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<strong>2017</strong><br />
6<br />
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16.– €<br />
378<br />
AMNT <strong>2017</strong>:<br />
Opening Address<br />
384 ı AMNT <strong>2017</strong>: Best Paper<br />
Emplacement Radiation Exposure Calculations<br />
for Generic Deep Geological Repositories<br />
392 ı Environment and Safety<br />
Retrofitting a Spent Fuel Pool Spray System<br />
396 ı Operation and New Build<br />
Cyber Security in Nuclear Power Plants<br />
402 ı Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
India Goes Nuclear<br />
Dear reader, India is a country of diversity and with its 1.3 billion citizens India is after China not only second most<br />
populous country in the world but also the most populous democracy in the world. This reflects the great responsibility<br />
for the countries’ politics, to create social conditions, which further maintain and strengthen the democracy. Economic<br />
growth is thereby unquestionably one of the most important components in order to expand social services and to create<br />
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<br />
global financial crisis as of 2007.<br />
With a view to the energy supply India is today, after China<br />
and the USA the third largest energy consumer in the<br />
world. Despite of extensive available energy resources,<br />
India developed to an important energy importer of fossil<br />
fuels. With a share of over 75 % of the energy generation<br />
the importance of coal for the energy and electricity supply<br />
is very dominant. In the year 2013 around 692 million tons<br />
of coal were used of which 159 million tons were imported.<br />
The 22 nuclear power plants with a gross capacity of<br />
6,780 MW have a share of around 2.2 % of the country’s<br />
total power generation capacity of 303,071 MWe and of<br />
the generation of around 3.5 % through the production of<br />
35 gigawatt hours in 2016. Due to the combination of a<br />
today comparatively low per capita rate of electricity<br />
consumption in the amount of 1,000 kWh per inhabitants<br />
and aspired growth, as well as the need to provide<br />
electricity to the approximately 240 million persons in<br />
India which do not have any access to electricity today, it<br />
will certainly further increase. Until the 2020s a doubling<br />
is expected. It should be noted for India, that the agriculture<br />
proportion on the energy consumption – especially for the<br />
irrigation of fields- up to one third, clarifies that a secure<br />
supply of energy is not a question of comfort put also of a<br />
primary care.<br />
This poses the country and its decision makers to great<br />
challenges. In order to manage this situation no options<br />
are excluded. Thus, growth for all energy carriers in India<br />
is expected and aimed for in the upcoming years until the<br />
middle of the 2020s with strong differing degrees of the<br />
single energy carrier. Ambitioned is the extension of<br />
renewables, with a target setting especially for wind of<br />
+60,000 MW and photovoltaic of +100,000 MW, which<br />
corresponds in total to a fourfold increase. But also the<br />
coal-fired generation will further increase.<br />
....and nuclear energy?<br />
Research and development of nuclear energy in India have<br />
a long national tradition. Today’s Bhaba Atomic Research<br />
Centre near Mumbai was established in the 1950s. A first<br />
light water reactor or rather heavy water moderated<br />
pressurised water reactor of the Canadian type CANDU<br />
was put in operation in 1969 or rather 1972 at the sites<br />
Tarapur and Rajasthan. The advancement of nuclear<br />
energy within the international network was then<br />
inhibited, as India, being a nuclear power, had not signed<br />
the Nuclear Non-Proliferation Treaty.<br />
India’s nuclear economy depended thus on its own<br />
development or rather further development of the<br />
expansion. Through standardising the heavy water reactor<br />
technology the possibility was given to establish an own<br />
productive reactor type and to commission until today<br />
18 plants. The long term perspective of nuclear energy will<br />
be underlined with the prototype establishment of a fast<br />
sodium-cooled 500-MW-reactor at the site Kalpakkam,<br />
whose commissioning is planned for autumn <strong>2017</strong>.<br />
Additionally the Indian Department for Atomic Energy just<br />
recently communicated, that two further 600-MW-breeder<br />
reactor shall follow in Kalpakkam<br />
With the end of the East-West conflict the relationships<br />
of many countries with India changed in the matter of<br />
nuclear technology. Through cooperation with Russia two<br />
WWER-reactors with each 1,000 MW of (output) were<br />
established and put in operation at the site Kudankulam<br />
as of the year 2002. An agreement with the Nuclear<br />
Suppliers Group in 2008 opened up the path for a couple<br />
bilateral agreements for the expansion of nuclear energy.<br />
Miscellaneous new-build projects are mentioned and<br />
negotiated repeatedly since this year in order to achieve<br />
the expansion target of +10 % capacity per year until the<br />
year 2025.<br />
India is getting serious also in the matter of nuclear<br />
energy. The Indian Prime Minister Narendra Mori<br />
announced in May <strong>2017</strong> as a first step for the government<br />
the initiation of a national nuclear energy expansion<br />
program. This programme shall bring a strong push<br />
to entire Indian economy: 10 nuclear power plant projects<br />
on the basis of the Indian heavy water reactor technology<br />
with an overall performance of 6,700 MW* – the<br />
same amount as the currently operated ones – shall be<br />
accommodated within the next 5 years.<br />
The full investment is mentioned with 11 b. $. Just in<br />
the country’s nuclear industry 33,400 new, qualified jobs<br />
shall be generated in this manner. With the experience<br />
made from the establishment and commissioning of<br />
today’s operating heavy water reactors and through the<br />
standardisation of 10 new plants, the „fleet construction<br />
program“ shall generate synergies through an „Economy<br />
by Number“ and thus electrically enable the aspired and<br />
comparatively low investment costs of 1,650 $ per installed<br />
kilowatt.<br />
India’s paths of a future energy supply are diverse and<br />
include the path of using nuclear energy; step by step and<br />
under the target set of further 80,000 MW until the end of<br />
the 2020’s.<br />
Christopher Weßelmann<br />
– Editor in Chief –<br />
* At the power<br />
generation 1 MW of<br />
installed power<br />
output corresponds<br />
due to a higher<br />
availability<br />
approximately to<br />
4 MW of installed<br />
wind power and<br />
8 MW of installed<br />
photovoltaic<br />
capacity – the often<br />
underestimated<br />
difference between<br />
labour and<br />
performance<br />
367<br />
EDITORIAL<br />
Editorial<br />
India Goes Nuclear
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
EDITORIAL 368<br />
*Bei der Stromerzeugung<br />
entspricht<br />
1 MW installierter<br />
Kernenergieleistung<br />
aufgrund höherer<br />
Arbeitsverfügbarkeit<br />
etwa 4 MW installierter<br />
Windkraft und<br />
8 MW installierter<br />
Fotovoltaikleistung –<br />
der häufig unterschätzte<br />
Unterschied<br />
von Leistung und<br />
Arbeit.<br />
Indiens Weg mit der Kernenergie<br />
Liebe Leserin, lieber Leser, Indien ist ein Land der Vielfalt und mit 1,3 Milliarden Menschen nicht nur zweitbevölkerungsreichstes<br />
Land der Welt nach China, sondern auch bevölkerungsreichste Demokratie der Welt. Dies<br />
spiegelt die große Verantwortung für die Politik des Landes wieder, gesellschaftliche und soziale Rahmenbedingungen<br />
zu schaffen, die diese Demokratie weiter zusammenhalten und festigen.<br />
Wirtschaftliches Wachstum ist dabei unzweifelhaft eine<br />
der wichtigsten Komponenten, um das Sozialwesen<br />
auszubauen und Lebensqualität zu schaffen. In den vergangenen<br />
10 Jahren ist Indiens Wirtschaft auch deshalb<br />
um durchschnittlich 11 % pro Jahr gewachsen und dies<br />
selbst in den Zeiten der globalen Finanzkrise ab 2007.<br />
Mit Blick auf die Energieversorgung ist Indien heute<br />
nach China und den USA drittgrößter Energieverbraucher<br />
der Welt. Trotz umfangreicher vorhandener Energie reserven<br />
hat sich Indien zum bedeutenden Energie importeur<br />
fossiler Energieträger entwickelt. Die Bedeutung von<br />
Kohle für die Energie- und Stromversorgung ist sehr dominant<br />
mit einem Anteil von über 75 % der Stromerzeugung.<br />
Hier wurden im Jahr 2013 rund 692 Millionen Tonnen<br />
Steinkohle eingesetzt, von denen 159 Millionen Tonnen<br />
importiert wurden. Die 22 Kernkraftwerke haben mit<br />
6.780 MW Bruttoleistung einen Anteil von rund 2,2 % an<br />
der Gesamtstromerzeugungsleistung des Landes von<br />
303.071 MWe und an der Erzeugung von rund 3,5 % durch<br />
die Produktion von 35 Gigawattstunden in 2016.<br />
Aufgrund der Kombination von heute vergleichsweise<br />
niedrigem pro Kopf Stromverbrauch in Höhe von 1.000<br />
kWh pro Einwohner und angestrebtem Wachstum sowie<br />
der Notwendigkeit den etwa 240 Millionen Menschen in<br />
Indien, die heute keinen Zugang zu Elektrizität haben,<br />
einen solchen zu verschaffen, wird sich dieser weiter<br />
erhöhen. Bis in die 2020er-Jahre wird eine Verdoppelung<br />
erwartet. Dabei ist für Indien zudem zu bemerken, dass<br />
der Anteil der Landwirtschaft am Stromverbrauch –<br />
wesent lich zur Bewässerung von Feldern – von bis zu<br />
einem Drittel verdeutlicht, dass eine gesicherte Stromversorgung<br />
keine Frage des Komforts, sondern auch der<br />
Grundversorgung ist.<br />
Dies stellt das Land und seine Entscheider vor enorme<br />
Herausforderungen. Um sie zu bewältigen, wird keine<br />
Option ausgeschlossen. Daher wird für alle Energieträger<br />
in Indien in den kommenden Jahren bis Mitte der 2020er<br />
ein Wachstum erwartet und angestrebt, mit unterschiedlich<br />
starker Ausprägung der einzelnen Energieträger.<br />
Ambitioniert ist der Ausbau der Erneuerbaren, mit einer<br />
Zielvorgabe vor allem beim Wind von +60.000 MW und<br />
der Fotovoltaik von +100.000 MW, was insgesamt einer<br />
Vervierfachung entspricht. Aber auch die Kohleverstromung<br />
wird sich weiter erhöhen.<br />
... und die Kernenergie?<br />
Forschung und Entwicklung der Kernenergie in Indien<br />
haben eine lange nationale Tradition. In den 1950er- Jahren<br />
wurde das heutige Bhaba Atomic Research Center nahe<br />
Mumbai errichtet. An den Standorten Tarapur und<br />
Rajasthan wurden 1969 bzw. 1972 ein erster Leichtwasserreaktor<br />
bzw. ein schwerwassermoderierter Druckwasserreaktor<br />
vom kanadischen CANDU-Typ in Betrieb<br />
genommen. Die Weiterentwicklung der Kernenergie im<br />
internationalen Verbund war dann aber gehemmt, da<br />
Indien als Atomwaffenmacht nicht den Atomwaffensperrvertrag<br />
unterzeichnet hatte. Indiens Nuklearwirtschaft<br />
war somit auf eigene Entwicklungen bzw.<br />
Weiterentwicklungen bei Ausbau angewiesen. Durch<br />
Standardisierung der Schwerwasser reaktortechnologie<br />
gelang es, eine eigene leistungsfähige Reaktorlinie zu<br />
etablieren und bis heute 18 solcher Anlagen in Betrieb zu<br />
nehmen. Die Langfristperspektive der Kern energie wird<br />
mit der prototypischen Errichtung eines schnellen natriumgekühlten<br />
500-MW-Reaktors am Standort Kalpakkam<br />
unterstrichen, dessen Inbetriebnahme für Herbst <strong>2017</strong><br />
angekündigt ist. Zudem wurde jüngst vom indischen<br />
Department for Atomic Energy mitgeteilt, dass weitere zwei<br />
600-MW-Brutreaktoren in Kalpakkam folgen sollen.<br />
Mit dem Ende des Ost-West-Konfliktes änderten<br />
sich auch die Beziehungen vieler Länder in Sachen<br />
Kernenergietechnologie mit Indien. Im Rahmen von<br />
Kooperationen mit Russland wurden so ab 2002 zwei<br />
WWER-Reaktoren mit jeweils 1.000 MW Leistung am<br />
Standort Kudankulam errichtet und in Betrieb genommen.<br />
Eine Vereinbarung mit der Nuclear Suppliers Group im<br />
Jahr 2008 eröffnete zudem den Weg für eine Reihe von<br />
bilateralen Vereinbarungen zum Ausbau der Kernenergie.<br />
Verschiedenste Neubauvorhaben werden seit diesem Jahr<br />
immer wieder genannt und verhandelt, um das Ausbauziel<br />
von +10 % Kapazität pro Jahr bis 2025 zu erreichen.<br />
So macht Indien jetzt ernst, auch in Sachen Kernenergie.<br />
Der indische Premierminister Narendra Mori gab<br />
im Mai <strong>2017</strong> für die Regierung als den ersten Schritt<br />
die Initiierung eines nationalen Kernenergieausbauprogramms<br />
bekannt. Dieser soll der indischen Wirtschaft<br />
insgesamt einen großen Schub verschaffen: Innerhalb der<br />
kommenden 5 Jahre sollen 10 Kernkraftwerksprojekte auf<br />
Basis der indischen Schwerwasserreaktortechnologie<br />
mit einer Gesamtleistung von 6.700 MW* – also in<br />
gleicher Höhe, wie die derzeit betriebenen – aufgenommen<br />
werden. Der Gesamtinvestitionsumfang wird<br />
mit 11 Mrd. $ angegeben. Allein in der kerntechnischen<br />
Industrie des Landes sollen so 33.400 neue, qualifizierte<br />
Arbeitsplätze geschaffen werden. Mit den Erfahrungen<br />
aus Errichtung und Inbetriebnahme der heute laufenden<br />
Schwerwasserreaktoren und durch die Standardisierung<br />
der 10 Neuanlagen soll das „Flottenbauprogramm“<br />
Synergien durch eine „Economy by Number“ erzeugen und<br />
so die angestrebten vergleichsweise niedrigen Investitionskosten<br />
von 1.650 $ pro installiertem Kilowatt elektrisch<br />
ermöglichen.<br />
Indiens Wege der zukünftigen Energieversorgung sind<br />
vielfältig und schließen den Weg der Kernenergienutzung<br />
mit ein; Schritt für Schritt und einer Zielvorgabe von<br />
weiteren 80.000 MW bis Ende der 2020er-Jahre.<br />
Christopher Weßelmann<br />
– Chefredakteur –<br />
Editorial<br />
India Goes Nuclear
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
370<br />
Issue 6<br />
June<br />
CONTENTS<br />
378<br />
AMNT <strong>2017</strong>:<br />
Opening Address<br />
| | Research, an inevitable part of know-how preservation and development. View of the FRM II upper head reactor components during<br />
commissioning. (Photo: FRM II)<br />
Editorial<br />
India Goes Nuclear. . . . . . . . . . . . . . . . . . . . 367<br />
Indiens Weg mit der Kernenergie . . . . . . . . . . 368<br />
Abstracts | English . . . . . . . . . . . . . . . . . . . 372<br />
Abstracts | German . . . . . . . . . . . . . . . . . . . 373<br />
AMNT <strong>2017</strong><br />
48 th Annual Meeting on Nuclear Technology<br />
(AMNT <strong>2017</strong>): Opening Address . . . . . . . . . . . 378<br />
Ralf Güldner<br />
48 th Annual Meeting on Nuclear Technology<br />
(AMNT <strong>2017</strong>): Impressions . . . . . . . . . . . . . . . 382<br />
Inside Nuclear with NucNet<br />
Q&A: Poland’s Progress on the Road<br />
to New Nuclear . . . . . . . . . . . . . . . . . . . . . . 374<br />
NucNet<br />
Calendar . . . . . . . . . . . . . . . . . . . . . . . 376<br />
DAtF Notes. . . . . . . . . . . . . . . . . . . . . .377<br />
384<br />
| | Different angles of the phantom with respect to POLLUX.<br />
AMNT <strong>2017</strong>: Best Paper<br />
Monte-Carlo Based Comparison of the<br />
Personal Dose for Emplacement Scenarios<br />
of Spent Nuclear Fuel Casks in Generic<br />
Deep Geological Repositories . . . . . . . . . . . . . 384<br />
378<br />
Héctor Saurí Suárez, Bo Pang, Frank Becker and Volker Metz<br />
| | Dr. Ralf Güldner delivering his Opening Address.<br />
Contents
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
371<br />
Spotlight on Nuclear Law<br />
The 15 th German Atomic Energy Act Amendment<br />
to the Implementation of the EURATOM Nuclear<br />
Safety Directive . . . . . . . . . . . . . . . . . . . . . . 391<br />
Die 15. AtG-Novelle zur Umsetzung<br />
der EURATOM-Sicherheits-Richtlinie. . . . . . . . . 391<br />
Christian Müller-Dehn<br />
Environment and Safety<br />
Retrofitting a Spent Fuel Pool Spray System<br />
for Alternative Cooling as a Strategy for Beyond<br />
Design Basis Events . . . . . . . . . . . . . . . . . . . 392<br />
Christoph Hartmann and Zoran Vujic<br />
392<br />
| | AP1000® Plant Spent Fuel Pool Spray System.<br />
Operation and New Build<br />
Cyber Security in Nuclear Power Plants and its<br />
Portability to Other Industrial Infrastructures . . . 396<br />
Sébastien Champigny, Deeksha Gupta, Venesa Watson<br />
and Karl Waedt<br />
|408<br />
418<br />
| | Nodalization for the primary system of AP1000.<br />
Research and Innovation<br />
Reliability Analysis on Passive Residual Heat<br />
Removal of AP1000 Based on Grey Model . . . . . 408<br />
Qi Shi, Zhou Tao, Muhammad Ali Shahzad, Li Yu and Jiang Guangming<br />
Experimental Investigation of a Two-Phase<br />
Closed Thermosyphon Assembly for Passive<br />
Containment Cooling System . . . . . . . . . . . . . 413<br />
Kyung Ho Nam and Sang Nyung Kim<br />
Displacement of Cryomodule<br />
in CADS Injector II . . . . . . . . . . . . . . . . . . . . 418<br />
Yuan Jiandong, Zhang Bin, Wang Fengfeng, Wan Yuqin, Sun Guozhen,<br />
Yao Junjie, Zhang Juihui and He Yuan<br />
| Model of the cryomudule.<br />
CONTENTS<br />
KTG Inside . . . . . . . . . . . . . . . . . . . . . . 422<br />
News . . . . . . . . . . . . . . . . . . . . . . . . . 424<br />
396<br />
| | Overview of cyber security portfolio.<br />
Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide<br />
Composition Data Using Percentage Differences<br />
and Detailed Analysis . . . . . . . . . . . . . . . . . . 402<br />
Man Cheol Kim<br />
Nuclear Today<br />
Clean Energy Proposals are Chance for Nuclear<br />
to have Rightful Place at Policy Table . . . . . . . . 430<br />
John Shepherd<br />
Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 375<br />
AMNT 2018: Call for Papers . . . . . . . . . . . . . Insert<br />
DAtF: Kernenergie in Zahlen <strong>2017</strong> . . . . . . . . . Insert<br />
Contents
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
372<br />
ABSTRACTS | ENGLISH<br />
Q&A: Poland’s Progress on the Road to New<br />
Nuclear<br />
NucNet | Page 374<br />
Conflicting reports have emerged from Poland<br />
about plans for its first reactors. Professor Grzegorz<br />
Wrochna of the Polish National Centre for Nuclear<br />
Research says the programme is on track and a<br />
business model is expected soon. The previous<br />
government’s programme calls for 6 GW in two<br />
locations. The government recently published a<br />
strategy for responsible development which calls<br />
for the nuclear programme to be sped up. However,<br />
no capacity figures were included. The industry<br />
should not be bound by a rigid number. In time,<br />
maybe we will speak of 4 GW or 12 GW, but it will<br />
depend on market needs and financial possibilities.<br />
The biggest risks do not come from cancellation or<br />
public opinion. They come from delays.<br />
48 th Annual Meeting on Nuclear Technology<br />
(AMNT <strong>2017</strong>): Opening Address<br />
Ralf Güldner | Page 378<br />
The past twelve months in German nuclear energy<br />
policy have been characterised mainly by legislative<br />
clearing up work which has been pending since<br />
the decision for an accelerated phase-out of nuclear<br />
power in 2011. This applies particularly to the<br />
reorganisation of financing in nuclear waste<br />
management. The other major political work<br />
package was the amendment to the Site Selection<br />
Act. Our real challenge though is nuclear expertise.<br />
This is important for research, for industry but<br />
above all for the state itself. The question of<br />
expertise covers the whole range of scientific and<br />
technical knowledge relating to nuclear technology:<br />
basic nuclear research, reactor safety research,<br />
radiochemistry, radiological protection, nuclear<br />
applications in medicine, industry and agriculture,<br />
to mention but a few examples.<br />
Monte-Carlo Based Comparison of the<br />
Personal Dose for Emplacement Scenarios<br />
of Spent Nuclear Fuel Casks in Generic Deep<br />
Geological Repositories<br />
Héctor Saurí Suárez, Bo Pang,<br />
Frank Becker and Volker Metz | Page 384<br />
In the operational phase of a deep geological<br />
disposal facility for high-level nuclear waste, the<br />
radiation field in the vicinity of a waste cask is<br />
influenced by the backscattered radiation of the<br />
surrounding walls of the emplacement drift. For a<br />
comparison of disposal of spent nuclear fuel in<br />
various host rocks, it is of interest to investigate the<br />
influence of the surrounding materials on the<br />
radiation field and the personal radiation exposure.<br />
In this generic study individual dosimetry of<br />
personnel involved in emplacement of casks with<br />
spent nuclear fuel in drifts in rock salt and in a clay<br />
formation was modelled.<br />
The 15th German Atomic Energy Act<br />
Amendment to the Implementation of the<br />
EURATOM Nuclear Safety Directive<br />
Christian Müller-Dehn | Page 391<br />
The 15th German Atomic Energy Act Amendment<br />
has now passed the parliamentary legislative<br />
procedure with the decision of the Bundestag in the<br />
third reading of 30 March <strong>2017</strong>. The publication in<br />
the Federal Law Gazette (Bundesgesetzblatt) is still<br />
pending. The background of the amendment is the<br />
addition to the Euratom safeguards directive<br />
adopted by the European Council in July 2014. This<br />
directive has to be implemented in the national<br />
regulations of the EURATOM Member States.<br />
However, since most of these supplements were<br />
already standard in German atomic law, the<br />
regulatory requirements for Germany were low.<br />
This is also explicitly stated in the statement to the<br />
act.<br />
Retrofitting a Spent Fuel Pool Spray System<br />
for Alternative Cooling as a Strategy for<br />
Beyond Design Basis Events<br />
Christoph Hartmann and Zoran Vujic | Page 392<br />
Due to requirements for nuclear power plants to<br />
withstand beyond design basis accidents, including<br />
events such as happened in 2011 in the Fukushima<br />
Daiichi Nuclear Power Plant in Japan, alternative<br />
cooling of spent fuel is needed. Alternative spent<br />
fuel cooling can be provided by a retrofitted spent<br />
fuel pool spray system based on the AP1000 plant<br />
design. As part of Krško Nuclear Power Plant’s<br />
Safety Upgrade Program, Krško Nuclear Power<br />
Plant decided on, and Westinghouse successfully<br />
designed a retrofit of the AP1000® plant spent fuel<br />
pool spray system to provide alternative spent fuel<br />
cooling.<br />
Cyber Security in Nuclear Power Plants<br />
and Its Portability to Other Industrial<br />
Infrastructures<br />
Sébastien Champigny, Deeksha Gupta,<br />
Venesa Watson and Karl Waedt | Page 396<br />
Power generation increasingly relies on decentralised<br />
and interconnected computerised systems.<br />
Concepts like “Industrial Internet of Things” of the<br />
Industrial Internet Consortium (IIC), and “Industry<br />
4.0” find their way in this strategic industry. Risk<br />
of targeted exploits of errors and vulnerabilities<br />
increases with complexity, interconnectivity<br />
and decentralization. Inherently stringent security<br />
requirements and features make nuclear<br />
computerised applications and systems a benchmark<br />
for industrial counterparts seeking to hedge<br />
against those risks. Consequently, this contribution<br />
presents usual cyber security regulations and<br />
practices for nuclear power plants. It shows how<br />
nuclear cyber security can be ported and used in an<br />
industrial context to protect critical infrastructures<br />
against cyber-attacks and industrial espionage.<br />
Validation of Spent Nuclear Fuel Nuclide<br />
Composition Data Using Percentage<br />
Differences and Detailed Analysis<br />
Man Cheol Kim | Page 402<br />
Nuclide composition data of spent nuclear fuels<br />
are important in many nuclear engineering<br />
applications. In reactor physics, nuclear reactor<br />
design requires the nuclide composition and the<br />
corresponding cross sections. In analyzing the<br />
radiological health effects of a severe accident on<br />
the public and the environment, the nuclide<br />
composition in the reactor inventory is among the<br />
important input data. Nuclide composition data<br />
need to be provided to analyze the possible<br />
environmental effects of a spent nuclear fuel<br />
repository. They will also be the basis for identifying<br />
the origin of unidentified spent nuclear fuels or<br />
radioactive materials.<br />
Reliability Analysis on Passive Residual Heat<br />
Removal of AP1000 Based on Grey Model<br />
Qi Shi, Zhou Tao, Muhammad Ali Shahzad,<br />
Li Yu and Jiang Guangming | Page 408<br />
It is common to base the design of passive systems<br />
on the natural laws of physics, such as gravity, heat<br />
conduction, inertia. For AP1000, a generation-III<br />
reactor, such systems have an inherent safety<br />
associated with them due to the simplicity of their<br />
structures. However, there is a fairly large amount<br />
of uncertainty in the operating conditions of these<br />
passive safety systems. In some cases, a small<br />
deviation in the design or operating conditions can<br />
affect the function of the system. The reliability of<br />
the passive residual heat removal is analysed.<br />
Experimental Investigation of a Two-Phase<br />
Closed Thermosyphon Assembly for Passive<br />
Containment Cooling System<br />
Kyung Ho Nam and Sang Nyung Kim | Page 413<br />
After the Fukushima accident, increasing interest<br />
has been raised in passive safety systems that<br />
maintain the integrity of the containment building.<br />
To improve the reliability and safety of nuclear<br />
power plants, long-term passive cooling concepts<br />
have been developed for advanced reactors.<br />
In a previous study, the proposed design was<br />
based on an ordinary cylindrical Two-Phase<br />
Closed Thermosyphon (TPCT). The exact assembly<br />
size and number of TPCTs should be elaborated<br />
upon through accurate calculations based on<br />
experiments. While the ultimate goal is to propose<br />
an effective MPHP design for the PCCS and experimentally<br />
verify its performance, a TPCT assembly<br />
that was manufactured based on the conceptual<br />
design in this paper was tested.<br />
Displacement of Cryomodule in CADS<br />
Injector II<br />
Yuan Jiandong, Zhang Bin, Wang Fengfeng,<br />
Wan Yuqin, Sun Guozhen, Yao Junjie,<br />
Zhang Juihui and He Yuan | Page 418<br />
As Cryomodule can easily reduce higher power<br />
consumption and length of an accelerator and the<br />
accelerator can be operated more continuously.<br />
The Chinese academy of sciences institute of<br />
modern physics is developing an accelerator driven<br />
subcritical system (CADS) Injector II. Cryomodules<br />
are extremely complex systems, and their design<br />
optimization is strongly dependent on the<br />
accelerator application for which they are intended.<br />
Clean Energy Proposals are Chance<br />
for Nuclear to Have Rightful Place at Policy<br />
Table<br />
John Shepherd | Page 430<br />
Foratom, the Brussels based trade association<br />
for the nuclear industry in Europe, published a<br />
position paper on the European Commission’s<br />
‘Clean Energy for All Europeans’ package of EU<br />
legislative proposals. The proposals seek to improve<br />
the functioning of the energy market and ensure all<br />
energy technologies can compete on a level-playing<br />
field without jeopardising climate and energy<br />
targets. If Europe seeks to have a coherent and<br />
inclusive energy policy, which encompasses all<br />
lowcarbon contributors, nuclear must be allowed a<br />
place at the policy table.<br />
Abstracts | English
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
Fortschritte in Polen auf dem Weg<br />
zur Nutzung der Kernenergie<br />
NucNet | Seite 374<br />
Zu den Bauplänen Polens für erste Kernkraftwerke<br />
gibt es widersprüchliche Meldungen. Professor<br />
Grzegorz Wrochna vom Polnischen Nationalen<br />
Zentrum für Kernenergieforschung sieht das Programm<br />
auf dem richtigen Weg und erwartet zeitnah<br />
die Vorlage eines geeigneten Geschäftsmodells. Das<br />
Programm der vorherigen polnischen Regierung<br />
sah 6 GW Leistung an zwei Standorten vor. Kürzlich<br />
stellte die jetzige Regierung eine Strategie für die<br />
Entwicklung vor und betonte, dass das Nuklearprogramm<br />
beschleunigt werden muss. Konkrete<br />
Ausbauzahlen, an die die Industrie gebunden wird,<br />
wurden nicht genannt. Der voraussichtliche Bedarf<br />
an Kapazitäten liegt zwischen 4 und 12 GW. Markt<br />
und Finanzierungsmöglichkeiten sollen diesen<br />
bestimmen. Die größten Risiken liegen nicht in<br />
einer Abkehr vom Programm oder der öffentlichen<br />
Meinung, sondern von Verzögerungen.<br />
48. Annual Meeting on Nuclear Technology<br />
(AMNT <strong>2017</strong>): Eröffnungsansprache<br />
Ralf Güldner | Seite 378<br />
In der deutschen Kernenergiepolitik waren die<br />
vergangenen zwölf Monate vor allem geprägt<br />
von gesetzgeberischen Arbeiten, die nach dem<br />
beschleunigten Ausstieg aus der Kernenergie 2011<br />
anstanden. Das gilt insbesondere für die<br />
Neu ordnung der Finanzierung in der nuklearen<br />
Ent sorgung. Das andere große politische Arbeitspaket<br />
war die Novelle des Standortauswahl gesetzes.<br />
Unsere eigentliche Herausforderung ist aber die<br />
kerntechnische Kompetenz. Dies gilt für die<br />
Forschung, die Industrie, aber vor allem auch für<br />
den Staat selbst. Die Frage der Kompetenz betrifft<br />
das gesamte Spektrum des wissenschaftlichen und<br />
technischen Wissens um die Kerntechnik: die<br />
grundlegende Kernforschung, die Reaktorsicherheitsforschung,<br />
die Radiochemie, den Strahlenschutz,<br />
kerntechnische Anwendungen in Medizin,<br />
Industrie und Landwirtschaft etc.<br />
Monte-Carlo basierter Vergleich der<br />
Personendosis in Szenarien zur Einlagerung<br />
von Behältern mit bestrahltem Kernbrennstoff<br />
in generischen Tiefenlagern<br />
Héctor Saurí Suárez, Bo Pang,<br />
Frank Becker and Volker Metz | Seite 384<br />
In der Betriebsphase eines Tiefenlagers für hochradioaktive<br />
Abfälle wird das Strahlenfeld um einen<br />
Lagerbehälter durch die Rückstreustrahlung von den<br />
Wänden der Einlagerungsstrecken verändert. Daher<br />
ist für einen Vergleich der Einlagerung von<br />
abgebranntem Kernbrennstoff in verschiedenen<br />
Wirtsgesteinen von Interesse, den Einfluss der unterschiedlichen<br />
Wandungsmaterialien auf die Strahlenexposition<br />
der dort Beschäftigten zu ermitteln. In<br />
dieser generischen Studie wurde die individuelle<br />
Dosimetrie von Beschäftigten bei Einlagerung von<br />
Behältern mit abgebranntem Kernbrennstoff in<br />
Steinsalz und einer Tonformation untersucht.<br />
Die 15. AtG-Novelle zur Umsetzung<br />
der EURATOM-Sicherheits-Richtlinie<br />
Christian Müller-Dehn | Seite 391<br />
Die 15. AtG-Novelle (AtG: Atomgesetz) hat das<br />
parlamentarische Gesetzgebungsverfahren mit dem<br />
Beschluss des Bundestages 30.3.<strong>2017</strong> nunmehr<br />
durchlaufen, harrt aber noch der Veröffentlichung<br />
im Bundesgesetzblatt. Hintergrund aller Regelungen<br />
sind die Ergänzungen der EURATOM- Sicherheits-<br />
Richtlinie, die der Europäische Rat im Juli 2014<br />
beschlossen hat und die bis spätestens August <strong>2017</strong><br />
in den nationalen Regelungen der EURATOM-<br />
Mitgliedsstaaten zu verankern sind. Da die meisten<br />
dieser Ergänzungen jedoch bereits geltender<br />
Standard im deutschen Atomrecht waren, waren die<br />
für Deutschland umsetzungsbedürftigen Regelinhalte<br />
gering. Dies wird ausdrücklich auch in der<br />
Gesetzesbegründung festgehalten.<br />
Nachrüstung eines Pool-Spraysystems eines<br />
Brenn elementlagerbeckens als alternative<br />
Strategie der Kühlung für auslegungsüberschreitende<br />
Ereignisse<br />
Christoph Hartmann und Zoran Vujic | Seite 392<br />
Aufgrund von Anforderungen an Kernkraftwerke,<br />
die über Auslegungsstörfälle hinausgehen, einschließlich<br />
derer, wie sie im Jahr 2011 im Kernkraftwerk<br />
Fukushima Daiichi in Japan auftraten, ist eine<br />
alternative Kühlungsmethode von Brennelementlagerbecken<br />
erforderlich. Diese alternative Kühlung<br />
kann durch ein nachgerüstetes abgebranntes<br />
Lagerbecken-Sprühsystem nach dem AP1000-<br />
Anlagendesign bereitgestellt werden. Im Rahmen<br />
des Sicherheits-Upgrade-Programms des Kernkraftwerks<br />
Krško entschied sich der Betreiber für die<br />
Nachrüstung mit dem Westinghouse-System des<br />
AP1000®.<br />
Cybersecurity in Kernkraftwerken und<br />
ihre Anwendung in weiteren industriellen<br />
Infrastrukturen<br />
Sébastien Champigny, Deeksha Gupta,<br />
Venesa Watson and Karl Waedt | Page 396<br />
Stromerzeugung ist verstärkt auf dezentralisierte<br />
und vernetzte Rechensysteme angewiesen. Begriffe<br />
wie „Industrial Internet of Things“ des Industrial<br />
Internet Consortium (IIC) und „Industrie 4.0“<br />
bahnen sich heute ihren Weg auch in diese<br />
bedeutende Industriebranche. Die Risiken einer<br />
gezielten Ausnutzung von Fehlern und Schwachstellen<br />
nehmen mit der Komplexität, mit dem<br />
Vernetzungsgrad und mit der Dezentralisierung zu.<br />
Die inhärent strengen Sicherheitsanforderungen<br />
der Kernenergiebranche und die langjährige<br />
Berücksichtigung von Anforderungen im Bereich<br />
Cybersecurity in der Entwicklung von Produkten<br />
und in projektbegleitenden Maßnahmen machen<br />
sie zum Gold-Standard der Risikovorbeugung.<br />
Die gewonnenen Erkenntnisse können für die<br />
Ableitung angepasster Sicherheitsvorkehrungen<br />
anderer Branchen dienen. Aus diesem Blickwinkel<br />
heraus wird das Thema Cybersecurity betrachtet.<br />
Der Artikel zeigt gängige Regularien und die Vorgehensweisen<br />
zum Schutz vor Cyberangriffen in<br />
Kernkraftwerken, sowie auch deren zahlreichen<br />
Übertragungsmöglichkeiten auf andere kritische<br />
Infrastrukturen, um sie gegen Cyberangriffe und<br />
Industriespionage zu wappnen.<br />
Validierung der Nuklid-Zusammensetzung<br />
von abgebranntem Kernbrennstoff unter<br />
Verwendung von Prozentsatzdifferenzen<br />
und einer detaillierten Analyse<br />
Mann Cheol Kim | Seite 402<br />
Die Informationen zur Nuklid-Zusammensetzung<br />
von abgebranntem Kernbrennstoff sind in vielen<br />
Anwendungen wichtig. In der Reaktorphysik<br />
erfordert die Konstruktion des Kernreaktors Informationen<br />
zur Nuklidzusammensetzung. Für die<br />
Analyse der radiologischen Folgen eines schweren<br />
Unfalls ist die Nuklidzusammensetzung eine<br />
wichtige Eingangsgröße. Ebenso ist sie eine<br />
Einganggröße und Grundlage für die Auslegung<br />
eines geologischen Endlagers. Ein weiteres Feld<br />
ist die Identifizierung der Herkunft von nicht<br />
identifizierten verbrauchten Kernbrennstoffen oder<br />
radioaktiven Stoffen.<br />
Zuverlässigkeitsanalyse für die passive<br />
Restwärmeabfuhr eines AP1000-Reaktors<br />
basierend auf einem „Grey model“<br />
Qi Shi, Zhou Tao, Muhammad Ali Shahzad,<br />
Li Yu und Jiang Guangming | Seite 408<br />
Passive Systeme stützen sich auf Naturgesetze<br />
der Physik, wie z.B. Schwerkraft, Wärmeleitung<br />
oder Trägheit. Beim Generation-III-Reaktor<br />
AP1000, einem Generation-III-Reaktor, bieten<br />
solche Systeme, verbunden mit einfachen Designstrukturen,<br />
inhärente Sicherheitsfunktionen. Unter<br />
Betriebsbedingungen werden für passive Sicherheitssysteme<br />
Unsicherheiten angegeben. In einigen<br />
wenigen Fällen kann eine geringe Abweichung von<br />
Konstruktions- oder Betriebsbedingungen die<br />
Funktion des Systems beeinträchtigen. Die Zuverlässigkeit<br />
eines solchen System wird analysiert.<br />
Experimentelle Untersuchung einer<br />
zweiphasigen geschlossenen Thermosyphon-<br />
Baugruppe für ein passives<br />
Containment-Kühlsystem<br />
Kyung Ho Nam und Sang Nyung Kim | Seite 413<br />
Nach dem Unfall von Fukushima stieg da Interesse<br />
an passiven Sicherheitssystemen, die die Integrität<br />
des Containments aufrechterhalten. Zur weiteren<br />
Erhöhung von Zuverlässigkeit und Sicherheit von<br />
Kernkraftwerken, wurden passive Kühlkonzepte für<br />
die Langfristkühlung fortgeschrittener Reaktoren<br />
entwickelt. In einer früheren Studie basierte ein<br />
vorgeschlagenes Design auf einem einfachen<br />
zylindrischen zweiphasigen geschlossenen Thermosyphon<br />
(TPCT). Baugröße und Anzahl der erforderlichen.<br />
Eine neue TPCT-Baugruppe wurde getestet,<br />
die auf der Grundlage des hier vorgestellten Designs<br />
entwickelt wurde.<br />
Auslegung eines Kryomoduls für den CADS<br />
Injektor II<br />
Yuan Jiandong, Zhang Bin, Wang Fengfeng,<br />
Wan Yuqin, Sun Guozhen, Yao Junjie,<br />
Zhang Juihui und He Yuan | Seite 418<br />
Durch den Einsatz von Cryomodule können die<br />
Leistungsaufnahme von Beschleunigern und ihre<br />
Länge reduziert sowie ein kontinuierlicher Betrieb<br />
unterstützt werden. An der Chinesischen Akademie<br />
der Wissenschaften, Institut für moderne Physik<br />
wird ein beschleunigergetriebenes subkritisches<br />
System (CADS) Injektor II entwickelt, bei dem<br />
Cryomodule zum Einsatz kommen. Cryomodule<br />
sind äußerst komplexe Systeme und ihre Designoptimierung<br />
hängt stark von der Beschleunigeranwendung<br />
ab, für die sie bestimmt sind.<br />
Vorschläge für „Saubere Energie“<br />
sind eine Chance für die Kernenergie<br />
John Shepherd | Seite 430<br />
Foratom, der Brüsseler Verband der kerntechnischen<br />
Industrie in Europa, hat ein Positionspapier<br />
zur EU-Gesetzgebung „Clean Energy for All<br />
Europea ns“ veröffentlicht. Die Vorschläge zielen<br />
darauf ab, das Funktionieren des Energiemarktes zu<br />
verbessern und sicherzustellen, dass alle Energietechnologien<br />
gleichermaßen verwendet werden<br />
können, ohne dass die Klima- und Energieziele<br />
gefährdet werden. Wenn Europa eine kohärente<br />
und integrative Energiepolitik anstrebt, die alle<br />
kohlenstoffarmen Optionen umfasst, muss die<br />
Kernenergie mit Berücksichtigung finden.<br />
373<br />
ABSTRACTS | GERMAN<br />
Abstracts | German
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
374<br />
INSIDE NUCLEAR WITH NUCNET<br />
Q&A: Poland’s Progress on the Road<br />
to New Nuclear<br />
NucNet<br />
Conflicting reports have emerged from Poland about plans for its first reactors, but Professor Grzegorz<br />
Wrochna of the Polish National Centre for Nuclear Research says the programme is on track and a business<br />
model is expected soon.<br />
NucNet: There have been various media reports<br />
in Poland about the country’s nuclear new-build<br />
project, with some saying it has been postponed.<br />
Could you tell us more about the current project status?<br />
Grzegorz Wrochna: If you depend on the media you will<br />
get a confused picture. The situation is rather straightforward.<br />
There is a delay. PGE EJ1, the company responsible<br />
for building the first nuclear plant in Poland, announced<br />
that a tender would be started in December 2015, but this<br />
has not happened. Based on this delay, some media has<br />
speculated that the programme is on hold, but that is not<br />
true. It is just the tendering procedure which has been<br />
suspended while all the other work – site surveys, preparation<br />
of the nuclear regulatory body, changing the nuclear<br />
law – is all going ahead and going well.<br />
The programme prepared by the previous government [in<br />
office from November 2011 until October 2015] is still valid.<br />
The cabinet accepted this programme, but asked the ministry<br />
of energy to present a new schedule and business model<br />
by spring <strong>2017</strong>. So I hope soon we will have a plan ready to<br />
be shown to the government by the minister of energy.<br />
NucNet: Have any of the conditions for the nuclear<br />
programme changed?<br />
Grzegorz Wrochna: The global economic situation has<br />
changed. When the previous government prepared the<br />
nuclear programme, it was difficult to get financing for this<br />
kind of investment. Therefore, the condition was that the<br />
tender should concern all elements, including reactor<br />
design, construction, the first few years of operation, fuel<br />
and, finally, financing, which was the most important part.<br />
The organisations pitching for the contract would be asked<br />
to present everything, including the financing.<br />
Now conditions are different. The cost of borrowing<br />
money has decreased and it is easier to find loans at low<br />
interest rates. The new government decided to split the<br />
tender into a technical part and a financial part, each to be<br />
considered separately. The detailed model has not been<br />
decided, but this, most probably, will be the new direction.<br />
NucNet: What about the schedule?<br />
Grzegorz Wrochna: The original plan assumed the project<br />
would take 10 years from the investment decision to the actual<br />
operation of the first reactor. This was based on International<br />
Atomic Energy Agency (IAEA) documents, which were<br />
in turn based on the experience of other nuclear countries.<br />
Many countries have managed to build nuclear power<br />
units in 10 years. In Poland, it turned out to be impossible<br />
under existing Polish laws, which did not allow many of<br />
the regulatory processes to run in parallel to each other. In<br />
other words, to get each consecutive decision, we first<br />
needed to get feedback from authorities on previous<br />
decisions. When PGE EJ1 and the NCBJ (Polish National<br />
Centre for Nuclear Research) recalculated the schedule, it<br />
turned out that Poland would need 16 years to have its<br />
nuclear programme operational, six years longer than<br />
originally anticipated. The media took this as a delay, but<br />
rather it was just a ‘procedural discovery’.<br />
We can now aim for commercial operation some time<br />
around 2028, but have to wait for the ministry of energy to<br />
officially present its schedule.<br />
NucNet: What is the government’s vision of the country’s<br />
energy mix? What roles do coal and nuclear play?<br />
Grzegorz Wrochna: In the not so distant past Poland was<br />
100 % independent concerning its sources of electric supply,<br />
but this was based almost entirely on coal. More than 90 %<br />
of electricity was produced from coal. This has changed a<br />
little, with the increased but still limited participation of<br />
renewables and gas. Coal will continue to dominate the<br />
energy landscape for many years, because it is our domestic<br />
resource, essential for our security of supply.<br />
But this is not enough, because we hope the Polish<br />
economy will grow along with the demand for energy. And<br />
we will not have any means other than nuclear and energy<br />
imports to meet this growing demand. If we want to<br />
maintain our energy independence, nuclear remains the<br />
only viable option. This does not mean there is competition<br />
between nuclear and coal. We do not have to choose<br />
between the two. We still need to build new coal-fired<br />
plants to replace old, inefficient ones. But the investment<br />
timeframe for coal-fired plants is a few years, while for<br />
nuclear it will be more than 10 years. Even if the government<br />
decides overnight to go for 100v% nuclear, nothing<br />
will change for coal for a few decades.<br />
NucNet: Is the introduction of nuclear energy a politicised<br />
issue in Poland or is there a sense of consensus among<br />
different parties and stakeholders?<br />
Grzegorz Wrochna: There is no consensus between the<br />
political parties. But there is a consensual understanding<br />
that our energy mix is too dependent on domestic resources.<br />
We do not have much wind or solar potential in Poland,<br />
hydro is being used but cannot be expanded much further,<br />
and we have some domestic gas, but it is far from sufficient<br />
to meet our needs. The only sufficient domestic resources<br />
are coal and then nuclear. We have no other choice.<br />
NucNet: How does the Polish public see the new-build<br />
programme? Are there concerns about safety?<br />
Grzegorz Wrochna: Public opinion is reasonably positive<br />
about nuclear. The most recent polls showed more than<br />
60 % in favour of nuclear in Poland and, surprisingly even<br />
for us, about 48 % said they would have a reactor close to<br />
their homes. People see this as an opportunity for economic<br />
prosperity.<br />
I think we, the scientists, have done a good job telling<br />
the public about nuclear energy. The way we communicated<br />
what happened in Fukushima was very important.<br />
People are now well aware that we do not have tsunamis or<br />
earthquakes of these magnitudes in Poland.<br />
NucNet: Poland has pledged to build four to five units with<br />
combined output of 6 GW, by the mid-2030s. Is this realistic?<br />
Grzegorz Wrochna: The previous government’s programme<br />
calls for 6 GW in two locations. The number of reactors<br />
per site would have depended on the technology choice.<br />
The government recently published a strategy for<br />
responsible development which calls for the nuclear<br />
programme to be sped up. However, no capacity figures<br />
were included. The industry should not be bound by a rigid<br />
number. In time, maybe we will speak of 4 GW or 12 GW,<br />
but it will depend on market needs and financial possibilities.<br />
The first reactor will be the most challenging. I believe<br />
Inside Nuclear with NucNet<br />
Q&A: Poland’s Progress on the Road to New Nuclear ı NucNet
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
it is possible to complete this first unit by 2027-2029 and<br />
then we could go for a total of 6 GW by the early-2030s.<br />
NucNet: The Polish Nuclear Roadmap includes a plan to<br />
deploy a high-temperature gas-cooled reactor (HTR). Can<br />
you elaborate on these plans.<br />
Grzegorz Wrochna: The Polish nuclear programme is in<br />
nature a light-water reactor (LWR) investment project.<br />
The Polish industry will be part of the supply chain, but not<br />
much will be gained in terms of intellectual property and<br />
technological know-how. Fundamentally, we will order<br />
existing reactor designs, pay for them and build them.<br />
But once we have spent so much money on building a<br />
nuclear plant it might be better to spend a little bit extra<br />
and make even greater gains for the economy. We could<br />
invest in R&D, which would have lasting benefits for us.<br />
Poland has an extensive chemical industry, which<br />
consumes a lot of heat, produced from coal or imported<br />
natural gas. If we want to become more independent, we<br />
need an alternative source of heat for industry. And it is<br />
here that high temperature reactor HTR nuclear technology<br />
could play a big part.<br />
HTRs produce high-temperature steam at about 550 °C.<br />
We could safely and easily replace an old gas or coal-fired<br />
boiler at a chemical plant with an HTR which would produce<br />
the same amount of heat. We are talking about 6 GW, but<br />
this time in heat rather than in electricity, distributed among<br />
10 or more sites. This is a parallel programme, but there are<br />
obvious synergies between the two – supply chain, regulation,<br />
and the scientific part. We would really like to see the<br />
HTR programme as a spin-off from the main LWR programme.<br />
What we plan is to build about 10 to 20 HTRs in<br />
Poland by 2050. We have the capacity for this. The first<br />
should be in operation by 2031-2032. The need of Europe<br />
we estimate about 100-200 of such reactors.<br />
The HTR programme is also mentioned in government<br />
policy. Last year the ministry of energy established a<br />
committee for HTR deployment. That committee is<br />
preparing an intermediate report and this year we are<br />
planning to establish a company to start designing an HTR,<br />
based on international experience. Preparation for the first<br />
demonstrator will be supported by the Gemini+ initiative,<br />
which is being funded by Euratom. Within the framework<br />
of the € 4 million project, NCBJ scientists will be coordinating<br />
international preliminary works aimed at<br />
implementing HTRs. This could eventually help the first<br />
European HTR become a reality in Poland.<br />
NucNet: Finally, what are the challenges and risks for the<br />
new-build programme?<br />
Grzegorz Wrochna: The biggest risks do not come from<br />
cancellation or public opinion. They come from delays. In<br />
Europe, all major investments, power stations and other<br />
infrastructure, experience cost overruns and take longer<br />
than expected. In the past, the designs were several<br />
thousand pages long and the investment agreements a few<br />
pages. Today it is the opposite – designs are general and<br />
often standardised, while investment agreements have<br />
become long and cumbersome. Nuclear is no exception.<br />
This is a malaise that has affected all major investments<br />
in Europe. I hope the time spent preparing the nuclear<br />
programme in Poland will help avoid delays.<br />
Author<br />
NucNet<br />
The Independent Global Nuclear News Agency<br />
Editor responsible for this story: Kamen Kraev<br />
Avenue des Arts 56<br />
1000 Brussels, Belgium<br />
www.nucnet.org<br />
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
376<br />
CALENDAR<br />
Calendar<br />
<strong>2017</strong><br />
04.<strong>06</strong>.-07.<strong>06</strong>.<strong>2017</strong><br />
37 th Annual Canadian Nuclear Society Conference.<br />
Niagara Falls, ON, Canada, www.cns-snc.ca<br />
<strong>06</strong>.<strong>06</strong>.-09.<strong>06</strong>.<strong>2017</strong><br />
International Conference on Topical Issues in<br />
Nuclear Installation Safety: Safety Demonstration<br />
of Advanced Water Cooled Nuclear Power Plants.<br />
Vienna, Austria. International Atomic Energy Agency<br />
(IAEA), www.iaea.org<br />
11.<strong>06</strong>.-17.<strong>06</strong>.<strong>2017</strong><br />
ENYGF <strong>2017</strong> – European Nuclear Young<br />
Generation Forum. Manchester, United Kingdom,<br />
ENS YGN, www.enygf.org<br />
11.<strong>06</strong>.-15.<strong>06</strong>.<strong>2017</strong><br />
ANS Annual Meeting. 10 th International Topical<br />
Meeting on Nuclear Plant Instrumentation,<br />
Control and Human Machine Interface Technology<br />
(embedded topical meeting). San Francisco, CA,<br />
USA, American Nuclear Society (ANS), www.ans.org<br />
13.<strong>06</strong>.-14.<strong>06</strong>.<strong>2017</strong><br />
Journees thematiques fusion — Journees<br />
thematiques fusion AFF CCS. Cadarche, France,<br />
Commission Cryogenie et Supraconductive (AFF),<br />
affccs.grenoble.fr<br />
19.<strong>06</strong>.-21.<strong>06</strong>.<strong>2017</strong><br />
ATOMEXPO <strong>2017</strong>. Moscow, Russia, <strong>2017</strong>.atomexpo.ru<br />
19.<strong>06</strong>.-20.<strong>06</strong>.<strong>2017</strong><br />
EURELECTRIC Annual Convention &<br />
Conference <strong>2017</strong>. Lisbon, Portugal, Eurelectric,<br />
www.eurelectric.org<br />
26.<strong>06</strong>.-30.<strong>06</strong>.<strong>2017</strong><br />
Third PETRUS-ANNETTE PhD and Early-Stage<br />
Researchers Conference <strong>2017</strong> – Radioactive<br />
Waste Management and Disposal. Lisboa,<br />
Portugal, Petrus and Annette (Euratom),<br />
petrus-annette-<strong>2017</strong>.strikingly.com/<br />
27.<strong>06</strong>.-28.<strong>06</strong>.<strong>2017</strong><br />
New Nuclear Build – NNB <strong>2017</strong>. London, UK,<br />
Nuclear Industry Association, www.niauk.org<br />
27.<strong>06</strong>.-29.<strong>06</strong>.<strong>2017</strong><br />
Power-Gen Europe <strong>2017</strong>. Cologne, Germany,<br />
PennWell, www.powergeneurope.com<br />
26.<strong>06</strong>.-30.<strong>06</strong>.<strong>2017</strong><br />
International Conference on Fast Reactors and<br />
Related Fuel Cycles. Yekaterinburg, Russia, International<br />
Atomic Energy Agency (IAEA), www.iaea.org<br />
27.<strong>06</strong>.-04.08.<strong>2017</strong><br />
World Nuclear University Summer Institute.<br />
Uppsala, Sweden, World Nuclear Association,<br />
www.world-nuclear.org<br />
31.07.-04.08.<strong>2017</strong><br />
AccApp'17 – 13 th International Topical Meeting<br />
on Nuclear Applications of Accelerators. Quebec<br />
City, Quebec, Canada, American Nuclear Society<br />
(ANS), www.ans.org, ccapp17.org<br />
<strong>06</strong>.08.-09.08.<strong>2017</strong><br />
Utility Working Conference and Vendor<br />
Technology Expo – The Nuclear Option – Clean,<br />
Safe, Reliable & Affordable. Amelias Island, FL,<br />
USA, American Nuclear Society (ANS), uwc.ans.org<br />
20.08.-25.08.<strong>2017</strong><br />
24 th International Conference on Structural<br />
Mechanics in Reactor Technology. Busan, Korea,<br />
SMIRT Organisation Committee, www.smirt24.org<br />
23.08.-01.09.<strong>2017</strong><br />
Frédéric Joliot/Otto Hahn (FJOH) Summer School<br />
FJOH-<strong>2017</strong> – Uncertainties in nuclear reactor<br />
systems analysis: Improving understanding,<br />
confidence and quantification. Karlsruhe, Germany,<br />
Nuclear Energy Division of Commissariat à l’énergie<br />
atomique et aux énergies alternatives (CEA) and Karlsruher<br />
Institut für Technologie (KIT), www.fjohss.eu<br />
27.08.-02.09.<strong>2017</strong><br />
INCC – 5 th International Nuclear Chemistry<br />
Congress. Gothenburg, Sweden. Chalmers<br />
University of Technology Division of Nuclear<br />
Chemistry (Organisation), www.chalmers.se<br />
03.09.-08.09.<strong>2017</strong><br />
NURETH 17 – 17 th International Topical Meeting<br />
on Nuclear Reactor Thermal Hydraulics. Xi’an,<br />
China, nureth17.com<br />
03.09.-<strong>06</strong>.09.<strong>2017</strong><br />
15 th IAEE European Conference Heading Towards<br />
Sustainability Energy Systems: by Evolution<br />
or Revolution? Vienna, Austria, AAEE/IAEE,<br />
www.iaee.org<br />
10.09.-14.09.<strong>2017</strong><br />
<strong>2017</strong> Water Reactor Fuel Performance Meeting.<br />
Jeju Island, Korea, Korean Nuclear Society, the<br />
Atomic Energy Society of Japan, the Chinese Nuclear<br />
Society, the American Nuclear Society and the<br />
European Nuclear Society, wrfpm<strong>2017</strong>.org<br />
10.09.-15.09.<strong>2017</strong><br />
<strong>2017</strong> Nuclear Criticality Safety Division Topical.<br />
Carlsbad, New Mexico, USA. American Nuclear<br />
Society (ANS), www.ans.org<br />
11.09.-14.09.<strong>2017</strong><br />
Nuclear Energy in New Europe – NENE <strong>2017</strong>.<br />
Bled, Slovenia, Nuclear Society of Slovenia,<br />
www.nss.si/nene<strong>2017</strong><br />
13.09.-14.09.<strong>2017</strong><br />
VGB CONGRESS <strong>2017</strong> – Generation in Competition.<br />
Essen, Germany, VGB PowerTech e.V., www.vgb.org<br />
13.09.-15.09.<strong>2017</strong><br />
World Nuclear Association Symposium <strong>2017</strong>.<br />
London, United Kingdom, World Nuclear Association<br />
(WNA), www.world-nuclear.org<br />
17.09.-20.09.<strong>2017</strong><br />
2 nd International CNS Conference on Fire Safety<br />
and Emergency Preparedness in the Nuclear<br />
Industry. Toronto, ON, Canada, Canadian Nuclear<br />
Society (CNS), www.cns-snc.ca<br />
18.09.-22.09.<strong>2017</strong><br />
61 st IAEA General Conference. Vienna, Austria,<br />
Inter national Atomic Energy Agency (IAEA),<br />
www.iaea.org<br />
24.09.-28.09.<strong>2017</strong><br />
PSA <strong>2017</strong> – <strong>2017</strong> International Topical Meeting<br />
on Probabilistic Safety Assessment and Analysis.<br />
Pittsburgh, Pennsylvania, USA, American Nuclear<br />
Society (ANS), www.ans.org, psa.ans.org<br />
01.10.-04.10.<strong>2017</strong><br />
11 th International Conference on CANDU<br />
Maintenance and Nuclear Component. Toronto,<br />
ON, Canada, Canadian Nuclear Society (CNS),<br />
www.cns-snc.ca<br />
01.10.-04.10.<strong>2017</strong><br />
SIEN <strong>2017</strong> – International Symposium for Nuclear<br />
Energy. Bucharest, Romania, www.sien.ro<br />
04.10.-05.10.<strong>2017</strong><br />
Fire Safety in Nuclear Power Plants. Bruges,<br />
Belgium, Bel V, Gesellschaft für Anlagen- und<br />
Reaktor sicherheit (GRS) gGmbH, Bundesamt<br />
für kerntechnische Entsorgungssicherheit (BfE),<br />
www.belv.be, www.grs.de<br />
10.10.-12.10.<strong>2017</strong><br />
4 th International Symposium on the System of<br />
Radiological Protection. Paris, France, IRSN,<br />
icrp-erpw<strong>2017</strong>.com<br />
17.10.-20.10.<strong>2017</strong><br />
27 th Atomic Energy Research (AER) Symposium.<br />
Munich, Germany, Contact: Gesellschaft für<br />
Anlagen- und Reaktorsicherheit (GRS) gGmbH,<br />
www.grs.de<br />
17.10.-18.10.<strong>2017</strong><br />
49. Kraftwerkstechnisches Kolloquium. Dresden,<br />
Germany, Technische Universität Dresden,<br />
www.kraftwerkskolloqium.de<br />
21.10.-28.10.<strong>2017</strong><br />
IEEE Nuclear Science Symposium and Medical<br />
Imaging Conference. Atlanta, Georgia, USA, IEEE,<br />
www.nss-mic.org<br />
23.10.-28.10.<strong>2017</strong><br />
Fourth International Conference on Nuclear Power<br />
Plant Life Management. Lyon, France, International<br />
Atomic Energy Agency (IAEA), www.iaea.org<br />
07.11.-09.11.<strong>2017</strong><br />
10 th International Symposium Release of<br />
Radioactive Materials Requirements for<br />
Exemption and Clearance. Berlin, Germany, TÜV<br />
Nord Akademie, www.tuev-nord.de/tk-rrm<br />
25.10.-26.10.<strong>2017</strong><br />
Chemistry in Power Plants. Koblenz, Germany,<br />
VGB PowerTech e.V., www.vgb.org<br />
29.10.-02.11.<strong>2017</strong><br />
<strong>2017</strong> ANS Winter Meeting and Nuclear<br />
Technology Expo. Washington, DC, USA, American<br />
Nuclear Society (ANS), www.ans.org<br />
26.11.-30.11.<strong>2017</strong><br />
International Symposium on Future I&C for Nuclear<br />
Power Plants. Gyeongiu, Korea, www.isofic.org<br />
27.11.-30.11.<strong>2017</strong><br />
ICOND <strong>2017</strong> – International Conference on<br />
Nuclear Decommissioning. Aachen, Germany,<br />
Aachen Institute for Nuclear Training GmbH,<br />
www.icond.de<br />
01.12.-02.12.<strong>2017</strong><br />
ThermAc 2016 – Aquatic Actinide Chemistry and<br />
Thermodynamics at elevated Temperatures.<br />
Dresden, Germany, HZDR, www.hzdr.de<br />
05.12.-07.12.<strong>2017</strong><br />
POWER-GEN International. Las Vegas, NV, USA.<br />
PennWell, www.power-gen.com<br />
2018<br />
26.02.-01.03.2018<br />
Nuclear and Emerging Technologies for Space<br />
2018. Las Vages, NV, USA. American Nuclear<br />
Society (ANS), www.ans.org<br />
08.04.-11.04.2018<br />
International Congress on Advances in Nuclear<br />
Power Plants – ICAPP 18. Charlotte, NC, USA,<br />
American Nuclear Society (ANS), www.ans.org<br />
22.04.-26.04.2018<br />
Reactor Physics Paving the Way Towords More<br />
Efficient Systems – PHYSOR 2018. Cancun, Mexico,<br />
www.physor2018.mx<br />
29.05.-30.05.2018<br />
49 th Annual Meeting on Nuclear Technology<br />
AMNT 2018 | 49. Jahrestagung Kerntechnik.<br />
Berlin, Germany, DAtF and KTG,<br />
www.nucleartech-meeting.com – Save the Date<br />
17.09.-20.09.<strong>2017</strong><br />
FONTEVRAUD 9. Avignon, France, Société<br />
Française d’Energie Nucléaire (SFEN),<br />
www.sfen-fontevraud9.org<br />
30.09.-05.10.2018<br />
Pacific Nuclear Basin Conferences – PBNC 2018.<br />
San Francisco, CA, USA, American Nuclear<br />
Society (ANS), www.ans.org<br />
14.10.-18.10.2018<br />
12 th International Topical Meeting on Nuclear<br />
Reactor Thermal-Hydraulics, Operation and<br />
Safety – NUTHOS-12. Qingdao, China<br />
Calendar
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
Notes<br />
Press Release 16 May <strong>2017</strong><br />
AMNT: President Warns Against<br />
Loss of Nuclear Expertise<br />
In his speech at the 48 th Annual Meeting on Nuclear<br />
Technology (AMNT <strong>2017</strong>), the President of the DAtF<br />
( German Atomic Forum), Dr. Ralf Güldner, warned against<br />
the loss of nuclear expertise and of nuclear research and<br />
industry in Germany. Güldner said that the challenge for<br />
nuclear technology in Germany lay in the long-term<br />
provision of expertise. He said this applied to research,<br />
industry and the state itself and that it was premised on<br />
using this expertise, for example, in industrial projects for<br />
upgrading plants or in development. He continued that<br />
the international demand for German safety expertise,<br />
which enjoys an excellent reputation, contributed<br />
significantly to maintaining it. He warned that the decision<br />
to phase out nuclear energy must not constitute a risk of<br />
losing this expertise.<br />
In his speech, Güldner said, “Nuclear safety research<br />
forms the basis for expertise in safety issues in which<br />
Germany intends to play a long-term role and exert its<br />
influence. If we want to continue participating in the international<br />
discussion of safety standards, then continuity in<br />
safety research is absolutely essential.” He complained<br />
that, especially in the case of innovative topics, reactor<br />
safety research was now being regarded as superfluous<br />
and that many federal state governments no longer wanted<br />
anything to do with it. He said that university chairs were<br />
not being refilled and universities and research institutes<br />
were shaped as to withdraw from areas that were not<br />
assigned to waste management or dismantling.<br />
Güldner therefore suggested a new beginning for safety<br />
research, “The solution might lie in a new Centre of<br />
Expertise for Nuclear Safety where current issues could be<br />
dealt with without the burden of past conflicts. Here, it<br />
may be possible to pool capacities, to network research,<br />
state and industry and to create an attractive hub for our<br />
international collaboration.”<br />
He drew attention to the fact that nuclear energy would<br />
continue to contribute to the security of the power supply<br />
in Germany. He indicated that the political consensus on<br />
the transformation of the German energy sector would<br />
also be implemented by operating the plants until 2022.<br />
And with regard to this he stated, “There must therefore be<br />
no factually unfounded complications to operation of the<br />
nuclear power plants in the last few years.” Güldner<br />
pointed out that the facilities for uranium enrichment and<br />
fuel assembly production were explicitly excluded from<br />
the phase out of nuclear energy use and he rejected any<br />
efforts to expand the phase out.<br />
DATF EDITORIAL NOTES<br />
377<br />
New Brochure<br />
The DAtF has published the new edition of<br />
its nuclear power statistics flyer with status April <strong>2017</strong>:<br />
“Kernenergie in Zahlen <strong>2017</strong>”<br />
3 The flyer can be downloaded and ordered<br />
at kernenergie.de under the headings<br />
Downloads and Shop.<br />
For further details<br />
please contact:<br />
Nicolas Wendler<br />
DAtF<br />
Robert-Koch-Platz 4<br />
10115 Berlin<br />
Germany<br />
E-mail: presse@<br />
kernenergie.de<br />
www.kernenergie.de<br />
DAtF Notes
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
378<br />
AMNT <strong>2017</strong><br />
48 th Annual Meeting<br />
on Nuclear Technology (AMNT <strong>2017</strong>):<br />
Opening Address<br />
16 to 17 May <strong>2017</strong>, Berlin<br />
Ralf Güldner<br />
Ladies and Gentlemen, Welcome to our 48 th Annual Meeting on Nuclear Technology in Berlin on behalf of the<br />
DAtF and the German Nuclear Society. It is my pleasure to see you again in Berlin. As in other years, we offer a<br />
comprehensive program, providing insights into many aspects of nuclear technology and contributing to the<br />
international exchange of knowledge and experience in industry, research, politics and administration.<br />
Ladies and Gentlemen,<br />
The content of our Meeting is already reflected in the<br />
Plenary Session with its fixed topics relating to politics,<br />
industry, expertise, communication and waste management.<br />
In the section on politics, Steffen Kanitz, rapporteur<br />
for nuclear energy of the CDU/CSU parliamentary group in<br />
the Bundestag, will provide us with an overview of a turbulent<br />
year for German nuclear energy policy. Guido Knott,<br />
Chairman of the Board of Management of PreussenElektra<br />
GmbH, will give us an understanding of the challenges<br />
involved in operating nuclear power plants cost-effectively<br />
in Germany. We are looking forward to the panel discussion<br />
on dismantling and I would also like to draw your attention<br />
to our workshop on the preservation of skills after the lunch<br />
break today and tomorrow morning.<br />
Special thanks are due to our partners in the exhibition<br />
and for the sponsoring without which our meeting would<br />
not even be possible. The exhibition provides you with the<br />
opportunity to make personal contact with a large number<br />
of companies and organisations in our industry with the<br />
chance for a direct exchange of ideas and information.<br />
We have further increased the number of international<br />
partners involved. I want to draw your attention to the<br />
Czech pavilion and also the UK’s pavilion. Our British<br />
colleagues are facing historic decisions in their own<br />
country and in respect of the future relationship with<br />
Europe. We hope that solutions will be found during the<br />
Brexit negotiations that enable constructive cooperation in<br />
nuclear technology to continue in the future. This applies<br />
not least to the new British construction projects.<br />
Upheavals, new beginnings and the travails of<br />
everyday business<br />
The past twelve months in German nuclear energy policy<br />
have been characterised mainly by, what I would call, late<br />
legislative clearing up work which has been pending since<br />
the decision for an accelerated phase-out of nuclear power<br />
in 2011. This applies particularly to the reorganisation of<br />
financing in nuclear waste management where a whole<br />
legislative package has been used to implement a change<br />
of system in many areas. This process is not yet quite<br />
complete. The laws themselves have not yet entered into<br />
force due to being examined for conformity with EU law,<br />
and the contractual arrangement sought between the<br />
nuclear power plant operators and the government has not<br />
yet been signed.<br />
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Of course, there is criticism of the arrangements.<br />
• On the one hand there are fundamental reservations<br />
regarding limitation of liability.<br />
• For the operators, however, the high risk premium on<br />
the waste management costs represents an additional<br />
burden that is unexpected and hard to bear and<br />
which is now likely to increase yet gain in the wake of<br />
recalculations.<br />
Overall, however, the reorganisation in waste management<br />
will satisfy the conditions of the phase-out. A<br />
situation with permanent separation between responsibility<br />
for action and financing, potentially unlimited<br />
secondary liability and a ban on using nuclear energy<br />
could not have existed in the long run.<br />
selection step, the localisation of subareas on the white<br />
map, actually be completed by 2021 as is currently the<br />
aim? The division and clear definition of tasks between BfE<br />
(Federal Office for the Safety of Nuclear Waste Management)<br />
and BGE (Federal Company for Final Disposal) is<br />
another issue. It applies particularly to final repository<br />
research which now has many new tasks. It has not yet<br />
been specified who will be responsible for final disposal<br />
research in the future.<br />
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The other major political work package – and Mr Kanitz<br />
will report on this in detail very shortly – was the<br />
amendment to the Site Selection Act (StandAG). Although<br />
political agreement was reached in spring 2013 on the<br />
search for a new site for a final repository for high active<br />
waste, in many details the law was still poorly conceived<br />
and left the Final Repository Commission with a number of<br />
unanswered questions along the way. Dr. Bernhard Fischer<br />
and Professor Gerd Jäger called on our industry’s expertise<br />
while working on this constructively and with tremendous<br />
dedication. As part of the practical implementation, transfer<br />
of the DBE (German Company for the Construction and<br />
Operation of Waste Repositories) to the government was<br />
completed here in Berlin yesterday. As an industry we have<br />
contributed to describing the path for a solution in the<br />
search for a new final repository. Now it is the politicians’<br />
task to implement the set framework consistently.<br />
Despite everything that has been achieved – here<br />
we should also mention reorganisation of the regulatory<br />
and institutional structure for final disposal, the 15 th<br />
amendment to the Atomic Energy Act and the first<br />
consolidated Radiological Protection Act in German legal<br />
history – there is still work to be done.<br />
Reorganisation in waste management also includes the<br />
transfer of responsibility for interim storage to the state.<br />
This is an even bigger change to the system than that for<br />
final disposal but it is not so much in the public eye. A<br />
whole series of operational challenges will arise when the<br />
operational responsibility changes. This change was set in<br />
motion when the Federal Company for Interim Storage<br />
was set up and the aim, for the central interim storage<br />
facilities, is for it to be completed during the course of this<br />
year, for the site-based high-level interim storage facilities<br />
early in 2019 and for the LLW/ILW storage facilities a year<br />
later. The first steps have been taken and the choice of<br />
Essen as the company’s headquarters will have a positive<br />
effect, particularly on preserving the necessary skills, as a<br />
result of GNS employees transferring over. Together, the<br />
nuclear power plant operators and the GNS are handing<br />
over a well-functioning system in which high safety<br />
requirements are applicable. They are thus making an<br />
important contribution to the reorganisation of responsibility<br />
in nuclear waste management.<br />
Now we come to the practical test of implementing the<br />
Site Selection Act. When will the Federal Company for<br />
Final Disposal, as the project developer, be capable of<br />
working operationally? What time frame must we<br />
realistically assume for the whole process? Will the first<br />
Due to the concentration on high active waste, another<br />
waste management issue has faded somewhat into the<br />
background: What exactly is happening with the Konrad<br />
facility? Are the plans for completing it by 2022 still valid?<br />
When will regular operation actually start? How is the outflow<br />
from the interim storage facilities to be prioritised?<br />
These questions are important for any region throughout<br />
Germany that has an interim storage facility, a state<br />
collecting facility or a dismantling project. By bundling the<br />
interim storage facilities in a federally-owned company,<br />
I see opportunities for bringing more common sense to the<br />
discussions and accelerating the processes.<br />
In the comments of the Federal Court of Auditors for<br />
2016, there is criticism that the Federal Government has<br />
not adequately exercised supervision of the Konrad project<br />
over the years. It recommends using the reorganisation of<br />
tasks, which is welcomed by the Federal Court of Auditors,<br />
to document the current situation, to make contractual<br />
agreements with the BGE and to implement closer<br />
monitoring. These considerations sound reasonable and<br />
the ongoing restructuring provides an excellent opportunity<br />
to get such project management off the ground; it<br />
could be the culmination, so to speak, of the many reforms<br />
in this legislative period.<br />
At the same time, of course, we have the operation of<br />
the nuclear power plants which we will safely continue<br />
with and which we would also like to continue costeffectively.<br />
Last January showed yet again that, particularly<br />
during the so-called “dark doldrums”, grid operators and<br />
reserve capacities are gradually reaching their limits. The<br />
reserve capacity requirement of 10,400 MW now specified<br />
by the Federal Network Agency for the coming winter<br />
speaks for itself.<br />
There is political consensus on the phased exit from<br />
nuclear energy which we will implement by 2022 with our<br />
expertise and also by safe operation. So there must be no<br />
factually unfounded complications to the operation of the<br />
nuclear power plants in the last few years.<br />
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Ladies and Gentlemen,<br />
The major issues of the future for nuclear technology in<br />
Germany are dismantling on the one hand and nuclear<br />
expertise on the other hand. These questions affect us all<br />
and are long-term issues.<br />
Dismantling by consensus<br />
Dismantling is on the right track. The first decommissioning<br />
and dismantling licences within the scope of<br />
phasing out nuclear energy have been issued – for Isar 1,<br />
Neckarwestheim 1, Biblis and Philippsburg 1. Important<br />
preliminary work has been carried out at all the sites; they<br />
are free of fuel or work is ongoing to ensure this. It is<br />
important here that the flask and storage licences still<br />
outstanding are issued on time. However, it is only in the<br />
coming years that the considerable breadth of the projects<br />
will become apparent.<br />
In recent years, when requesting factual information<br />
about dismantling and when consulting with citizens,<br />
cooperation with the authorities has been good and the<br />
support of politicians has been helpful. It is important for<br />
speedy dismantling to maintain the consensus which now<br />
exists between state and operators and to push the projects<br />
forward efficiently on this basis. I have little sympathy here<br />
with the traditional adversaries of nuclear energy who for<br />
ideological reasons are now fighting the dismantling<br />
process as well. In Germany dismantling is being carried<br />
out in compliance with the highest safety standards and,<br />
just like the construction and operation of a nuclear power<br />
plant, it is subject to constant inspection by the authorities<br />
and their experts.<br />
Preserving and developing nuclear expertise<br />
Our real challenge though is nuclear expertise. This is<br />
important for research, for industry but above all for the<br />
state itself. Many people may simply not be aware of this.<br />
The topic of preserving and building up nuclear<br />
expertise by shifting operational responsibility to federallyowned<br />
companies will gain relevance particularly in the<br />
waste management sector. Taking into account all the<br />
authorities and public companies that operate in the waste<br />
management sector, we could soon be talking about up to<br />
4,000 employees. Together with the civil servants and<br />
government employees in other areas of nuclear technology,<br />
in expert appraisal and in research, it may be assumed<br />
that in the future at least a sixth of the more than 30,000<br />
employees in the industry will be assigned to the public<br />
sector. In the long term, this will require appropriate<br />
training of skilled staff and targeted human resources<br />
planning. It can only be successful if there are positive<br />
prospects for young people who employers would like to<br />
win over for the important work ahead. It also needs to<br />
include appropriate public discussion of the subject.<br />
The question of expertise covers the whole range of<br />
scientific and technical knowledge relating to nuclear<br />
technology: basic nuclear research, reactor safety research,<br />
radiochemistry, radiological protection, nuclear applications<br />
in medicine, industry and agriculture, to mention but<br />
a few examples.<br />
Let’s take reactor safety research which is closely linked<br />
to the operation of nuclear power plants. Reactor development<br />
in particular is now subject to the accusation of being<br />
redundant; sometimes it is regarded as outmoded or even<br />
illegitimate. Nuclear safety research, however, forms the<br />
basis for expertise in safety issues in which Germany has<br />
stated its intention to play a long-term role and exert<br />
its influence. If we want to continue participating in<br />
the international discussion of safety standards, then<br />
continuity in safety research is absolutely essential.<br />
Our nuclear expertise, however, can only develop in<br />
collaboration with scientifically attractive partners in<br />
other countries. To win them over for this purpose requires<br />
appropriate facilities and experts who are able to offer<br />
added scientific value. This applies to all topics, especially<br />
innovations and new design concepts. After all, we need to<br />
be able to knowledgeably have a say too. Consider, for<br />
example, a development in fuel assemblies, such as that<br />
which Seth Grae, CEO of the Lightbridge Corporation from<br />
the USA, will be presenting later. In the long run, scepticism<br />
about research or even a ban on research has never done<br />
any industrialised country any good.<br />
In practice, however, we see that teaching and research<br />
are being thinned out, that university chairs are not being<br />
refilled and, under political pressure or for image reasons<br />
and in a spirit of anticipatory obedience, whole institutes<br />
are withdrawing from those areas that are not assigned to<br />
waste management or dismantling.<br />
Centre of Expertise for Nuclear Safety?<br />
The question here is: what can we do? On the one hand,<br />
the Federal Government wants and needs to access the<br />
appropriate expertise and it also has the funds for this. On<br />
the other hand, many federal state governments want<br />
nothing more to do with the subject and are thus shaping<br />
the orientation of universities and research institutes. The<br />
solution might lie in a new Centre of Expertise for Nuclear<br />
Safety where current issues could be dealt with without<br />
the burden of past conflicts. Here, it may be possible to<br />
pool capacities, to network research, state and industry<br />
and to create an attractive hub for our international<br />
collaboration. A new start such as this might provide<br />
young people who want to become involved in nuclear<br />
technology with credibly fascinating tasks, good prospects,<br />
respect and appreciation. Perhaps such a project would not<br />
require the very broad general consensus but rather a<br />
viable coalition of people with insight.<br />
Nuclear energy – long-term reality in Europe<br />
Insight also includes the realisation that other countries<br />
are not following our path. Now, after many years of delay,<br />
the new construction projects of Olkiluoto and Flamanville<br />
have reached the preparations for commissioning and are<br />
no longer merely a mirage. The Hinkley Point C project has<br />
received its first partial permit. By the way, all four reactors<br />
will be constructed using instrumentation and control<br />
equipment made in Germany. In the United Kingdom, in<br />
addition to the EPR by Areva, the AP 1000 by Westinghouse<br />
has also received confirmation in the Generic Design<br />
Assessment and the ABWR by Hitachi will follow by the<br />
end of the year.<br />
Things are also happening east of Germany: a few<br />
months ago unit 1 of the Novovoronezh II nuclear power<br />
plant went online – with German instrumentation and<br />
control equipment and a planned operating period up to<br />
2077. Unit 1 of the Leningrad II nuclear power plant, which<br />
is set to replace the old Chernobyl-type plants, is in start-up<br />
commissioning. Construction of the first nuclear power<br />
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plant in Belarus is scheduled and the projects in Paks and<br />
Hanhikivi are also being pushed forward consistently. Our<br />
Czech partners also have expansion plans, not least with a<br />
view to preventing CO2. There will be no shortage of<br />
interested parties as no less than six suppliers have already<br />
expressed an interest. In Poland, the site selection process<br />
for the first nuclear power plant has entered the concrete<br />
phase within the defined area. If safety is also going to be a<br />
concern for us in the coming decades then it must be<br />
Germany’s goal to count permanently as a partner in safety<br />
with recognised expertise. However, the repetition of<br />
demands for phase out is not sufficient, what is needed in<br />
fact is a constructive attitude.<br />
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Nuclear technology – part of the location for<br />
industry and science<br />
And let’s not forget that Germany will also benefit from<br />
nuclear technology in many respects and in the long term.<br />
The research reactors in Munich, Berlin and Mainz are not<br />
only used for basic research, they also do a great deal for<br />
applied research and industrial development. They are<br />
also indispensable for direct applications in industry and<br />
medicine. Nuclear technology is also found elsewhere:<br />
such as in non-destructive material testing, plant breeding,<br />
in medical diagnosis and therapy. Nuclear technology is<br />
directly linked to our status as a country of science and<br />
technology.<br />
And let’s not forget economic value creation. Many<br />
internationally recognised nuclear technology companies<br />
are both important employers and taxpayers. This<br />
industrial value chain made up of manufacturers, suppliers<br />
and service providers also requires nuclear expertise,<br />
especially in safety engineering. Germany has a good<br />
reputation in this field and German products and services<br />
related to nuclear safety are in great demand. Obstructing<br />
export will not increase nuclear safety for Germany, for our<br />
neighbours or for the world. And vital expertise can only<br />
develop while it’s in use, e.g. in industry, and therefore in<br />
the medium term largely in exports.<br />
This also applies to companies involved in the fuel cycle<br />
in Germany which are now frequently becoming the target<br />
of political debate. These facilities are explicitly excluded<br />
from the phase out of nuclear energy use and we reject any<br />
efforts to expand the phase out. The Federal Government<br />
may well profess uranium enrichment and fuel assembly<br />
manufacturing in Germany as centres of expertise. When it<br />
comes to using the expertise of these companies for<br />
operational and waste management safety, for the<br />
subject of non-proliferation and for security-policy risk<br />
assessments, then it is not so distant. In this field too, Germany<br />
would like to have its own knowledge and it’s the<br />
same here as with reactor safety. Those who want to<br />
perfect the phase out will also perfect the loss of expertise.<br />
This cannot and must not be our aim.<br />
of our meeting. I would like to thank you all very much for<br />
your contribution to the AMNT, which in <strong>2017</strong> has once<br />
again become our industry’s most important platform for<br />
exchanging knowledge and experience in Germany.<br />
I would also like to thank all those taking part who make<br />
our AMNT so diverse and enriching.<br />
The German Atomic Forum’s traditional reception,<br />
which you are cordially invited to attend, will take place<br />
this evening from 7 pm. It will flow seamlessly into the<br />
usual social evening which we are all looking forward to.<br />
As in previous years, our exhibitors hope you will accept<br />
their invitation to join them.<br />
Ladies and Gentlemen,<br />
I wish everyone a successful meeting with lively discussions<br />
and valuable insights. And please don’t forget to enjoy your<br />
participation here and your stay in the vibrant city of<br />
Berlin.<br />
Author<br />
Dr. Ralf Güldner<br />
President of the DAtF<br />
(German Atomic Forum)<br />
Robert-Koch-Platz 4<br />
10115 Berlin, Germany<br />
Successful AMNT<br />
Ladies and Gentlemen,<br />
Maintaining and developing expertise in addition to<br />
national and international networking are ultimately the<br />
key tasks of the AMNT. In this case, the commitment and<br />
expertise of those who participate in designing the<br />
programme, who are responsible for the sessions and give<br />
presentations in their specialist fields, form the backbone<br />
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Monte-Carlo Based Comparison of<br />
the Personal Dose for Emplacement<br />
Scenarios of Spent Nuclear Fuel Casks in<br />
Generic Deep Geological Repositories<br />
Héctor Saurí Suárez, Bo Pang, Frank Becker and Volker Metz<br />
The paper “Monte<br />
Carlo Based Comparison<br />
of the Personal<br />
Dose for Emplacement<br />
Scenarios of<br />
Spent Nuclear Fuel<br />
Casks in Generic Deep<br />
Geological Repositories”<br />
by Héctor Saurí<br />
Suárez, Bo Pang,<br />
Frank Becker and<br />
Volker Metz has been<br />
awarded as Best<br />
Paper of the 48 th<br />
Annual Meeting on<br />
Nuclear Technology<br />
(AMNT <strong>2017</strong>), Berlin,<br />
16 and 17 May <strong>2017</strong>.<br />
1 Introduction When a high-level nuclear waste cask is transported to its final position in a deep geological<br />
disposal facility, the radiation exposure received by the workers in such a facility is expected to be significantly influenced<br />
by the materials of the surrounding layers. Moreover, the question arises if there is an enhanced directional dependent<br />
influence on the personal radiation exposure in such facilities since certain amount of backscattered radiation comes<br />
from the back and lateral sides. Hence, it is of interest to study the influence of the worker’s position and its orientation<br />
on the personal dose assessment.<br />
In the current study, the generalpurpose<br />
Monte-Carlo N-Particle code<br />
MCNP6 [Pelowitz et al., 2013] was<br />
employed to calculate the radiation<br />
field around POLLUX® type shielding<br />
casks [Janberg and Spilker, 1998;<br />
Filbert et al., 2011] loaded with spent<br />
nuclear fuel (SNF), which were<br />
emplaced in horizontal drifts of deep<br />
geological repositories. Furthermore,<br />
a simplified mathematical phantom<br />
was used to represent a worker inside<br />
the facility, in order to calculate<br />
the personal radiation exposure for<br />
working scenarios with MCNP6.<br />
Emplacement in two different geological<br />
disposal facilities was considered,<br />
i.e. a horizontal drift in rock salt (from<br />
now on in short as “rock salt drift” or<br />
RSD) and a horizontal drift in a clay or<br />
shale formation (from now on in short<br />
as “clay drift” or CLD). In contrast to a<br />
repository in rock salt, drifts and<br />
access galleries of a repository in soft<br />
rock, such as clay and shale, have<br />
to be reinforced with concrete lining<br />
with a thickness of several decimetres<br />
[e.g. Chen et al., 2014; Leon Vargas<br />
et al., <strong>2017</strong>]. The radiation field was<br />
calculated in terms of ambient dose<br />
equivalent for both drifts at different<br />
positions to the shielding cask, which<br />
is disposed on the ground of the drift.<br />
In order to study the dependence of<br />
the worker’s orientation towards the<br />
cask on the personal exposure,<br />
simulations with different angles<br />
between phantom and POLLUX® cask<br />
were performed in RSD. Finally, a<br />
comparison between the calculated<br />
personal dose rate during a working<br />
scenario in RSD and in CLD was<br />
conducted.<br />
2 Methodology<br />
2.1 Waste inventories considered<br />
for POLLUX® type casks<br />
Based on the average inventory of<br />
used fuel elements discharged from<br />
pressurized water reactors (PWR)<br />
in Germany [Peiffer et al. 2011], a<br />
representative waste loading of 90 %<br />
uranium oxide (UOX) fuel and 10 %<br />
mixed-oxide (MOX) fuel with a burnup<br />
of 55 GWd/t(HM) was considered.<br />
The POLLUX® self-shielding cask<br />
[ Janberg and Spilker, 1998; Filbert et<br />
al., 2011], designed for deep geological<br />
disposal in RSD, was employed for<br />
both RSD and CLD. In our model for<br />
disposal in RSD, a POLLUX® cask,<br />
loaded with fuel rods of ten PWR fuel<br />
assemblies was numerically simulated.<br />
This corresponds to a waste load of<br />
about 5.45 metric tons heavy metal<br />
(tHM). The cask with fuel rods of one<br />
MOX and nine UOX fuel assemblies is<br />
herein after referred to as POLLUX-10.<br />
Due to temperature restrictions<br />
regard ing emplacement of casks with<br />
heat-generating waste in clay and<br />
shale formations [Leon Vargas et al,<br />
<strong>2017</strong>], for the emplacement in a CLD<br />
the maximum amount of fuel assemblies<br />
per cask was set to three. Therefore,<br />
one POLLUX® type cask with an<br />
homogeneous mixture of two thirds<br />
PWR-UOX and one third PWR-MOX<br />
fuel corresponding to one MOX and<br />
two UOX fuel assemblies, comprising<br />
1.64 tHM (herein after referred to as<br />
POLLUX-3M), and two POLLUX® casks<br />
with an homogeneous mixture of<br />
PWR-UOX corresponding to three<br />
UOX fuel assemblies, comprising<br />
1.64 tHM (herein after referred to as<br />
POLLUX-3U), were employed for the<br />
emplacement in a CLD. In general, the<br />
MOX fuel rods are supposed to be<br />
placed in the center of the cask<br />
surrounded by the UOX fuel rods. This<br />
arrangement provides an additional<br />
shielding for neutrons coming from<br />
MOX fuel rods. Hence, a homogeneous<br />
mixture will give conservative results<br />
since the MOX fuel is homogeneously<br />
distributed and supposed to be less<br />
shielded. Two zones were defined in a<br />
fuel rod, i.e. an “active zone” which<br />
contains the fuel pellets and an<br />
“ inactive zone” which corresponds to<br />
the top and bottom of the fuel rod and<br />
it is mainly composed of Zircaloy<br />
cladding [Janberg and Spilker, 1998].<br />
The effective density in these zones<br />
was calculated according to the equation:<br />
where m zone is the mass of the corresponding<br />
zone and V canisterzone is the<br />
total volume available in the POLLUX®<br />
type cask for that zone.<br />
An average burnup of 55 gigawattdays<br />
per metric ton of heavy metal<br />
(GWd/tHM) was assumed for both<br />
UOX and MOX SNF. Before emplacement<br />
a cooling time of the SNF was<br />
assumed to be 50 years after discharge<br />
from the reactor core. This duration<br />
corresponds to an assumed interim<br />
storage time before disposal of SNF in<br />
a deep geological disposal facility to<br />
be built in 2050 according to BMUB<br />
[2015]. Isotope mass of the SNF in dependence<br />
of the cooling time was taken<br />
from [Peiffer et al., 2011]. The SNF<br />
inventory is composed of hundreds of<br />
different isotopes, but many of them<br />
have negligibly small activities. As<br />
investigated in a previous study [Pang<br />
et al., 2016], for the waste inventory<br />
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| | Fig. 1.<br />
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<br />
direction and the number the distance in meters to the cask surface.<br />
AMNT <strong>2017</strong><br />
considered in this study, neutrons<br />
dominate the radiation field and<br />
exposure outside the shielding cask.<br />
Hence, only those isotopes that<br />
contrib ute significantly to neutron<br />
activity were considered when<br />
defining the radiation sources for<br />
simulations with MCNP6.<br />
For the considered fuel inventory<br />
the main contributor to neutron emissions<br />
is the spontaneous fission of<br />
244 Cm (90 % of the total emission) and<br />
246 Cm (5 % of the total emission),<br />
while the contribution due to (α,n)<br />
reactions, mainly stemming from<br />
interactions with 18 O, is less than 5 %.<br />
The total neutron source strength<br />
for the POLLUX-10 inventory is<br />
1.66 · 10 +9 neutrons/sec (n/s),<br />
while those for the POLLUX-3M<br />
and POLLUX-3U inven tory are<br />
9.02 · 10 +8 n/s and 3.24 · 10 +8 n/s,<br />
respectively.<br />
2.2 Calculation of the ambient<br />
dose equivalent rate Ḣ*(10)<br />
In the generic model for a repository<br />
for heat generating waste of Stahlmann<br />
et al. [2015], an emplacement<br />
drift has a length of 57 m (RSD) and<br />
63 m (CLD), respectively. The drift is<br />
surrounded by a host rock layer of at<br />
least 100 m thickness and several<br />
decimetres of concrete lining in the<br />
case of CLD. To simplify the calculations,<br />
the thickness of the drift walls,<br />
i.e. rock salt for POLLUX-10 and<br />
concrete lining for POLLUX-3M and<br />
POLLUX-3U, was set to 1 m in the<br />
MCNP6 model, which is sufficient to<br />
account for possible interactions of<br />
the radiation with the drift wall<br />
materials. With respect to interactions<br />
of neutrons and photons with clay and<br />
concrete, both materials are characterized<br />
by similar densities and elemental<br />
/ oxidic compositions, dominated<br />
by SiO 2 , CaO, Al 2 O 3 and H 2 O.<br />
Figure 1 shows the modelling of the<br />
deep geological disposal facility and<br />
the POLLUX® type cask with MCNP6.<br />
As a simplification, only one cask<br />
(cylindrical form with a length of<br />
5.5 m and an outer diameter of<br />
1.56 m) was placed on the ground of<br />
the drift with its bottom surface at<br />
2.63 m distance to the drift end side.<br />
Detailed geometrical information of<br />
POLLUX® type cask and the generic<br />
emplacement drifts can be found in<br />
[Janberg and Spilker, 1998; Filbert et<br />
al., 2011] and [Stahlmann et al.,<br />
2015], while the respective detailed<br />
MCNP6 models were already<br />
described in [Pang et al., 2016], hence<br />
they are not shown here. Since the<br />
radiation scattered by the drift layers<br />
might have an important impact on<br />
the radiation field, a third drift was<br />
also modelled. This one has the same<br />
geometry as the ones described above<br />
but the surrounding wall layers were<br />
replaced by air, representing a cask<br />
free in air (from now on in short as<br />
FIA).<br />
As denoted by black dots in Fig. 1,<br />
twelve MCNP6 point detector F5 tallies<br />
[Pelowitz et al., 2013] were employed<br />
to calculate the neutron fluence rate<br />
and the ambient dose equivalent rate<br />
Ḣ*(10) at different positions inside the<br />
drift. Tallies X1, Y1 and Z1 (see Fig. 1)<br />
were defined to compare Ḣ*(10) at 1 m<br />
distance to the cask surface in the<br />
respective directions. To study the<br />
change of Ḣ*(10) with the distance to<br />
the cask, the tallies X1 to X10 (see<br />
Fig. 1) were also employed. The neutron<br />
fluence- to-ambient-dose-equivalent<br />
conversion coefficients given by<br />
ICRP [1996] were employed to convert<br />
the F5 tally results into Ḣ*(10). To<br />
assess the precision of the result,<br />
MCNP6 produces a wealth of information<br />
about a simulation, which is<br />
represented by ten statistical checks<br />
[see Pelowitz et al. (2013)]. To pass the<br />
ten statistical checks, 2 · 10 +7 particles<br />
were required per simulation.<br />
2.3 Calculation of the personal<br />
dose equivalent rate Ḣ p (d)<br />
To obtain the personal dose equivalent<br />
rate Ḣ p (d), a worker inside the<br />
drift was represented in this study<br />
with a simplified anthropomorphic<br />
phantom. This phantom is a virtual<br />
representation of the BOMAB (BOttle<br />
MAnnikin ABsorber) phantom, which<br />
models the head, neck, chest,<br />
abdomen, thighs, calves, and arms<br />
with cylinders or elliptical cylinders.<br />
A detailed description of its components<br />
can be found in [U.S.<br />
Department of Energy, 2016]. Figure 2<br />
shows the MCNP6 model of the<br />
phantom used in the current study.<br />
| | Fig. 2.<br />
MCNP6 representation of the BOMAB phantom with a cylindrical detector<br />
at the front side (chest dosimeter) and at the back side (back dosimeter).<br />
As recommended by ICRP, [2007]<br />
the personal dose equivalent rate<br />
Ḣ p (d) at a depth d=10 mm gives a<br />
conservative assessment of the effective<br />
dose rate under most irradiation<br />
conditions. However, this requires the<br />
personal dosimeter to be worn at a<br />
position on the body which is representative<br />
with respect to the exposure.<br />
In general it is recommended to wear<br />
a dosimeter in front of the chest,<br />
where Ḣ p (d) is supposed to give a conservative<br />
estimation of the effective<br />
dose even in cases of lateral or isotropic<br />
radiation incidence on the body<br />
[ICRP, 2007]. However, in cases of<br />
exposure from the back, the question<br />
arises if a dosimeter worn at the front<br />
still appropriately assesses the effective<br />
dose. For a worker inside an<br />
emplacement drift, as investigated in<br />
the current study, a certain amount of<br />
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radiation incidents on the backside<br />
due to the backscattered radiation by<br />
the surrounding drift layers. Therefore,<br />
in order to study the influence<br />
of the backscattered radiation, two<br />
cylindrical detectors (2 cm radius and<br />
0.2 cm length) were modelled in the<br />
phantom (see Fig. 2): one on the front<br />
side (representing a dosimeter worn<br />
in front of the chest) and another one<br />
on the back side (representing a<br />
dosimeter worn at the back side), both<br />
at 10 mm depth to calculate Ḣ p (d).<br />
The personal dose equivalent rate<br />
Ḣ p (d) is calculated as:<br />
Where Ḋ n (E n ) is the neutron absorbed<br />
dose rate, and Ḋ γ (E γ ) is the gamma<br />
absorbed dose rate. The quality factor<br />
for photons (Q γ ) is equal to 1; while<br />
for neutrons (Q n ) it is dependent on<br />
the neutron energy (E n ) and the linear<br />
energy transfer (L) according to:<br />
The MCNP6 energy deposition tally<br />
F6 [Pelowitz et al., 2013] was used to<br />
calculate the absorbed dose rate (D).<br />
Tabulated values for Q n (E n ) were<br />
taken from [Siebert and Schuh macher,<br />
1995] to convert the F6 tally results to<br />
dose equivalents.<br />
| | Fig. 3.<br />
Different angles of the phantom with respect to POLLUX.<br />
To study the effect the orientation<br />
of the worker with respect to the<br />
shielding cask, five simulations with<br />
the phantom at angles of 0°, 15°,<br />
45°, 60°, and 90° with respect to<br />
POLLUX-10 symmetrical axis (see<br />
Figure 3) were performed in RSD. For<br />
each simulation, the Ḣ p (10) obtained<br />
with the front dosimeter and the sum<br />
of the Ḣ p (10) obtained with the front<br />
and back dosimeter were compared to<br />
check if the use of only one dosimeter<br />
may underestimate the received dose<br />
rate. To reduce the calculation time,<br />
the MCNP6 variance reduction technique<br />
“geometry splitting” [Pelowitz et<br />
al., 2013] was applied. Using geometry<br />
splitting, a weighting is assigned in<br />
the following way: regions near the<br />
tallies (cylindrical detectors in the<br />
phantom) are assigned with a greater<br />
importance than regions farther away.<br />
When a particle leaves a region it is<br />
split/killed according to the importance<br />
ratio adjusting the weight of the<br />
remaining particles to leave the tally<br />
unbiased. A total of 1 · 10 +8 particles<br />
were required per simulation to pass<br />
the ten MCNP6 statistical checks.<br />
2.4 Comparison of Ḣ p (10) in<br />
the rock salt and clay<br />
formation drifts during a<br />
typical working scenario<br />
The above explained methodology,<br />
i.e. the use of two dosimeters for the<br />
estimation of Ḣ p (10) was applied to<br />
the working scenario of POLLUX®<br />
| | Fig. 4.<br />
MCNP6 model of the four steps for a POLLUX® disposal scenario (for details see text).<br />
disposal in the emplacement drift<br />
based on the proposal of DBE TECH-<br />
NOLOGY GmbH [Filbert et al., 1995].<br />
Figure 4 shows the MCNP6 models of<br />
four main working steps in the cask<br />
disposal procedure as well as the main<br />
components. A description of their<br />
geometry can be found in [Bollingerfehr<br />
et al., 2011]. The four steps are:<br />
first the cask is transported on a<br />
carriage through the drift with an<br />
electric locomotive with a driver<br />
sitting inside the cabin (see Fig. 4a).<br />
Once it arrives at the disposal position,<br />
as shown in Fig. 4b, the cask is slowly<br />
positioned under a storage equipment<br />
which elevates the cask from the<br />
carriage to allow locomotive and<br />
carriage to drive back. Once the locomotive<br />
is driven back, the storage<br />
equipment places the cask on the<br />
ground (see Fig. 4c). Finally, as shown<br />
in Fig. 4d, the locomotive moves<br />
the storage equipment to the next<br />
disposal position.<br />
Since the cask geometry of<br />
POLLUX-3M and POLLUX-3U are<br />
equal to that of POLLUX-10, the same<br />
steps as described in Fig. 4 were simulated<br />
for the disposal of POLLUX-10,<br />
POLLUX-3M and POLLUX-3U cask. To<br />
compare the radiation exposure, the<br />
same or a similar amount of SNF<br />
should be disposed in both emplacement<br />
drifts, i.e. one POLLUX-10 cask<br />
in RSD containing fuel rods of one<br />
MOX and nine UOX fuel assemblies<br />
(5.45 tHM) and three casks in CLD<br />
(one POLLUX-3M and two POLLUX-<br />
3U, in total 4.92 tHM). As the working<br />
steps are the same for the disposal<br />
of POLLUX-10, POLLUX-3M and<br />
POLLUX- 3U, Ḣ p (10) was employed to<br />
compare the radiation exposure in the<br />
different working steps as described<br />
above.<br />
The driver sitting inside the cabin<br />
was represented in this study by<br />
the phantom (see Fig. 2). Since the<br />
worker stays all the time inside the<br />
cabin and faces the shielding cask,<br />
the angle between phantom and cask<br />
is always 0°. However, the amount<br />
of backscattered radiation may be<br />
further increased due to the reflection<br />
by the cabin walls and backscattered<br />
neutrons from the drift walls. To<br />
perform the MCNP6 simulations,<br />
geometry splitting was employed in<br />
the drift and inside the locomotive<br />
cabin to reduce the number of transported<br />
particles. To pass the ten<br />
MCNP6 statistical checks, 4 · 10 +8<br />
particles were required for simulation<br />
of the transport and location under<br />
the storage equipment. For the placement<br />
and retreat of the storage<br />
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a) Inside the gallery b) Free in air<br />
| | Fig. 5.<br />
Spectral neutron fluence rate calculated with MCNP6 (for details see text).<br />
equipment 1.5 · 10 +9 particles were<br />
required per simulation, since the<br />
distance between the cask and the<br />
phantom is larger.<br />
3 Results and discussion<br />
3.1 The ambient dose<br />
equivalent rate Ḣ*(10) in<br />
the emplacement drift<br />
Figure 5a and Figure 5b show the<br />
spectral fluence rate calculated with<br />
MCNP6 at 1 m distance to the cask<br />
surface in X direction inside the RSD<br />
and CLD as well as FIA. The relative<br />
error of the fluence rates in each<br />
energy bin is in general less than 4 %,<br />
except for some bins with fluence<br />
rates lower than 0.005 cm −2 s −1 . The<br />
effect of the backscattered radiation<br />
can be observed for the RSD with the<br />
local minimum of the spectral fluence<br />
rate between 2 · 10 -3 to 3 · 10 -3 MeV<br />
(Fig. 5a), which is caused by elastic<br />
neutron scattering on 23 Na (one of the<br />
main isotopes of the surrounding rock<br />
salt), which has a peak in the crosssection<br />
at 2.8 · 10 -3 MeV. In the CLD,<br />
the maximum between 1 · 10 -8 and<br />
1 · 10 -6 MeV (Fig. 5a) shows the presence<br />
of moderated neutrons mainly<br />
due to interactions with 16 O content of<br />
the concrete lining (mainly composed<br />
of CaO, SiO 2 , Al 2 O 3 , H 2 O).<br />
Figure 6 shows Ḣ*(10) at 1 meter<br />
distance to cask surface in the<br />
different drifts as well as FIA. Since<br />
the cask shielding in the X direction is<br />
the thickest, Ḣ*(10) outside the cask<br />
in the X direction is lower than that<br />
in Y and Z directions. For the FIA<br />
scenarios, Ḣ*(10) in the X direction<br />
outside the POLLUX-10 cask is 24 %<br />
lower than that outside the POLLUX-<br />
3M, while in the Y and Z directions<br />
Ḣ*(10) outside the POLLUX-10 cask is<br />
24 % and 20 % higher than outside<br />
the POLLUX-3M, respectively. This<br />
can be explained due to the influence<br />
of the inactive zone at the top and<br />
bottom of the fuel rods (X direction).<br />
Figure 7 shows the neutron spectra<br />
before and after the Zircaloy layer in<br />
the inactive zone for POLLUX-3M and<br />
POLLUX-10. The total neutron fluence<br />
rate before the inactive zone is<br />
higher for POLLUX-10 (1538 cm −2 s −1 )<br />
than for POLLUX-3M (861 cm −2 s −1 ).<br />
According to the calculated density of<br />
the SNF (see Equation 1), which is for<br />
POLLUX-10 (active and inactive zone)<br />
3 times larger than for POLLUX-3M,<br />
neutrons emitted in the X direction<br />
are stronger shielded by the Zircaloy<br />
layer in POLLUX-10 (total neutron<br />
fluence after the inactive zone<br />
331 cm −2 s −1 ) than in POLLUX-3M<br />
( total neutron fluence after the<br />
inactive zone 340 cm −2 s −1 ).<br />
| | Fig. 6.<br />
Ḣ*(10) at 1 meter distance from the cask calculated with MCNP6 (for details see text).<br />
| | Fig. 7.<br />
Spectral neutron fluence rate for the SNF calculated with MCNP6 (for details see text).<br />
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| | Fig. 8.<br />
Ḣ*(10) at different distances in the X direction for POLLUX-10 in rock salt, POLLUX-3M and POLLUX-3U<br />
in clay formation and POLLUX-3U in free air.<br />
Figure 8 shows Ḣ*(10) at different<br />
distances in the X direction<br />
for POLLUX- 10, POLLUX-3M and<br />
POLLUX- 3U as well as also for their<br />
respective FIA cases. Ḣ*(10) for<br />
POLLUX- 10 in RSD is between 30 %<br />
(at the cask surface) and 80 % (at<br />
10 m distance) higher than for<br />
POLLUX- 10 FIA. For POLLUX-3M and<br />
POLLUX-3U, Ḣ*(10) in CLD is between<br />
15 and 75 % higher than FIA. This<br />
reveals the important role of the<br />
backscattered radiation on the radiation<br />
field in a geological disposal<br />
facility. Figure 8 shows further that up<br />
to 1 m distance, Ḣ*(10) is higher (up<br />
to 9 %) for POLLUX-3M in CLD than<br />
for POLLUX-10 in RSD. From this<br />
point on, the higher moderation of the<br />
concrete layers and the larger reflection<br />
of the salt layers lead to a higher<br />
Ḣ*(10) in the RSD (between 10 % and<br />
40 %). Since only spent UOX was<br />
loaded in a POLLUX-3U cask, Ḣ*(10)<br />
for the case of POLLUX-3U is in<br />
general 63 % lower than that for<br />
POLLUX-3M.<br />
Angle<br />
[degrees]<br />
Ḣ p (10) Total<br />
[µSv/h]<br />
Ḣ p (10) Chest<br />
[µSv/h]<br />
3.2 Influence of the angle<br />
between phantom and<br />
disposal cask on Ḣ p (10)<br />
Table 1 shows Ḣ p (10) obtained with<br />
the detector at the chest Ḣ p (10) Chest<br />
and at the back Ḣ p (10) Back of the<br />
phantom (see Fig. 2). Also included in<br />
the table are the sum of both detectors<br />
Ḣ p (10) Total and the contribution of<br />
each detector to Ḣ p (10) Total . When the<br />
frontal body part of the phantom is<br />
facing the cask, corresponding to an<br />
angle of 0°, the main contribution to<br />
Ḣ p (10) Total comes from the detector at<br />
the chest. However, as the angle between<br />
phantom and cask increases,<br />
the contribution of the detector at the<br />
back increases. This phenomenon<br />
arrives its maximum when the phantom<br />
has an angle 90° with the cask. In<br />
this case, the dose rate obtained with<br />
each detector represents approximately<br />
50 % of Ḣ p (10) Total . Hence, the<br />
addition of Ḣ p (10) Chest and Ḣ p (10) is a<br />
simple way to account for the angular<br />
dependence.<br />
% Chest<br />
[%]<br />
Ḣ p (10) Back<br />
[µSv/h]<br />
% Back<br />
[%]<br />
0 1.4 1.2 86 0.19 14<br />
15 1.5 1.3 87 0.20 13<br />
45 1.2 0.97 82 0.22 18<br />
60 1 0.75 75 0.26 25<br />
90 0.89 0.45 50 0.44 50<br />
| | Tab. 1.<br />
Total Ḣ p (10) values and the contribution of the chest and back detectors with different angles between<br />
phantom and disposal cask. Calculations were performed in a RSD at 5 meter distance to POLLUX-10<br />
surface.<br />
3.3 Comparison of Ḣ p (10) in<br />
the rock salt and clay<br />
formation drifts during<br />
a working scenario<br />
Table 2 shows the calculated dose<br />
rate Ḣ p (10) Chest and Ḣ p (10) Back for the<br />
working steps of the disposal scenario<br />
shown in Fig. 4. Ḣ p (10) Total in the table<br />
refers to the sum of Ḣ p (10) Chest and<br />
Ḣ p (10) Back while % Chest and % Back are<br />
their percentage contribution to<br />
Ḣ p (10) Total , respectively. In the table,<br />
POLLUX-10 refers to the calculated<br />
Ḣ p (10) for each working step of the<br />
disposal in a RSD, while POLLUX-3<br />
refers to the sum of the calculated<br />
Ḣ p (10) for two POLLUX-3U and one<br />
POLLUX-3M. The calculated dose rate<br />
for the disposal of only one POLLUX-<br />
3M and one POLLUX-3U is also<br />
included in Tab. 2.<br />
For the simulated working steps,<br />
the angle between phantom and<br />
source is always 0°. Hence, the main<br />
contribution to Ḣ p (10) Total is coming<br />
from the chest detector. However,<br />
comparing with the results at 0° given<br />
in Tab. 1, the contribution of the back<br />
dosimeter to Ḣ p (10) Total for the worker<br />
inside the cabin is higher than that for<br />
the worker standing alone in the drift.<br />
This effect is attributed to additional<br />
backscattered radiation due to the<br />
cabin walls and locomotive elements.<br />
The calculated dose rate for each<br />
working step is similar for POLLUX-3M<br />
and for POLLUX-10, while that for<br />
POLLUX-3U is 60 % lower that for<br />
POLLUX-3M, since no spent MOX fuel<br />
was stored in POLLUX-3U. However,<br />
to dispose the same amount of waste<br />
as in a POLLUX-10, one POLLUX-3M<br />
and two POLLUX-3U have to be employed.<br />
Therefore, Ḣ p (10) Total is 30 %<br />
higher for the transport and location<br />
in a CLD that in a RSD. For the<br />
placement and retreat in CLD the<br />
Ḣ p (10) Total is more than a 40 % higher<br />
than that in the RSD. This reveals that<br />
the selection of the host rock can play<br />
an important role in the radiation<br />
exposure of the workers in such facilities.<br />
The developed methodology can<br />
be applied to assess the exposure<br />
during the different steps of nuclear<br />
waste disposal. In this work the same<br />
geometrical parameter were considered<br />
for both emplacement drifts.<br />
However, due to the lower loading<br />
capacity of the cask in CLD, a larger<br />
disposal space is required resulting in<br />
a larger repository compared to a<br />
repository in rock salt [e.g. DBE-Tec,<br />
2016]. This leads to a longer transport<br />
distance and also longer exposure<br />
duration. Since the transport of the<br />
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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
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Step Cask Ḣ p (10) Total Ḣ p (10) Chest % Chest Ḣ p (10) Back % Back<br />
Transport POLLUX-3M 2.8 2.4 87 0.35 13<br />
POLLUX-3U 0.72 0.63 88 0.09 12<br />
POLLUX-3 4.2 3.7 88 0.52 12<br />
POLLUX-10 2.8 2.3 83 0.49 17<br />
Location POLLUX-3M 2.6 2.2 85 0.38 15<br />
POLLUX-3U 0.72 0.63 87 0.09 13<br />
POLLUX-3 4.1 3.5 86 0.57 14<br />
POLLUX-10 2.6 2.1 80 0.50 20<br />
Placement POLLUX-3M 0.18 0.15 86 0.02 14<br />
POLLUX-3U 0.04 0.03 85 0.01 15<br />
POLLUX-3 0.26 0.22 86 0.04 14<br />
POLLUX-10 0.19 0.16 82 0.03 18<br />
Retreat POLLUX-3M 0.07 0.05 73 0.02 27<br />
POLLUX-3U 0.02 0.01 85 0.002 15<br />
POLLUX-3 0.10 0.08 77 0.02 23<br />
POLLUX-10 0.<strong>06</strong> 0.05 88 0.01 12<br />
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| | Tab. 2.<br />
Total Ḣ p (10) values and the contribution of the chest and back detectors for the different disposal steps in RSD and CLD.<br />
cask is one of the steps with the<br />
highest personal dose rate, it is<br />
plausible to assume that the total<br />
personal exposure for disposal in clay<br />
formation is higher than for disposal<br />
in rock salt.<br />
Since a precise description of the<br />
duration of each working step is<br />
still unknown, only a dose rate<br />
comparison is conducted in this study.<br />
The following example will try to<br />
demonstrate the importance of this<br />
description. Assuming that five hours<br />
are required to dispose a POLLUX-10,<br />
where four hours are for transport (to<br />
simplify, only transport in a drift is<br />
considered) and the fifth hour is<br />
equally divided amongst the other<br />
three steps (20 min/step). For the<br />
disposal in CLD the transport of<br />
each cask will take longer since the<br />
needed space within the drift is larger<br />
(assuming 7 h). This will lead to a<br />
dose of 12.26 μSv and 30.79 μSv for<br />
RSD and CLD, respectively. However<br />
as more and more casks are disposed<br />
in the drift, the transport time will<br />
reduce. Assuming that only 1 h is<br />
required for the transport when the<br />
drift is almost full, and since the time<br />
for the other three steps will be the<br />
same, it will lead to a dose of 3.77 μSv<br />
and 5.65 μSv for RSD and CLD,<br />
respectively.<br />
As illustrated in the example<br />
above, the duration of the working<br />
steps (especially the transport) plays a<br />
decisive role in the personal dose.<br />
Therefore, a precise description of<br />
the different steps is necessary for<br />
a proper comparison between the<br />
different disposal options and to<br />
provide recommendations for minimizing<br />
the occupational radiation<br />
exposure.<br />
4 Summary and<br />
conclusions<br />
In the current study, the ambient<br />
dose equivalent rate Ḣ*(10) and the<br />
personal dose equivalent rate Ḣ p (10)<br />
were calculated for emplacement of<br />
casks with spent UOX / MOX fuel<br />
within two generic deep geological<br />
disposal facilities. In the rock salt drift<br />
a POLLUX-10 was placed, while for<br />
the clay drift with concrete lining a<br />
POLLUX-3M and two POLLUX-3U<br />
were disposed. In addition, casks free<br />
in air were also investigated. Results<br />
show that the backscattered radiation<br />
of the host rock layers or the concrete<br />
lining increases Ḣ*(10) in the disposal<br />
drift in comparison with a cask FIA.<br />
Ḣ*(10) for POLLUX-10 in RSD is<br />
between 30 % (at the cask surface)<br />
and 80 % (at 10 m distance) higher<br />
than for POLLUX-10 free in air. For<br />
POLLUX-3M and POLLUX-3U, Ḣ*(10)<br />
in CLD is between 15 and 75 % higher<br />
than FIA. The higher increase for<br />
POLLUX-10 is caused by the neutron<br />
reflection of the rock salt layers, while<br />
in the clay drift the presence of oxygen<br />
in the concrete lining moderates the<br />
neutrons resulting in a lower increase.<br />
For the calculation of Ḣ p (10) a<br />
mathematical phantom was modelled<br />
with two detectors, one at the front<br />
side of the chest and another one<br />
at the back side. Calculations with<br />
different angles between the phantom<br />
and the cask show that there is an<br />
angular dependence of the registered<br />
dose rate values. This effect is<br />
enhanced if the dose rate is obtained<br />
with only one dosimeter. Therefore, it<br />
was proposed to sum up the dose rate<br />
obtained with both dosimeters. This<br />
methodology was applied to the<br />
working scenario for the disposal of a<br />
POLLUX® type cask in an emplacement<br />
drift. The results of the investigated<br />
scenario, where the worker is<br />
sitting inside the locomotive cabin<br />
and always facing the cask, show that<br />
the main contribution to Ḣ p (10) comes<br />
from the front detector. However, due<br />
to the additional neutron scatterings<br />
at the cabin, the contribution of the<br />
back detector to Ḣ p (10) is up to 10 %<br />
higher that with the worker just<br />
standing alone in the drift. Therefore,<br />
a study of the effective dose under<br />
this irradiation conditions should be<br />
performed to verify if Ḣ p (d) is still<br />
a conservative assessment.<br />
The calculated personal dose rate<br />
for each working step is similar for<br />
POLLUX-3M and for POLLUX-10 but is<br />
40 % higher than that for POLLUX-3U.<br />
However, to dispose the same amount<br />
of waste as in the RSD, three casks<br />
have to be placed in the CLD. Therefore<br />
each disposal step has to be<br />
carried out three times (one POLLUX-<br />
3M and two POLLUX-3U, in Tab. 2<br />
summarized as POLLUX-3), which<br />
leads to a higher dose rate (between<br />
35 % and 40 % depending of the<br />
working step) for the disposal in CLD.<br />
In this study the same geometrical<br />
parameter where considered for both<br />
galleries. However, due to the lower<br />
loading capacity of the cask in CLD, a<br />
larger emplacement drift is required.<br />
AMNT <strong>2017</strong><br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
390<br />
AMNT <strong>2017</strong><br />
This leads to a longer transport<br />
distance and also longer exposure<br />
durations. Since the transport of the<br />
canister is one of the steps with the<br />
highest personal dose rate, it is<br />
plausible to assume that the total<br />
personal exposure for disposal in a<br />
clay formation drift is higher than for<br />
disposal in a rock salt drift.<br />
Acknowledgements<br />
The authors would like to thank our<br />
colleagues of DBE TECHNOLOGY<br />
GmbH for fruitful discussions regarding<br />
the emplacement of POLLUX®<br />
casks and for providing a movie showing<br />
the working scenario in a drift in<br />
rock salt. This study was financially<br />
supported by the German Federal<br />
Ministry of Education and Research<br />
(BMBF; grant number 15S9082E)<br />
as part of the joint research project<br />
ENTRIA – Disposal Options for Radioactive<br />
Residues: Interdisciplinary<br />
Analyses and Development of Evaluation<br />
Principles.<br />
| | M. Sc. Héctor Saurí Suárez, winner of the “AMNT <strong>2017</strong> Best Paper Award”<br />
during his presentation of the paper “Monte-Carlo Based Comparison of the<br />
Personal Dose for Emplacement Scenarios of Spent Nuclear Fuel Casks in<br />
Generic Deep Geological Repositories” at the 48 th AMNT in Berlin, Germany.<br />
| | “AMNT <strong>2017</strong> Best Paper Award” ceremony: Dr. Alexander Zulauf, NUKEM<br />
Technologies Engineering Services GmbH; Dr. Ralf Güldner, President<br />
of DAtF; M. Sc. Héctor Saurí Suárez; Frank Apel, Chairperson of KTG;<br />
Dr. Ron Dagan, Karlsruhe Institute of Technology (KIT) (f.l.t.r.)<br />
References<br />
| | Bollingerfehr, W., Filbert, W., Lerch, C.,<br />
& Tholen, M. (2011): Endlagerkonzepte.<br />
Bericht zum Arbeitspaket 5 Vorläufige<br />
Sicherheitsanalyse für den Standort<br />
Gorleben. Gesellschaft für Anlagen und<br />
Reaktorsicherheit (GRS) mbH. GRS-272.<br />
| | BMUB (2015): National programme for<br />
the responsible and safe management<br />
of spent fuel and radioactive waste.<br />
Bundesministerium für Umwelt, Naturschutz,<br />
Bau und Reaktorsicherheit<br />
(BMUB) Berlin, Germany, August 2015.<br />
| | Chen, L., Duveau, G., Poutrel, A., Jia, Y.,<br />
Shao, J.F., & Xie, N. (2014): Numerical<br />
study of the interaction between<br />
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Journal of Rock Mechanics and Mining<br />
Sciences Vol. 71, pp. 405–417.<br />
| | DBE-Tec (2016): Flächenbedarf für<br />
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GmbH, Gutachten für Kommission<br />
Lagerung hoch radioaktiver Abfall stoffe,<br />
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| | Filbert, W., Engelmann, H.J., Heda, M.,<br />
& Neydek, J. (1995): Direkte Endlagerung<br />
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(DEAB) – Handhabungsversuche zur<br />
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von Endlagern für Abfallstoffe mbH<br />
(DBE), T60, Peine.<br />
| | Filbert, W., Tholen, M., Engelmann, H.J.,<br />
Graf, R., & Brammer, K.-J. (2011).:<br />
Disposal of Spent Fuel from German<br />
Nuclear Power Plants: The Third Option<br />
- Disposal of Transport and Storage<br />
Casks (Status). Proceedings of the<br />
WM2011 Conference, February 27 -<br />
March 3, 2011, Phoenix, USA<br />
| | International Commission on Radiological<br />
Protection (1991): 1990<br />
Recommendations of the International<br />
Commission on Radiological Protection.<br />
ICRP Publication 60. Ann. ICRP 21 (1-3).<br />
| | International Commission on Radiological<br />
Protection (1996): Conversion<br />
Coefficients for use in Radiological<br />
Protection against External Radiation.<br />
ICRP Publication 74. Ann. ICRP 26 (3-4).<br />
| | International Commission on Radiological<br />
Protection (2007): The 2007<br />
Recommendations of the International<br />
Commission on Radiological Protection.<br />
ICRP Publication 103. Ann. ICRP 37 (2-4).<br />
| | Janberg, K. & Spilker, H. (1998): Status of<br />
the development of final disposal casks<br />
and prospects in Germany. Nuclear<br />
Technology, Vol. 121, pp 136-147.<br />
| | Leon Vargas, R. Stahlmann, J. &<br />
Mintzlaff, V. (<strong>2017</strong>): Thermal impact in<br />
the geo metrical settings in deep<br />
geological repositories for HLW with<br />
retrievability and monitoring.<br />
Proceedings of the International<br />
High-Level Radioactive Waste Management<br />
Conference, IHLRWMC<strong>2017</strong>,<br />
April 9 -13, <strong>2017</strong>, Charlotte, USA,<br />
pp. 664-670.<br />
| | Pang, B., Saurí Suárez, H., & Becker, F.<br />
(2016): Individual dosimetry in disposal<br />
repository of heat-generating nuclear<br />
waste. Radiation Protection Dosimetry,<br />
first published online May 5. 2016<br />
| | Peiffer, F., McStocker, B., Gründler, D.,<br />
Ewig, F., Thomauske, B., Havenith, A., &<br />
Kettler, J. (2011): Abfallspezifikation<br />
und Mengengerüst, Basis Ausstieg aus<br />
der Kernenergienutzung. Bericht zum<br />
Arbeitspaket 3 Vorläufige Sicherheitsanalyse<br />
für den Standort Gorleben.<br />
Gesellschaft für Anlagen und Reaktorsicherheit<br />
(GRS) mbH. GRS-278.<br />
| | Pelowitz, D.B., Goorley, J.T., James, M.R.,<br />
Booth, T.E., Brown, F.B, Bull, J.S., ...<br />
Zukaitis, A. (2013): MCNP6 TM User’s<br />
Manual Version 1.0. Los Alamos National<br />
Security, LA-CP-13-0<strong>06</strong>34, Rev. 0.<br />
| | Siebert, B.R.L., & Schuhmacher, H.<br />
(1995): Quality factors, ambient and<br />
personal dose equivalent for neutrons,<br />
based on the new ICRU stopping power<br />
data for protons and alpha particles.<br />
Radiation Protection Dosimetry,<br />
Vol. 58, pp 177-183.<br />
| | Stahlmann, J., Mintzlaff, V. & Leon<br />
Vargas, R. (2015): Generische Tiefenlagermodelle<br />
mit Option zur Rückholung<br />
der radioaktiven Reststoffe:<br />
Geologische und Geotechnische<br />
Aspekte für die Auslegung. ENTRIA-<br />
Arbeitsbericht-03. TU Braunschweig,<br />
Institut für Grundbau und Bodenmechanik,<br />
Germany.<br />
| | U.S. Department of Energy (data last<br />
retrieved July 1. 2016): Bottle Manikin<br />
Absorption (BOMAB) Phantoms.<br />
Retrieved from http://www.id.energy.<br />
gov/resl/phantom/bomab.html<br />
Authors<br />
M. Sc. Héctor Saurí Suárez a<br />
Dr. Bo Pang a,b<br />
Dr. Frank Becker a<br />
Dr. Volker Metz a<br />
(a) Institute for Nuclear Waste<br />
Disposal (INE)<br />
Karlsruhe Institute of Technology<br />
(KIT)<br />
Hermann-von-Helmholtz-Platz 1<br />
76344, Eggenstein-Leopoldshafen,<br />
Germany<br />
(b) College of Physics and Energy<br />
Shenzhen University<br />
Nanhai Avenue 3688<br />
518<strong>06</strong>0, Nanshan District,<br />
Shenzhen, China<br />
AMNT <strong>2017</strong><br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
Die 15. AtG-Novelle zur Umsetzung der<br />
EURATOM-Sicherheits- Richtlinie<br />
391<br />
Christian Müller-Dehn<br />
Die 15. AtG-Novelle (AtG: Atomgesetz) hat das parlamentarische Gesetzgebungsverfahren mit dem Beschluss des<br />
Bundestages in der dritten Lesung vom 30.3.<strong>2017</strong> und der Befassung im Bundesrat vom 12.5.<strong>2017</strong> nunmehr vollständig<br />
durchlaufen, harrt aber noch der Veröffentlichung im Bundesgesetzblatt. Hintergrund aller Regelungen sind die<br />
Ergänzungen der EURATOM-Sicherheits-Richtlinie, die der Europäische Rat im Juli 2014 beschlossen hat und die<br />
bis spätestens August <strong>2017</strong> in den nationalen Regelungen der EURATOM-Mitgliedsstaaten zu verankern sind. Da<br />
die meisten dieser Ergänzungen jedoch bereits geltender Standard im deutschen Atomrecht waren, waren die für<br />
Deutschland umsetzungsbedürftigen Regelinhalte gering. Dies wird ausdrücklich auch in der Gesetzesbegründung<br />
festgehalten.<br />
Die in Deutschland danach noch umzusetzenden Regelungen<br />
lassen sich drei Regelungskreisen zuordnen: Der<br />
Einführung eines periodischen Topical Peer Reviews<br />
für die kerntechnischen Anlagen, der Erweiterung der<br />
Betreiber pflichten sowie der Etablierung von Informations-<br />
und sonstigen Pflichten für die atomrechtlich<br />
zuständigen Behörden.<br />
Unstreitig neu und regelungsbedürftig ist das Topical<br />
Peer Review, das jetzt ausführlich in § 24b Abs. 2 AtG<br />
geregelt wird. Danach soll, beginnend im Jahr <strong>2017</strong>, für in<br />
Betracht kommende und sich im Geltungsbereich dieses<br />
Gesetzes befindliche kerntechnische Anlagen mindestens<br />
alle 6 Jahre eine Selbstbewertung hinsichtlich ausgewählter<br />
technischer Themen vorgenommen werden. Das<br />
bereits begonnene Topical Peer Review für <strong>2017</strong>, für<br />
das nun auch die gesetzliche Legitimation geschaffen<br />
wird, hat gemäß der europaweiten Vorgabe technische<br />
Fragen zum Alterungsmanagement zum Gegenstand.<br />
Wenn und soweit ein EU-weit abgestimmtes Thema zum<br />
nächsten oder einem späteren Topical Peer Review nur<br />
Anlagen im Leistungsbetrieb betreffen sollte, wäre das<br />
Topical Peer Review in Deutschland dann entbehrlich.<br />
§ 7 c AtG, also die Norm, die die Pflichten des Genehmigungsinhaber<br />
regelt, wächst und wächst. Die Norm<br />
wird in dreierlei Hinsicht ergänzt:<br />
So hat der Genehmigungsinhaber sicher zu stellen, dass<br />
auch Auftragnehmer und Unterauftragnehmer über die<br />
zur Erfüllung der atomrechtlichen Pflichten erforderlichen<br />
personellen Mittel verfügen. Freilich wird hierzu in der<br />
Gesetzesbegründung festgehalten, dass dies nur der Klarstellung<br />
dient und bereits zuvor materiell galt.<br />
Außerdem wird in § 7c Abs. 2 AtG eine neue Nummer 4<br />
eingefügt, die den Betreiber im Rahmen seiner Kommunikationspolitik<br />
zur Information der Öffentlichkeit,<br />
insbesondere der lokalen Bevölkerung und von Interessenträgern<br />
verpflichtet. Rechtspolitisch mag eine<br />
solche detaillierte Regelung zur Öffentlichkeitsarbeit eines<br />
Industrieunternehmens befremden, aufgrund der zahlreichen<br />
nationalen Regelungen, insbesondere in der AtVfV<br />
(Verordnung über das Verfahren bei der Genehmigung<br />
von Anlagen nach § 7 des Atomgesetzes (Atomrechtliche<br />
Verfahrensverordnung)), der AtSMV (Verordnung über<br />
den kerntechnischen Sicherheitsbeauftragten und über<br />
die Meldung von Störfällen und sonstigen Ereignissen<br />
(Atomrechtliche Sicherheitsbeauftragten- und Meldeverordnung<br />
– AtSMV)) und den Sicherheitsanforderungen<br />
an Kernkraftwerken, besteht freilich insoweit bereits eine<br />
so große Regelungsdichte, dass sich hier nur noch<br />
nachdrücklich die Frage aufdrängt, ob überhaupt noch<br />
ein Umsetzungsbedarf bestand.<br />
Drittens wird in § 7c AtG ein neuer Absatz 3 eingefügt,<br />
der den Genehmigungsinhaber verpflichtet, angemessene<br />
Verfahren und Vorkehrungen für den anlageninternen<br />
Notfallschutz vorzusehen. Aufgrund der bestehenden<br />
gesetzlichen Verpflichtungen gemäß §§ 7d, 19a Abs. 4 AtG<br />
und §§ 51, 53 StrlSchV sowie weiteren Konkretisierungen<br />
im untergesetzlichen Regelwerk, nämlich den Sicherheitsanforderungen<br />
an Kernkraftwerke, den Leitfäden zur<br />
periodischen Sicherheitsüberprüfung und einschlägigen<br />
RSK- und SSK-Empfehlungen (RSK: Reaktor-Sicherheitskommission,<br />
SSK: Strahlenschutzkommission) bestand<br />
insoweit allerdings keine ausfüllungsbedürfte Rechtslücke.<br />
Die entsprechende Regelung dient somit im<br />
Ergebnis lediglich der Transparenz gegenüber europäischen<br />
Organen. Der sehr hohe Standard hinsichtlich der<br />
mit der 15. AtG geregelten Betreiberpflichten spiegelt sich<br />
in der Stellungnahme des nationalen Normen-Kontrollrates<br />
wieder, der hierfür nur einen sehr geringen Aufwand<br />
pro kerntechnischer Anlage wiedergibt.<br />
Abgerundet werden die Neuregelungen durch Verpflichtungen,<br />
die sich an Behörden richten. Dies ist<br />
zum einen die Verpflichtung der zuständigen Behörden<br />
nach § 24 a Abs. 1 AtG, die Öffentlichkeit über den bestimmungsgemäßen<br />
Betrieb kerntechnischer Anlagen<br />
sowie über meldepflichtige Ereignisse und Unfälle zu<br />
informieren. Weiterhin wird das für die kerntechnische<br />
Sicherheit und den Strahlenschutz zuständige Bundesministerium<br />
verpflichtet, unverzüglich zu einer internationalen<br />
Überprüfung einzuladen, falls es zu einem<br />
Unfall in einer kerntechnischen Anlagen käme, der Maßnahmen<br />
des externen Notfallschutzes erforderte. Beide<br />
Pflichten sind neu und daher auch umsetzungsbedürftig.<br />
Die Änderungen der EURATOM-Sicherheits-Richtlinie<br />
waren darauf gerichtet, die kerntechnische Sicherheit<br />
in Europa weiter zu erhöhen. Vor diesem Hintergrund<br />
belegt der hier aufgezeigte sehr geringe Regelungsbedarf<br />
zur Umsetzung der EURATOM-Sicherheits-Richtlinie<br />
nochmals und sehr nachdrücklich das hohe Niveau der<br />
Anforderungen an die kerntechnischen Anlagen in<br />
Deutschland, das bereits zuvor bestanden hatte und von<br />
diesen erfüllt wird.<br />
Author<br />
Dr. Christian Müller-Dehn<br />
Senior Vice President Nuclear Regulation and Policy<br />
PreussenElektra GmbH<br />
Tresckowstraße 5<br />
30457 Hannover, Deutschland<br />
SPOTLIGHT ON NUCLEAR LAW<br />
Spotlight on Nuclear Law<br />
The 15 th German Atomic Energy Act Amendment to the Implementation of the EURATOM Nuclear Safety Directive ı Christian Müller-Dehn
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
392<br />
ENVIRONMENT AND SAFETY<br />
Retrofitting a Spent Fuel Pool Spray<br />
System for Alternative Cooling as a<br />
Strategy for Beyond Design Basis Events<br />
Christoph Hartmann and Zoran Vujic<br />
Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as<br />
happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed.<br />
Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant<br />
design. As part of Krško Nuclear Power Plant’s Safety Upgrade Program, Krško Nuclear Power Plant decided on, and<br />
Westinghouse successfully designed a retrofit of the AP1000® plant spent fuel pool spray system to provide alternative<br />
spent fuel cooling.<br />
1 Introduction<br />
Following the tsunami and resulting<br />
events in 2011 at the Fukushima<br />
Daiichi Nuclear Power Plant in<br />
Japan, the Western European Nuclear<br />
Regulators Association (WENRA) updated<br />
the safety reference levels in<br />
its report “WENRA Reactor Safety<br />
Reference Levels,” [1] to incorporate<br />
lessons learned from the event.<br />
The update includes establishing an<br />
independent heat removal system<br />
for the spent fuel pool to maintain<br />
the integrity of used fuel assemblies<br />
being temporarily stored there in the<br />
unlikely event of a beyond design<br />
basis accident. The AP1000® nuclear<br />
power plant design foresees provisions<br />
for beyond design basis events.<br />
This includes failure of the spent fuel<br />
pool walls or floor, which would result<br />
in the spent fuel pool draining and<br />
fuel assemblies being uncovered. This<br />
design is also in agreement with the<br />
Nuclear Energy Institute (NEI) issue<br />
of NEI <strong>06</strong>-12, Revision 2, “B.5.b Phase<br />
2 & 3 Submittal Guideline” [2], where<br />
an external spent fuel pool makeup<br />
and spray strategy is recommended.<br />
For events with extended loss of AC<br />
power, that is, station blackout, and/<br />
or loss of heat sink due to the spent fuel<br />
pool draining or partially draining,<br />
spent fuel cooling can be provided by<br />
a spent fuel pool spray system. A spent<br />
fuel pool spray system based on the<br />
AP1000® plant design can be retrofitted<br />
for existing nuclear power<br />
plants. In the case of an uncontrolled<br />
spent fuel pool water level drop to<br />
such an extent that the spent fuel pool<br />
would be completely dried out, an<br />
emergency spray system is the best<br />
practical solution that can be applied<br />
for sufficient cooling of the spent fuel<br />
assemblies.<br />
2 AP1000 Plant Spent Fuel<br />
Pool Cooling<br />
The AP1000® plant design features<br />
multiple, diverse lines of defense to<br />
ensure spent fuel cooling can be<br />
maintained for design basis and<br />
beyond design basis events.<br />
During normal and abnormal<br />
conditions, defense-in-depth and duty<br />
systems provide highly reliable spent<br />
fuel pool cooling. These systems are<br />
driven by offsite AC power or the<br />
onsite standby diesel generators.<br />
For unlikely events with extended<br />
loss of AC power, that is, station<br />
blackout, and/or loss of heat sink,<br />
passive systems provide spent fuel<br />
pool cooling. These passive systems<br />
require minimal or no operator action<br />
and are sufficient for at least 72 hours<br />
under all possible loading conditions.<br />
After 72 hours, several different<br />
means are provided to continue spent<br />
fuel pool cooling using installed plant<br />
equipment, as well as off-site equipment.<br />
Even for beyond design basis<br />
events with postulated spent fuel pool<br />
damage and multiple failures in the<br />
passive safety-related systems and<br />
active defense-in-depth systems, the<br />
AP1000® plant spent fuel pool spray<br />
system provides an additional line of<br />
defense to prevent spent fuel damage.<br />
The spent fuel pool is located in a<br />
hardened section of the Auxiliary<br />
Building and contains used fuel that<br />
has been removed from the nuclear<br />
reactor core. Typically, 64 fuel assemblies<br />
are removed from the reactor<br />
| | Fig. 1.<br />
AP1000® Plant Spent Fuel Pool Spray System. Spray headers and nozzles (left) and section view of spray pattern from nozzle (right).<br />
Environment and Safety<br />
Retrofitting a Spent Fuel Pool Spray System for Alternative Cooling as a Strategy for Beyond Design Basis Events ı Christoph Hartmann and Zoran Vujic
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 2.<br />
Pipe routing of spent fuel pool alternate cooling with mobile heat exchanger (MHX) and spray system.<br />
during refueling every 18 months and<br />
stored in the spent fuel pool. The<br />
AP1000® plant’s spent fuel pool has<br />
the capacity to cool up to 889 spent or<br />
used fuel assemblies, which are<br />
continuously submerged beneath<br />
approximately 7.6 m of water.<br />
The spent fuel assemblies continue<br />
to generate decay heat naturally even<br />
when they are removed from the<br />
reactor and are placed in the spent<br />
fuel pool. This decay heat will<br />
decrease significantly over time so<br />
that older spent fuel produces less<br />
heat than spent fuel that has recently<br />
been removed from the reactor.<br />
The spent fuel in the spent fuel<br />
pool is cooled by transferring the<br />
decay heat from the used fuel to the<br />
water in the spent fuel pool. The spent<br />
fuel pool water is, in turn, pumped<br />
through a loop with a heat exchanger<br />
where it is cooled and decay heat is<br />
transferred to a second water cooling<br />
system. The cooled water is then<br />
returned from the second water cooling<br />
system to the spent fuel pool and<br />
the decay heat is transferred to the<br />
environment. There are two identical<br />
spent fuel pool cooling trains, though<br />
only one pump and heat exchanger in<br />
one of the two trains are in operation<br />
in most circumstances.<br />
3 AP1000 plant spent fuel<br />
pool spray system<br />
The AP1000® spent fuel pool spray<br />
system is designed to cool the spent<br />
fuel during a beyond design basis<br />
event in accordance with the B.5.b<br />
guideline [2].<br />
The AP1000® plant spent fuel pool<br />
spray system has two redundant spray<br />
headers located on either side of the<br />
spent fuel pool. There are 16 spray<br />
nozzles on each header (Figure 1,<br />
left). One header receives water<br />
through either gravity-fed draining of<br />
the passive containment cooling water<br />
storage tank, which is located on top<br />
the Shield Building, or from a flanged<br />
connection located in the truck bay,<br />
which is used with an onsite portable<br />
pump. The other header receives<br />
water from the fire protection water<br />
tanks and the diesel-driven or electric<br />
motor-powered fire protection system<br />
water pumps. Spray nozzles distribute<br />
water spray in the form of a hollow<br />
spray cone over the fuel assemblies.<br />
Only one spray header is required<br />
to assure sufficient cooling of the<br />
exposed spent fuel due to sensible<br />
heat and latent heat from water spray<br />
vaporization (Fig. 1, right).<br />
The spray system used to cool the<br />
spent fuel pool during a postulated<br />
loss-of-large-area event is sized to<br />
provide an adequate amount of<br />
spray to the hottest fuel assembly that<br />
will enter the spent fuel pool. The<br />
analytical basis for determining the<br />
minimum amount of spray needed to<br />
cool a fuel assembly is adapted from<br />
the calculation used in Section 3.3<br />
of the Sandia report, “Mitigation of<br />
Spent Fuel Pool Loss-of-Coolant<br />
Inventory Accidents And Extension of<br />
Reference Plant Analyses to Other<br />
Spent Fuel Pools” [3]. Further, to<br />
prevent pressurization inside the<br />
Fuel Handling Building, the system<br />
includes a relief panel to release steam<br />
that is produced during the cooling<br />
process.<br />
4 Krško nuclear power<br />
plant safety upgrade<br />
program<br />
Krško Nuclear Power Plant was<br />
already in the process of making<br />
significant upgrades as a result of<br />
applying for a license extension in<br />
2009 to operate beyond 2023. The<br />
Krško Safety Upgrade Program<br />
was designed in response to the<br />
Slovenian Nuclear Safety Administration’s<br />
re gulations and interpretation<br />
of reference safety levels<br />
from the report, “ WENRA Reactor<br />
Safety Reference Levels” [1], concerning<br />
reasonable measures to prevent<br />
and mitigate severe accidents in<br />
preparation for the possibility of<br />
extending original plant operating<br />
licenses. The reference safety levels<br />
within the report were updated in<br />
2014 to incorporate lessons learned<br />
from the event at the Fukushima site.<br />
The measures defined in the frame of<br />
the Krško Safety Upgrade Program<br />
are in agreement with the nuclear<br />
industry’s response to the Fukushima<br />
accident and the resulting update of<br />
the safety reference levels proposed<br />
by WENRA. This includes plant upgrades<br />
and design changes to address<br />
design extension conditions defined<br />
in the report and beyond design basis<br />
accidents.<br />
Krško’s Safety Upgrade Program is<br />
divided into various projects being<br />
carried out during three phases. The<br />
Spent Fuel Pool Alternative Cooling<br />
Project is in the scope of Phase 2. The<br />
project is scheduled to be completed<br />
by the end of <strong>2017</strong>.<br />
The Spent Fuel Pool Alternate<br />
Cooling Project shall assure alternate<br />
cooling of used fuel assemblies<br />
by using a mobile heat exchanger<br />
or spray system (see Figure 2).<br />
Furthermore, it shall assure depressurization<br />
of the Fuel Handling<br />
Building by using relief panels to<br />
release steam produced during the<br />
cooling process.<br />
The systems of the Spent Fuel Pool<br />
Alternate Cooling Project are designed<br />
to assure that heat is removed from<br />
the spent fuel during Design Extension<br />
Conditions A and B and to mitigate<br />
spent fuel damage. The operational<br />
conditions for the systems of the Spent<br />
Fuel Pool Alternate Cooling Project<br />
are classified according to the plant’s<br />
severe accident scenarios, following<br />
the WENRA guidance document “ Issue<br />
F: Design Extension of Existing<br />
Reactors” [1].<br />
ENVIRONMENT AND SAFETY 393<br />
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ENVIRONMENT AND SAFETY 394<br />
| | Fig. 3.<br />
Spent fuel pool spray system simplified process flow diagram.<br />
5 Spent fuel pool spray<br />
system function and<br />
design<br />
The spent fuel pool spray system will<br />
be used in in the event of scenarios<br />
described in Design Extension Conditions<br />
A and B, which postulate a<br />
highly unlikely drainage of the spent<br />
fuel pool. In such scenarios, cooling<br />
water can be provided by a spray<br />
header with fixed spray nozzles<br />
installed along the spent fuel pool<br />
walls. Westinghouse conducted<br />
experi mental testing to determine the<br />
exact number of nozzles required and<br />
their optimal positions. Water can be<br />
supplied by two diverse sources: the<br />
fire protection system and the Sava<br />
River.<br />
The fire protection system can be<br />
flexibly connected using fire hoses.<br />
The Sava River water can be pumped<br />
using a mobile pump unit and fire hoses.<br />
The mobile pump unit is powered<br />
directly by a diesel motor. Either water<br />
source is connected to the spent fuel<br />
pool spray system with hose connections<br />
that are installed inside and<br />
outside of the Fuel Handling Building.<br />
A simplified process flow diagram of<br />
the spent fuel pool spray system is<br />
shown in Figure 3.<br />
The Krško Nuclear Power Plant’s<br />
spent fuel pool spray system is<br />
designed to perform its designated<br />
safety function under Design Extension<br />
Conditions. This includes being<br />
designed to meet the seismic performance<br />
requirements for operation<br />
and mitigation during and after a<br />
design extension condition earthquake,<br />
which is equal to twice the<br />
design requirements for a safe shutdown<br />
earthquake for the existing<br />
systems, structures and components<br />
of the Krško Nuclear Power Plant.<br />
The spent fuel pool spray system’s<br />
permanently installed equipment will<br />
be protected against flood events with<br />
additional margin, even in the case<br />
of a Sava River bank failure. The<br />
system’s permanently installed equipment<br />
is also designed to withstand<br />
extreme winds and tornados. In order<br />
to keep the design of the system as<br />
simple as possible only local indicators<br />
are used and there are no electrically<br />
driven components.<br />
As in the AP1000® spent fuel pool<br />
spray system, the Krško Nuclear Power<br />
Plant’s spent fuel pool spray system:<br />
• Will provide a sufficient amount of<br />
cooling water to maintain the<br />
spent fuel cladding temperature at<br />
lower than 400 °C for a long term<br />
during the loss of ultimate heat<br />
sink,<br />
• Is sized to provide an adequate<br />
amount of spray to the hottest fuel<br />
assembly that will enter the spent<br />
fuel pool during a postulated lossof-large-area<br />
event,<br />
• Has an analytical basis for determining<br />
the minimum amount<br />
of spray needed to cool a fuel<br />
assembly adapted from the calculation<br />
used in Section 3.3 of the<br />
Sandia Letter Report “Mitigation of<br />
Spent Fuel Pool Loss-of-Coolant<br />
Inventory Accidents And Extension<br />
of Reference Plant Analyses to<br />
Other Spent Fuel Pools” [3].<br />
An example of the water spray coverage<br />
and distribution through the<br />
nozzles throughout the spent fuel<br />
pool is shown in Figure 4. The red<br />
circles show the area inside of the<br />
spent fuel pool covered by water<br />
through the spray nozzles, which are<br />
installed at the side walls of the spent<br />
fuel pool. Darker red areas show<br />
possible overlap, whereas water that<br />
does not fit the spent fuel pool geometry<br />
hits the spent fuel pool walls.<br />
Westinghouse conducted experi mental<br />
testing to optimize the spray configuration<br />
around the spent fuel pool edge<br />
and to verify the spray water distribution<br />
and overlap.<br />
Inside the spent fuel pool is a two<br />
region rack design. Recently offloaded<br />
fuel, which is comprised of the hottest<br />
fuel assemblies, is placed in the<br />
Region 1 racks. The Region 2 racks<br />
provide high density storage for<br />
| | Fig. 4.<br />
Example of the spent fuel pool spray system’s nozzle coverage and distribution of water spray.<br />
Environment and Safety<br />
Retrofitting a Spent Fuel Pool Spray System for Alternative Cooling as a Strategy for Beyond Design Basis Events ı Christoph Hartmann and Zoran Vujic
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
cooled, irradiated fuel assemblies. Per<br />
the initial cooling time defined in<br />
plant procedures (typically 4.5 years<br />
or three fuel cycles), the most recently<br />
cooled, irradiated fuel assemblies are<br />
transferred from the Region 1 racks to<br />
the Region 2 racks.<br />
6 Experimental determination<br />
of spray nozzle<br />
coverage and distribution<br />
A specific flow density and a uniform<br />
coverage of cooling water over a large<br />
rectangular area are required to cool<br />
fuel assemblies in the unlikely event<br />
that they become uncovered during a<br />
beyond design basis accident. The<br />
experimental testing used to determine<br />
the optimal coverage area, spray<br />
flow density and distribution of the<br />
spent fuel pool spray nozzles for cooling<br />
uncovered fuel assemblies in case<br />
of Design Condition Extensions A and<br />
B include testing different types of<br />
nozzles with different volume flow<br />
rates as the decay heat decreases with<br />
cooling time (see Region 1 and Region<br />
2 in Fig. 4). The measurement setup<br />
of the experimental testing at the<br />
Lechler GmbH Technology Center<br />
Metzingen is shown in Figure 5.<br />
• Volume Flow:<br />
variable adjustable, depending on<br />
spray nozzle type and pressure<br />
The coverage area measurements<br />
were performed by using adequate<br />
collecting canisters with the same<br />
dimensions as the rack cell canisters<br />
at specified positions (see Fig. 5).<br />
Pictures and video recording documented<br />
the testing process and resulting<br />
water distribution. An example of<br />
the measurement setup is shown in<br />
Figure 6.<br />
| | Fig. 6.<br />
Example of water distribution from a spray<br />
nozzle [4].<br />
7 Managed challenges<br />
Retrofitting the AP1000® spent fuel<br />
pool spray system to Krško Nuclear<br />
Power Plant’s existing systems and<br />
structures posed a few challenges.<br />
These included:<br />
• Planning the pipe routing and<br />
meeting the space requirements<br />
for pipe supports (see Figure 7),<br />
• Assuring pipe routing had minimal<br />
impact to equipment already existing<br />
inside of the Fuel Handling<br />
Building,<br />
• Planning the installation process<br />
for the system to meet the space<br />
requirements in the Fuel Handling<br />
Building,<br />
• Assuring the number of fuel assembly<br />
racks that cannot be used for<br />
spent fuel storage due to nozzles<br />
and/or supports which protrude<br />
into the pool are as low as possible,<br />
• Designing the retrofit system with<br />
consideration of the fuel assembly<br />
loading pattern at Krško Nuclear<br />
Power Plant to assure that the<br />
spray nozzles provide the required<br />
water spray amount, coverage and<br />
distribution to cool the spent<br />
fuel pool during Design Extension<br />
Conditions A and B.<br />
Each of these challenges was successfully<br />
resolved.<br />
8 Summary<br />
Due to requirements for nuclear<br />
power plants to withstand beyond<br />
design basis accidents, including<br />
events such as happened in 2011 in<br />
the Fukushima Daiichi Nuclear Power<br />
Plant in Japan, alternative cooling of<br />
spent fuel is needed. Alternative spent<br />
fuel cooling can be provided by a<br />
retrofitted spent fuel pool spray<br />
system based on the AP1000® plant<br />
design. As part of Krško Nuclear Power<br />
Plant’s Safety Upgrade Program,<br />
Krško Nuclear Power Plant decided<br />
on, and Westinghouse successfully<br />
designed a retrofit of the AP1000®<br />
plant spent fuel pool spray system to<br />
provide alternative spent fuel cooling.<br />
The spent fuel pool spray system<br />
will be installed inside and outside the<br />
Fuel Handling Building. For diverse<br />
water supply, sources such as the fire<br />
protection system and river water<br />
were considered and chosen by Krško<br />
Nuclear Power Plant. The system has a<br />
robust design that employs local<br />
measurements and indicators and<br />
ENVIRONMENT AND SAFETY 395<br />
| | Fig. 5.<br />
Exemplary measurement setup for experimental<br />
testing of spray nozzle coverage and<br />
distribution [4].<br />
The following conditions were<br />
applied to determine the flow density<br />
and the coverage area:<br />
• Spray height:<br />
variable adjustable, depending on<br />
local conditions around the spent<br />
fuel pool<br />
• Setting angle horizontal:<br />
variable adjustable, depending on<br />
local conditions around spent fuel<br />
pool and spray nozzle type<br />
• Pressure:<br />
variable adjustable, depending<br />
on spray nozzle type and hydraulic<br />
design of spent fuel pool spray<br />
system<br />
| | Fig. 7.<br />
Depiction of the local conditions around the Krško nuclear power plant spent fuel pool.<br />
Environment and Safety<br />
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396<br />
OPERATION AND NEW BUILD<br />
does not require electrically driven<br />
components. Westinghouse determined<br />
the required amount of water<br />
spray needed to cool the spent fuel<br />
pool during Design Extension Conditions<br />
A and B analytically; designed<br />
the system’s hydraulics to provide<br />
sufficient flow rates for the volume;<br />
defined the corresponding pipe<br />
diameters; and determined the pipe<br />
routing, based on the on the local<br />
space restrictions around Krško<br />
Nuclear Power Plant’s spent fuel pool.<br />
Westinghouse also conducted experimental<br />
testing at the Lechler GmbH<br />
Technology Center Metzingen to determine<br />
the coverage and distribution of<br />
the spent fuel pool spray nozzles and<br />
to confirm the spray height, setting<br />
angle horizontal orientation, pressure<br />
and volume of water flow.<br />
The spent fuel pool spray system is<br />
planned to be installed by the end of<br />
<strong>2017</strong>.<br />
References<br />
[1] Western European Nuclear Regulators<br />
Association (WENRA): WENRA Safety<br />
Reference Levels for Existing Reactors,<br />
September 2014.<br />
[2] Nuclear Energy Institute (NEI): B.5.b<br />
Phase 2 & 3 Submittal Guideline,<br />
Revision 2, NEI <strong>06</strong>-12, December 20<strong>06</strong>.<br />
[3] Sandia National Laboratories:<br />
Mitigation of Spent Fuel Pool Loss-of<br />
Coolant Inventory Accidents And<br />
Extension of Reference Plant Analyses<br />
to Other Spent Fuel Pools, Sandia Letter<br />
Report, Rev. 2, November 20<strong>06</strong>.<br />
[4] Lechler GmbH, Technology Center<br />
Metzingen.<br />
Authors<br />
Dipl.-Ing. Christoph Hartmann<br />
Project Engineer Safety<br />
Engineering<br />
Dr.-Ing. Zoran Vujic<br />
Marketing Manager Business<br />
Development<br />
Westinghouse Electric Germany<br />
GmbH<br />
Dudenstraße 6<br />
68167 Mannheim, Germany<br />
Cyber Security in Nuclear Power Plants<br />
and its Portability to Other Industrial<br />
Infrastructures<br />
Sébastien Champigny, Deeksha Gupta, Venesa Watson and Karl Waedt<br />
Introduction This technical contribution provides a snapshot of the current cyber security efforts in different<br />
industry domains. We argue that stringent security controls (countermeasures) that are already in place for nuclear<br />
power plants (NPP) can be ported to other industry domains. A reason for this is that the nuclear domain is more<br />
formally regulated, thus graded security requirements were already mandated long before the critical infrastructure<br />
debates started and before gradual enforcement of the European and national legislation.<br />
Note: Generally, in the nuclear and<br />
industrial automation domain, the<br />
term “control” is used mainly to<br />
denote Instrumentation and Control<br />
(I&C), Industrial Automation and<br />
Control Systems (IACS) or SCADA<br />
(Supervisory Control and Data Acquisition)<br />
referring to control theory<br />
tasks. However, in the security context,<br />
the term “Security Control” is<br />
ubiquitous, and means any countermeasure<br />
that can reduce the systems<br />
risk due to security threats. Countermeasures<br />
are not limited to add-on<br />
provisions at the components or systems<br />
level. For example, they also include<br />
provisions at the software<br />
source code level.<br />
In Section 1, we will provide an<br />
overview of current international and<br />
national cyber security guidance, and<br />
how this guidance evolved for International<br />
Atomic Energy Agency (IAEA),<br />
Nuclear IEC and selected countries.<br />
Section 2 summarises the increasing<br />
cyber security efforts for Industrial<br />
Automation and Industry 4.0 as well<br />
as its Chinese “Manufactured in China<br />
2025” and US “Industrial Internet of<br />
Things” counterparts. Section 3<br />
provides reasons for the portability<br />
of Security Controls from Nuclear<br />
to other industrial infrastructure.<br />
Summary provides an outlook on the<br />
newest cyber security-related activities<br />
in the different domains, and<br />
concludes with a summary of the<br />
main steps that are necessary for<br />
achieving and maintaining a target<br />
security level.<br />
1 Cyber security and safety<br />
requirements for NPPs<br />
In the nuclear domain, for Safety,<br />
Human Factors Engineering, Physical<br />
Security, Radiation Protection and<br />
Cyber Security, the international<br />
top-level guidance is provided by the<br />
IAEA. The IAEA guidance is regularly<br />
updated based on priorities set by<br />
yearly or bi-yearly meetings of representatives<br />
of all IAEA member states.<br />
The overall IAEA Cyber Security<br />
guidance is refined, e.g. for Instrumentation<br />
& Control (I&C) and<br />
Electrical Systems (ES), by the<br />
Nuclear IEC subcommittees. However,<br />
each country may supersede the<br />
international guidance by providing a<br />
mandatory higher priority regulation,<br />
as will be addressed in section 1.4 for<br />
selected countries.<br />
1.1 Stringent and graded<br />
security requirements for<br />
I&C already since 1986<br />
Safety and security grading are<br />
essential when addressing critical<br />
industrial infrastructures. Grading by<br />
Safety Categories in IEC 61226 and<br />
Safety Classes in IEC 61513, were<br />
already in place since the first editions<br />
of these standards. The softwarespecific<br />
requirements for software<br />
implementing Category A or Category<br />
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Cyber Security in Nuclear Power Plants and its Portability to Other Industrial Infrastructures ı Sébastien Champigny, Deeksha Gupta, Venesa Watson and Karl Waedt
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
B and C I&C functions, are also graded<br />
by the respective standards IEC<br />
60880:1986 and IEC 62138. The first<br />
edition of IEC 60880:1986 already<br />
contained explicit requirements on<br />
security during software development<br />
and security during software deployment,<br />
two essential phases in the software<br />
development lifecycle.<br />
1.2 Overall IAEA Cyber security<br />
Guidance<br />
The IAEA Cyber security Guidance is<br />
published in the IAEA Nuclear Security<br />
Series (NSS). Currently the top-level<br />
guidance is IAEA NSS 17 from 2011.<br />
Developing this guidance took several<br />
years with considerable input by<br />
member states provided since 20<strong>06</strong>,<br />
and essential agreements being<br />
achieved during the first major IAEA<br />
cyber security conference in summer<br />
2011. IAEA NSS 17 introduces a graded<br />
security approach with 5 security<br />
levels and recommendations on<br />
security zones.<br />
IAEA NSS 17 is complemented by<br />
IAEA NSS 8 on preventive and protective<br />
measures against insider<br />
threats, and further IAEA NSS guidance,<br />
including IAEA NSS 12, on a<br />
comprehensive educational program<br />
in nuclear security.<br />
1.3 Nuclear IEC Cyber security<br />
Standards<br />
Subsequently, the three major Nuclear<br />
IEC cyber security standards will be<br />
introduced.<br />
1.3.1 The Top-level Nuclear IEC<br />
Cyber security Standard<br />
After initial attempts to structure<br />
the top-level nuclear IEC standard<br />
according to nuclear safety and other<br />
criteria, finally, a core-team devised<br />
the alignment with the most popular<br />
information security standard ISO/<br />
IEC 27001:2005 then in place. This<br />
structuring was proposed mainly in<br />
order to reduce the initial training<br />
needs of security staff already familiar<br />
with the mainstream standards, and<br />
in order to avoid annexes with cumbersome<br />
mappings.<br />
While ISA99 experts were involved<br />
in the development of the first toplevel<br />
nuclear IEC 62645:2013 cyber<br />
security standard, an alignment<br />
with ISA99 industrial cyber security<br />
standards or the corresponding IEC<br />
62443-x-x was ultimately not attempted,<br />
as several planned parts of the IEC<br />
62443-x-x series were not yet available<br />
and because the Security grading<br />
follows a different approach, as will be<br />
addressed in a subsequent section.<br />
| | Fig. 1.<br />
Safety functions, process functions and I&C functions.<br />
1.3.2 Coordinating safety and<br />
cyber security by IEC 62859<br />
Whether safety and cyber security<br />
should be considered jointly or subsequently,<br />
is a part of ongoing debates<br />
in different industry domains. For<br />
nuclear, the security grading is<br />
directly related to the potential impact<br />
of a security attack on nuclear safety.<br />
Figure 1 shows the hierarchical<br />
refinement from Safety Objectives<br />
(level 1) down to I&C Functions (level<br />
3 and 4). Main safety objectives are<br />
control of reactivity, residual heat<br />
removal and confinement of radioactive<br />
material.<br />
Cyber security is applied at the<br />
level of I&C and IT equipment while<br />
considering the potential impact of<br />
manipulations on Safety Functions<br />
and Safety Objectives.<br />
IEC 62859:2016 [1] specifies the<br />
main requirements for coordinating<br />
safety and cyber security. In other<br />
industries, work on this important<br />
topic was just started, e.g. by the new<br />
working group WG20 of IEC TC65.<br />
1.3.3 Detailed security controls<br />
for nuclear by IEC 63096<br />
Similar to the alignment of IEC 62645<br />
with ISO/IEC 27001, the new working<br />
draft IEC 63096 is being aligned<br />
with the ISO/IEC JTC1/SC27 WG1<br />
standard ISO/IEC 27002:2013. This<br />
nuclear IEC standard extends the<br />
generic security controls of ISO/IEC<br />
27002 by recommendations for each<br />
security level: BR (Baseline Requirements),<br />
S3, S2 and S1 (highest<br />
security level). It also provides<br />
guidance for the main I&C and ES<br />
(Electrical Systems) lifecycle phases:<br />
Product & Platform Development,<br />
Engineering and Operation & Maintenance.<br />
Additionally, it provides<br />
security control specific guidance for<br />
legacy I&C and ES systems.<br />
As a sector-specific standard,<br />
similar to ISO/IEC 27009 [3], for<br />
non-nuclear utilities, IEC 63096<br />
provides guidance that is structured<br />
and formatted in principle in line with<br />
ISO/IEC 27009 which provides<br />
common guidance on the elaboration<br />
of sector-specific security controls and<br />
Information Security Management<br />
Systems (ISMS) standards.<br />
1.4 International and national<br />
nuclear cyber security<br />
regulations<br />
Table 1 lists the international standards<br />
discussed above, along with the<br />
national standards for Germany, USA<br />
and the UK. In Germany, SEWD<br />
(Schutz gegen Störmaßnahmen oder<br />
sonstige Einwirkungen Dritter/Protection<br />
against Disruptive Acts or Other<br />
Intervention of Third Parties) is a requirement<br />
found in §6 para. 2 no. 4<br />
Atomic Energy Act. released in 1959.<br />
[7]. Licenses for the storage of nuclear<br />
fuels are only granted once risks and<br />
threats, as a result of SEWD, can be<br />
considered as negligible. Created<br />
by Congress in 1974, the USA’s NRC<br />
regulates commercial nuclear power<br />
plants and other uses of nuclear<br />
materials. NRC RG 5.71 [4] provides<br />
guidelines for the protection of digital<br />
computer and communication systems<br />
and networks from cyberattacks,<br />
against which licensees should provide<br />
assurance. The Nuclear Energy Institute<br />
(NEI) 08-09 “Cyber Security<br />
Plan for Nuclear Power Reactors” provides<br />
a generic template for a cyber<br />
security plan, which must be used<br />
by licensees to develop their cyber<br />
security plans to be submitted to the<br />
NRC [8]. The HMG IA (Information<br />
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OPERATION AND NEW BUILD 398<br />
International and national nuclear laws and regulations<br />
| | Tab. 1.<br />
Examples of international and national nuclear cyber security regulations.<br />
Assurance) Standard is intended for us<br />
by IA practitioners, working especially<br />
with UK Government ICT systems, as<br />
the foundation for their Information<br />
Risk Management Policy. This standard<br />
provides a methodology by which<br />
these practitioners can “identify, assess<br />
and determine the level of risk to an ICT<br />
system and a framework for the selection<br />
of appropriate risk treatments.”<br />
Requirements from these international<br />
nuclear Cyber Security<br />
standards are applicable for the whole<br />
nuclear power plant. Figure 2 shows<br />
the scope of applicability of these<br />
requirements using the example of a<br />
typical nuclear I&C architecture.<br />
In Figure 3, the relationships<br />
between safety standards (in purple)<br />
and security standards (in orange)<br />
from different industries are indicated.<br />
All the individual fields have<br />
their own specific standards for safety<br />
and security. For example, IEC 6<strong>06</strong>01<br />
and IEC 62304 are the safety standards<br />
referred in medical field.<br />
| | Fig. 2.<br />
An example of a nuclear I&C architecture (© AREVA).<br />
| | Fig. 3.<br />
Safety and Security Interface at the Standards Level (© IEC TC65).<br />
2 Gradual consideration<br />
of information security<br />
in Industry 4.0 and IoT<br />
Industry 4.0 and “Manufactured in<br />
China 2025” are governed by a “Reference<br />
Architecture Model Industry 4.0”<br />
(RAMI) or similar which are typically<br />
represented by cubes subdivided as<br />
6x6x6 or 5x5x5. The 3 axis of the cube<br />
are “Layers”, “Hierarchy Levels” and<br />
“Value Streams”. None of the 6 Layers<br />
(Business, Functional, Information,<br />
Communication, Integration and<br />
Asset) explicitly contains cyber<br />
security. Similarly along the other two<br />
axes, cyber security is not explicitly<br />
included. This is due to the fact that<br />
security and interoperability are<br />
considered as integral components in<br />
multiple of the 3D elements that built<br />
up the complete cube, see Figure 4.<br />
2.1 Generic information<br />
security<br />
One purpose of generic security standards<br />
is to be applicable by any size of<br />
an organization, e.g. a one- employee<br />
service provider or a multinational organization.<br />
The ISO/IEC 27000 series<br />
takes credit on meeting this criterion.<br />
Still, beyond these generic information<br />
security standards in the 27000 to<br />
27021 range, additional standards in<br />
the 27031 to 27050 and other ranges<br />
provide more in-depth guidance.<br />
2.2 IT security for power<br />
generating plants<br />
VGB-S-175 addresses generic security<br />
requirements, Defense-in-Depth<br />
Operation and New Build<br />
Cyber Security in Nuclear Power Plants and its Portability to Other Industrial Infrastructures ı Sébastien Champigny, Deeksha Gupta, Venesa Watson and Karl Waedt
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 4.<br />
Reference Architectural Model Industry 4.0 by ZVEI (© Plattform Industrie 4.0).<br />
principles, redundancy and diversity,<br />
risk management, risk analysis and<br />
security countermeasures for both,<br />
new built and power plant modernization<br />
projects.<br />
Furthermore, VGB provides guidance<br />
on intrusion detection and<br />
prevention (addressed in more detail<br />
by ISO/IEC 27039), patch management<br />
(addressed in more detail by<br />
IEC 62443-2-3), security gateways<br />
(addressed in more detail in ISO/<br />
IEC 27033-4), wireless (ISO/IEC<br />
27033-6), documentation of security<br />
incidents (ISO/IEC 27035-3) and<br />
additional countermeasures.<br />
2.3 Emerging industrial<br />
automation security<br />
Cyber security for Industrial Automation<br />
mainly builds on the ISA99<br />
specific standards which are published<br />
as IEC 62443-x-x. The 13 parts<br />
of this series are not yet complete. The<br />
security grading is based on the risk<br />
an attacker imposes and on its<br />
strength. This regularly leads to controversy,<br />
as the strength of an attacker<br />
can change over time, e.g. today’s<br />
“script kiddies” have other malicious<br />
tools as compared to 10 years earlier.<br />
2.4 Initial Industry 4.0 and IoT<br />
proposals<br />
Despite its current incompleteness,<br />
IEC 62443-x-x builds a solid basis for<br />
cyber security in the Industry 4.0<br />
RAMI framework. Interoperability is a<br />
key component of Industry 4.0. The<br />
multipart IEC 62541 defines the Open<br />
Connectivity Unified Architecture<br />
(OPC UA) not just as a communications<br />
protocol, but as a communication<br />
architecture that supports<br />
among other services, interoperability<br />
between digital technologies from<br />
different vendors. The services, as<br />
provided by the layers of the platformindependent<br />
OPC UA, include the<br />
semantics of an information model,<br />
address spaces, discovery services,<br />
alarm functions, etc.<br />
AREVA NP implements Embedded<br />
OPC UA, for example, in its SIPLUG®<br />
family of monitoring sensors, as<br />
shown on Figure 5. Hence, it can<br />
directly be connected to reporting and<br />
trend surveillance systems. This<br />
feature drastically reduces the costs<br />
for interconnecting the respective<br />
sensor devices with equipment from<br />
different vendors, as deployed worldwide<br />
at NPP sites [6].<br />
Part 2 of IEC 62541 provides the<br />
security framework for OPC UA, the<br />
main aim of which is to provide<br />
security for the data exchanges facilitated<br />
by this architecture.<br />
While there seems to be general<br />
acceptance on OPC UA as a part of<br />
Industry 4.0 and IoT, the final hard<br />
real-time communication protocols<br />
and the respective security solutions<br />
are still to emerge.<br />
| | Fig. 5.<br />
SIPLUG® OPC UA based example.<br />
3 Portability of cyber<br />
security knowledge and<br />
features from nuclear to<br />
other industrial infrastructures<br />
The subsequent sections exemplify<br />
some domains where solutions from<br />
the nuclear domain can be adapted<br />
and applied to other domains.<br />
3.1 Joint functional safety and<br />
cyber security consideration<br />
One benefit of IEC 62859:2016, as<br />
compared to generic safety & security<br />
related solution, is its well delimited<br />
context of the applicability for NPPs.<br />
The grading is well defined based on<br />
the maximum impact in the nuclear<br />
context. The transition between the<br />
safety states is also well understood<br />
due to comprehensive deterministic<br />
and probabilistic safety analyses.<br />
These results from the functional<br />
safety experts can directly be leveraged<br />
by the security staff. This<br />
approach can be transferred and<br />
adjusted for other business domains.<br />
The security grading has to be<br />
adjusted to the possible impact levels<br />
in the respective business domain.<br />
Similarly, an analysis is needed and<br />
feasible on which security events can<br />
lead to a similar impact as the respective<br />
safety events (like equipment<br />
faults, failures of supporting assets,<br />
spurious actuations). Based on this<br />
mapping, a risk management process<br />
can be modified in order to adjust<br />
and justify the criticality assignment<br />
(assignment of security degrees to<br />
systems) and to apply complementary<br />
security controls.<br />
3.2 Security grading<br />
The generic information security<br />
standards like ISO/IEC 2700x define<br />
no security grading- also called<br />
security levels or levels of trust. Unfortunately,<br />
in some industries the<br />
grading may be defined based on<br />
criteria that may change over time.<br />
Thus, the strength of an attacker may<br />
change while the impact will not<br />
change or only in well-justified (and<br />
easily identifiable) circumstances, e.g.<br />
after power up-rating of an NPP.<br />
As for nuclear, in implementing a<br />
long-term stable impact-based grading<br />
approach, the overall risk management<br />
and security control adjustment<br />
requirements could be considerably<br />
reduced.<br />
3.3 Security awareness<br />
training<br />
Safety Culture and Security Culture<br />
have a long tradition in nuclear, see<br />
e.g. IAEA NSS 7 “Nuclear Security<br />
Culture” from 2008. With humans as<br />
the strongest and also as the weakest<br />
link in the security chain, specific<br />
security training is essential. Such<br />
training can be adapted for other<br />
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OPERATION AND NEW BUILD 400<br />
business domains and for different<br />
staff roles, like operators, service<br />
engineers, physical security staff,<br />
cyber security staff and management.<br />
3.4 Strong preventive security<br />
controls<br />
Often mimicking the activities of their<br />
counterparts in the office IT world,<br />
cyber security safety staff deploy<br />
network or host monitoring systems,<br />
like Network and Host Intrusion<br />
Detection Systems (IDS). These detective<br />
security controls may be the only<br />
option in an office IT environment,<br />
where the exact content and frequency<br />
and destination of messages<br />
sent via communication networks<br />
cannot be predicted. However, for<br />
nuclear and for many other industries,<br />
like process automation and discrete<br />
manufacturing, the data exchange is<br />
of a periodic nature, e.g. with fixed<br />
communication cycle times.<br />
This allows the implementation of<br />
strong Preventive Security controls<br />
beyond baseline firewall filtering. In<br />
many cases, the network architecture<br />
may be adjusted to include Data<br />
Diodes as (preferably optical) Physically<br />
Unidirectional Security Gateways.<br />
Applying these network architecture<br />
level improvements ensures<br />
reaching and maintaining the required<br />
target security degree. An<br />
example of preventive security control<br />
is provided in Figure 6. On the left<br />
half of the figure, an automation<br />
system is shown in its standard configuration.<br />
On the right half of the<br />
figure, the automation system is protected<br />
by patented software called<br />
OPANASec. OPANASec is both a<br />
preventive and a detective measure<br />
against cyber-attacks on the automation<br />
system. It protects the system’s<br />
integrity by detecting any read or<br />
write access to the automation system<br />
and announces it to the operator in<br />
the main control room, by means of a<br />
red traffic light for example. It also<br />
prevents information retrieval and<br />
any modifications of the automation<br />
system by locking read and write<br />
access.<br />
3.5 Forensic readiness<br />
Reports on system intrusions and<br />
manipulations without a trace to the<br />
identity and location of hackers or<br />
threat agents are, in general, frequently<br />
reported in technical magazines,<br />
but also more and more by<br />
commercial media. Typically, the<br />
reason for this is that no forensic<br />
readiness specific security controls are<br />
in place. Also, the implementation of<br />
the forensic readiness security controls<br />
(e.g. log files related) may not be<br />
adequate for the target security level.<br />
As for nuclear, this can be improved<br />
by systematically performing attack<br />
tree analyses and assigning appropriate<br />
forensic readiness security<br />
controls in line with the security<br />
grading.<br />
3.6 Incident response<br />
While incident response on Safety<br />
related incidents has a long tradition<br />
with nuclear, cyber security incident<br />
management is currently in the focus<br />
of the first IAEA financed cyber<br />
security R&D with 14 international<br />
partners.<br />
As one of the major partners in the<br />
IAEA Coordinated Research Proposal<br />
(CRP) J02008, AREVA NP, together<br />
with one of its German partner Universities,<br />
can leverage the results for<br />
other business domains.<br />
3.7 Security testing<br />
The appropriate assignment of<br />
security controls based on a continuous<br />
risk management, is essential<br />
for achieving a high security posture.<br />
However, the implementation or configuration<br />
of some security controls<br />
may be flawed. Even more important,<br />
the implementation and configuration<br />
of the software and FPGA-based<br />
systems may include vulnerabilities,<br />
some of which may be security<br />
relevant.<br />
This mandates a selective, prioritised,<br />
in-depth penetration and<br />
fuzz-testing. We are currently working<br />
on an extensive R&D together with<br />
multiple German partner universities<br />
and several Cyber security PhD candidates,<br />
as part of the partially BMWi<br />
Ministry funded the SMARTEST R&D<br />
project on “smart” (model based)<br />
cyber security testing. The respective<br />
results can be leveraged, as most<br />
of the six (6) Industrial Automation<br />
platforms deployed in NPPs and<br />
analyzed by the project, are also<br />
deployed in other industries.<br />
3.8 Security modelling<br />
The I&C and ES Architecture of NPPs<br />
comprises multiple distributed I&C<br />
systems that are built-up from several<br />
subsystems and components. Modelling<br />
these systems together with<br />
models of the physical process (including<br />
pumps, valves …) is common<br />
practices for several decades. Typically,<br />
this includes simulators which<br />
run in real-time or faster than realtime.<br />
There are modelling approaches<br />
which include the security control<br />
definitions into existing 3D models<br />
(for physical security related security<br />
controls) and 2D models, e.g. for<br />
network architectures. These models<br />
support the systematic generation and<br />
analysis of attack trees, far beyond<br />
any paper-based manual analysis.<br />
This approach can be leveraged by<br />
using the same modelling framework<br />
(e.g. AutomationML from the Industry<br />
4.0 context) for other business<br />
domains. The initial investment in<br />
defining the models is compensated<br />
not only by the more comprehensive<br />
analysis, but also by the opportunities<br />
that the models provide for training of<br />
different staff and even for advertising<br />
security features of the customer<br />
products.<br />
| | Fig. 6.<br />
Security control using patented software OPANASec.<br />
3.9 Security asset management<br />
Implicit asset identification is unavoidable<br />
in order to purchase and install<br />
the equipment. However, an asset<br />
management in line with ISO 55000<br />
and ISO/IEC 19770-x (4 layers of<br />
maturity) is needed in order to leverage<br />
the relevant knowledge about<br />
assets. This is a precondition for<br />
Operation and New Build<br />
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
correct patch management. It can be<br />
well applied in many industries.<br />
3.10 Secure human-machine<br />
interaction<br />
Main control rooms and I&C maintenance<br />
rooms equipped with HMI<br />
equipment are common for power<br />
plants and stringently regulated for<br />
NPPs, e.g. with regard to the explicit<br />
documentation for plant operators.<br />
Different solutions exist for secure<br />
human-machine interaction. An example<br />
of it is a Qualified Display<br />
System (QDS), which limits functionalities<br />
accessible to the operator. The<br />
respective security provision may be<br />
transferred or adapted for other HMI<br />
related user activities.<br />
3.11 Domain specific application<br />
security controls<br />
The semi-formal approach of the upcoming<br />
ISO/IEC 27034-x is applied to<br />
the nuclear context. A key concept is<br />
the Application Security Controls<br />
(ASCs). An ASC provides a semiformal<br />
definition of a security control.<br />
It also includes the indication of the<br />
security grade that the ASC can meet,<br />
the status of the ASC implementation<br />
(e.g. whether verification and validation<br />
were completed), the role assignment<br />
according to RACI (Responsible,<br />
Accountable, Consulted, Informed)<br />
and the specification of links to other<br />
ASCs. AREVA NP even considers advanced<br />
features, like ASC inheritance,<br />
not yet included in the current ISO/<br />
IEC 27034-x standard versions.<br />
As an example of the adaptation of<br />
the ASCs concept, the default grading<br />
of 10 levels of trust has to be adjusted<br />
to the domain specific grading, or a<br />
grading has to be introduced for<br />
the target domain. Additionally, the<br />
accompanying concepts of an Organization<br />
Normative Framework and an<br />
Application Normative Framework<br />
can be adapted.<br />
This is in line with the key concepts<br />
of ASCs, which promote the development<br />
and delivery of high-quality<br />
specialised ASCs by standardsconforming<br />
sub-suppliers.<br />
3.12 Advanced persistent<br />
threats<br />
Targeted Advanced Persistent Threats<br />
(APT), like Stuxnet, are the most<br />
feared attack scenarios in any business<br />
domain. The combination of several<br />
of the aforementioned approaches,<br />
including a comprehensive asset<br />
management, semi-formal modelling<br />
of the assets and supporting assets,<br />
semi-formal description of the<br />
| | Fig. 7.<br />
Overview of cyber security portfolio.<br />
Application Security Controls, targeted<br />
security testing, Forensic<br />
Readiness Security Controls and<br />
further security controls related to the<br />
secure software development will<br />
support in systematically increasing<br />
the security posture and thus, the<br />
effort needed to be spent by an APT<br />
agent.<br />
Similar APT analysis can be performed<br />
for other business domain,<br />
provided the above listed preparations,<br />
like asset management and<br />
semi-formal modelling are already in<br />
place or are implemented.<br />
The knowledge areas described<br />
above should be organised in different<br />
products and services offered to<br />
selected critical industries for efficient<br />
application. An example of how to<br />
implement this is shown in Figure 7.<br />
Summary<br />
Monitoring agencies like the “US<br />
Industrial Control Systems Cyber<br />
Emergency Response Team (ISC-<br />
CERT)”, the “French National Agency<br />
for Information systems’ security” and<br />
the “German Federal Office for Information<br />
Security” (BSI) all record steep<br />
increases in cyberattacks on companies<br />
and institutions in general, and<br />
on critical infrastructures in particular.<br />
For example, the BSI reported an<br />
increase of 20% in the number of<br />
known malicious program versions,<br />
from 2015 to 2016, up to 560 million a<br />
year. Hence, overall public awareness<br />
of cyber security threats, as well as of<br />
legislators, of power plant operators<br />
and of their owners, is also on a steep<br />
rise.<br />
Preemptive cyber security measures<br />
not only avoid loss of revenues,<br />
costs of crisis management, costs of<br />
reimbursements and higher insurance<br />
premiums. They also avoid upcoming<br />
legal penalties for infringement of<br />
an increasingly intransigent legislation.<br />
AREVA NP’s long-standing expertise<br />
in nuclear cyber security relies on<br />
in-depth knowledge of industrial and<br />
legislative requirements and of the<br />
corresponding companies’ protection<br />
needs. As shown above, it applies to a<br />
great extent to any industrial infrastructure<br />
using control systems. Not<br />
only the energy sector, but also the<br />
manufacturing sector, the water and<br />
wastewater systems sector and the<br />
defense industrial base sector benefit<br />
from such an expertise.<br />
References<br />
[1] IEC 62859:2016, Nuclear Power Plants –<br />
I&C Systems – Requirements for Coordinating<br />
Safety and Cyber security.<br />
[2] IEC 62443-3-3:2013, Industrial communication<br />
networks – Network and system<br />
security – Part 3-3: System security<br />
requirements and security levels.<br />
[3] ISO/IEC 27009:2016, Information<br />
technology – Security techniques –<br />
Sector- specific application of ISO/IEC<br />
27001 – Requirements.<br />
[4] U.S. Nuclear Regulatory Commission<br />
(2010). Regulatory Guide 5.71 Cyber<br />
Security Programs for Nuclear Facilities.<br />
Available at: https://www.nrc.gov/<br />
docs/ML0903/ML090340159.pdf.<br />
[5] Th. Poussier, S. Gomes-Augusto, K.<br />
Waedt: Cyber security Aspects of a<br />
Safety Display System. IAEA Inter national<br />
Conference on Computer<br />
Security in a Nuclear World: Expert<br />
Discussion and Exchange, Vienna,<br />
2015-<strong>06</strong>.<br />
[6] OPC Foundation (2016), Unified Architecture:<br />
Interoperability for Industrie 4.0<br />
and the Internet of Things. Available at:<br />
https://opcfoundation.org/wpcontent/uploads/2016/05/OPC-UA-<br />
Interoperability-For-Industrie4-and-IoT-<br />
EN-v5.pdf.<br />
[7] (BMUB) Federal Ministry for the<br />
Environ ment, Nature Conservation,<br />
Building and Nuclear Safety (2015):<br />
Constitution and Laws. Available at:<br />
http://www.bmub.bund.de/en/topics/<br />
nuclear-safety-radiological-protection/<br />
nuclear-safety/legal-provisionstechnical-rules/constitution-and-laws/.<br />
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402<br />
DECOMMISSIONING AND WASTE MANAGEMENT<br />
[8] Department of Homeland Security<br />
(2015). Cyber security Framework<br />
Implementation Guidance for U.S.<br />
Nuclear Power Reactors. Available at:<br />
https://www.us-cert.gov/sites/default/<br />
files/c3vp/framework_guidance/<br />
nuclear-framework-implementationguide-2015-508.pdf.<br />
Validation of Spent Nuclear Fuel Nuclide<br />
Composition Data Using Percentage<br />
Differences and Detailed Analysis<br />
Man Cheol Kim<br />
1 Introduction Nuclide composition data of spent nuclear fuels are important in many nuclear engineering<br />
applications. In reactor physics, nuclear reactor design requires the nuclide composition and the corresponding cross<br />
sections. In analyzing the radiological health effects of a severe accident on the public and the environment, the nuclide<br />
composition in the reactor inventory is among the important input data. Nuclide composition data need to be provided<br />
to analyze the possible environmental effects of a spent nuclear fuel repository. They will also be the basis for identifying<br />
the origin of unidentified spent nuclear fuels or radioactive materials.<br />
The Spent Fuel Isotopic Composition<br />
(SFCOMPO) database [1–3], which<br />
was originally developed by the Japan<br />
Atomic Energy Research Institute and<br />
is now managed by the Organization<br />
for Economic Co-operation and Development/Nuclear<br />
Energy Agency (OECD/<br />
NEA), provides measured nuclide composition<br />
data of spent nuclear fuels.<br />
The SFCOMPO database has been<br />
widely used to validate computer codes<br />
and nuclear data libraries for spent<br />
fuel and fuel cycle applications. For<br />
example, Lee [4] validated TR4PEP, a<br />
depletion code combining the continuous-energy<br />
Monte Carlo transport<br />
code TRIPOLI-4.3 and the point<br />
depletion code PEPIN-2, using the<br />
Takahama-3 post-irradiation examination<br />
results provided in the SFCOMPO<br />
database. Fast et al. [5] compared the<br />
code calculation results obtained using<br />
SCALE 6.0 and 6.1 with measurement<br />
data from Obrigheim nuclear power<br />
plant (NPP) to investigate the validity<br />
of the correlations for burnup calculations<br />
involving key nuclides such as<br />
Cs-134, Cs-137, and Eu-154. Nicolaou<br />
[6] tested the potential of isotopic<br />
fingerprinting for nuclear forensics<br />
purposes using the measurement data<br />
provided in the SFCOMPO database.<br />
With the recognition of the importance<br />
of extending the SFCOMPO<br />
database, the Expert Group on<br />
Assay Data of Spent Nuclear Fuel<br />
Authors<br />
Sébastien Champigny<br />
MBA, Dipl.-Phys., M.Eng.<br />
Product manager cyber security for<br />
critical infrastructures<br />
Deeksha Gupta<br />
M.Sc. in Nuclear Sci. & Tech.<br />
Cyber security PhD Candidate<br />
(EGADSNF), which is in charge of<br />
main taining the OECD/NEA SFCOMPO<br />
database, is trying to obtain new assay<br />
data that are not open to the public or<br />
are open but not widely available.<br />
Suyama et al. [7] mentioned that the<br />
use of the OECD/NEA framework was<br />
intended to facilitate the collection of<br />
new data from member countries.<br />
Possible candidates for newly added<br />
data are summarized in Gauld and<br />
Rugama [8] and the state-of-the-art<br />
report by EGADSNF [9]. For this<br />
purpose, Suyama et al. [10] provided<br />
additional measurement data from<br />
Ohi-1 and Ohi-2 with detailed information<br />
and specifications so that they can<br />
be added to the SFCOMPO database.<br />
Raap et al. [11] reported on the expansion<br />
of the SFCOMPO database by<br />
the addition of measured data from<br />
CANDU reactors, MAGNOX reactors,<br />
VVERs, and RBMKs for use in developing<br />
isotopic signatures for nuclear<br />
forensics purposes.<br />
As the EGADSNF admitted in the<br />
state-of-the-art report [9], measurement<br />
data were added to the SFCOMPO<br />
database as reported by laboratories,<br />
without peer review. Validation of<br />
the data to assess the quality of the<br />
measurements is con sidered a priority<br />
task for improvement of the SFCOMPO<br />
database. The measurement data in<br />
the SFCOMPO database have been<br />
validated in several recent studies.<br />
Venesa Watson<br />
Master in Computer Forensics<br />
Cyber security PhD Candidate<br />
Dr. Karl Waedt<br />
Senior expert Cyber Security<br />
Concepts & Architecture<br />
AREVA GmbH<br />
Paul-Gossen-Straße 100<br />
91052 Erlangen, Germany<br />
Gauld et al. [12] described the recent<br />
experience of Oak Ridge National<br />
Laboratory in validating the measured<br />
isotopic composition data of spent<br />
nuclear fuel. Among the 118 PWR<br />
experimental assay data, 87 (73.7 %)<br />
were from the SFCOMPO database.<br />
Gauld et al. [12] reported problems<br />
such as highly erratic Am-241<br />
measurement data from Takahama-3<br />
due to possible errors in the adjustment<br />
of the time of discharge and<br />
physically impossible measurement<br />
results due to possible typographical<br />
errors. However, the details on how the<br />
SFCOMPO database should be revised<br />
are not clearly described. Okumura et<br />
al. [13] described how the measurements<br />
of Se-79, Tc-99, Sn-126, and<br />
Cs-135 for the Cooper, Calvert Cliffs-1,<br />
and H. B. Robinson-2 reactors in the<br />
SFCOMPO database should be revised<br />
by applying the latest nuclear data,<br />
especially the half-lives of the four<br />
nuclides. For example, the calculatedto-experimental<br />
value for Se-79<br />
changed from 5.5 to 0.92 after the<br />
application of the latest half-life of<br />
Se-79 provided by Bienvenu et al. [14].<br />
This paper proposes a simple<br />
method for analysis and validation of<br />
nuclide composition data of spent<br />
nuclear fuels such as those found in<br />
the SFCOMPO database. The proposed<br />
method consists of a simplified<br />
code calculation, the assumption of a<br />
Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 1.<br />
Distribution of U-235 initial enrichment and burnup of 246 samples in OECD/NEA SFCOMPO database.<br />
constant power history, comparison of<br />
the measured data and code calculation<br />
results, and detailed analysis of<br />
those data that deviate significantly<br />
from other data to identify the causes<br />
of the deviations. During such a crosscheck<br />
process, many, if not all, of the<br />
errors in either the measured data<br />
or the code calculations could be<br />
identified. The proposed method is<br />
described in Section 2. The application<br />
of the proposed method to<br />
nuclide composition data of spent<br />
nuclear fuels from Obrigheim NPP is<br />
described in Section 3, and the identified<br />
data and associated findings are<br />
analyzed in detail. Section 4 presents<br />
the conclusion of this paper.<br />
2 Approach to validation<br />
of nuclide composition<br />
data<br />
2.1 Overview of the data in<br />
the SFCOMPO database<br />
The OECD/NEA SFCOMPO database<br />
provides 10,282 measurement data<br />
from 246 samples of nuclear fuels<br />
irradiated in 14 reactors. Figure 1<br />
shows the distribution of the U-235<br />
initial enrichment and burnup of 246<br />
samples in the SFCOMPO database.<br />
The initial enrichment ranges from<br />
1.45 wt % for the samples from<br />
Fukushima-Daiichi-3 and Monticello<br />
to 4.11 wt % for those from<br />
Takahama-3, and is concentrated<br />
in the range from 2.5 to 3.5 wt %.<br />
The burnup ranges from 2.21 to<br />
71.84 GWd/MTU. Note, however, that<br />
the samples from Monticello show<br />
exceptionally high burnup (exceeding<br />
40 MWd/MTU) despite the low initial<br />
U-235 enrichment. Owing to this<br />
abnormal overburn of low-enriched<br />
fuels, Hermann et al. [15] rated<br />
the data from Monticello as ‘not<br />
recommended.’ If the samples from<br />
Monticello are excluded, the burnup<br />
ranges from 2.21 to 47.3 GWd/MTU.<br />
2.2 Approach to validation<br />
Nuclide composition data of spent<br />
nuclear fuels such as those in the<br />
OECD/NEA SFCOMPO database can<br />
be validated using the percentage<br />
differences between the computed<br />
and measured compositions of each<br />
isotope in each sample. Here, the<br />
percentage difference is defined as the<br />
calculated value minus the measured<br />
value, divided by the measured<br />
quantities or ratios, as follows:<br />
(1)<br />
The percentage difference has been<br />
used in many code validation studies<br />
such as those of Hermann et al. [16],<br />
DeHart and Hermann [17], and Jang<br />
et al. [18]. By reviewing the percentage<br />
differences of nuclide compositions<br />
or ratios, candidates for<br />
detailed analysis can be identified.<br />
The identified candidates can be<br />
subjected to detailed analyses such as<br />
consistency checks as a function of<br />
burnup or parent–daughter pairs, as<br />
performed by Gauld and Rugama [8],<br />
and the original data sources can be<br />
reviewed.<br />
2.3 Code calculations<br />
As mentioned in the state-of-the-art<br />
report by EGADSNF [9], evaluation of<br />
the measured data in the SFCOMPO<br />
database requires significant effort<br />
and is therefore a significant but<br />
challenging objective of future activities<br />
of the EGADSNF. Thus, using<br />
simplified models in code calculations<br />
would be more efficient for performing<br />
a large number of code<br />
calculations in a manageable and unified<br />
way than using detailed models.<br />
For this reason, code calculations<br />
were performed using ORIGEN-ARP<br />
[19,20]. Consequently, the code<br />
calculations were performed mainly<br />
based on important parameters such<br />
as the fuel assembly type, initial<br />
enrichment, burnup, operation<br />
history, and cooling time of the samples.<br />
Fast et al. [5] compared the code<br />
calculation results of the ARP model<br />
(simple and fast) and the NEWT<br />
model (compli cated and precise) for a<br />
sample from Obrigheim NPP and<br />
found that the percentage differences<br />
for most nuclides are within 20 % for<br />
the APR model and within 10 % for<br />
the NEWT model. To make code<br />
calculations for a large number of<br />
samples, the use of ORIGEN-ARP<br />
provides an efficient way of calculating<br />
nuclide compositions with sufficient<br />
precision.<br />
2.4 Consideration of operation<br />
history<br />
The degree of detail in the irradiation<br />
history varies among the 14 reactors<br />
in the OECD/NEA SFCOMPO database.<br />
Very detailed information on the<br />
cycle number, elapsed time, time<br />
interval, core power density, bundle<br />
power density, and so on is provided<br />
for Cooper NPP. On the other hand, no<br />
information is available for JPDR-I,<br />
Tsuruga-1, and Takahama-3 NPPs in<br />
the SFCOMPO database. The operation<br />
history of Takahama-3 NPP is<br />
available in NUREG/CR-6798 [21].<br />
Research has found that code<br />
calculations of nuclide compositions<br />
are not significantly affected by the<br />
detailed power history. Chabert et al.<br />
[22] and Nakahara et al. [23] compared<br />
the nuclide compositions of<br />
spent fuels obtained using the accurate<br />
power history and an assumed<br />
constant power history, and found<br />
that the principal uranium and<br />
plutonium isotopes and other fission<br />
products are not significantly affected<br />
by the power history. Based on these<br />
findings, a constant power history was<br />
assumed as a reasonable approximation<br />
instead of the accurate power<br />
history, which was available for only a<br />
limited number of samples.<br />
3 Application to Obrigheim<br />
NPP Nuclide Composition<br />
Data<br />
A total of 1,153 nuclide composition<br />
or nuclide ratio data were provided<br />
for 23 samples from Obrigheim NPP.<br />
The samples were analyzed at two<br />
different laboratories, Ispra and<br />
Karlsruhe. Ispra analyzed 17 samples,<br />
and Karlsruhe analyzed 10 samples.<br />
DECOMMISSIONING AND WASTE MANAGEMENT 403<br />
Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
DECOMMISSIONING AND WASTE MANAGEMENT 404<br />
| | Fig. 2.<br />
Percentage differences between measured and computed nuclide compositions for Obrigheim NPP (measured at Ispra).<br />
Four samples were analyzed in both<br />
laboratories for cross-checking.<br />
The proposed method was applied<br />
to a total of 728 nuclide composition<br />
or nuclide ratio data for 17 samples<br />
measured at Ispra to identify those<br />
data that deviate significantly from<br />
other data and therefore became<br />
the candidates for detailed analysis.<br />
Various detailed analysis appropriate<br />
to identify the root causes of the<br />
significant deviation from other data<br />
were applied and described below.<br />
3.1 Burnup for GEROBRPWR-9<br />
Figure 2 shows the percentage differences<br />
between the calculated and<br />
measured nuclide composition data for<br />
the samples from Obrigheim NPP measured<br />
at the Ispra laboratory. Relatively<br />
high percentage differences between<br />
the calculated and measured data were<br />
observed for one sample, which<br />
was found to be the GEROBRPWR-9<br />
sample. According to Barbero et al.<br />
[24], the burnup for the GEROBRPWR-9<br />
sample measured at Karlsruhe using<br />
the Nd-148 method (22,700 MWd/<br />
MTU) was abnormally high compared<br />
to the burnup measured at Ispra using<br />
the Nd-148, non-destructive Cs-137,<br />
and destructive Cs-137 methods<br />
(17,130, 16,970, and 17,490 MWd/<br />
MTU, respectively). Because the<br />
burnup in the OECD/NEA SFCOMPO<br />
database was based on the measurement<br />
at Karlsruhe, the code calculation<br />
was performed again using the burnup<br />
measured at Ispra with the Nd-148<br />
method (17,130 MWd/MTU). Figure 3<br />
shows the percentage differences<br />
after the burnup correction for the<br />
GEROBRPWR-9 sample. It can be<br />
seen that the deviations of the<br />
GEROBRPWR-9 sample from other<br />
samples were properly corrected.<br />
Figure 3 also identifies those<br />
nuclide composition or nuclide ratio<br />
data that need to be analyzed in more<br />
detail. The Pu-241/Pu-239 ratio<br />
shows relatively high percentage<br />
differences. The percentage differences<br />
of Cs-137/U-238 ratio are<br />
divided into two groups. A large<br />
uncertainty and a large deviation in<br />
percentage differences are observed<br />
for Am-241 and Am-242, respectively.<br />
Detailed analysis on each of the<br />
identified nuclide composition or<br />
nuclide ratio data are described in the<br />
following sections.<br />
3.2 Cooling time of plutonium<br />
isotopes<br />
As indicated in Fig. 3, the Pu-241/<br />
Pu-239 ratio was found to have<br />
higher percentage differences among<br />
samples than other nuclide ratios.<br />
Barbero et al. [24], the original source<br />
of the measurement data, reported<br />
the nuclide ratios of uranium and<br />
plutonium with the dates of the<br />
measurements. The actual cooling time<br />
of the samples can be calculated from<br />
the date of discharge (August 16, 1974)<br />
and the date of measurement (e.g.,<br />
April 12, 1978 for GEROBRPWR-3).<br />
However, the cooling times of the<br />
data were specified as zero in the<br />
OECD/NEA SFCOMPO database,<br />
which means that the data were<br />
adjusted to the time of discharge. The<br />
code calculations were performed<br />
assuming that the cooling times of<br />
the samples were zero. Because the<br />
half-life of Pu-241 (14.325 years) is<br />
comparable with the cooling time of<br />
the samples, a non-negligible amount<br />
of Pu-241 decayed out; therefore, the<br />
calculated Pu-241/Pu-239 ratios were<br />
higher than the measured ones.<br />
| | Fig. 3.<br />
Percentage differences between measured and computed nuclide compositions for Obrigheim NPP (measured at Ispra) after burnup<br />
correction for GEROBRPWR-9 sample.<br />
3.3 Possible errors during<br />
Cs-137/U-238 ratio calculation<br />
from measured data<br />
As indicated in Fig. 3, the Cs-137/<br />
U-238 ratios were found to fall into<br />
two groups. One group consists of 5<br />
samples with very small percentage<br />
differences, and the other group consists<br />
of 12 samples with percentage<br />
differences of up to −35 %. The existence<br />
of two distinct groups motivated<br />
a detailed analysis considering possible<br />
systematic errors in the data.<br />
Figure 4 shows the amounts of<br />
Cs-137 buildup and the remaining<br />
U-238 in the spent nuclear fuel per<br />
metric ton of final uranium. As a<br />
burnup monitor, the amount of Cs-137<br />
buildup shows a linear relationship<br />
with the burnup. Although the<br />
amount of U-238 remaining shows a<br />
Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 4.<br />
Cs-137 buildup and remaining U-238 as functions of burnup for Obrigheim NPP<br />
(measured at Ispra).<br />
very slight decreasing trend as the<br />
burnup increases, it can be considered<br />
as constant at around 950 kg/MTU.<br />
Therefore, the Cs-137/U-238 ratio is<br />
expected to show a linear relationship<br />
with the burnup.<br />
Figure 5 shows the relationship<br />
between the Cs-137/U-238 ratio and<br />
the burnup for the data in the OECD/<br />
NEA SFCOMPO database for<br />
Obrigheim NPP measured at Ispra. It<br />
was first confirmed that the data in<br />
the SFCOMPO database are identical<br />
with those in the original source,<br />
Barbero et al. [24]. It can be seen that<br />
two distinct relationships exist in<br />
Fig. 5. The lower part is the 5 samples<br />
with very small percentage differences.<br />
The upper part is the 12<br />
samples with percentage differences<br />
of up to −35 %, which means that the<br />
measured quantities are about 50 %<br />
higher than the calculated quantities.<br />
Because the half-life of Cs-137 (30.08<br />
years) is much shorter than that of<br />
U-238 (4.47E+09 years), the Cs-137/<br />
U-238 ratio is expected to decrease<br />
gradually as time passes; therefore,<br />
the cooling time would not make the<br />
measured quantities much higher<br />
than the calculated quantities, which<br />
assumed zero cooling time.<br />
Table 1 compares the Cs-137/<br />
U-238 data obtained directly from the<br />
OECD/NEA SFCOMPO database, the<br />
data calculated from the Cs-137 and<br />
U-238 quantities provided in the database,<br />
and the data calculated using the<br />
code. From Fig. 4, the isotopic quantities<br />
of Cs-137 and U-238 from the<br />
database seem to be free of error;<br />
hence, the Cs-137/U-238 ratio from<br />
| | Fig. 5.<br />
Cs-137/U-238 as a function of burnup for Obrigheim NPP (measured at Ispra).<br />
the two isotopic quantities is expected<br />
to provide correct results. In Tab. 1, the<br />
‘Percentage difference I’ column lists<br />
the difference between the Cs-137/<br />
U-238 values obtained directly from<br />
the database and those calculated from<br />
the nuclide composition data for the<br />
two nuclides (Cs-137, U-238) divided<br />
by the Cs-137/U-238 values taken<br />
directly from the database. The nuclide<br />
composition data of the two nuclides in<br />
Tab. 1 are given in kilograms per<br />
metric ton of final uranium (kg/<br />
MTU final). The 12 data points with<br />
percentage differences of up to −35 %<br />
also appear in the ‘Percentage difference<br />
I’ column.<br />
Code calculation results obtained<br />
using ORIGEN-ARP are also provided<br />
in Tab. 1, with the associated percentage<br />
difference denoted as Percentage<br />
DECOMMISSIONING AND WASTE MANAGEMENT 405<br />
Burnup<br />
(MWd/<br />
MTU)<br />
Cs-137<br />
(SFCOMPO)<br />
U-238<br />
(SFCOMPO)<br />
Cs-137/<br />
U-238<br />
(SFCOMPO)<br />
Cs-137/<br />
U-238<br />
(Calculation)<br />
Percentage<br />
difference I<br />
(Calculation)<br />
Cs-137/<br />
U-238<br />
(Code)<br />
Percentage<br />
difference II<br />
(Code)<br />
Percentage<br />
difference III<br />
(Calc. vs. Code)<br />
GEROBRPWR-2 27,900 1.88.E-03 1.88.E-03 0.24 %<br />
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 %<br />
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 %<br />
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 %<br />
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 %<br />
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 %<br />
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 %<br />
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 %<br />
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 %<br />
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 %<br />
GEROBRPWR-15 29,400 1.<strong>06</strong>.E+00 948 3.01.E-03 1.94.E-03 −35.47 % 1.98.E-03 −34.09 % 2.14 %<br />
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 %<br />
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 %<br />
GEROBRPWR-19 30,200 1.12.E+00 951 3.20.E-03 2.05.E-03 −36.<strong>06</strong> % 2.<strong>06</strong>.E-03 −35.68 % 0.60 %<br />
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 %<br />
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 %<br />
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 %<br />
| | Tab. 1.<br />
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.<br />
Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
DECOMMISSIONING AND WASTE MANAGEMENT 4<strong>06</strong><br />
| | Fig. 6.<br />
Correlation of measured and calculated Am-241 data with burnup for BE124 and BE210 fuel assemblies<br />
of Obrigheim NPP.<br />
difference II. The definition of<br />
Percentage difference II is similar to<br />
that of Percentage difference I, except<br />
that the code calculation results<br />
replace the Cs-137/U-238 values calculated<br />
from the nuclide composition<br />
data for the two nuclides (Cs-137,<br />
U-238). The 12 data points with percentage<br />
differences of up to −35 %<br />
can also be seen in the ‘Percentage<br />
difference II’ column.<br />
The ‘Percentage difference III’<br />
column lists the differences between<br />
the code calculation results and the<br />
Cs-137/U-238 values calculated from<br />
the nuclide composition data for the<br />
two nuclides (Cs-137, U-238) divided<br />
by the calculated Cs-137/U-238<br />
values. The magnitudes of the percentage<br />
differences in this column<br />
are less than 5 %, except for one<br />
case. Therefore, in the OECD/NEA<br />
SFCOMPO database, the nuclide<br />
composition data for these two<br />
nuclides (Cs-137, U-238) seem to be<br />
trustworthy, whereas the Cs-137/<br />
U-238 data taken directly from<br />
the database can be suspected<br />
of including calculation errors.<br />
Recalculation of the Cs-137/U-238<br />
data in the OECD/NEA SFCOMPO<br />
database using the measured data<br />
for the two nuclides (Cs-137, U-238)<br />
is recommended.<br />
in the percentage differences for<br />
Am-241 (up to 85 %) and Cm-242<br />
(70 % for Karlsruhe and 45 % for<br />
Ispra). Gauld et al. [12] also indicated<br />
possible problems in back-calculating<br />
isotopic compositions to a reference<br />
date, using the measurement data<br />
of Am-241 for Takahama-3 NPP as<br />
an example. Note that most of<br />
the Am-241 (half-life = 432.6 years)<br />
at the time of measurement is from<br />
the decay of Pu-241 (half-life =<br />
14.325 years).<br />
Figure 6 shows the correlation of<br />
the measured and calculated Am-241<br />
data with the burnup for the BE124<br />
and BE210 fuel assemblies at<br />
Obrigheim NPP. First, it was confirmed<br />
that the data provided in the OECD/<br />
NEA SFCOMPO database are identical<br />
with those provided in the original<br />
data source, Barbero et al. [24].<br />
Although a gradual increase in<br />
Am-241 quantities is expected as the<br />
burnup increases, as can be seen in<br />
the calculated data, a large variation<br />
is observed in the measured data<br />
as the burnup increases, especially<br />
when the burnup is 31.5 GWd/MTU<br />
( GEROBRPWR-11). Some measured<br />
data, especially those within the<br />
ellipse in Fig. 6, appear to be inconsistent<br />
with other measured data.<br />
The measured Am-241 data were<br />
obtained by mass spectrometry or<br />
alpha spectrometry at Ispra or alpha<br />
spectrometry at Karlsruhe. All of the<br />
measured data that were found to be<br />
larger (sometimes significantly larger)<br />
than the calculated data, which are<br />
indicated by an ellipse in Fig. 6, were<br />
obtained by alpha spectrometry at<br />
Ispra. In addition, 8 of the 10 values<br />
measured by alpha spectrometry at<br />
Ispra are found in the ellipse. The<br />
measurement data obtained by mass<br />
spectrometry at Ispra and alpha<br />
spectrometry at Karlsruhe are found<br />
to be relatively close to the calculated<br />
data. Owing to the relatively high<br />
inaccuracy of the measured data<br />
obtained by alpha spectrometry at<br />
Ispra compared to the measured data<br />
obtained by mass spectrometry<br />
at Ispra and alpha spectrometry at<br />
Karlsruhe, as shown in Fig. 6, a<br />
detailed review of the measured data<br />
obtained by alpha spectrometry at<br />
Ispra seems to be necessary.<br />
Figure 7 shows the correlation of<br />
the measured and calculated Am-242<br />
data from Obrigheim NPP with the<br />
burnup. Note that the scale of the<br />
measured data (left-hand Y-axis) is 10<br />
times larger than that of the calculated<br />
data (right-hand Y-axis). Figure 7<br />
shows that the measured Am-242<br />
data are about 10 times larger than<br />
the calculated data for the seven<br />
samples from Obrigheim NPP.<br />
Because Cm-242 is produced by the<br />
decay of Am-242, the ratio of the halflives<br />
of Am-242 and Cm-242 provides<br />
information on the atomic ratio of the<br />
3.4 Large uncertainty in<br />
Am-241 and large<br />
deviation in Am-242<br />
As indicated in Fig. 3, a large uncertainty<br />
between the measured and<br />
calculated data for Am-241 and a<br />
large deviation between the measured<br />
and calculated data for Am-242 were<br />
observed. In a comparison of the<br />
measured and computed data for a<br />
sample from Obrigheim NPP, Fast et<br />
al. [5] also observed a high deviation<br />
| | Fig. 7.<br />
Correlation of measured and calculated Am-242 data from Obrigheim NPP with burnup.<br />
Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 8.<br />
Ratios of measured Cm-242 data to measured and calculated Am-242 data for the seven samples from<br />
Obrigheim NPP.<br />
two nuclides in equilibrium. The halflives<br />
of Am-242 and Cm-242 are<br />
16.02 hours and 162.8 days, respectively,<br />
and 82.70 % of Am-242 goes<br />
through beta decay to form Cm-242.<br />
Therefore, the measured values for<br />
Cm-242 are expected to be about<br />
200 times larger than the measured<br />
values for Am-242 in equilibrium.<br />
Figure 8 shows the ratios of the measured<br />
Cm-242 values to the measured<br />
and calculated Am-242 values for the<br />
seven samples from Obrigheim NPP.<br />
Although the ratio of the Cm-242<br />
values to the calculated Am-242 values<br />
is around 150, the ratio of the Cm-242<br />
values to the measured Am-242 values<br />
is generally less than 20. Therefore, it<br />
is likely that the measured Am-242<br />
values overestimate the actual quantity<br />
of Am-242 by about 10 times. Detailed<br />
analysis of how the measured Am-242<br />
values were derived from the raw<br />
experimental data to the data provided<br />
in the OECD/NEA SFCOMPO database<br />
seems to be necessary.<br />
4 Conclusions<br />
Nuclide composition data of spent<br />
nuclear fuels such as those provided in<br />
the OECD/NEA SFCOMPO database<br />
are important in many fields including<br />
reactor physics, fuel cycle applications,<br />
radiological consequence<br />
analysis, and nuclear forensics. To<br />
reduce unnecessary uncertainties<br />
associated with nuclide composition<br />
data, the validation of such data is a<br />
high-priority task.<br />
As one of the first steps, a simple<br />
method is proposed for identifying<br />
the nuclide composition data that<br />
may include errors and therefore<br />
require detailed analysis or further<br />
investigation. The proposed method<br />
is based on the ORIGEN-ARP code<br />
calculation, the assumption of a<br />
constant power history, the percentage<br />
differences of the calculated and<br />
measured composition data, and<br />
detailed analysis of the identified<br />
data. The application of the proposed<br />
method to the nuclide composition<br />
data of spent nuclear fuels from<br />
Obrigheim NPP demonstrated that<br />
the method can effectively identify<br />
various possible errors or data that<br />
need to be further investigated. Errors<br />
identified during detailed analysis<br />
or possible errors that require further<br />
investigation include:<br />
• Errors in burnup measurement<br />
(e.g., GEROBRPWR-9)<br />
• Errors in properly considering the<br />
cooling time (e.g., Pu-241/Pu-239)<br />
• Errors in the ratio calculation from<br />
measured data (e.g., Cs-137/<br />
U-238)<br />
• Possible systematic errors in<br />
measurements of isotopic composition<br />
(e.g., Am-241 and Am-242<br />
measurements at Ispra)<br />
Although the nuclide composition<br />
data that were not identified as<br />
needing detailed analysis or further<br />
investigation cannot necessarily be<br />
considered as definitively validated, it<br />
is believed that the proposed method<br />
can identify a significant portion of<br />
the errors in the nuclide composition<br />
data. Despite the simplicity of the<br />
proposed method, it is believed to be a<br />
very efficient method of identifying<br />
those nuclide composition data that<br />
require detailed analysis or further<br />
investigation. For this reason, the proposed<br />
method is expected to be useful<br />
as the first step in validation of nuclide<br />
composition data of spent nuclear<br />
fuels such as those in the OECD/NEA<br />
SFCOMPO database.<br />
Acknowledgements<br />
This work was supported by a grant<br />
from the Nuclear Safety Research<br />
Program of the Korea Foundation of<br />
Nuclear Safety, with funding from the<br />
Korean government's Nuclear Safety<br />
and Security Commission. This work<br />
was also supported by a grant from<br />
the Nuclear Research & Development<br />
Program of the National Research<br />
Foundation of Korea, which is funded<br />
by the Korean government's Ministry<br />
of Science, ICT & Future Planning<br />
(Grant Code: <br />
NRF-2016M2B2A9A02945211).<br />
References<br />
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[2] Masayoshi Kurosawa, Yoshitaka Naito,<br />
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[3] H. Mochizuki, K. Suyama, Y. Nomura,<br />
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[6] G. Nicolaou. Discrimination of spent<br />
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[7] Kenya Suyama, Ali Nouri, Hirold<br />
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[8] Ian C. Gauld, Yolanda Rugama. Activities<br />
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Decommissioning and Waste Management<br />
Validation of Spent Nuclear Fuel Nuclide Composition Data Using Percentage Differences and Detailed Analysis ı Man Cheol Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
408<br />
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[9] Expert Group on Assay Data of Spent<br />
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[10] Kenya Suyama, Minoru Murazaki,<br />
Kiyoshi Ohkubo, Yoshinori Nakahara,<br />
Gunzo Uchiyama. Re-evaluation of<br />
Assay Data of Spent Nuclear Fuel<br />
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Research Institute for validation of<br />
burnup calculation code systems,<br />
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[11] M.C. Brady Raap, B.A. Collins, J.A. Lyons,<br />
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on the Augmentation of the Spent Fuel<br />
Composition Dataset for Nuclear<br />
Forensics: SFCOMPO/NF, PNNL-23225,<br />
Pacific Northwest National Laboratory,<br />
Richland, Washington, March 2014.<br />
[12] Ian C. Gauld, Georgeta Radulescu,<br />
Germina Ilas. SCALE Validation<br />
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Assay Database for Spent Nuclear Fuel,<br />
Proceedings of the International<br />
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[13] Keisuke Okumura, Shiho Asai, Yukiko<br />
Hanzawa, Hideya Suzuki, Masaaki<br />
Toshimitsu, Jun Inagawa, Tsutomu<br />
Okamoto, Nobuo Shinohara, Satoru<br />
Kaneko, Kensuke Suzuki. Analyses of<br />
Assay Data of LWR Spent Nuclear Fuels<br />
with a Continuous-Energy Monte Carlo<br />
Code MVP and JENDL-4.0 for Inventory<br />
Estimation of 79Se, 99Tc, 126Sn and<br />
135Cs, Progress in NUCLEAR SCIENCE<br />
and TECHNOLOGY, Vol. 2, pp.369-374,<br />
2011.<br />
[14] Philippe Bienvenu, Philippe Cassette,<br />
Gilbert Andreoletti, Marie-Martine Bé,<br />
Jérôme Comte, Marie-Christine Lépy. A<br />
new determination of 79 Se half-life,<br />
Applied Radiation and Isotopes, vol.65,<br />
355-364, 2007.<br />
[15] O. W. Hermann, M. D. DeHart, and B. D.<br />
Murphy. Evaluation of measured LWR<br />
spent fuel composition data for use in<br />
code validation, ORNL/M-6121, Oak<br />
Ridge National Laboratory, Oak Ridge,<br />
Tennessee, February 1998.<br />
[16] O. W. Hermann, S. M. Bowman, M. C.<br />
Brady, C. V. Parks. Validation of the<br />
SCALE System for PWR Spent Fuel<br />
Nuclide composition Analyses, ORNL/<br />
TM-12667, Oak Ridge National Laboratory,<br />
Oak Ridge, TN, March 1995.<br />
[17] M. D. DeHart, O. W. Hermann. An<br />
Extension of the Validation of SCALE<br />
(SAS2H) Isotopic Predictions for PWR<br />
Spent Fuel, ORNL/TM-13317, Oak<br />
Ridge National Laboratory, Oak Ridge,<br />
TN, September 1996.<br />
[18] Jeong-nam Jang, Hyung-moon Kwon,<br />
Jung-suk Kim, Yong-bum Chun.<br />
Validation of SCALE SAS2H Isotopic<br />
Predictions for high burnup PWR spent<br />
fuels, Transactions of the 2009 Korean<br />
Nuclear Society Spring Meeting, Jeju,<br />
Korea, May 22, 2009.<br />
[19] M. J. Bell. ORIGEN – the ORNL isotope<br />
generation and depletion code,<br />
ORNL-4628, Oak Ridge National Laboratory,<br />
Oak Ride, Tennessee, May 1973.<br />
[20] http://scale.ornl.gov/origen-arp.shtml<br />
[21] C. E. Sanders, L C. Gauld, R. Y. Lee.<br />
Isotopic Analysis of High-Burnup PWR<br />
Spent Fuel Samples From the<br />
Takahama-3 Reactor, NUREG/CR-6798,<br />
ORNL/TM-2001/259, United States<br />
Nuclear Regulatory Commission,<br />
Washington, DC, January 2003.<br />
[22] Christine Chabert, Alain Santamarina,<br />
Robin Dorel, Didier Biron, Christine<br />
Poinot-Salanon. Qualification of the<br />
APOLLO 2 assembly code using PWR-<br />
UO2 isotopic assays – the importance of<br />
irradiation history and thermomechanics<br />
onfuel inventory prediction,<br />
Proceedings of the American Nuclear<br />
Society International Topical Meeting<br />
on Advances in Reactor Physics, and<br />
Mathematics and Computation Into<br />
the Next Millennium (PHYSOR-2000),<br />
Pittsburgh, Pennsylvania, May 7-11,<br />
2000.<br />
[23] Yoshinori Nakahara, Kenya Suyama,<br />
and Takenori Suzaki. Technical<br />
Development on Burn-up Credit for<br />
Spent LWR Fuels, (Eds.), JAERI-Tech<br />
2000-071, Japan Atomic Energy<br />
Research Institute (JAERI), 2000 (in<br />
Japanese). Translation published as<br />
ORNL/TR-2001/01, Oak Ridge National<br />
Laboratory, 2002.<br />
[24] P.Barbero et.al. Post Irradiation Analysis<br />
of The Obrigheim PWR Spent Fuel.<br />
Nuclear Science and Technology, 1980.<br />
Figure captions<br />
Author<br />
Man Cheol Kim<br />
School of Energy Systems<br />
Engineering<br />
Chung-Ang University<br />
84 Heukseok-ro<br />
Dongjak-gu, Seoul <strong>06</strong>974, Korea<br />
Reliability Analysis on Passive Residual<br />
Heat Removal of AP1000 Based on Grey<br />
Model<br />
Qi Shi, Zhou Tao, Muhammad Ali Shahzad, Li Yu and Jiang Guangming<br />
1 Introduction It is common to base the design of passive systems [1, 2] on the natural laws of physics, such<br />
as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated<br />
with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating<br />
conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can<br />
affect the function of the system, and the failure to achieve its desired aim is termed as function failure [3].<br />
In the reliability analysis of the passive<br />
systems, the main sources of the<br />
uncertainty [4] are the numerical<br />
errors in the calculation program such<br />
as RELAP5 and the reactor parameters.<br />
However, a lot of experience is required<br />
to analyze the error propagation in<br />
such system codes. The difficult is<br />
increased by the fact that AP1000 has<br />
not been connected to the grid yet. In<br />
this paper, more focus has been placed<br />
on the uncertainties of design and<br />
operation parameters of the reactor.<br />
The analytic hierarchy process (AHP)<br />
[5, 6] and artificial neural network<br />
(ANN) [7] have been applied, in order<br />
to perform a sensitivity analysis on different<br />
parameters of the passive safety<br />
systems. However, there are large<br />
subjective qualitative considerations in<br />
the AHP. On the other hand, ANN has a<br />
large amount of randomness, thus<br />
requiring a large amount of data for its<br />
training. Hence, these methods have<br />
many limitations. The grey correlation<br />
method [8]-[9], which has been<br />
applied in many fields, can make up for<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 1.<br />
Passive Residual Heat Removal System.<br />
the limitations of above statistical<br />
methods. It does not require a large<br />
amount of data and its results<br />
are consistent with the qualitative<br />
implications. For sensitivity analysis on<br />
PRHRS, it is rare to find the application<br />
of Grey correlation method in literature.<br />
The grey derivative and differential<br />
equations are defined in the grey<br />
system model [10] to establish the<br />
dynamic prediction, based on concepts<br />
of space relevance and smooth discrete<br />
function. It is used to forecast the<br />
passive system parameters.<br />
For AP1000, the loss of normal<br />
feedwater accident has been taken as<br />
an example in this paper, which<br />
involves the drop of control rod, operation<br />
of PRHRS, coolant pump outage<br />
and so on. Grey correlation is used to<br />
analyze the importance of influencing<br />
factors on PRHRS, and the Grey model<br />
plays an important role for predicting<br />
the data. This provides a new viewpoint<br />
for studying the PRHRS of<br />
AP1000.<br />
2 Research object<br />
2.1 System description<br />
As an important part of AP1000<br />
passive core cooling system, PRHRS<br />
[11] is used to remove the decay heat<br />
of the core for ensuring the safety of<br />
reactor during the accident operating<br />
conditions. It consists of a ‘C’ type<br />
heat exchanger, an in-containment<br />
refueling water storage tank (IRWST)<br />
and the corresponding pipes and<br />
valves.<br />
Figure 1 shows the layout of<br />
PRHRS in AP1000. The PRHRS heat<br />
exchanger is located at a higher elevation<br />
than the reactor core. Its upper<br />
head is connected to the hot leg of<br />
reactor coolant system (RCS) and its<br />
lower head is connected to the lower<br />
head of the steam generator. There is<br />
an electro valve normally in the open<br />
state located on the inlet line. Two<br />
parallel pneumatic valves normally in<br />
the close state are located on the<br />
outlet line. Once the accident takes<br />
place, the pneumatic valves are<br />
opened by pneumatic signal, and the<br />
electro valve also receives a signal to<br />
confirm its open state. Corresponding,<br />
a completely natural circulation loop<br />
is established and the decay heat is<br />
removed by the IRWST using density<br />
difference.<br />
It is complex to gage the factors,<br />
affecting the heat transfer capacity of<br />
PRHRS, and its interactions with RCS.<br />
Hence, the parametric uncertainties<br />
have different effects on PRHRS under<br />
different accident conditions. Loss of<br />
normal feedwater accident belongs to<br />
the second type accidents, meaning<br />
medium frequency accident. When<br />
the accident takes place, the decay<br />
heat needs to be adequately removed<br />
from the reactor core. Otherwise, the<br />
reactor core may be damaged. In this<br />
paper, the influence of parametric<br />
uncertainties on PRHRS has been<br />
studied under the loss of normal feedwater<br />
accident.<br />
X 1<br />
X 2<br />
| | Tab. 1.<br />
Parameters and corresponding distribution.<br />
2.2 Identification and quantification<br />
of uncertainties<br />
PRHRS relies on natural circulation,<br />
which has a much weaker driving<br />
force than the active systems. A small<br />
deviation from the design and operating<br />
conditions can lead to function<br />
failure. According to the criterion of<br />
International Atomic Energy Agency<br />
(IAEA) [13], the PRHRS is considered<br />
to fail in providing its safety function,<br />
once the maximum coolant temperature<br />
at the reactor outlet exceeds<br />
beyond 350 °C, in order to avoid fuel<br />
cladding damage. Only the epistemic<br />
uncertainties have been considered in<br />
the present analysis. In accordance<br />
with the previous studies [14, 15],<br />
some design and operating parameters<br />
have been selected for analysis<br />
together with the corresponding<br />
distribution as shown in Table. 1.<br />
2.3 Thermal hydraulic model<br />
and verification<br />
According to the current research<br />
results [12], RELAP5 program can be<br />
used to analyze the loss of normal<br />
feedwater accident in AP1000. It has<br />
been developed by Idaho National<br />
Engineering and Enviromental Laboratory<br />
for transient thermal hydraulic<br />
analysis in light water reactors. In this<br />
paper, RELAP5/MOD3.4 has been<br />
used for the calculation. The nodalization<br />
scheme of AP1000 is shown in<br />
Figure 2.<br />
As depicted in Fig. 2, the model<br />
consists of reactor core, two steam<br />
generators, pressurizer, reactor coolant<br />
pump, PRHRS core makeup tank<br />
(CMT) and so on. Some preliminary<br />
calculations are performed to determine<br />
the steady-state parameters of<br />
AP1000 using RELAP5, in order to<br />
ensure their consistency with the<br />
reference standards. According to<br />
the sequence events in transient<br />
con ditions, some factors have been<br />
Variable Description Average value Standard deviation Distribution<br />
T<br />
(K)<br />
D<br />
(mm)<br />
Temperature<br />
of IRWST<br />
Diameter<br />
of PRHR HX<br />
X 3 Kin<br />
Resistance<br />
coefficient of inlet<br />
X 4<br />
X 5<br />
X 6<br />
H<br />
(m)<br />
P<br />
(MPa)<br />
Q<br />
(MW)<br />
Height of ascending<br />
Pipeline<br />
300 10 Normal<br />
0.0162 0.002 Normal<br />
50 17 Normal<br />
9.0 0.33 Normal<br />
Initial pressure 15.5 0.5 Normal<br />
Initial power level 3415 11.6 Normal<br />
RESEARCH AND INNOVATION 409<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
RESEARCH AND INNOVATION 410<br />
| | Fig. 2.<br />
Nodalization for the primary system of AP1000.<br />
compared with the results of<br />
LOFTRAN, such as pressure, flow rate<br />
etc. The comparison results show that<br />
RELAP5 has the capability to predict<br />
the system parameters [11] correctly.<br />
Among above these parameters, the<br />
reactor coolant temperature is the<br />
most important parameter for the<br />
PRHRS loop, as shown in Figure 3.<br />
As depicted in Fig. 3, the results of<br />
RELAP5 and LOFTRAN exhibit the<br />
same trend. A similar value of maximum<br />
temperature has been observed<br />
in the two results. During the accident,<br />
however, a difference in the sequence<br />
of events and response of the reactor<br />
control system can lead to a slight<br />
difference in the temperature trend.<br />
This has no significant influence on<br />
the analysis.<br />
3 Calculation methods<br />
3.1 Latin hypercube sampling<br />
Latin Hypercube Sampling (LHS) [16,<br />
17] is an improvement over the traditional<br />
Monte Carlo sampling method.<br />
It can overcome its drawbacks, in that<br />
most of sampled results lie near the<br />
average value. In order to improve the<br />
accuracy of the parameters, therefore,<br />
the method covers the upper and<br />
lower limits of the distributions.<br />
Hence, this method has the ability to<br />
determine any value, as long as the<br />
parameters are known. The steps are<br />
as follows.<br />
(1) For each variable, the probability<br />
distribution is divided into N nonoverlapping<br />
equal probability<br />
interval [0, 1/N], [1/N, 2/N],…,<br />
[(N-1)/N, 1]. This ensures that<br />
the degree of the correlation of<br />
LHS is small.<br />
(2) The random standard normal<br />
sample matrix Z N×n is used to<br />
represent the order of sample<br />
points, and the integer matrix<br />
R N×n is used to record information<br />
regarding the ordering of the<br />
above sample points. Hence,<br />
R ij = k shows that the sequence<br />
of the j th variable in the i th sampling<br />
is k.<br />
(3) According to in each interval, the<br />
cumulative probability function<br />
of each sample point in the Latin<br />
hypercube can be obtained randomly,<br />
as shown in Eq. (1).<br />
(1)<br />
Here, i = 1,..., N and j = 1,..., n.<br />
The function r a n d (0,1) represents<br />
a random number, which is<br />
uniformly distributed in the [0,1]<br />
interval.<br />
(4) In the Latin hypercube, the sampling<br />
points are obtained by the<br />
method of equal probability<br />
change, as shown in Eq. (2).<br />
(2)<br />
Here, φ –1 (.) is the inverse normal<br />
distribution function.<br />
3.2 Grey Relation Method<br />
The Grey Relational Method [8] is a<br />
quantitative technique for comparative<br />
analysis. The basic idea is to determine<br />
the exponent of each factor in<br />
the correlation, according to their<br />
degree of similarity with the geometry<br />
of sequence curve. If the curve is close<br />
for a particular factor, it would have a<br />
high exponent in the correlation. X 0 is<br />
defined as the target parameter, with<br />
k referring to the sequence of the<br />
parameter, denoted as {X 0 (k)}. It is<br />
assumed that there are a total of m<br />
control parameters, and a parameter j<br />
in the same sequence k is called the<br />
comparison sequence, denoted as<br />
{X j (k)} (k = 1,…, N)(j = 1,…, m). In<br />
this correlation, the exponent of each<br />
factor can be obtained by comparing<br />
the tendency of development between<br />
the target and influence parameters.<br />
These steps are shown as follows [18].<br />
(1) The reference and comparison sequences<br />
are normalized, as shown<br />
in Eq. (3).<br />
(3)<br />
(2) The absolute value of difference<br />
between the reference and the<br />
comparison sequences is calculated<br />
as Eq. (4) based on above normalization<br />
results.<br />
(4)<br />
a) Results of RELAP5. b) Results of LOFTRAN.<br />
| | Fig. 3.<br />
Reactor coolant temperature.<br />
(3) Maximum and minimum absolute<br />
values are calculated shown as<br />
Eq. (5)-(6).<br />
(5)<br />
(6)<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
Where, the Δ max is the maximum<br />
value of the absolute difference<br />
from m control parameters in<br />
accordance with j = 1,…, m after<br />
finding N number of maximum<br />
absolute differences in the j-type<br />
control parameter. The procedure<br />
of determining Δ min is similar<br />
to Δ max .<br />
(4) The correlation coefficients<br />
between reference and comparison<br />
sequences are calculated by using<br />
Eq. (7).<br />
(7)<br />
Where, ρ is the resolution ratio in<br />
the (0,1) interval.<br />
(5) The degree of correlation γ 0j is<br />
determined, as shown in Eq. (8).<br />
(8)<br />
The corresponding differential equation<br />
of GM(1,h) is given as<br />
(13)<br />
For determining the â and YN parameters,<br />
the general equations are given<br />
as follows.<br />
(14)<br />
(15)<br />
(16)<br />
(17)<br />
Hence, Eq. (12) can be solved by using<br />
Eq. (15), (16), (17)<br />
| | Fig. 4a.<br />
Natural circulation mass flow.<br />
RESEARCH AND INNOVATION 411<br />
3.3 Grey Model Method<br />
In the grey system theory [10],<br />
the grey derivative and differential<br />
equations are defined to establish<br />
dynamic prediction model based on<br />
the concepts of space relevance and<br />
smooth discrete function. This model<br />
(GM) is used to forecast the passive<br />
system parameters. The general<br />
equation of grey model is written as<br />
GM(n, h), where the variable h is<br />
expressed by n th –order differential<br />
equation. The procedure is discussed<br />
as follows.<br />
The correlation sequence is calculated<br />
as follows.<br />
(9)<br />
Corresponding accumulative value<br />
sequence is calculated by<br />
Where<br />
(10)<br />
The generated sequences of corresponding<br />
means value is calculated<br />
by<br />
(11)<br />
The algebraic equation of GM(1,h) is<br />
written as<br />
(12)<br />
4 Results and analysis<br />
(18)<br />
4.1 Parameter uncertainties<br />
For correlation analysis, Mendenhall<br />
[19] reports that the sample sizes are<br />
5 to 10 times larger than the variable.<br />
In this paper, the key parameters,<br />
which affect the coolant temperature,<br />
have been shown in Tab. 1 sampled by<br />
LHS with a size of 100. After identification<br />
of 100 combinations for the key<br />
parameters, RELAP5 model as verified<br />
in Section 2.3 is used for the analysis.<br />
Figue 4 shows 25 groups of parametric<br />
uncertainties, which influence<br />
the natural circulation flow rate and<br />
coolant temperature at the outlet of<br />
the reactor core.<br />
As seen from Fig. 4a, the PRHRS<br />
actuates at 400 s with an initial mass<br />
flow rate of 250 kg/s. At 1400 s, the<br />
reactor coolant pimp stops and the<br />
mass flow decreases to 125 kg/s. To<br />
drive the natural circulation system,<br />
PRHRS relies on the difference of<br />
density between cold and hot sections<br />
to take away the decay heat from<br />
the reactor core. After actuation of<br />
the reactor safety systems, the core<br />
temperature decreases and there is a<br />
rise in the IRWST temperature. During<br />
this time, there is a lesser density<br />
difference which leads to a decrease in<br />
the mass flow of PRHRS. The mass<br />
flow is maintained until the decay<br />
heat power is balanced with the<br />
| | Fig. 4b.<br />
Coolant outlet temperature.<br />
cooling capacity. There are great<br />
changes in the mass flow during<br />
natural circulation due to uncertainty<br />
in the parameters. This has an effect<br />
on the coolant temperature.<br />
As seen from Fig. 4b, the PRHRS<br />
starts at 400 s with an initial temperature<br />
of 560 K. The CMT is actuated at<br />
1,500 s together with the corresponding<br />
systems, limiting the outlet<br />
temperature to 550 K. At 6,000 s, this<br />
temperature decreases to about 530 K.<br />
There is a 50 K variation in the outlet<br />
temperature of the coolant, owing to<br />
uncertainty in the parameters.<br />
4.2 Grey correlation analysis<br />
As mentioned before, maximum outlet<br />
temperature of the coolant should<br />
always be kept below 350 °C, in order<br />
to avoid function failure of PRHRS.<br />
Assuming X 0 as the maximum outlet<br />
temperature of the coolant, X 1 as the<br />
temperature of IRWST, X 2 as the<br />
diameter of PRHR HX, X 3 as the<br />
resistance coefficient of inlet, X 4 as<br />
height of ascending pipe, X 5 as initial<br />
pressure, and X 6 as Initial power level,<br />
the correlation is created using<br />
RELAP5. Eq. (3)-(8) are used for<br />
the grey correlation with resolution<br />
coefficients of 0.1, 0.2, and 0.5. The<br />
results are shown in Figure 5.<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
RESEARCH AND INNOVATION 412<br />
| | Fig. 5.<br />
Correlation degree in different resolution.<br />
As seen from Fig. 5, using different<br />
resolutions, the factors have different<br />
effect on the maximum outlet temperature<br />
of the coolant. There is a<br />
little difference between the effects of<br />
different factors, when the resolution<br />
is 0.5. It shows that the above<br />
parameters are not well distributed.<br />
The individual characteristics of<br />
the above parameters are gradually<br />
distinguished, once the resolution is<br />
reduced from 0.5 to 0.1. The order of<br />
these parameters is established, based<br />
on their degree of the importance, X 6 ,<br />
X 1 , X 4 , X 5 , X 2 , X 3 . The initial power has<br />
the strongest influence on the decay<br />
heat after shutdown. This is followed<br />
by temperature of IRWST, which is<br />
the cooling source for the core. As<br />
the height of the ascending pipe is<br />
| | Fig. 6.<br />
Error analysis.<br />
increased, there is a greater density<br />
difference between cold and hot<br />
sections, further increasing the mass<br />
flow rate of PRHRS. Other parameters<br />
X5, X2 and X3 have a less pronounced<br />
effect on the maximum outlet temperature<br />
of the coolant.<br />
4.3 GM(1,6) model<br />
The Grey correlation has been used<br />
to determine the influence of each<br />
parameter on the maximum coolant<br />
temperature at the reactor’s output.<br />
Considering the relationship among<br />
the parameters, the coolant fluid<br />
temperature (X 0 ) is considered to<br />
represent the main behavior of the<br />
system. The temperature of IRWST<br />
(X 1 ), the diameter of PRHR HX(X 2 ),<br />
resistance coefficient (X 3 ), height of<br />
ascending pipe(X 4 ), initial pressure(X<br />
5 ) and the initial power level(X 6 )<br />
are considered to represent the correlated<br />
behavior factors. According to<br />
Eq. (9)-(18), an in-house code has<br />
been used to build GM (1,6) model.<br />
The code randomly selects 90 groups<br />
and the other 10 groups are used to<br />
validate. The corresponding differential<br />
equation is shown as Eq. (19).<br />
(19)<br />
Figure 6 shows the errors in the<br />
coolant temperature as determined<br />
by the results from GM(1,6) and<br />
RELAP5.<br />
As seen in Fig. 6, the results of<br />
GM(1,6) agree well with RELAP5,<br />
and the errors fall within 15%. The<br />
Grey model can adequately predict<br />
maximum coolant temperature at the<br />
outlet of the reactor core using a small<br />
amount of data, making up for the<br />
deficiency of artificial neural network<br />
(ANN), which becomes unstable with<br />
a small amount of data. This is a<br />
new way to replace thermal-hydraulic<br />
model.<br />
5 Conclusion<br />
By taking the loss of normal feedwater<br />
in AP1000 as an example, the behavior<br />
of PRHRS has been analyzed with the<br />
help of RELAP5, and the Grey system<br />
method has been applied for calculating<br />
the maximum coolant temperature<br />
at the outlet of the reactor<br />
core. The following conclusions are<br />
drawn.<br />
(1) The degree of Grey correlation is<br />
used to analyze the importance of<br />
influencing factors. Smaller the<br />
resolution, more obvious is the<br />
difference among these factors.<br />
The behavior of the factors can be<br />
distinguished very easily, when the<br />
resolution is 0.1.<br />
(2) The initial reactor power has the<br />
greatest influence on the maximum<br />
coolant temperature at the<br />
reactor outlet. And the sequences<br />
is followed by the temperature of<br />
IRWST, height of ascending pipe,<br />
initial pressure. Correspondingly,<br />
the diameter and resistance coefficient<br />
of PRHRS-HX have a lesser<br />
effect.<br />
(3) The GM(1,6) model is built to<br />
predict maximum coolant temperature<br />
at the reactor outlet. All<br />
errors fall within 15 % range.<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
Acknowledgments<br />
The research has been funded by<br />
Science and Technology on Reactor<br />
System Design Technology Laboratory<br />
Funds (2015BJ0151).<br />
References<br />
[1] Zhou Tao, Li Jingjing, Ru Xiaolong, et al.<br />
Application and development of<br />
passive technology in nuclear power<br />
units [J]. Proceedings of the CSEE, 2013,<br />
33(8):81-89.<br />
[2] Zhou Tao. Passive concept and<br />
technology [M]. Beijing, Tsinghua<br />
university press, 2016.<br />
[3] Burgazzi L. Evaluation of uncertainties<br />
related to passive systems performance<br />
[J]. Nuclear Engineering and Design,<br />
2004, 230(1):93-1<strong>06</strong>.<br />
[4] Zhang Shunxiang, Liang Guoxing.<br />
Application of status uncertainty<br />
analysis methods for AP1000 LBLOCA<br />
calculation [J]. Atomic Energy Science<br />
and Technology, 2012,S1:330-334.<br />
[5] Zio E, Cantarella M, Cammi A. The<br />
analytic hierarchy process as a<br />
systematic approach to the<br />
identification of important parameters<br />
for the reliability assessment of passive<br />
systems [J]. Nuclear Engineering and<br />
Design, 2003, 226(3):311-336.<br />
[6] Ma G, Yu Y, Huang X, et al. Screening<br />
key parameters related to passive<br />
system performance based on Analytic<br />
Hierarchy Process [J]. Annals of Nuclear<br />
Energy, 2015, 85:1141-1151.<br />
[7] Zio E, Apostolakis G E, Pedroni N.<br />
Quantitative functional failure analysis<br />
of a thermal–hydraulic passive system<br />
by means of bootstrapped Artificial<br />
Neural Networks [J]. Annals of Nuclear<br />
Energy, 2010, 37(37):639-649.<br />
[8] Liu Sifeng. Grey system theory and<br />
application [M]. Beijing, Science Press,<br />
2008<br />
[9] Liu Ping, Zhou Tao, Zhang Ming et al.<br />
Study on grey correlation degree of<br />
influence factors on ONB in narrow<br />
channel under natural circulation [J].<br />
Nuclear Power Engineering, 2011,<br />
32(4):29-32.<br />
[10] Zhou Tao, Yang Ruichang, Qin Shiwei<br />
et al. Study on grey model in ONB of<br />
nature circulation [J]. Nuclear Power<br />
Engineering, 2005, 26(2):121-124.<br />
[11] Sun Hanhong. Third generation nuclear<br />
power technology AP1000 [M]. Beijing,<br />
China Power Press, 2010.<br />
[12] Li Yankai, Lin Meng, Hou Dong et al.<br />
Qualitative accident analysis on loss of<br />
normal feedwater for AP1000 [J].<br />
Atomic Energy Science and<br />
Technology, 2012, S1:295-300.<br />
[13] IAEA. Natural Circulation in Water<br />
Cooled Nuclear Power Plants<br />
Phenomena Models, and Methodology<br />
for System Reliability Assessment. IAEA<br />
(TEC-DOC-1474).<br />
[14] Baosheng Wang, Dongqing Wang et<br />
al. Efficient estimation of the functional<br />
reliability of a passive system by means<br />
of an improved Line Sampling method<br />
[J]. Annals of Nuclear Energy, 2013,55:<br />
9-17.<br />
[15] Liu Qiang. Reliability analysis of AP1000<br />
passive system based on artificial neural<br />
networks [D]. Beijing: Tsinghua<br />
University, 2014.<br />
[16] Jiang Shuihua, Li Dianqing, Zhou<br />
Chuangbing. Non-instrusive stochastic<br />
finite element method for slope<br />
reliability analysis based on Latin<br />
hypercube sampling [J]. Chinese journal<br />
of Geotechnical enginerring, 2013,<br />
35(S2):70-76.<br />
[17] Wu Guojun, Chen Weizhong, Tan<br />
Xiaojun, et al. Program development of<br />
finite element reliability method and its<br />
application based on Latin Hypercube<br />
sampling [J]. Rock and Soil Mechanics,<br />
2015(2):550-554.<br />
[18] Zhang Xiaolian, Hao Sipeng, Li Jun, et<br />
al. Grey correlation based analysis on<br />
impacting factors of maximum power<br />
point tracking control of wind power<br />
generating unit [J]. Power system<br />
Technology, 2015, 39(2):445-449.<br />
[19] W Mendenhall. Statistics for engineers<br />
and the sciences [M]. Beijing, China<br />
Machine Press, 2009.<br />
Authors<br />
Qi Shi<br />
Zhou Tao<br />
Muhammad Ali Shahzad<br />
Li Yu<br />
School of Nuclear science and<br />
Engineering<br />
North China Electric Power<br />
University<br />
Beijing, 1022<strong>06</strong>, China<br />
Beijing Key Laboratory of Passive<br />
Safety Technology for Nuclear<br />
Energy<br />
Beijing, 1022<strong>06</strong>,China<br />
Jiang Guangming<br />
Science and Technology on Reactor<br />
System Design Technology<br />
Laboratory<br />
Nuclear Power Institute of China<br />
Chengdu, 610041, China<br />
RESEARCH AND INNOVATION 413<br />
Experimental Investigation of a Two-<br />
Phase Closed Thermosyphon Assembly<br />
for Passive Containment Cooling System<br />
Kyung Ho Nam and Sang Nyung Kim<br />
1 Introduction After the Fukushima accident, increasing interest has been raised in passive safety systems that<br />
maintain the integrity of the containment building. The conventional containment building is a thick, airtight reinforced<br />
concrete structure the design of which is highly unfavorable for removing heat from the containment atmosphere to the<br />
environment following an accident. Therefore, the sprays and/or fan coolers are installed to control the containment<br />
pressure and temperature for maintaining the integrity of the containment. However, either sprays or fan coolers are<br />
dependent on the power supply, which is unreliable if Design Basis Accidents (DBAs) are coupled with a station blackout<br />
(SBO) or Extended Loss of AC Power (ELAP). Therefore, to improve the reliability and safety of Nuclear Power Plants<br />
(NPPs), long-term passive cooling concepts have been developed for advanced reactors. In a previous study, The<br />
proposed design was based on an ordinary cylindrical Two-Phase Closed Thermosyphon (TPCT).[1] The exact assembly<br />
size and number of TPCTs should be elaborated upon through accurate calculations based on experiments. While the<br />
ultimate goal is to propose an effective MPHP design for the PCCS and experimentally verify its performance, a TPCT<br />
assembly that was manufactured based on the conceptual design in this paper was tested. Figure 1.<br />
Research and Innovation<br />
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
RESEARCH AND INNOVATION 414<br />
| | Fig. 1.<br />
Schematic of the Passive Containment Cooling System using the Multi-Pod Heat Pipe.<br />
2 Experiment procedure<br />
2.1 Experimental apparatus<br />
design<br />
As illustrated in Figure 2, an experimental<br />
facility was designed and<br />
installed to acquire various types of<br />
information related to the heat transfer<br />
capacity of MPHP. The facility consists<br />
of three major parts: a pressure vessel,<br />
a coolant tank, and an experimental<br />
TPCT assembly. An experimental TPCT<br />
assembly is a key part of the experimental<br />
apparatus used in this study.<br />
It conducts a heat transfer from the<br />
heater in the pressure tank to the<br />
coolant in the coolant tank. This<br />
assembly is made up of seven TPCTs,<br />
which are a 1-m long boiling region<br />
and condensation region, respectively,<br />
and has a hexagonal array.<br />
2.1.1 Design of TPCT assembly<br />
The operation of TPCT is based on the<br />
force of gravity and the temperature<br />
differences between its parts; one<br />
side is heated while the other side<br />
is cooled. Heat transfer occurs in<br />
TPCT due to these temperature<br />
differences. The thermal resistance<br />
(or the heat transfer coefficient) is<br />
calculated for each region. These<br />
are combined in a thermal resistance<br />
circuit, as shown in Figure 3, to<br />
calculate the total thermal resistance<br />
between the pressure tank inside and<br />
the coolant water inside the coolant<br />
tank (the heat transfer coefficient) for<br />
one TPCT. If R tot and ∆T are the total<br />
thermal resistance and the temperature<br />
difference between the pressure<br />
tank inside, which heats the boiling<br />
region, and the water cooling the<br />
condensation region, respectively,<br />
then it holds that:<br />
(1)<br />
where, ˙Q is the heat removal rate for a<br />
TPCT.<br />
To obtain an explicit expression for<br />
R tot , the heat transfer coefficient of<br />
each region was calculated first, and<br />
then the heat removal of one TPCT<br />
was calculated from this.<br />
The total heat transfer coefficient<br />
(resistance) for a given value of T h , T c<br />
and ∆T bc was calculated by summing<br />
up the aforementioned thermal resistances<br />
in each region. We first assumed<br />
that the temperature distribution was<br />
uniform, that is, there was no temperature<br />
difference between the air in the<br />
assembly center and the air inside the<br />
containment, as mentioned above. In<br />
fact, a considerable temperature drop<br />
is expected, and it is difficult to predict<br />
the specific value. This needs to be<br />
researched through additional experiments<br />
or a review of the literature. For<br />
convenience, we denote the overall<br />
number of pipes in the boiling and<br />
condensation regions as N b and N c ,<br />
respectively, and the heat removal<br />
rate per TPCT as ˙Q i . The following<br />
equation holds for the temperature<br />
drop at the inner boundary of the<br />
boiling region (where the resultant<br />
thermal resistance R5 can also be<br />
determined):<br />
(2)<br />
| | Fig. 2.<br />
Experimental apparatus for heat transfer performance test of MPHP.<br />
| | Fig. 3.<br />
The thermal resistance circuit in a TPCT.<br />
Research and Innovation<br />
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
Similarly, R 3 can be determined by<br />
dividing the thermal resistance of a<br />
single pipe by the number of pipes in<br />
the condensation region.<br />
(3)<br />
The resulting thermal resistance of<br />
the pipe walls can be determined by<br />
dividing the value for one TPCT by the<br />
number of TPCTs. At the same time,<br />
the heat transfer increases by the<br />
same factor.<br />
(4)<br />
(5)<br />
(6)<br />
where, ˙Q i is the heat transfer rate per<br />
pipe. If we assume that there is no<br />
heat transfer or temperature drop in<br />
the adiabatic region, then R 4 = 0 and<br />
R 8 = 0.<br />
The total thermal power extracted<br />
from the containment by one assembly<br />
can be calculated by multiplying the<br />
reciprocal of the sum of the thermal<br />
resistance values found above by<br />
the temperature difference between<br />
the containment atmosphere and the<br />
cooling water on the top of the containment<br />
dome shell and taking into<br />
account the difference ∆T bc as well.<br />
Here R MPHP is the total thermal resistance<br />
of the MPHP.<br />
(7)<br />
(8)<br />
where, R i is the thermal resistance<br />
component of a single TPCT. Therefore,<br />
the total heat transfer coefficient<br />
of the MPHP assembly is:<br />
(9)<br />
Section Material Height, m Diameter, m P/D ratio Thickness, m FR<br />
Pipe Stainlesssteel 1 0.03 2 0.0005<br />
Adiabatic 304<br />
0.3 0.2 − 0.013<br />
| | Tab. 1.<br />
Specifications of experimental thermosyphon assembly.<br />
This equation will be applied to compare<br />
the theoretical and experimental<br />
results in Chapter 3 of this paper.<br />
The heat transfer mechanisms in<br />
the boiling region occur in various patterns,<br />
which are the natural convection,<br />
evaporation, nucleate boiling, and a<br />
combination of each to the fill charge<br />
ratio (FR), heat flux, etc. The fill charge<br />
ratio indicates the percentage of the<br />
boiling region volume that is filled by<br />
the working fluids. In this study, it<br />
is determined that a fill charge ratio<br />
is 30 % of the boiling region based on<br />
a previous study performed by Imura<br />
[2, 3, 4]. Thus, an experimental TPCT<br />
assembly is partially filled with distilled<br />
water and then sealed. The specifications<br />
of the experimental TPCT assembly<br />
are presented in Table 1.<br />
2.1.2 Pressure tank and Coolant<br />
tank<br />
A pressure tank simulates the inner<br />
containment building, and steam is<br />
generated at the bottom of the vessel<br />
by two horizontally mounted immersion<br />
electric heaters with a total<br />
capacity of 30 kW. Air and makeup<br />
water can be injected into the vessel,<br />
and a drain line is located at the<br />
bottom of the vessel. The maximum<br />
rated operating pressure for the tank<br />
is 1 MPa, which is insured by a safety<br />
relief valve. The coolant tank is an<br />
open type, and the height is sufficiently<br />
high so that the condensation<br />
region of the pipes can be submerged<br />
in the coolant. The pressure tank<br />
is fully insulated with fiberglass to<br />
reduce heat loss so that sufficient<br />
power can be delivered to the TPCT<br />
assembly. Measurements have shown<br />
that the vessel heat loss is less than<br />
1 kW, which can be easily compensated<br />
by the heaters, which have a<br />
capacity of 30 kW. The capacity of a<br />
heater is determined by considering<br />
the predicted performance limitation<br />
of the TPCT assembly. The maximum<br />
heat transfer rate owing to entrainment<br />
limitations can be calculated<br />
through flooding correlations that are<br />
expressed in terms of the Kutateladze<br />
dimensionless groups [5].<br />
(10)<br />
where,<br />
(11)<br />
(12)<br />
The maximum heat transfer rate is<br />
predicted to be about 4 kW per pipe.<br />
Therefore, an experimental TPCT<br />
assembly that has seven pipes is<br />
considered, in which the maximum<br />
heat transfer rate is about 28 kW.<br />
However, this limitation is a conservative<br />
value, and thus this value is<br />
only considered to determine the<br />
capacity of an electrical heater.<br />
The coolant tank is an open type<br />
and the height is sufficiently high so<br />
that the condensation region of pipes<br />
can be submerged in coolant as shown<br />
in figure 2. As shown in figure 3, the<br />
heat generated in pressure tank transfers<br />
and this coolant in coolant tank<br />
will be external heat sink during the<br />
operation.<br />
2.3 Experiment procedure<br />
The experiment was conducted in<br />
the following manners. Two types of<br />
instrumentation devices were used<br />
in the experiment: thermocouples for<br />
temperature measurement, and pressure<br />
gauges/transducers for pressure<br />
measurement. K-type thermocouples<br />
were placed at each point to provide<br />
temperature readings. Temperature<br />
data were continuously recorded<br />
during operation with the thermocouples.<br />
The temperatures at the<br />
pressure tank, pipe wall surface, inner<br />
pipe, and coolant tank were recorded.<br />
One pressure transducer was installed<br />
to measure the overall pressure of the<br />
vessel. A communication-based Data<br />
Acquisition System (DAS) was set up<br />
for this experiment. All thermocouple<br />
leads and pressure transducer output<br />
cables were wired into the measurement<br />
channels. Additionally, a Silicon<br />
Controlled Rectifier (SCR) was installed<br />
to control the electrical power<br />
of the heater so that the electrical power<br />
was consistently fixed to the heater.<br />
0.3<br />
RESEARCH AND INNOVATION 415<br />
Research and Innovation<br />
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
RESEARCH AND INNOVATION 416<br />
Test case<br />
Heat<br />
input,<br />
kW<br />
Initial<br />
absolute<br />
pressure,<br />
MPa<br />
The predicted Air<br />
weight fraction<br />
at steady state,<br />
w/o<br />
10-#1 10 0.1 0.4<br />
10-#2 0.14 0.45<br />
10-#3 0.21 0.5<br />
15-#1 15 0.1 0.35<br />
15-#2 0.15 0.4<br />
15-#3 0.22 0.45<br />
20-#1 20 0.1 0.3<br />
20-#2 0.16 0.35<br />
20-#3 0.27 0.4<br />
25-#1 25 0.1 0.25<br />
25-#2 0.18 0.3<br />
25-#3 0.29 0.35<br />
30-#1 30 0.1 0.2<br />
30-#2 0.14 0.25<br />
30-#3 0.2 0.3<br />
Area of<br />
interest<br />
| | Tab. 2.<br />
Test matrix.<br />
• Temperature profile according to the heat<br />
input and initial air pressure<br />
• Pressure profile in the pressure tank<br />
Region Correlation Author<br />
Inside<br />
pressure tank<br />
Inside<br />
boiling region<br />
of pipe<br />
Inside<br />
condensation region<br />
of pipe<br />
Inside<br />
coolant tank<br />
<br />
<br />
<br />
| | Tab. 3.<br />
Heat transfer correlations for predicted value compared with the experiment results.<br />
Uchida<br />
Tagami<br />
Kataoka<br />
Murase<br />
Imura<br />
Nusselt<br />
Rohsenow<br />
The test conditions are listed in<br />
Table 2. Test cases 10-# to 30-#<br />
were performed to evaluate the heat<br />
removal performance in the pressure<br />
vessel. Heat input flowed from 10 kW<br />
to 30 kW in each case. As mentioned<br />
above, non-condensable gases greatly<br />
affect the heat transfer inside the<br />
containment because they depend<br />
only on natural circulation, and no<br />
power supply or fan operation for<br />
forced circulation is possible. Furthermore,<br />
as the MPHP assembly consists<br />
of a multitude of long pipes, the<br />
lengths and radial locations of<br />
the pipe in the assembly are expected<br />
to have a significant effect on the<br />
passage of steam to pipes. Therefore,<br />
the increase in concentration of noncondensable<br />
gases owing to steam<br />
condensation in the pipe array was<br />
considered. For this reason, the<br />
weight fraction range of air is determined<br />
to be from 0.2 to 0.5 w/o.<br />
3 Results and discussions<br />
The empirical correlations compared<br />
with the experimental results are<br />
presented in Table 3. The four correlations<br />
based on steam condensation<br />
with non-condensable gas were chosen<br />
for a comparison with the experimental<br />
data. These correlations are only<br />
dependent on the con centration of<br />
non- condensable gas, and thus are selected<br />
to compare with the data. [7, 8]<br />
As shown in Figure 4, all correlations<br />
compared with the experimental<br />
data tend to under predict the<br />
measured values. These correlations<br />
were developed for steam condensation<br />
with non-condensable gas on a<br />
long vertical plate. In this study, the<br />
geometry of the condensation area<br />
making contact with a steam and air<br />
mixture is of a cylinder type, and it<br />
shows that air weight accumulated<br />
on the pipe surface is lower than<br />
pre dicted. Thus, steam condensates<br />
better than the predicted models.<br />
The correlation reported by Imura<br />
et al. was compared with the experimental<br />
data, as shown in Figure 5.<br />
The Imura et al. correlation tends to<br />
under predict the measured values<br />
though the heat transfer coefficients<br />
in the boiling region, and predictions<br />
generally show reasonable agreement<br />
with the majority of the points being<br />
within the 35 % band.<br />
The correlation reported by Nusselt<br />
was compared with the experimental<br />
data. This correlation covers all data<br />
within ±10 %, as shown in Figure 6,<br />
and shows very good agreement with<br />
the measurements.<br />
| | Fig. 4.<br />
Predicted and experimentally determined heat transfer coefficients in the<br />
pressure tank for steam condensation with non-condensable gas.<br />
| | Fig. 5.<br />
Predicted and experimentally determined heat transfer coefficients in the<br />
boiling region for full pool boiling mode with distilled water.<br />
Research and Innovation<br />
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 6.<br />
Predicted and experimentally determined heat transfer coefficients in the<br />
condensation region for film condensation.<br />
| | Fig. 7.<br />
Comparison of the overall heat transfer rates between the predicted and<br />
experimentally determined results (left), and variation of total thermal<br />
resistance of a TPCT assembly vs. air weight fraction in a pressure tank (right).<br />
RESEARCH AND INNOVATION 417<br />
| | Fig. 8.<br />
Comparison of the overall heat transfer rates between the predicted and experimentally determined results (Left), Variation in total thermal resistance of a TPCT<br />
assembly vs air weight fraction in the pressure tank (Right).<br />
The correlation reported by<br />
Rohsenow was compared with the<br />
experimental data, as shown in<br />
Figure 7. Most heat transfer coefficients<br />
of the experimental results are<br />
much lower than those obtained by<br />
the correlations. In this study, city<br />
water was used as a coolant in the<br />
experiment, and the scale generated<br />
on the pipe surface effected as insulation.<br />
If the accident sequence determines<br />
that the designed water sources<br />
are not available in sufficient quantity,<br />
or at a sufficient rate, any water source<br />
should be used without delay. It is<br />
quite possibly that the city water may<br />
be used during an accident when the<br />
water sources are not available. For<br />
conservatism, city water was used as a<br />
coolant, and this condition causes the<br />
experimental data to be much lower<br />
than the predicted values.<br />
The input heat transfer rates versus<br />
the temperature difference between<br />
the inner pressure tank and coolant<br />
are plotted in Figure 8, and total<br />
thermal resistance is obtained from<br />
Eq. 9 and also presented in Fig. 8.<br />
It shows that the concentration of<br />
non-con densable gas in the containment<br />
is key factor which affects total<br />
thermal resistance and the performance<br />
of the MPHP when the MPHP is<br />
implemented in actual NPPs.<br />
4 Conclusion<br />
An analysis of experimental data and<br />
comparison to existing widely used<br />
correlations lead to the following<br />
conclusions:<br />
1. Measured heat transfer coefficients<br />
in each region and the overall heat<br />
transfer rate are higher than the<br />
predicted values. This shows that<br />
the theoretical results are conservative<br />
when a MPHP is implemented<br />
in an actual NPP. Additionally,<br />
the key factor that affects the<br />
total thermal resistance of a MPHP<br />
assembly is non-condensable gas<br />
concentration in the containment.<br />
2. The experiment results show that<br />
a TPCT consists of a 1-m long<br />
boiling and condensation region,<br />
respectively, and can transfer at<br />
least 45 kW/m 2 of heat flux.<br />
3. Based on the measured heat flux<br />
and heat transfer capacity, a MPHP<br />
assembly consists of 1-m long<br />
boiling and condensation pipes,<br />
respectively, and has about 2,000<br />
pipes with an overall diameter of<br />
about 1.75 m to provide 50 % heat<br />
removal capacity. In the case of<br />
100 % heat removal capacity, it<br />
has 4,500 pipes and the overall<br />
diameter is about 2.4 m.<br />
4. Precise calculations using computer<br />
code simulate the behavior<br />
(pressure, temperature) of the<br />
containment atmosphere when<br />
the novel PCCS is in operation and<br />
to account for other heat sources<br />
than the decay power.<br />
5. The development of average parameters<br />
(lumped parameter method)<br />
and performing param etric studies<br />
to account for the effects of increasing<br />
heat pipe length and array size<br />
(steam access from the containment<br />
to pipes). The air weight fraction<br />
was con sidered to be up to 0.5 w/o<br />
in this experiment, and thus this<br />
effect was considered roughly in<br />
this experiment.<br />
6. Because of the large added mass<br />
(cylindrical wall extension and/<br />
or water tanks on the dome top,<br />
cooling water, MPHP assemblies<br />
with water, and their accessories),<br />
an additional seismic evaluation of<br />
the containment (concrete walls<br />
and dome) is necessary.<br />
Research and Innovation<br />
Experimental Investigation of a Two-Phase Closed Thermosyphon Assembly for Passive Containment Cooling System ı Kyung Ho Nam and Sang Nyung Kim
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
RESEARCH AND INNOVATION 418<br />
Acknowledgement<br />
This study was sponsored by the<br />
Ministry of Trade, Industry and Energy<br />
(MOTIE) and was supported by<br />
Nuclear Convergence and Original<br />
Technology Development Program<br />
Grant funded by the Korea Institute<br />
of Energy Technology Evaluation and<br />
Planning (KETEP) (Grant code:<br />
20111520100030)<br />
Nomenclature<br />
P pressure (Pa)<br />
h heat transfer coefficient (W/m2°C)<br />
W weight fraction (w/o)<br />
T temperature (°C)<br />
k thermal conductivity (W/m°C)<br />
N number of pipe ( - )<br />
L length (m)<br />
D diameter (m)<br />
ρ density(kg/m 3 )<br />
c specific heat (J/kg-°C)<br />
q heat flux (W/m 2 )<br />
g gravity (m/s 2 )<br />
h latent heat (J/kg)<br />
μ viscosity (kg/m-s)<br />
Q heat transfer rate (W)<br />
Subscripts<br />
nc non-condensable gas ( - )<br />
b boiling region of pipe ( - )<br />
c condensation region of pipe ( - )<br />
hot outside boiling region of pipe ( - )<br />
cold outside condensation region of pipe( - )<br />
P/D Pitch-to-Diameter ratio ( - )<br />
References<br />
[1] G.H. Nam, J.S. Park, S.N. Kim.<br />
Conceptual Design of Passive Containment<br />
Cooling System for APR-1400<br />
using Multi-Pod Heat Pipe, Nuclear<br />
Technology. 189 (2015) 278–293.<br />
[2] H. Imura. Heat Transfer in the Two-<br />
Phase Closed Thermosiphon, Trans.<br />
JSME, Vol. 45, pp.712-722, 1979.<br />
[3] H. Imura. Critical Heat Flux in a Closed<br />
Two-Phase Thermosyphon, Int. J. Heat<br />
Mass Transfer, Vol26, No.8,<br />
pp. 1181-1188, 1983.<br />
[4] I. Khazaee, R. Hosseini, S.H. Noie.<br />
Experimental investigation of effective<br />
parameters and correlation of geyser<br />
boiling in a two-phase closed thermosyphon,<br />
Applied Thermal Engineering,<br />
Vol. 30, pp. 4<strong>06</strong>-412, 2010.<br />
[5] S. Khandekar, et. al. Thermal performance<br />
of closed two-phase thermosyphon<br />
using nanofluids, Int. J. Thermal<br />
Science, Vol. 47, 659-667, 2008.<br />
[6] Y.G. Lee, et. al. An experimental study<br />
of air-steam condensation on the<br />
exterior surface of a vertical tube under<br />
natural convection conditions, Int. J.<br />
Heat and Mass Transfer, Vol. 104,<br />
pp. 1034-1047, <strong>2017</strong>.<br />
[7] A. Dehbi. A generalized correlation for<br />
steam condensation rates in the<br />
presence of air under turbulent free<br />
convection, Int. J. Heat and Mass<br />
Transfer, Vol. 86, pp.1-15, 2015.<br />
[8] J.C. de la Rosa, A, Escriva. Review on<br />
condensation on the containment<br />
structures, Nuclear Energy, Vol. 51,<br />
pp. 33-36, 2009.<br />
Authors<br />
Kyung Ho Nam<br />
Korea Atomic Energy Research<br />
Institute<br />
111, Daedeok-daero 989beon-gil<br />
Yuseong-gu, Daejeon, Korea<br />
Sang Nyung Kim<br />
Kyunghee University<br />
1732, Deogyeong-daero<br />
Giheung-gu, Yongin-si,<br />
Gyeonggi-do, Korea<br />
Displacement of Cryomodule<br />
in CADS Injector II<br />
Yuan Jiandong, Zhang Bin, Wang Fengfeng, Wan Yuqin, Sun Guozhen, Yao Junjie, Zhang Juihui and He Yuan<br />
1 Introduction As Cryomodule can easily reduce higher power consumption and length of an accelerator,<br />
make the accelerator can be run continuously, it is becoming increasingly important in the superconducting linac [1].<br />
Due to the invisibility and coupled with ultra-low temperature characteristics (4 k), Cryomodule is the key points and<br />
difficulties for a superconducting linear accelerator. The Chinese academy of sciences institute of modern physics is<br />
developing an accelerator driven subcritical system (CADS) Injector II [2].CADS will accelerate protons with a beam<br />
current of 10mA to about 1.5 GeV to produce neutrons for the transmutation of nuclear waste [3]. To avoid generating<br />
beam orbit distortion, the magnet magnetic center must be on the beam axis, so the displacement of cold components<br />
has extremely requirements [4]. From the theoretical point, there are generally three approaches to deal with the<br />
displacement on the working condition [5]. One is to maintain the alignment upon the cooldown. In this approach, the<br />
structure is designed so that the cooldown is absolutely symmetric. The other is to allow realignment once cold. In this<br />
approach, components must be realigned after they reached their final cryogenic temperature. As we all know that both<br />
the above two situations cannot easily be reached.<br />
The last approach is to allow the<br />
components to change in a predict able<br />
and repeated way. There are four<br />
different methods to realize this<br />
objective currently. The European<br />
organization for nuclear research<br />
developed a double-sided Brandeis<br />
CCD Angle Monitor (BCAM) [6]. The<br />
Japanese high-energy accelerator<br />
research organization adopted white<br />
light interferometer (WLI) [7]. German<br />
electron synchrotron [8], the institute<br />
of high energy physics Chinese academy<br />
of sciences [9] and Fermi national<br />
accelerator laboratory [10] employed<br />
a Wire Position Monitor (WPM) to<br />
monitor the contraction. The France<br />
large national heavy-ion accelerator<br />
adopted a micro-alignment telescope<br />
to align Cryomodule intuitively [11].<br />
However, these above methods only<br />
investigated the cryo-displacement,<br />
did not concern the effect of the negative<br />
pressure of the vacuum. Ref [12]<br />
(D. Passarelli) have estimated the<br />
pressure distribution inside the cavity<br />
string used a mathematical model.<br />
Ref [13] analyzed the displacement<br />
induced by temperature differences,<br />
but did not correlate the cryo-vacuum<br />
displacement.<br />
In this paper, we present a detailed<br />
description of the principle of the<br />
vacuum cryo-environments firstly;<br />
and then we take out the simulation<br />
of vacuum and cryo-displacement<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
respectively; in the section IV, we compared<br />
the measured results with the<br />
simu lated ones; At last, we correlated<br />
the cryo-vacuum displacement. The<br />
deep investigation will benefit for the<br />
optimization [14] and upgrade [15-16]<br />
of Cryomodule design for the CIADS.<br />
2 Heat transfer principle<br />
CADS injector II project includes 4<br />
cells 6 cavities cryostats. Most cavities<br />
and magnets are working in the<br />
cryostat at liquid helium (LHE) temperature.<br />
The vacuum jacket of each<br />
cryomodule is rectangular box shaped<br />
of dimension 4.3 m × 1.7 m × 2.1 m.<br />
Their alignment will be carried out at<br />
room temperature first, and then after<br />
the compensation, the position error<br />
of the cavities and magnets shall be<br />
within ±0.1 mm. Finite element<br />
method was used here to analysis the<br />
thermal stress and displacement: we<br />
used Solid Works to construct model<br />
first; and then imported in ANSYS,<br />
meshing with four surfaces unit;<br />
finally simulated the thermal stress<br />
and cold displacement.<br />
Generally, heat transfer includes<br />
the sum of thermal radiation, convection,<br />
and sometimes conduction<br />
transfer. Usually, more than one of<br />
these processes occurs in a given<br />
situation. Since the Cryomodules are<br />
operated in a cryo-vacuum environment,<br />
there is no convective heat<br />
transfer in the static heat loads [17,<br />
18]. There are only thermal radiation<br />
[19] from the “hotter” environment<br />
and direct thermal conduction<br />
through the cold mass supports,<br />
power coupler and the feedthrough<br />
[20]. Thermal conduction is the<br />
transfer of heat (internal energy) by<br />
microscopic collisions of particles and<br />
movement of electrons within a body<br />
[21]. According to Fourier's law,<br />
the heat flux resulting from thermal<br />
conduction is proportional to the<br />
magnitude of the temperature gradient,<br />
the thermal conductivity and the<br />
cross-sectional surface area. However,<br />
it is inversely proportional to the<br />
length of a conduction path and<br />
opposite to it in sign. In many cases,<br />
the analysis may be simplified by the<br />
use of thermal conductivity integrals.<br />
In this approach, the conduction heat<br />
transfer in one dimension is given by<br />
[22] (eq. 1):<br />
(1)<br />
Thermal radiation is an electromagnetic<br />
radiation generated by the<br />
thermal motion of charged particles in<br />
matter. All matter with a temperature<br />
greater than absolute zero emits<br />
thermal radiation. When the temperature<br />
of a body is greater than absolute<br />
zero, inter-atomic collisions cause<br />
the kinetic energy of the atoms or<br />
molecules to change. This results in<br />
charge-acceleration and/or dipole<br />
oscillation which produces electromagnetic<br />
radiation, and the wide<br />
spectrum of radiation reflects the<br />
wide spectrum of energies and accelerations<br />
that occur even at a single<br />
temperature. During the cool-down,<br />
the thermal shield receives radiative<br />
heat flux from both the external vacuum<br />
vessel and from the internal cold<br />
mass. In fact, the heat flux density due<br />
to radiation transport between two<br />
surfaces at different temperatures can<br />
be written as following [22] (eq. 2):<br />
(2)<br />
Where [22] the vector Q is the heat<br />
flux (in W/m 2 ) in the positive direction;<br />
λ is known as the conductivity<br />
constant or conduction coefficient<br />
(in w/m k); A is the total crosssectional<br />
area of conducting surface<br />
(in m 2 ); L is the thickness of conducting<br />
surface (in m). Where [22] σ =<br />
5.67 10 is the Stefan-Boltzmann constant<br />
and ε is the effective emissivity<br />
of the king into account the view<br />
factor and surface emissivities. For a<br />
simple “feeling” of the order of<br />
magnitudes, it is useful to refer to the<br />
simple case of black body radiation<br />
(unitary emissivity) intercepted by a<br />
unitary surface at 2 K from a parallel<br />
plate at different temperatures.<br />
2 Simulation<br />
A Vacuum displacement<br />
To guarantee an effective cool-down<br />
process for the Cryomodule, a<br />
high-vacuum level must be achieved<br />
[12]. The thorough stress and Strain<br />
analysis of the vacuum chamber<br />
under atmospheric pressure and selfgravity<br />
must be carried out. According<br />
to the Hooke's Law, the displacement<br />
ΔL (eq. 3) is proportional to the above<br />
composite forces F n , is inversely proportional<br />
to the cross section size A<br />
and the Young's Modulus E.<br />
(3)<br />
The transfer of energy to the gravitational<br />
field results in the deformation<br />
of vacuum. The Finite element model<br />
of the cold mass and its support are<br />
shown in Figure 1. The generalized<br />
Hooke's law can be used to predict<br />
the displacement (Formula 4) caused<br />
in a given material by an arbitrary<br />
combination of stresses. Where v is<br />
the Poisson ratio, F x , F y and F z are the<br />
combined forces of the x, y, z-direction<br />
respectively.<br />
<br />
(4)<br />
Given that 1.5 tons of gravity were<br />
exerted on the two insulating<br />
RESEARCH AND INNOVATION 419<br />
| | Fig. 1.<br />
Model of the cryomudule.<br />
| | Fig. 2.<br />
Vacuum simulation.<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
RESEARCH AND INNOVATION 420<br />
| | Fig. 3.<br />
Cryo-Simulation<br />
supports, one atmospheric pressure<br />
(0.1Mpa) was applied to the six<br />
surfaces of the vacuum chamber; the<br />
four bottom supports were fixed. As<br />
seen from Figure 2, the vacuum<br />
displacement occurs mainly in the<br />
central area around the horizontal<br />
and vertical zone of support are 0.42<br />
and 0.62 mm respectively; and the<br />
central region is larger than the lateral<br />
area, with pot-shaped.<br />
B Cryo-displacement<br />
The two cooling experiments were<br />
cooled using Radiation cooling firstly<br />
(300 to 210 K); then the cryomodule<br />
is precooled with liquid nitrogen<br />
(210 to 150 K). After that, the<br />
cooldown of the cryomodule begins<br />
with liquid helium (150 to 4.2 K). The<br />
helium reservoir and cavity can be<br />
filled with liquid helium once temperature<br />
becomes 4.2 K. Liquid helium is<br />
pumped to reduce the vapor pressure<br />
of the liquid helium in the helium<br />
reservoir and cavity in order to make<br />
2 K. Internal stresses are also developed<br />
in the statically indeterminate<br />
structure if the free movement of the<br />
joint is prevented. According to the<br />
analysis of displacement in a statically<br />
determinate structures induced by<br />
temperature changes [24-25] (eq. 5),<br />
if the temperature of the member is<br />
decreased uniformly throughout its<br />
length, a denotes the coefficient of<br />
thermal expansion of the material<br />
(mm/K); ΔT denotes the temperature<br />
change (K); L denotes the length of<br />
the structure (mm).<br />
(5)<br />
In the model, Magnets, the helium<br />
tank, and its welding bracket adopted<br />
316 L stainless steel materials, HWR<br />
cavity and its welding bracket<br />
used titanium material, cold quality<br />
support components used titanium<br />
material, collimation bracket and<br />
cross hair targets used G11 materials.<br />
The contact surface of support was<br />
operated at 300 K. The thermal conductivity<br />
of materials varies strongly<br />
with a temperature between 300 and<br />
77 K [17]. The surface heat load of<br />
77 K (BPMs) was 1 W/m 2 [26-27].<br />
The boundary conditions and load<br />
[28] contain a self-gravity of the cold<br />
mass assembly, a distributive load of<br />
temperature, a force of the cold mass<br />
assemblies and the top suspending<br />
support. The force of the cold mass<br />
assemblies acting on the Ti support<br />
frame is decided by the gravity of each<br />
cold mass assembly. According to the<br />
mechanical characteristics of cold<br />
mass [28], the simulated results of the<br />
solenoid and HWR cavity were contracted<br />
0.77 mm in Horizontal and<br />
risen 2.98 mm in Vertical direction,<br />
respectively (As shown in Figure 3).<br />
4 Results and Analysis<br />
A Vacuum displacement<br />
As shown in Figure 4, the two processes<br />
of vacuum pump started at<br />
18:00 on November 30 and 16:00 on<br />
December 2, 2016, respectively. The<br />
vacuum level reached 0.1 Pa about<br />
3 hours later and 10-3 Pa about<br />
11 hours later. The two processes of<br />
vacuum release started at 8:00 on<br />
December 2 and 7:00 on December 4,<br />
2016, respectively. The vacuum level<br />
returned to 0:1 Pa about 8 hours later.<br />
Figure 5 illustrates the vertical and<br />
horizontal displacements monitored<br />
by the Laser Tracker from the top and<br />
left of the vacuum vessel. The Laser<br />
Tracker system is able to compensate<br />
for temperature and humidity effects<br />
based on the measurement conditions.<br />
The monitored displacements<br />
are 0.42 mm in the vertical direction<br />
and 0.62 mm in horizontal direction<br />
respectively. The monitored and<br />
simulated displacements are matching<br />
very well: the differences are<br />
0.02 mm in the vertical direction and<br />
0.04 mm in the horizontal direction.<br />
B Cryo-displacement<br />
The optical instrument microalignment<br />
telescope (MAT) was<br />
adopted to monitor the displacement.<br />
Fig. 5 shows the monitor results of<br />
Hori zontal (Vertical) direction during<br />
one thermal cycling: as the target was<br />
located on the right (below) of<br />
solenoid and HWR cavity, A plus sign<br />
means that it is close to center(rise<br />
up); A minus sign means that it is<br />
off center (go down). After cooled<br />
down 24 hours, cold mass has<br />
contracted 0.8 mm in horizontal and<br />
2.27 mm in vertical direction respectively<br />
on average(with respect to the<br />
pumped). And then cold mass has<br />
warmed up to 290 K after warmed up<br />
24 hours, and has expanded 0.5 mm<br />
horizontal and 1 mm vertical respectively<br />
on average.<br />
| | Fig. 4.<br />
Vacuum displacement (DX: horizontal; DY: vertical).<br />
| | Fig. 5.<br />
Cryo-displacement.<br />
Research and Innovation<br />
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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | Fig. 6.<br />
Stress Analysis.<br />
C Discussion<br />
As shown in Figure 6 (a), due to the<br />
supports were located on the outer<br />
surface of the bottom vacuum chamber,<br />
the direct vertical stress exerted<br />
on the inside cold mass (cavity, solenoid<br />
and so on) comprised the negative<br />
atmospheric pressure (the force F<br />
normal to the surface per area A),<br />
gravity and thermal stress during the<br />
process of cooling down. Therefore,<br />
the complete displacements in the<br />
vertical direction were the superposition<br />
effect of the above stress. However,<br />
the horizontal displacement<br />
resulted only from the thermal stress.<br />
Root mean square value of the<br />
measurements was 0.03 mm in the<br />
vertical direction and 0.02 mm in the<br />
horizontal direction. As shown in Fig.<br />
6 (b), the direct vertical stress exerted<br />
on the inside cold mass (cavity, solenoid<br />
and so on) comprised the negative<br />
atmospheric pressure, gravity and<br />
positive thermal stress during the<br />
process of warming up. Therefore, the<br />
complete displacements in the vertical<br />
direction were the subtraction of the<br />
above stress. As shown in Fig. 5, the<br />
differences of cryo displacement<br />
( released) with respect to nominal<br />
zero resulting from plastic displacement<br />
[29] and the not fully released<br />
pressure of the vacuum. The above<br />
results indicate that the reproducibility<br />
of the horizontal and<br />
vertical position is 0.3 mm and<br />
0.6 mm respectively.<br />
5 Conclusion<br />
Cryomodules are extremely complex<br />
systems, and their design optimization<br />
is strongly dependent on the<br />
accelerator application for which they<br />
are intended. We have demonstrated<br />
that the simulated vacuum and<br />
cryo-displacement shows a good<br />
agreement with the measured values.<br />
The above data provides information<br />
not only on the nature of the heat<br />
exchange phenomena and their effect<br />
on the structural stability of the internal<br />
components of the Cryomodule,<br />
but also benefit to an optimization<br />
for future Cryomodules design. The<br />
analysis procedure will be helpful for<br />
the estimation of displacements in<br />
working conditions like mechanical<br />
and thermal loads. We will study the<br />
on-line continuous monitoring system<br />
in the future, which will further reveal<br />
low-temperature displacement mechanism<br />
of the cryostat.<br />
Acknowledgment<br />
This work was supported by the<br />
National Natural Science Foundation<br />
of China (No.11605262). This work<br />
could not have been accomplished<br />
without the advice and support of our<br />
colleagues: Juihui Zhang and Bin<br />
Zhang.<br />
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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
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
422<br />
KTG INSIDE<br />
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(in Chinese).<br />
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Architectural Mechanics [M]. Dongbei<br />
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[27] R. Ge, R.X. Han, L. Bian, et al. Design of<br />
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[29] N.V. Isaev, T.V. Grigorova, O.V. Mendiuk,<br />
et al. Plastic deformation mechanisms<br />
of ultrafine-grained copper in the<br />
temperature range of 4.2–300 K.<br />
Low Temperature Physics.<br />
2016,42,9:825-835.<br />
Authors<br />
Yuan Jiandong<br />
Zhang Bin<br />
Wang Fengfeng<br />
Wan Yuqin<br />
Sun Guozhen<br />
Yao Junjie<br />
Zhang Juihui<br />
He Yuan<br />
Institute of Modern Physics<br />
Chinese Academy of Sciences<br />
509#, Nan chang Road, Lanzhou,<br />
China, 730000<br />
Inside<br />
KTG Sektion Süd und Fachgruppe Kernfusion<br />
Vortragsveranstaltung<br />
zur Fusionsforschung<br />
| | Prof. Dr. Zohm (rechts) – u.a. 2014 mit dem<br />
John Dawson Award der Amerikanischen<br />
Physikalischen Gesellschaft und 2016 mit<br />
dem Hannes Alfvén-Preis der Europäischen<br />
Physikalischen Gesellschaft ausgezeichnet –<br />
hier im Gespräch mit dem Sprecher der FG<br />
Kernfusion Dr. Thomas Mull (links) und der<br />
Sprecherin der Sektion Süd Yvonne Broy (Mitte).<br />
Erstmals luden für den 3. Mai <strong>2017</strong> die KTG Sektion Süd<br />
und die Fachgruppe Kernfusion zu einer gemeinsamen<br />
Veranstaltung nach Erlangen ein.<br />
Referent Prof. Dr. Hartmut Zohm – Leiter des Bereichs<br />
Tokamak-Szenario-Entwicklung am Max-Planck-Institut<br />
für Plasmaphysik – beantwortete die Frage „Wo steht die<br />
Fusionsforschung?“ und ging dabei auf den Tokamak, ITER<br />
und internationale Perspektiven ein.<br />
Die Forschungen zum magnetischen Einschluss von<br />
Wasserstoffplasmen mit Temperaturen von mehr als<br />
100 Millionen °C zur Energiegewinnung aus Kernfusion<br />
haben in den letzten Jahrzehnten große Fortschritte<br />
gemacht. Dabei werden unterschiedliche<br />
Fragen der Plasmaphysik, wie<br />
z.B. Wärmetransport oder Stabilität,<br />
experimentell und theoretisch untersucht.<br />
Parallel dazu werden spezielle<br />
Technologien, wie etwa der Bau großer<br />
supraleitender Spulen, vorangetrieben.<br />
Das Max-Planck- Institut für Plasmaphysik<br />
betreibt dazu in Garching das<br />
Groß experiment ASDEX Upgrade und<br />
hat eine weitere Großanlage, Wendelstein<br />
7-X, in Greifswald in Betrieb<br />
genommen.<br />
Im Vortrag ging Prof. Zohm ebenfalls<br />
auf den Test reaktor ITER ein, der<br />
zurzeit in einer weltweiten Zusammen<br />
arbeit in Cadarache, Frankreich,<br />
entsteht und eine Schlüsselrolle auf<br />
dem Weg zur Nutzung von Kernfusionsenergie<br />
spielen wird. Der ITER<br />
wird in Cadarache (Frankreich) durch<br />
| | 80 Teilnehmer verfolgten interessiert den hochinteressanten Vortrag,<br />
der zunächst mit einer Einführung in die Tiefen der Kernfusion begann.<br />
China, EU, Indien, Japan, Korea, Russland und die USA gebaut,<br />
wobei jeder der Partner sogenannte „In-kind“ – Leistungen<br />
erbringt, was die Projektsteuerung sehr komplex<br />
macht.<br />
Die EU-Roadmap zum Fusionskraftwerk sieht vor, das<br />
spätestens 2050 ein erstes derartiges Kraftwerk in Betrieb<br />
gehen soll. Der Weg vom ITER zu einem ersten DEMO ist<br />
allerdings noch weit. Neben dem Nachweis zuverlässiger<br />
Energieerzeugung mit abgeschlossenem Brennstoffkreislauf<br />
sind auch Verbesserungen in Physik und Technologie<br />
notwendig, um ein attraktiveres DEMO Design anbieten zu<br />
können.<br />
Alle ITER Partner haben starke nationale Aktivitäten:<br />
die Roadmap für China sieht dabei bereits 2030 die<br />
Inbetriebnahme des CFETR vor – China Fusion Engineering<br />
Test Reactor.<br />
Fazit des kurzweiligen Vortrages: Die Fusionsforschung<br />
hat in den letzten Jahren große Fortschritte erzielt, die<br />
nun im nächsten Schritt eine Realisierung des ITER<br />
ermöglicht. Fusionskraftwerke könnten ab 2050 Baustein<br />
der Energieversorgung sein, was immer noch rechtzeitig<br />
wäre, um eine weltweite Energiewende zu vollziehen, aber<br />
diese Entwicklung wird kontinuierliche Unterstützung<br />
benötigen. Die deutsche Fusionsforschung ist dabei ebenso<br />
von großer Bedeutung – nicht zuletzt mit W7-X wird auch<br />
der Stellarator eine wichtige Rolle spielen.<br />
Yvonne Broy<br />
KTG Inside
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Net-based values<br />
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Repair<br />
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Stretch-out-operation<br />
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of NEI, said: “America’s 99 nuclear<br />
reactors have a vital role to play when<br />
it comes to powering our economy,<br />
protecting the environment and supporting<br />
our nation’s influence around<br />
the world. GNI’s recommendations<br />
reflect a common interest in finding<br />
policy solutions to help keep our<br />
plants running, advance new designs<br />
and promote the role our nuclear<br />
suppliers play in generating jobs at<br />
home while strengthening America’s<br />
hand in global governance in the face<br />
of challenges abroad. We look forward<br />
to continuing our work with GNI<br />
on these important issues as we chart<br />
the future of our industry.”<br />
Kenneth Luongo, President of the<br />
Partnership for Global Security, said:<br />
“GNI has responded to the realities of<br />
the complex global environment<br />
where the linkages between critical<br />
issues including climate change,<br />
nuclear power and international<br />
security require new responses and<br />
innovative partnerships. Nuclear<br />
power has an important role to play<br />
in tackling climate change, but there<br />
are governance and geopolitical<br />
challenges that need to be addressed.<br />
The GNI report focuses attention on<br />
the nexus of these issues and provides<br />
an actionable agenda for progress that<br />
will benefit the global community.”<br />
Richard Meserve, former Chairman<br />
of the Nuclear Regulatory Commission<br />
and a member of the GNI working<br />
group, said: “This report draws attention<br />
to nuclear power’s geopolitical dimension,<br />
which often is overlooked in<br />
the debate. The nuclear rules are<br />
shaped by the countries with the<br />
largest market share, and traditional<br />
leaders like the U.S. will soon be<br />
overtaken by China and Russia. There<br />
is a danger that the U.S. will lose the<br />
capacity to influence the global norms<br />
for safety, security and nonproliferation.<br />
There thus are national<br />
security issues at stake.”<br />
Armond Cohen, Executive Director<br />
of the Clean Air Task Force and a<br />
member of the GNI working group,<br />
said: “Nuclear energy has increasingly<br />
come forward as a climate change<br />
management tool as we realize how<br />
deep and fast carbon cuts need to happen.<br />
Nuclear can be part of a portfolio<br />
approach – along with renewables,<br />
carbon capture and sequestration<br />
and improvements in efficiency –<br />
that gives us multiple options to<br />
decarbonize the electricity sector and<br />
sustain economic growth. But developing<br />
nuclear energy at sufficient<br />
scale and speed will require both<br />
technical innovation and close cooperation<br />
among industry, international<br />
regulatory bodies, civil society, and<br />
public and private investors. I look<br />
forward to the GNI’s involvement in<br />
that process in the months ahead.”<br />
Recommendations<br />
The report, “Nuclear Power for the<br />
Next Generation: Addressing Energy,<br />
Climate and Security Challenges,”<br />
addresses critical issues around<br />
climate policy, nuclear technology<br />
and global security. Its principal<br />
recommendations are:<br />
• Nuclear power is necessary to<br />
address climate change.<br />
Operating Results December 2016<br />
Plant name Country Nominal<br />
capacity<br />
Type<br />
gross<br />
[MW]<br />
net<br />
[MW]<br />
Operating<br />
time<br />
generator<br />
[h]<br />
Energy generated. gross<br />
[MWh]<br />
Month Year Since<br />
commissioning<br />
Time availability<br />
[%]<br />
Energy availability<br />
[%] *) Energy utilisation<br />
[%] *)<br />
Month Year Month Year Month Year<br />
OL1 Olkiluoto BWR FI 910 880 744 677 943 7 3<strong>06</strong> 048 247 231 855 100.00 92.73 99.26 91.61 100.13 91.40<br />
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<br />
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<br />
KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0<br />
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<br />
KKG Gösgen 7) PWR CH 1<strong>06</strong>0 1010 744 795 765 8 668 128 296 610 635 100.00 93.72 99.98 93.33 100.90 93.10<br />
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<br />
CNT-I Trillo PWR ES 1<strong>06</strong>6 1003 744 791 255 8 552 866 230 493 717 100.00 92.38 100.00 92.24 99.41 90.85<br />
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.<strong>06</strong><br />
Dukovany B2 PWR CZ 500 473 0 0 2 521 816 101 322 628 0 59.04 0 58.69 0 57.58<br />
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<br />
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<br />
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<br />
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<br />
Doel 1 PWR BE 454 433 744 338 530 3 169 852 - 100.00 80.57 99.89 79.64 100.04 79.26<br />
Doel 2 PWR BE 454 433 744 342 566 3 207 325 - 100.00 80.33 99.99 79.90 100.70 79.85<br />
Doel 3 PWR BE 1056 10<strong>06</strong> 580 598 227 7 689 354 - 77.87 83.03 74.76 82.40 75.72 82.46<br />
Doel 4 PWR BE 1084 1033 744 811 914 9 270 685 - 100.00 98.92 100.00 98.03 100.23 96.78<br />
Tihange 1 PWR BE 1009 962 0 0 3 005 326 - 0 34.45 0 33.85 0 33.98<br />
Tihange 2 PWR BE 1055 1008 744 789 428 8 954 388 - 100.00 97.01 99.98 96.42 101.23 97.17<br />
Tihange 3 PWR BE 1089 1038 744 809 473 8 226 519 - 100.00 86.78 99.99 86.19 99.83 85.94<br />
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
• Nuclear governance needs significant<br />
strengthening.<br />
• Evolving nuclear suppliers impact<br />
geopolitics.<br />
• Innovative nuclear policy requires<br />
“break the mold” partnerships.<br />
The full report and additional information<br />
can be found on the GNI<br />
website: www.globalnexusinitiative.<br />
org. A webcast of the media briefing<br />
also is available.<br />
| | www.nei.org, 8345<br />
World<br />
OECD figure show slight<br />
decrease for nuclear share of<br />
net electricity production<br />
(nucnet) Net electricity production in<br />
the 35 Organisation for Economic Cooperation<br />
and Development (OECD)<br />
member countries grew by 0.9 % in<br />
2016 compared to 2015 with nuclear’s<br />
share falling by 0.1% to 18.1 % figures<br />
released by the International Energy<br />
Agency show. Total OECD cumulative<br />
production of nuclear electricity in<br />
2016 was 1,873.6 TWh, a decrease<br />
of 2.7 TWh. Europe was the only<br />
region which decreased its nuclear<br />
pro duction, by 19.6 TWh, or 2.4 %, to<br />
790 TWh led by the continued<br />
phaseout of nuclear electricity in<br />
Germany as well as decreases in<br />
the Czech Republic and France caused<br />
by extended outages. There were<br />
also operational outages in Slovenia<br />
and Switzerland. There was a large<br />
increase of 9.5 % in renewable<br />
generation and a smaller, but still<br />
significant, increase of 2.2 % for<br />
hydro. Combustible fuels fell by<br />
0.2 % and 0.1 %. Non-combustible<br />
renew ables accounted for 22.4 %<br />
of all generation compared to 21.6 %<br />
in 2015.<br />
| | www.oecd.org, 9345<br />
Europe<br />
Foratom: EU Energy Proposals<br />
must take nuclear industry’s<br />
views into account<br />
(nucnet) Legislative proposals in the<br />
European Commission’s ‘Clean Energy<br />
for All Europeans’ package could<br />
ensure a coherent and optimal approach<br />
towards meeting energy and<br />
climate objectives, provided they take<br />
into account the views of the nuclear<br />
energy industry, Foratom, the Brusselsbased<br />
trade association for the industry<br />
in Europe, said in a position paper.<br />
The position paper said the goal of<br />
the EU to decarbonise the economy by<br />
more than 80% by 2050 cannot be<br />
achieved without nuclear power.<br />
The EC’s legislative proposals aim<br />
to improve the functioning of the<br />
energy market and make sure that all<br />
energy technologies compete on a<br />
level- playing field.<br />
| | www.foratom.org, 3845<br />
UK Nuclear Industry Study –<br />
steps required to avoid Brexit<br />
Euratom cliff edge<br />
(nia) The Government needs to work<br />
closely with industry in order to bring<br />
about replacement arrangements for<br />
Euratom in a timely manner to avoid a<br />
cliff edge for the nuclear industry, is<br />
the main message from a new position<br />
paper, Exiting Euratom, published<br />
today by the UK Nuclear Industry<br />
Association (NIA).<br />
The paper, prepared by the NIA<br />
following detailed consultation and<br />
discussion with its members, sets out<br />
the priority areas for negotiations<br />
with the European Commission as the<br />
UK ceases to be a full member of the<br />
Euratom community alongside the<br />
process to leave the EU. The paper<br />
also sets out the steps the UK Government<br />
need to take to avoid serious<br />
disruption to normal nuclear business<br />
in the UK and across the European<br />
Union.<br />
The key steps for government include:<br />
• Agreeing a replacement Voluntary<br />
Offer Agreement with the IAEA for<br />
a new UK safeguards regime<br />
• Replacing the Nuclear Cooperation<br />
Agreements (NCA) with<br />
key nuclear markets; the Euratom<br />
Community, United States,<br />
Canada, Australia, Kazakhstan and<br />
South Korea<br />
• Clarifying the validation of the<br />
UK’s current bilateral Nuclear<br />
Co-operation Agreements with<br />
Japan and other nuclear states<br />
• Setting out the process for the<br />
movement of nuclear material,<br />
goods, people and services<br />
• Agreeing a new funding arrangement<br />
for the UK’s involvement in<br />
Fusion 4 Energy and wider European<br />
Union nuclear R&D programme<br />
• Maintaining confidence in the<br />
industry and securing crucial<br />
investment<br />
Addressing these priority areas will<br />
enable the nuclear sector to continue<br />
its work with other countries, both<br />
within and outside the continuing EU,<br />
as the UK ceases to be a member of the<br />
European Union.<br />
However, given the amount to be<br />
concluded within the next 22 months,<br />
there is a risk that new arrangements<br />
will not be in place. The NIA is urging<br />
the Government to begin these negotiations<br />
by seeking an agreement with<br />
the EU that existing arrangements<br />
will continue to apply until the process<br />
of agreeing new arrangements is<br />
concluded, and avoiding the cliff edge<br />
scenario that is not in the interests of<br />
the industry, consumers, the UK or the<br />
EU.<br />
Tom Greatrex, Chief Executive of<br />
the Nuclear Industry Association,<br />
said:<br />
“The UK civil nuclear industry is<br />
ready and willing to work with the<br />
Government as it begins the process of<br />
putting replacement arrangements for<br />
Euratom in place. The clock is ticking,<br />
and this is a priority of increasing<br />
urgency.<br />
“This new report demonstrates<br />
that without new arrangements in<br />
place by the time the UK leaves the<br />
Euratom community, there is scope<br />
for real and considerable disruption.<br />
The industry has not only set out the<br />
priority areas to be addressed, but<br />
also the steps we think the Government<br />
needs to take to address those<br />
issues.<br />
“Government Ministers have stated<br />
their desire to both work with industry<br />
and to ensure the same high standards<br />
will continue to apply as the UK leaves<br />
the EU – there is no disagreement on<br />
that principle.<br />
“The Government now need to get<br />
down to the work of putting such<br />
arrangements in place, including a<br />
prudent approach to ensuring there<br />
are transitional arrangements in place,<br />
to avoid a gap in regulation. That<br />
would not be in the interests of the EU,<br />
the UK or the industry globally.”<br />
The NIA has called for a joint<br />
industry and Government working<br />
group to be created to help develop a<br />
plan to preserve the essential benefits<br />
of Euratom membership. This was<br />
also a key recommendation by the<br />
House of Lords Science and Technology<br />
Committee in its report published<br />
earlier this week.<br />
| | www.niauk.org, 3856<br />
Reactors<br />
Kansai Electric to begin<br />
restart process for Japan’s<br />
Takahama-3 and -4<br />
(nucnet) Kansai Electric Power Company<br />
(Kepco) said it plans to begin the<br />
425<br />
NEWS<br />
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
426<br />
NEWS<br />
restart process for the Takahama-3<br />
and -4 nuclear reactor units in Fukui<br />
Prefecture, western Japan.<br />
The company said it had received<br />
consent from local authorities to<br />
restart the two 830-MW pressurised<br />
water reactor units. Kepco said<br />
it is scheduled to load fuel into<br />
Takahama -4 later this week and plans<br />
to connect the reactor to the grid in<br />
late May and to start commercial<br />
operation in mid-June.<br />
The utility expects to load fuel into<br />
Takahama-3 in mid-May and to connect<br />
the unit to the grid in early June.<br />
| | www.kepco.co.jp, 9345<br />
Site work at Turkey’s Akkuyu<br />
to begin in July <strong>2017</strong><br />
(nucnet) Site Work at Turkey’s first<br />
nuclear power plant at Akkuyu is<br />
scheduled to begin this summer and<br />
run for almost two years, Russian<br />
state nuclear corporation Rosatom<br />
said, according to the state-operated<br />
domestic news agency RIA Novosti.<br />
Earthworks will begin in July<br />
with construction of the reactor pits<br />
scheduled to begin in January 2019,<br />
RIA Novosti said. Akkuyu, near Mersin<br />
on the Turkey’s southern Mediterranean<br />
coast, is to be built in<br />
cooperation with Rosatom under a<br />
contract signed in 2010. The station<br />
will have four 1,200-MW VVER units.<br />
On 3 March <strong>2017</strong>, Akkuyu Nuclear, the<br />
joint stock company in charge of the<br />
project, applied for a construction<br />
licence to the Turkish Atomic Energy<br />
Authority.<br />
| | www.rosatom.ru, 8345<br />
Hungary: Construction of<br />
initial facilities at Paks 2 to<br />
begin in autumn<br />
(nucnet) Construction of auxiliary<br />
facilities for the planned two-unit<br />
Paks 2 nuclear power station in<br />
Hungary will begin in the autumn of<br />
<strong>2017</strong>, Alexei Likhachev, head of<br />
Russian state nuclear corporation<br />
Rosatom, said in a statement.<br />
Rosatom said auxiliary facilities<br />
include a number of production,<br />
storage and other buildings to be used<br />
by contractors during the project’s<br />
construction phase.<br />
| | www.atomeromu.hu, 9345<br />
Research<br />
In ‘anti-nuclear’ Denmark:<br />
How a reactor startup is helping<br />
to change opinions<br />
(nucnet) Seaborg Technologies of<br />
Copenhagen is developing an<br />
advanced thorium-based molten salt<br />
reactor (MSR) and has received a<br />
grant from the public funding agency<br />
Innovation Fund Denmark, a move<br />
that marks the first Danish investment<br />
into nuclear fission research since a<br />
1985 ban on nuclear energy.<br />
The decision to fund the reactor,<br />
known as the Seaborg CUBE-100<br />
(short for Compact Used Fuel BurnEr),<br />
is the beginning of the first Danish<br />
venture into the development of novel<br />
fission reactor concepts, Seaborg said.<br />
NucNet editor-in-chief David<br />
Dalton spoke to Seaborg’s co-founders<br />
about the significance of the funding,<br />
the next steps on the road to commercialisation,<br />
and how attitudes towards<br />
nuclear in traditionally anti-nuclear<br />
Denmark are changing.<br />
Full story for NucNet subscribers:<br />
http://bit.ly/2oKDzCp<br />
| | seaborg.dk, 3452<br />
United States announces<br />
€ 1 million pledge for modernization<br />
of IAEA Nuclear<br />
Applications Laboratories<br />
(iaea) The United States announced a<br />
pledge of €1 million to support the<br />
modernization of the International<br />
Atomic Energy Agency (IAEA) Nuclear<br />
Applications Laboratories in Seibersdorf,<br />
outside Vienna. These facilities<br />
opened their doors in 1962 and play a<br />
key role in the peaceful uses of nuclear<br />
science and technology to assist<br />
countries in areas such as human and<br />
animal health, food security and the<br />
protection of the environment.<br />
The announcement was made<br />
during the first day of the first session<br />
| | Tentative schematic of the SWaB reactor.<br />
(Illustration: Seaborg Technologies, Denmark).<br />
of the Preparatory Committee for the<br />
2020 Review Conference of the<br />
Parties to the Treaty on the Non-<br />
Proliferation of Nuclear Weapons<br />
(NPT), May 2–12 in Vienna, Austria.<br />
The contribution will go towards<br />
the construction of a new Animal<br />
Production and Health Laboratory,<br />
one of eight laboratories that will<br />
be upgraded under the Agency’s<br />
Renovation of the Nuclear Applications<br />
Laboratories (ReNuAL) and<br />
ReNuAL Plus initiatives.<br />
IAEA Director General Yukiya<br />
Amano, addressing the Preparatory<br />
Committee meeting, said the modernization<br />
of the eight IAEA Nuclear<br />
Applications Laboratories was proceeding<br />
well.<br />
“The laboratories train scientists,<br />
support research in human health,<br />
food and other areas, and provide<br />
analytical services to national laboratories,”<br />
Amano said. “I thank donor<br />
countries for their generous contributions<br />
and I hope that Member<br />
States will continue to provide strong<br />
support for further work on this<br />
important modernization project.”<br />
U.S. Ambassador Robert Wood, the<br />
country’s Permanent Representative<br />
to the Conference on Disarmament in<br />
Geneva, said the IAEA plays a key<br />
part in helping countries realize the<br />
practical benefits of the NPT.<br />
“I am pleased to announce a U.S.<br />
pledge of € 1 million to support the<br />
IAEA’s project to renovate its Nuclear<br />
Applications Laboratories, in addition<br />
to the nearly € 8.9 million we have<br />
provided to date. This ReNuAL project<br />
aims to renew the infrastructure<br />
needed to sustain the IAEA’s programmes<br />
for peaceful uses of nuclear<br />
energy. We also urge other IAEA<br />
Member States to join us in meeting<br />
this year’s ReNuAL Plus fundraising<br />
goals.”<br />
“The U.S. pledge brings us halfway<br />
to the funding target of € 2 million<br />
that we need to reach by June to start<br />
building this important laboratory<br />
on time and to maximize our cost<br />
efficiencies, so it is significant both<br />
in terms of its size and timing,” said<br />
IAEA Deputy Director General Aldo<br />
Malavasi, who heads the IAEA’s<br />
Department of Nuclear Sciences and<br />
Applications.<br />
“The IAEA’s work in helping<br />
countries to apply nuclear technologies<br />
to quickly detect and control<br />
animal diseases posing threats to food<br />
and economic security and to health is<br />
increasingly in demand,” Malavasi<br />
said. “This week, for example, in<br />
Seibersdorf the Agency is training<br />
News
<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
| | The Animal Production and Health Laboratory<br />
is one of eight facilities that will be upgraded<br />
under the IAEA’s ReNuAL and ReNuAL Plus<br />
initiatives. The photo shows the Agency's<br />
training of veterinary experts in diagnosing<br />
MERS-CoV in camels – a zoonotic disease that<br />
is very dangerous to humans. (Photo: IAEA)<br />
16 veterinary experts from seven<br />
Member States in diagnosing<br />
Middle East Respiratory Syndrome<br />
Coronavirus (MERS-CoV) in camels –<br />
a zoonotic respiratory disease that<br />
is very dangerous to humans. This<br />
contribution is very welcome.”<br />
| | www.iaea.org, 9538<br />
Company News<br />
Nuclear safety: AREVA NP to<br />
support international OECD<br />
research program<br />
(areva) The research program of the<br />
Nuclear Energy Agency (NEA), as part<br />
of the Organization for Economic<br />
Co-operation and Development<br />
(OECD), is being continued at AREVA<br />
NP’s PKL test facility until mid-2020,<br />
through a fourth 4-years contract<br />
agreement (PKLIII-i). During this<br />
period, the focus is to systematically<br />
investigate thermal hydraulic<br />
phenomena in pressurized water<br />
reactors (PWR). Another main topic<br />
within the program is the experimental<br />
verification of cool-down<br />
procedures for operational and<br />
emergency manuals of such plants.<br />
These efforts aim to enhance safety of<br />
nuclear power plants worldwide.<br />
AREVA NP’s unique PKL test facility<br />
is part of AREVA NP’s Technical Center<br />
in Erlangen (Germany) and models<br />
the nuclear steam supply system of a<br />
PWR in full scale height. Test series<br />
can therefore be performed and<br />
evaluated under realistic conditions.<br />
The results obtained allow experts to<br />
develop recommendations for plant<br />
operation under accident situations.<br />
”It is a great honor for us to further<br />
contribute to safety research within<br />
the international frame of OECD/<br />
NEA. We have adapted our test facility<br />
to the requirements of the new program<br />
during the last months. Recently,<br />
we started a first test series together<br />
with our international partners“, said<br />
Klaus Umminger, who is responsible<br />
for the project at AREVA NP.<br />
Under the umbrella of the OECD,<br />
the project is funded jointly by<br />
the German Federal Ministry for Economic<br />
Affairs and Energy and other<br />
contributors like Safety technical<br />
support organizations, research<br />
institutes and PWR operating utilities<br />
from 14 OECD/NEA countries. The<br />
budget of the project is about one<br />
million euros per year.<br />
| | www.areva.com, 7345<br />
ROSATOM and the French<br />
National Institute for Nuclear<br />
Science and Technology held a<br />
seminar on nuclear education<br />
and training<br />
(rosatom) On 27 April, <strong>2017</strong>,<br />
ROSATOM State Atomic Energy<br />
Corporation and the National Institute<br />
for Nuclear Science and Technology<br />
administered by the French Atomic<br />
Energy and Alternative Energies<br />
Commission (INSTN) held a seminar<br />
“Human capital issues facing nuclear<br />
energy education and training today”<br />
in the Russian Cultural Centre<br />
in Paris. The event was attended<br />
both by representatives of leading<br />
education and research institutions,<br />
and by business representatives, who<br />
discussed the major and most central<br />
issues of the nuclear energy education<br />
and training today.<br />
The sessions focused on global<br />
trends and best innovative practices,<br />
experience and potential of international<br />
nuclear education programs<br />
in partner countries, including internships<br />
as well as further education and<br />
instructor training courses.<br />
During the seminar, the Superior<br />
National School of Advanced Techniques<br />
(ENSTA ParisTech), INSTN,<br />
the National Research Nuclear University<br />
MEPhI and Lomonosov MSU<br />
presented their extensive expertise<br />
in implementing education and<br />
training programs. Representatives of<br />
ROSATOM State Corporation and JSC<br />
Rusatom Service highlighted the<br />
importance of training highly qualified<br />
specialists in the nuclear industry.<br />
In her welcome speech Anne Lazar-<br />
Sury, Governor for France to the IAEA,<br />
Director of the Division for International<br />
Affairs of the French Alternative<br />
Energies and Atomic Energy<br />
Commission called for increased<br />
cooperation between France and<br />
Russia in nuclear education.<br />
Philippe Corréa, Director of the<br />
INSTN noted that “education and<br />
training is a pillar between research<br />
and industry. We must anticipate<br />
needs of our business partners and<br />
consequently offer a challenging<br />
professional development to our<br />
students”, - he added.<br />
Speaking about the needs of<br />
nuclear education and training,<br />
Evgeny Salkov, Director General<br />
of JSC Rusatom Service stressed<br />
that “development of cooperation<br />
with INSTN is of great importance<br />
for Rusatom Service. Considering<br />
ambitious goals set out in our roadmap<br />
we are interested in partnership<br />
with those, who can help us prepare<br />
personnel for nuclear facilities in time<br />
and of quality, given that safety is our<br />
top priority.<br />
Andrey Rozhdestvin, Director of<br />
Rosatom in Western Europe reminded<br />
that «nowdays Europe is missing the<br />
highly qualified human resources.<br />
This question is even more crucial for<br />
the newcomer countries».<br />
In her turn, Tatiana Leonova,<br />
Vice-Principal of MEPhI – partner of<br />
ROSATOM State Corporation in the<br />
sphere of nuclear education – introduced<br />
the implemented programs to<br />
the audience. “The best universities<br />
develop successfully joint education<br />
and research programs, it raises the<br />
general level of education in newcomer<br />
countries”,- she said.<br />
During the seminar the participants<br />
discussed current issues of<br />
distance education and possible applications<br />
of experimental equipment<br />
and immersive 3D technologies in<br />
educational process. Furthermore,<br />
experience in implementing education<br />
projects in foreign countries,<br />
including newcomer countries embarking<br />
on nuclear power programmes<br />
was presented.<br />
In conclusion, experts of both<br />
countries underscored the importance<br />
of further exchange of experience<br />
and development of Russian-French<br />
cooperation.<br />
| | ROSATOM and the French National Institute for Nuclear Science and<br />
Technology held a seminar on nuclear education and training.<br />
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
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In the framework of the event, the<br />
participants visited INSTN laboratories<br />
where they got acquainted with<br />
the ISIS training reactor as well as a<br />
new virtual tool for studying radiotherapy<br />
VERT (virtual environment<br />
for radiotherapy training).<br />
The French National Institute<br />
for Nuclear Science and Technology<br />
( INSTN) is a higher educational<br />
institution administered by the French<br />
Atomic Energy and Alternative<br />
Energies Commission founded in<br />
1956. The INSTN is under the joint<br />
authority of the Ministry of National<br />
Education, Higher Education and<br />
Research, the Ministry of the Economy<br />
and Finance and the Ministry of<br />
the Environment. The Institute is the<br />
main nuclear education centre in<br />
France.<br />
ROSATOM State Atomic Energy<br />
Corporation brings together more<br />
than 320 enterprises and scientific<br />
organizations, including all civil<br />
nuclear companies of Russia’s nuclear<br />
industry, research centers and the<br />
world’s only nuclear icebreaker fleet.<br />
ROSATOM holds leading positions in<br />
the global market of nuclear technologies<br />
and is currently implementing<br />
projects to build 42 nuclear power<br />
units both in Russia and abroad.<br />
| | www.rosatom.ru, 3845<br />
Organisations<br />
NEA and China’s National<br />
Energy Administration<br />
sign MOU to strengthen<br />
co‐operation<br />
(oecd-nea) On 28 April <strong>2017</strong>, the NEA<br />
and the National Energy Administration<br />
of China (C/NEA) signed a<br />
Memorandum of Understanding<br />
(MOU) in the Field of Peaceful Uses<br />
of Nuclear Energy, enhancing co‐operation<br />
between both parties. An official<br />
ceremony was held in Beijing, China,<br />
at which C/NEA Deputy Adminis trator<br />
Li Fanrong signed the MoU on behalf<br />
of the C/NEA and NEA Director-<br />
General William D. Magwood, IV,<br />
signed on behalf of the NEA. The<br />
agreement foresees co‐operation in a<br />
number of fields, including nuclear<br />
energy development, nuclear safety<br />
research and radiological protection.<br />
The memorandum of understanding<br />
between the NEA and the C/NEA<br />
represents further progress in the<br />
growing collaboration between China<br />
and the Agency, and complements<br />
the memorandum of understanding<br />
signed by the NEA and the National<br />
Nuclear Safety Administration ( NNSA)<br />
of China in 2014 and the Joint Declaration<br />
on Co‐operation signed by the<br />
NEA and the China Atomic Energy<br />
Authority (CAEA) in 2013.<br />
| | www.oecd-nea.ogr, 7349<br />
People<br />
Russ Brian announced as new<br />
WANO Atlanta Centre director<br />
(wano) David Garchow, Atlanta<br />
Centre Director of the World Association<br />
of Nuclear Operators (WANO)<br />
and Vice President, International at<br />
the Institute of Nuclear Power Operations<br />
(INPO), has announced his<br />
retirement effective 14 July <strong>2017</strong>. He<br />
joined INPO in 2005 as a Team Leader,<br />
was elected Vice President of Plant<br />
Technical Support in 2010 and<br />
assigned to his current role in 2013.<br />
Succeeding Garchow will be Russ<br />
Brian, INPO’s Director of Plant Evaluations.<br />
Effective 12 July <strong>2017</strong>, he will<br />
be promoted to WANO Atlanta Centre<br />
Director and INPO Vice President,<br />
International. Russ joined INPO<br />
in 2010 and has served as a Team<br />
Leader, Deputy Director of Corporate<br />
Evaluations, and Director of Plant<br />
Evaluations.<br />
WANO Chief Executive Officer,<br />
Peter Prozesky said, “We are grateful<br />
for the substantive contributions Dave<br />
Garchow has made to our industry<br />
and wish him well in his retirement.<br />
We also welcome Russ Brian to his<br />
new role and look forward to him<br />
continuing this strong legacy of<br />
leadership.”<br />
| | www.wano.info, 7345<br />
NEA expert receives award<br />
for international co-operation<br />
from Korea<br />
(oecd-nea) Dr Henri Paillère, NEA’s<br />
Senior Nuclear Analyst and Acting<br />
Head of the Division of Nuclear<br />
Development, has been honoured with<br />
the Award for Person of Merit for<br />
International Co-operation in Nuclear<br />
Industry by the Korean Ministry of<br />
Science, ICT and Future Planning. The<br />
honour was awarded in recognition<br />
of Dr Paillère’s dedication and service<br />
for the promotion of co-operation<br />
between Korea and the NEA, including<br />
through his work in support of the<br />
Generation IV International Forum<br />
(GIF) and the International Framework<br />
for Nuclear Energy Cooperation<br />
(IFNEC). “We are very pleased to see<br />
Dr Paillère’s accomplishments being<br />
acknowledged,” NEA Director- General<br />
Mr Magwood said. “We are very fortunate<br />
to have outstanding people like<br />
Henri at the Agency.”<br />
| | www.oecd-nea.org, 7345<br />
Market data<br />
(All information is supplied without<br />
guarantee.)<br />
Nuclear fuel supply<br />
market data<br />
Information in current (nominal)<br />
U.S.-$. No inflation adjustment of<br />
prices on a base year. Separative work<br />
data for the formerly “secondary<br />
market”. Uranium prices [US-$/lb<br />
U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =<br />
0.385 kg U]. Conversion prices [US-$/<br />
kg U], Separative work [US-$/SWU<br />
(Separative work unit)].<br />
January to December 2013<br />
• Uranium: 34.00–43.50<br />
• Conversion: 9.25–11.50<br />
• Separative work: 98.00–127.00<br />
January to December 2014<br />
• Uranium: 28.10–42.00<br />
• Conversion: 7.25–11.00<br />
• Separative work: 86.00–98.00<br />
January to June 2015<br />
• Uranium: 35.00–39.75<br />
• Conversion: 7.00–9.50<br />
• Separative work: 70.00–92.00<br />
June to December 2015<br />
• Uranium: 35.00–37.45<br />
• Conversion: 6.25–8.00<br />
• Separative work: 58.00–76.00<br />
2016<br />
January to June 2016<br />
• Uranium: 26.50–35.25<br />
• Conversion: 6.25–6.75<br />
• Separative work: 58.00–62.00<br />
July 2016<br />
• Uranium: 26.50–27.80<br />
• Conversion: 6.00–6.50<br />
• Separative work: 58.00–62.00<br />
August 2016<br />
• Uranium: 22.25–26.40<br />
• Conversion: 5.50–5.75<br />
• Separative work: 58.00–62.00<br />
September 2016<br />
• Uranium: 22.25–22.75<br />
• Conversion: 5.50–5.75<br />
• Separative work: 52.00–55.00<br />
October 2016<br />
• Uranium: 19.60–22.90<br />
• Conversion: 5.50–5.75<br />
• Separative work: 49.00–53.00<br />
November 2016<br />
• Uranium: 18.50–18.90<br />
• Conversion: 5.50–5.75<br />
• Separative work: 48.00–51.00<br />
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<strong>atw</strong> Vol. 62 (<strong>2017</strong>) | Issue 6 ı June<br />
December 2016<br />
• Uranium: 18.75–21.50<br />
• Conversion: 5.50–5.75<br />
• Separative work: 47.00–50.00<br />
<strong>2017</strong><br />
January <strong>2017</strong><br />
• Uranium: 20.25–25.50<br />
• Conversion: 5.50–6.75<br />
• Separative work: 47.00–50.00<br />
February <strong>2017</strong><br />
• Uranium: 23.50–26.50<br />
• Conversion: 5.50–6.75<br />
• Separative work: 48.00–50.00<br />
March <strong>2017</strong><br />
• Uranium: 24.00–26.00<br />
• Conversion: 5.50–6.75<br />
• Separative work: 47.00–50.00<br />
| | Source: Energy Intelligence<br />
www.energyintel.com<br />
| | Uranium spot market prices from 1980 to <strong>2017</strong> and from 2007 to <strong>2017</strong>. The price range is shown.<br />
In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.<br />
429<br />
NEWS<br />
Cross-border price<br />
for hard coal<br />
Cross-border price for hard coal in<br />
[€/t TCE] and orders in [t TCE] for<br />
use in power plants (TCE: tonnes of<br />
coal equivalent, German border):<br />
2012: 93.02; 27,453,635<br />
2013: 79.12, 31,637,166<br />
2014: 72.94, 30,591,663<br />
2015: 67.90; 28,919,230<br />
2016: 67.07; 29,787,178<br />
I. quarter: 56.87; 8,627,347<br />
II. quarter: 56.12; 5,970,240<br />
III. quarter: 65.03, 7.257.041<br />
IV. quarter: 88.28; 7,932,550<br />
| | Source: BAFA, some data provisional<br />
www.bafa.de<br />
EEX Trading Results<br />
in March <strong>2017</strong><br />
(eex) In March <strong>2017</strong>, the European<br />
Energy Exchange (EEX) reached a<br />
volume of 311.2 TWh on its power<br />
derivatives markets, representing a<br />
year-on-year increase of 22% (March<br />
2016: 255.8 TWh).<br />
In particular, power products for<br />
the German-Austrian market contributed<br />
to this result. At 250.5 TWh,<br />
volumes in this market increased by<br />
42 % (March 2016: 176.6 TWh).<br />
This includes 225.9 TWh from<br />
Phelix Futures and 24.6 TWh from<br />
Phelix Options.<br />
The March volumes comprised also<br />
158.7 TWh registered at EEX for<br />
clearing. Clearing and settlement<br />
of all transactions was executed by<br />
European Commodity Clearing (ECC).<br />
The Settlement Price for base load<br />
contract (Phelix Futures) with<br />
delivery in 2018 amounted to<br />
29.77 €/MWh. The Settlement Price<br />
for peak load contract (Phelix Futures)<br />
with delivery in 2018 amounted to<br />
37.49 €/MWh.<br />
| | Separative work and conversion market price ranges from 2007 to <strong>2017</strong>. The price range is shown.<br />
)1<br />
In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.<br />
On the EEX Market for emission<br />
allowances, a total volume of<br />
117.5 million tonnes of CO 2 was<br />
traded in March which represents a<br />
year-on-year increase of 50 % (March<br />
2016: 78.4 million tonnes of CO 2 ). On<br />
the EUA secondary market, volumes<br />
have doubled to 33.6 million tonnes<br />
of CO 2 (March 2016: 15.7 million<br />
tonnes of CO 2 ). The primary market<br />
auctions contributed 83.5 million<br />
tonnes of CO 2 to the total volume.<br />
The E-Carbix amounted to<br />
5.09 €/EUA, the EUA price with<br />
delivery in December 2016 amounted<br />
to 4.63/5.91 €/ EUA (min./max.).<br />
| | www.eex.com<br />
MWV crude oil/product prices<br />
in March <strong>2017</strong><br />
(mwv) According to information and<br />
calculations by the Association of the<br />
German Petroleum Industry MWV e.V.<br />
in March <strong>2017</strong> the prices for super fuel<br />
and heating oil noted lower for fuel oil<br />
sligthly lower compared with the<br />
previous month February <strong>2017</strong>. The<br />
average gas station prices for Euro<br />
super consisted of 136.28 €Cent<br />
( February <strong>2017</strong>: 139.39 €Cent, approx.<br />
-3.11 % in brackets: each information<br />
for previous month or rather previous<br />
month comparison), for diesel fuel of<br />
116.56 €Cent (118.27; -1.45 %) and<br />
for heating oil (HEL) of 56.81 €Cent<br />
(59.28, -4.17 %).<br />
The tax share for super with a<br />
consumer price of 136.28 €Cent<br />
(139.39 €Cent) consisted of<br />
65.45 €Cent (48.03 %, 65.45 €Cent)<br />
for the current constant mineral oil<br />
tax share and 21.76 €Cent (current<br />
rate: 19.0 % = const., 22.26 €Cent)<br />
for the value added tax. The product<br />
price (notation Rotterdam) consisted<br />
of 36.05 €Cent (26.45 %, 39.59 €Cent)<br />
and the gross margin consisted of<br />
13.02 €Cent (9.5 %; 12.09 €Cent).<br />
Thus the overall tax share for super<br />
results of 67.0 % (66.0 %).<br />
Worldwide crude oil prices<br />
(monthly average price OPEC/Brent/<br />
WTI, Source: U.S. EIA) were approx.<br />
-6.48 % (+1.41 %) lower in March<br />
compared to February <strong>2017</strong> also<br />
despite the decision of the OPEC<br />
to restrict and lower the crude oil<br />
production. The market showed a<br />
stable development with lower<br />
prices; each in<br />
US-$/ bbl: OPEC basket: 50.32<br />
(53.37); UK-Brent: 51.59 (54.87);<br />
West Texas Intermediate (WTI):<br />
49.33 (53.47)<br />
| | www.mwv.de<br />
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NUCLEAR TODAY<br />
Links to reference<br />
sources:<br />
European Commission<br />
announcement on<br />
Clean Energy for All:<br />
http://bit.ly/2fQbVQk<br />
Foratom position<br />
paper: http://bit.ly/<br />
2oI2Sna<br />
Clean Energy Proposals are Chance for<br />
Nuclear to have Rightful Place at Policy<br />
Table<br />
John Shepherd<br />
Every New Year’s Eve in Germany, there is a tradition that has become as fixed on the calendar as the ringing of<br />
church bells at midnight and the clinking of Champagne-filled glasses to toast the year ahead.<br />
This tradition, curiously, takes the form of a number of<br />
television stations broadcasting an 18-minute black-andwhite<br />
1963 TV recording of an English-language comedy<br />
sketch called ‘Dinner for One’ – also known as the ‘90th<br />
Birthday’.<br />
The show features the late British comedians Freddie<br />
Frinton and May Warden. For those of you who might be<br />
unfamiliar with the sketch – although its popularity has<br />
since spread to other European nations – it centres on the<br />
annual birthday dinner of upper-class Englishwoman, Miss<br />
Sophie. Every year, she hosts a celebration dinner for her<br />
friends. The problem is that, due to Miss Sophie’s considerable<br />
age, she has outlived all of her friends.<br />
The only ones at the annual celebration are Miss Sophie<br />
herself and her equally-aged manservant, James. His task<br />
it is to make his way around the dining table, impersonating<br />
each of the absent guests in turn. As he carries out his<br />
duties, James asks Miss Sophie: “Same procedure as every<br />
year?” To which she nods affirmatively. Poor James is also<br />
required to drink the copious glasses of alcohol on the<br />
guests’ behalf in toasts ordered by Miss Sophie throughout<br />
the evening until, inevitably, he becomes inebriated with<br />
hilarious consequences.<br />
I cannot do justice to the sketch here – you must see it<br />
for yourselves! But this anniversary event reminds me of a<br />
ritual that we frequently see in attempts to guide energy<br />
policy towards a sustainable future and to combat the<br />
effects of climate change.<br />
In April, Foratom, the Brussels-based trade association<br />
for the nuclear industry in Europe, published a position<br />
paper on the European Commission’s ‘Clean Energy for All<br />
Europeans’ package of EU legislative proposals.<br />
The proposals seek to improve the functioning of the<br />
energy market and ensure all energy technologies can<br />
compete on a level-playing field without jeopardising<br />
climate and energy targets.<br />
Foratom has called for “cost-efficient decarbonisation,<br />
an effective power market leading to competitive and<br />
affordable electricity prices for the end consumers and the<br />
promotion of investments in low carbon technologies”.<br />
Foratom also underlined the importance of the EU<br />
Emissions Trading Scheme (ETS) and “of protecting it<br />
from conflicting policy overlaps, in particular from the proposed<br />
new 30% energy efficiency binding target”.<br />
The European Commission’s proposals were unveiled<br />
towards the end of last year in a move designed to show the<br />
clean energy transition “is the growth sector of the future”.<br />
“Clean energies in 2015 attracted global investment of<br />
over EUR300 billion,” the Commission said. “The EU is<br />
well-placed to use our research, development and innovation<br />
policies to turn this transition into a concrete industrial<br />
opportunity. By mobilising up to EUR177bn of public and<br />
private investment per year from 2021, this package can<br />
generate up to 1% increase in gross domestic product over<br />
the next decade and create 900,000 new jobs.”<br />
The EU’s Commissioner for Climate Action and Energy<br />
Miguel Arias Canete said: "Our proposals provide a strong<br />
market pull for new technologies, set the right conditions<br />
for investors, empower consumers, make energy markets<br />
work better and help us meet our climate targets. I'm<br />
particularly proud of the binding 30% energy efficiency<br />
target, as it will reduce our dependency on energy imports,<br />
create jobs and cut more emissions.”<br />
The Commissioner said last year that Europe was “on<br />
the brink of a clean energy revolution”. He added: “We can<br />
only get this right if we work together. With these proposals,<br />
the Commission has cleared the way to a more<br />
competitive, modern and cleaner energy system. Now we<br />
count on the European Parliament and our member states<br />
to make it a reality.”<br />
Despite the Commissioner’s rallying call, can such an<br />
ambitious policy agenda ever come to fruition? And can<br />
“new technologies” also really encompass support for new<br />
nuclear technologies? Yes, the proposals may well be<br />
adopted by the parliament and EU nations, but I would<br />
suggest that while countries such as Austria and Germany<br />
are dead set against the further development and deployment<br />
of nuclear (within their own borders at least), how<br />
can the clean energy package ever work for the benefit of<br />
the EU as a whole?<br />
It is for this reason that I started this article by referring<br />
to Miss Sophie’s annual birthday dinner ritual. There are<br />
rather amusing parallels in the regular unveiling of various<br />
proposals at European and international level. It is the<br />
“same procedure”, if not quite every year, but as regular as<br />
clockwork. A new policy is drawn up that aims to be<br />
“ inclusive” and encourage all low-carbon energy technologies<br />
to compete to offer the best deals for electricity<br />
consumers and the environment.<br />
However, many of the EU nations who are called<br />
together to sit around the policy table at each of these<br />
ritual initiatives are, like Miss Sophie’s guests, not really<br />
there. Yes, nations are represented in a physical sense, but<br />
many “go through the motions”, to coin a phrase. They<br />
nod, toast the initiatives, then go back to what suits their<br />
political objectives at home.<br />
This is frequently the case in relation to the benefits of<br />
nuclear energy. European and international bodies recognise<br />
the benefits of nuclear as part of a mix of energy<br />
technologies, everyone agrees, then it is the “same procedure”<br />
as before, as James would say. Everyone removes<br />
from the mix what does not suit their domestic policies –<br />
and the result is an inebriated, incoherent performance, in<br />
this case an unbalanced energy policy.<br />
If Europe seeks to have a coherent and inclusive energy<br />
policy, which encompasses all low-carbon contributors,<br />
nuclear must be allowed a place at the policy table. If not,<br />
it will be a charade only worthy of the comic antics of<br />
James and Miss Sophie.<br />
Author<br />
John Shepherd<br />
nuclear 24<br />
41a Beoley Road West, St George’s<br />
Redditch B98 8LR, United Kingdom<br />
Nuclear Today<br />
Clean Energy Proposals are Chance for Nuclear to have Rightful Place at Policy Table ı John Shepherd
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The International Expert Conference on Nuclear Technology<br />
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