ReNuAL Plus initiatives.
IAEA Director General Yukiya
Amano, addressing the Preparatory
Committee meeting, said the modernization
of the eight IAEA Nuclear
Applications Laboratories was proceeding
well.
“The laboratories train scientists,
support research in human health,
food and other areas, and provide
analytical services to national laboratories,”
Amano said. “I thank donor
countries for their generous contributions
and I hope that Member
States will continue to provide strong
support for further work on this
important modernization project.”
U.S. Ambassador Robert Wood, the
country’s Permanent Representative
to the Conference on Disarmament in
Geneva, said the IAEA plays a key
part in helping countries realize the
practical benefits of the NPT.
“I am pleased to announce a U.S.
pledge of € 1 million to support the
IAEA’s project to renovate its Nuclear
Applications Laboratories, in addition
to the nearly € 8.9 million we have
provided to date. This ReNuAL project
aims to renew the infrastructure
needed to sustain the IAEA’s programmes
for peaceful uses of nuclear
energy. We also urge other IAEA
Member States to join us in meeting
this year’s ReNuAL Plus fundraising
goals.”
“The U.S. pledge brings us halfway
to the funding target of € 2 million
that we need to reach by June to start
building this important laboratory
on time and to maximize our cost
efficiencies, so it is significant both
in terms of its size and timing,” said
IAEA Deputy Director General Aldo
Malavasi, who heads the IAEA’s
Department of Nuclear Sciences and
Applications.
“The IAEA’s work in helping
countries to apply nuclear technologies
to quickly detect and control
animal diseases posing threats to food
and economic security and to health is
increasingly in demand,” Malavasi
said. “This week, for example, in
Seibersdorf the Agency is training
News

atw Vol. 62 (2017) | Issue 6 ı June
| | The Animal Production and Health Laboratory
is one of eight facilities that will be upgraded
under the IAEA’s ReNuAL and ReNuAL Plus
initiatives. The photo shows the Agency's
training of veterinary experts in diagnosing
MERS-CoV in camels – a zoonotic disease that
is very dangerous to humans. (Photo: IAEA)
16 veterinary experts from seven
Member States in diagnosing
Middle East Respiratory Syndrome
Coronavirus (MERS-CoV) in camels –
a zoonotic respiratory disease that
is very dangerous to humans. This
contribution is very welcome.”
| | www.iaea.org, 9538
Company News
Nuclear safety: AREVA NP to
support international OECD
research program
(areva) The research program of the
Nuclear Energy Agency (NEA), as part
of the Organization for Economic
Co-operation and Development
(OECD), is being continued at AREVA
NP’s PKL test facility until mid-2020,
through a fourth 4-years contract
agreement (PKLIII-i). During this
period, the focus is to systematically
investigate thermal hydraulic
phenomena in pressurized water
reactors (PWR). Another main topic
within the program is the experimental
verification of cool-down
procedures for operational and
emergency manuals of such plants.
These efforts aim to enhance safety of
nuclear power plants worldwide.
AREVA NP’s unique PKL test facility
is part of AREVA NP’s Technical Center
in Erlangen (Germany) and models
the nuclear steam supply system of a
PWR in full scale height. Test series
can therefore be performed and
evaluated under realistic conditions.
The results obtained allow experts to
develop recommendations for plant
operation under accident situations.
”It is a great honor for us to further
contribute to safety research within
the international frame of OECD/
NEA. We have adapted our test facility
to the requirements of the new program
during the last months. Recently,
we started a first test series together
with our international partners“, said
Klaus Umminger, who is responsible
for the project at AREVA NP.
Under the umbrella of the OECD,
the project is funded jointly by
the German Federal Ministry for Economic
Affairs and Energy and other
contributors like Safety technical
support organizations, research
institutes and PWR operating utilities
from 14 OECD/NEA countries. The
budget of the project is about one
million euros per year.
| | www.areva.com, 7345
ROSATOM and the French
National Institute for Nuclear
Science and Technology held a
seminar on nuclear education
and training
(rosatom) On 27 April, 2017,
ROSATOM State Atomic Energy
Corporation and the National Institute
for Nuclear Science and Technology
administered by the French Atomic
Energy and Alternative Energies
Commission (INSTN) held a seminar
“Human capital issues facing nuclear
energy education and training today”
in the Russian Cultural Centre
in Paris. The event was attended
both by representatives of leading
education and research institutions,
and by business representatives, who
discussed the major and most central
issues of the nuclear energy education
and training today.
The sessions focused on global
trends and best innovative practices,
experience and potential of international
nuclear education programs
in partner countries, including internships
as well as further education and
instructor training courses.
During the seminar, the Superior
National School of Advanced Techniques
(ENSTA ParisTech), INSTN,
the National Research Nuclear University
MEPhI and Lomonosov MSU
presented their extensive expertise
in implementing education and
training programs. Representatives of
ROSATOM State Corporation and JSC
Rusatom Service highlighted the
importance of training highly qualified
specialists in the nuclear industry.
In her welcome speech Anne Lazar-
Sury, Governor for France to the IAEA,
Director of the Division for International
Affairs of the French Alternative
Energies and Atomic Energy
Commission called for increased
cooperation between France and
Russia in nuclear education.
Philippe Corréa, Director of the
INSTN noted that “education and
training is a pillar between research
and industry. We must anticipate
needs of our business partners and
consequently offer a challenging
professional development to our
students”, - he added.
Speaking about the needs of
nuclear education and training,
Evgeny Salkov, Director General
of JSC Rusatom Service stressed
that “development of cooperation
with INSTN is of great importance
for Rusatom Service. Considering
ambitious goals set out in our roadmap
we are interested in partnership
with those, who can help us prepare
personnel for nuclear facilities in time
and of quality, given that safety is our
top priority.
Andrey Rozhdestvin, Director of
Rosatom in Western Europe reminded
that «nowdays Europe is missing the
highly qualified human resources.
This question is even more crucial for
the newcomer countries».
In her turn, Tatiana Leonova,
Vice-Principal of MEPhI – partner of
ROSATOM State Corporation in the
sphere of nuclear education – introduced
the implemented programs to
the audience. “The best universities
develop successfully joint education
and research programs, it raises the
general level of education in newcomer
countries”,- she said.
During the seminar the participants
discussed current issues of
distance education and possible applications
of experimental equipment
and immersive 3D technologies in
educational process. Furthermore,
experience in implementing education
projects in foreign countries,
including newcomer countries embarking
on nuclear power programmes
was presented.
In conclusion, experts of both
countries underscored the importance
of further exchange of experience
and development of Russian-French
cooperation.
| | ROSATOM and the French National Institute for Nuclear Science and
Technology held a seminar on nuclear education and training.
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In the framework of the event, the
participants visited INSTN laboratories
where they got acquainted with
the ISIS training reactor as well as a
new virtual tool for studying radiotherapy
VERT (virtual environment
for radiotherapy training).
The French National Institute
for Nuclear Science and Technology
( INSTN) is a higher educational
institution administered by the French
Atomic Energy and Alternative
Energies Commission founded in
1956. The INSTN is under the joint
authority of the Ministry of National
Education, Higher Education and
Research, the Ministry of the Economy
and Finance and the Ministry of
the Environment. The Institute is the
main nuclear education centre in
France.
ROSATOM State Atomic Energy
Corporation brings together more
than 320 enterprises and scientific
organizations, including all civil
nuclear companies of Russia’s nuclear
industry, research centers and the
world’s only nuclear icebreaker fleet.
ROSATOM holds leading positions in
the global market of nuclear technologies
and is currently implementing
projects to build 42 nuclear power
units both in Russia and abroad.
| | www.rosatom.ru, 3845
Organisations
NEA and China’s National
Energy Administration
sign MOU to strengthen
co‐operation
(oecd-nea) On 28 April 2017, the NEA
and the National Energy Administration
of China (C/NEA) signed a
Memorandum of Understanding
(MOU) in the Field of Peaceful Uses
of Nuclear Energy, enhancing co‐operation
between both parties. An official
ceremony was held in Beijing, China,
at which C/NEA Deputy Adminis trator
Li Fanrong signed the MoU on behalf
of the C/NEA and NEA Director-
General William D. Magwood, IV,
signed on behalf of the NEA. The
agreement foresees co‐operation in a
number of fields, including nuclear
energy development, nuclear safety
research and radiological protection.
The memorandum of understanding
between the NEA and the C/NEA
represents further progress in the
growing collaboration between China
and the Agency, and complements
the memorandum of understanding
signed by the NEA and the National
Nuclear Safety Administration ( NNSA)
of China in 2014 and the Joint Declaration
on Co‐operation signed by the
NEA and the China Atomic Energy
Authority (CAEA) in 2013.
| | www.oecd-nea.ogr, 7349
People
Russ Brian announced as new
WANO Atlanta Centre director
(wano) David Garchow, Atlanta
Centre Director of the World Association
of Nuclear Operators (WANO)
and Vice President, International at
the Institute of Nuclear Power Operations
(INPO), has announced his
retirement effective 14 July 2017. He
joined INPO in 2005 as a Team Leader,
was elected Vice President of Plant
Technical Support in 2010 and
assigned to his current role in 2013.
Succeeding Garchow will be Russ
Brian, INPO’s Director of Plant Evaluations.
Effective 12 July 2017, he will
be promoted to WANO Atlanta Centre
Director and INPO Vice President,
International. Russ joined INPO
in 2010 and has served as a Team
Leader, Deputy Director of Corporate
Evaluations, and Director of Plant
Evaluations.
WANO Chief Executive Officer,
Peter Prozesky said, “We are grateful
for the substantive contributions Dave
Garchow has made to our industry
and wish him well in his retirement.
We also welcome Russ Brian to his
new role and look forward to him
continuing this strong legacy of
leadership.”
| | www.wano.info, 7345
NEA expert receives award
for international co-operation
from Korea
(oecd-nea) Dr Henri Paillère, NEA’s
Senior Nuclear Analyst and Acting
Head of the Division of Nuclear
Development, has been honoured with
the Award for Person of Merit for
International Co-operation in Nuclear
Industry by the Korean Ministry of
Science, ICT and Future Planning. The
honour was awarded in recognition
of Dr Paillère’s dedication and service
for the promotion of co-operation
between Korea and the NEA, including
through his work in support of the
Generation IV International Forum
(GIF) and the International Framework
for Nuclear Energy Cooperation
(IFNEC). “We are very pleased to see
Dr Paillère’s accomplishments being
acknowledged,” NEA Director- General
Mr Magwood said. “We are very fortunate
to have outstanding people like
Henri at the Agency.”
| | www.oecd-nea.org, 7345
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)].
January to December 2013
• Uranium: 34.00–43.50
• Conversion: 9.25–11.50
• Separative work: 98.00–127.00
January to December 2014
• Uranium: 28.10–42.00
• Conversion: 7.25–11.00
• Separative work: 86.00–98.00
January to June 2015
• Uranium: 35.00–39.75
• Conversion: 7.00–9.50
• Separative work: 70.00–92.00
June to December 2015
• Uranium: 35.00–37.45
• Conversion: 6.25–8.00
• Separative work: 58.00–76.00
2016
January to June 2016
• Uranium: 26.50–35.25
• Conversion: 6.25–6.75
• Separative work: 58.00–62.00
July 2016
• Uranium: 26.50–27.80
• Conversion: 6.00–6.50
• Separative work: 58.00–62.00
August 2016
• Uranium: 22.25–26.40
• Conversion: 5.50–5.75
• Separative work: 58.00–62.00
September 2016
• Uranium: 22.25–22.75
• Conversion: 5.50–5.75
• Separative work: 52.00–55.00
October 2016
• Uranium: 19.60–22.90
• Conversion: 5.50–5.75
• Separative work: 49.00–53.00
November 2016
• Uranium: 18.50–18.90
• Conversion: 5.50–5.75
• Separative work: 48.00–51.00
News

atw Vol. 62 (2017) | Issue 6 ı June
December 2016
• Uranium: 18.75–21.50
• Conversion: 5.50–5.75
• Separative work: 47.00–50.00
2017
January 2017
• Uranium: 20.25–25.50
• Conversion: 5.50–6.75
• Separative work: 47.00–50.00
February 2017
• Uranium: 23.50–26.50
• Conversion: 5.50–6.75
• Separative work: 48.00–50.00
March 2017
• Uranium: 24.00–26.00
• Conversion: 5.50–6.75
• Separative work: 47.00–50.00
| | Source: Energy Intelligence
www.energyintel.com
| | Uranium spot market prices from 1980 to 2017 and from 2007 to 2017. The price range is shown.
In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.
429
NEWS
Cross-border price
for hard coal
Cross-border price for hard coal in
[€/t TCE] and orders in [t TCE] for
use in power plants (TCE: tonnes of
coal equivalent, German border):
2012: 93.02; 27,453,635
2013: 79.12, 31,637,166
2014: 72.94, 30,591,663
2015: 67.90; 28,919,230
2016: 67.07; 29,787,178
I. quarter: 56.87; 8,627,347
II. quarter: 56.12; 5,970,240
III. quarter: 65.03, 7.257.041
IV. quarter: 88.28; 7,932,550
| | Source: BAFA, some data provisional
www.bafa.de
EEX Trading Results
in March 2017
(eex) In March 2017, the European
Energy Exchange (EEX) reached a
volume of 311.2 TWh on its power
derivatives markets, representing a
year-on-year increase of 22% (March
2016: 255.8 TWh).
In particular, power products for
the German-Austrian market contributed
to this result. At 250.5 TWh,
volumes in this market increased by
42 % (March 2016: 176.6 TWh).
This includes 225.9 TWh from
Phelix Futures and 24.6 TWh from
Phelix Options.
The March volumes comprised also
158.7 TWh registered at EEX for
clearing. Clearing and settlement
of all transactions was executed by
European Commodity Clearing (ECC).
The Settlement Price for base load
contract (Phelix Futures) with
delivery in 2018 amounted to
29.77 €/MWh. The Settlement Price
for peak load contract (Phelix Futures)
with delivery in 2018 amounted to
37.49 €/MWh.
| | Separative work and conversion market price ranges from 2007 to 2017. 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.
On the EEX Market for emission
allowances, a total volume of
117.5 million tonnes of CO 2 was
traded in March which represents a
year-on-year increase of 50 % (March
2016: 78.4 million tonnes of CO 2 ). On
the EUA secondary market, volumes
have doubled to 33.6 million tonnes
of CO 2 (March 2016: 15.7 million
tonnes of CO 2 ). The primary market
auctions contributed 83.5 million
tonnes of CO 2 to the total volume.
The E-Carbix amounted to
5.09 €/EUA, the EUA price with
delivery in December 2016 amounted
to 4.63/5.91 €/ EUA (min./max.).
| | www.eex.com
MWV crude oil/product prices
in March 2017
(mwv) According to information and
calculations by the Association of the
German Petroleum Industry MWV e.V.
in March 2017 the prices for super fuel
and heating oil noted lower for fuel oil
sligthly lower compared with the
previous month February 2017. The
average gas station prices for Euro
super consisted of 136.28 €Cent
( February 2017: 139.39 €Cent, approx.
-3.11 % in brackets: each information
for previous month or rather previous
month comparison), for diesel fuel of
116.56 €Cent (118.27; -1.45 %) and
for heating oil (HEL) of 56.81 €Cent
(59.28, -4.17 %).
The tax share for super with a
consumer price of 136.28 €Cent
(139.39 €Cent) consisted of
65.45 €Cent (48.03 %, 65.45 €Cent)
for the current constant mineral oil
tax share and 21.76 €Cent (current
rate: 19.0 % = const., 22.26 €Cent)
for the value added tax. The product
price (notation Rotterdam) consisted
of 36.05 €Cent (26.45 %, 39.59 €Cent)
and the gross margin consisted of
13.02 €Cent (9.5 %; 12.09 €Cent).
Thus the overall tax share for super
results of 67.0 % (66.0 %).
Worldwide crude oil prices
(monthly average price OPEC/Brent/
WTI, Source: U.S. EIA) were approx.
-6.48 % (+1.41 %) lower in March
compared to February 2017 also
despite the decision of the OPEC
to restrict and lower the crude oil
production. The market showed a
stable development with lower
prices; each in
US-$/ bbl: OPEC basket: 50.32
(53.37); UK-Brent: 51.59 (54.87);
West Texas Intermediate (WTI):
49.33 (53.47)
| | www.mwv.de
News

atw Vol. 62 (2017) | Issue 6 ı June
430
NUCLEAR TODAY
Links to reference
sources:
European Commission
announcement on
Clean Energy for All:
http://bit.ly/2fQbVQk
Foratom position
paper: http://bit.ly/
2oI2Sna
Clean Energy Proposals are Chance for
Nuclear to have Rightful Place at Policy
Table
John Shepherd
Every New Year’s Eve in Germany, there is a tradition that has become as fixed on the calendar as the ringing of
church bells at midnight and the clinking of Champagne-filled glasses to toast the year ahead.
This tradition, curiously, takes the form of a number of
television stations broadcasting an 18-minute black-andwhite
1963 TV recording of an English-language comedy
sketch called ‘Dinner for One’ – also known as the ‘90th
Birthday’.
The show features the late British comedians Freddie
Frinton and May Warden. For those of you who might be
unfamiliar with the sketch – although its popularity has
since spread to other European nations – it centres on the
annual birthday dinner of upper-class Englishwoman, Miss
Sophie. Every year, she hosts a celebration dinner for her
friends. The problem is that, due to Miss Sophie’s considerable
age, she has outlived all of her friends.
The only ones at the annual celebration are Miss Sophie
herself and her equally-aged manservant, James. His task
it is to make his way around the dining table, impersonating
each of the absent guests in turn. As he carries out his
duties, James asks Miss Sophie: “Same procedure as every
year?” To which she nods affirmatively. Poor James is also
required to drink the copious glasses of alcohol on the
guests’ behalf in toasts ordered by Miss Sophie throughout
the evening until, inevitably, he becomes inebriated with
hilarious consequences.
I cannot do justice to the sketch here – you must see it
for yourselves! But this anniversary event reminds me of a
ritual that we frequently see in attempts to guide energy
policy towards a sustainable future and to combat the
effects of climate change.
In April, Foratom, the Brussels-based trade association
for the nuclear industry in Europe, published a position
paper on the European Commission’s ‘Clean Energy for All
Europeans’ package of EU legislative proposals.
The proposals seek to improve the functioning of the
energy market and ensure all energy technologies can
compete on a level-playing field without jeopardising
climate and energy targets.
Foratom has called for “cost-efficient decarbonisation,
an effective power market leading to competitive and
affordable electricity prices for the end consumers and the
promotion of investments in low carbon technologies”.
Foratom also underlined the importance of the EU
Emissions Trading Scheme (ETS) and “of protecting it
from conflicting policy overlaps, in particular from the proposed
new 30% energy efficiency binding target”.
The European Commission’s proposals were unveiled
towards the end of last year in a move designed to show the
clean energy transition “is the growth sector of the future”.
“Clean energies in 2015 attracted global investment of
over EUR300 billion,” the Commission said. “The EU is
well-placed to use our research, development and innovation
policies to turn this transition into a concrete industrial
opportunity. By mobilising up to EUR177bn of public and
private investment per year from 2021, this package can
generate up to 1% increase in gross domestic product over
the next decade and create 900,000 new jobs.”
The EU’s Commissioner for Climate Action and Energy
Miguel Arias Canete said: "Our proposals provide a strong
market pull for new technologies, set the right conditions
for investors, empower consumers, make energy markets
work better and help us meet our climate targets. I'm
particularly proud of the binding 30% energy efficiency
target, as it will reduce our dependency on energy imports,
create jobs and cut more emissions.”
The Commissioner said last year that Europe was “on
the brink of a clean energy revolution”. He added: “We can
only get this right if we work together. With these proposals,
the Commission has cleared the way to a more
competitive, modern and cleaner energy system. Now we
count on the European Parliament and our member states
to make it a reality.”
Despite the Commissioner’s rallying call, can such an
ambitious policy agenda ever come to fruition? And can
“new technologies” also really encompass support for new
nuclear technologies? Yes, the proposals may well be
adopted by the parliament and EU nations, but I would
suggest that while countries such as Austria and Germany
are dead set against the further development and deployment
of nuclear (within their own borders at least), how
can the clean energy package ever work for the benefit of
the EU as a whole?
It is for this reason that I started this article by referring
to Miss Sophie’s annual birthday dinner ritual. There are
rather amusing parallels in the regular unveiling of various
proposals at European and international level. It is the
“same procedure”, if not quite every year, but as regular as
clockwork. A new policy is drawn up that aims to be
“ inclusive” and encourage all low-carbon energy technologies
to compete to offer the best deals for electricity
consumers and the environment.
However, many of the EU nations who are called
together to sit around the policy table at each of these
ritual initiatives are, like Miss Sophie’s guests, not really
there. Yes, nations are represented in a physical sense, but
many “go through the motions”, to coin a phrase. They
nod, toast the initiatives, then go back to what suits their
political objectives at home.
This is frequently the case in relation to the benefits of
nuclear energy. European and international bodies recognise
the benefits of nuclear as part of a mix of energy
technologies, everyone agrees, then it is the “same procedure”
as before, as James would say. Everyone removes
from the mix what does not suit their domestic policies –
and the result is an inebriated, incoherent performance, in
this case an unbalanced energy policy.
If Europe seeks to have a coherent and inclusive energy
policy, which encompasses all low-carbon contributors,
nuclear must be allowed a place at the policy table. If not,
it will be a charade only worthy of the comic antics of
James and Miss Sophie.
Author
John Shepherd
nuclear 24
41a Beoley Road West, St George’s
Redditch B98 8LR, United Kingdom
Nuclear Today
Clean Energy Proposals are Chance for Nuclear to have Rightful Place at Policy Table ı John Shepherd

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