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3Energy Economy: Under the Banner of Jobs<br />
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Dear reader, Truly, these are no good news the two large integrated tech companies General Electric (GE) and Siemens<br />
released to their employees right before Christmas: GE announced to cut 12,000 jobs worldwide in its power plant division,<br />
6,800 in Europe alone. Siemens plans to cut 6,900 jobs, 6,100 of them in the power plant division. Thus, both groups<br />
continue job cutback in this field. GE has already massively cut jobs in Europe when having acquired the power plant<br />
division of the French company Alstom. And at Siemens job cutbacks in the energy sector is almost an ongoing theme.<br />
Both companies quote that this is ultimately a reaction to<br />
the changed worldwide investment landscape. The<br />
number of projects for these division-driving conventional<br />
gas and coal-fired power plants would decline distinctly so<br />
that with a constant offer a further decline in prices is<br />
inevitable not leaving enough margins anymore.<br />
With few exceptions blanket accusations, as common in<br />
the past, with the hint “one has not early enough and not<br />
sufficiently focused on other products such as renewable<br />
energies” stayed out. Possibly, political protagonists were<br />
aware of recent facts, which acknowledge renewable<br />
energies a rather moderate perspective as alleged job<br />
creator. Insolvency of the German Solarworld AG, decline of<br />
the Chinese “Solar Valley” as well as only few weeks earlier<br />
announced downsizing of around 6,000 out of 27,000 jobs<br />
at Siemens wind energy subsidiary Siemensgamesa are<br />
individual examples.<br />
Therefore, rather the overall situation should be<br />
considered which; according to the “World Energy Investment<br />
2017” report of the International Energy Agency has<br />
changed in the past years. For the first time since the year<br />
2000 the major share of all investments worldwide in the<br />
energy sector of 1,700 bn. $ were made in electricity<br />
production, superseding considerably higher investments<br />
in oil and gas in the prior years. This represents 2.2 % of<br />
the Gross World Product (GWP), however, a decrease by<br />
12 % from 1,900 bn. $ in the past year.<br />
The total of roughly 724 bn. US $ consists of 277 bn. US<br />
$ for grids (39 %), 297 bn. US $ for renewables (41 %) and<br />
143 bn. US $ for conventional production and as a total are<br />
situated clearly in the upper range of the past 20 years. The<br />
total investments inflation-adjusted were 750 bn. US $ in<br />
the year 2000. It has to be taken into account that these<br />
numbers do not display the actual contribution to the<br />
electricity supply and the supply security. Measured in<br />
production capacity these are 165,000 MW of newly<br />
installed plants in the renewables and around 95,000 MW<br />
of conventional plants.<br />
Eventually, the actual possible contribution to power<br />
supply reflects an entirely different relation. Natural availability<br />
of the renewables and with the technical avail ability<br />
of the conventional these 165,000 MW would approximately<br />
adequate 25,000 MW of conventional power.<br />
Regarding the distribution of investments and<br />
mentioned further high investments the question remains<br />
for the reasons of job downsizing.<br />
Here, my unloved, because rather nondescript term of<br />
globalisation plays a role.<br />
The market for all facilities and establishments in<br />
electricity production has transformed massively. For one,<br />
the investments have significantly shifted regionally.<br />
Nowadays it is invested in countries other than Europe or<br />
North America, in Asia, in Africa and in South America.<br />
And, even more essential, also the landscape of manufacturing<br />
has shifted towards Asia.<br />
Additionally, there is a noticeable deterioration of the<br />
political investment setting for conventional electricity<br />
production in western countries, even though it is about<br />
exports and therefore own local jobs. Lacking loan<br />
guarantees and for instance prohibition initiated by French<br />
politicians for any governmental subsidies for the export of<br />
conventional technology aggravates the situation. By the<br />
generalising term “Green Investments” it is hoped for<br />
popularity. It should be questioned if the stakeholders<br />
know that “new” players from Asia promptly close such a<br />
gap by not only bringing along the technology but also<br />
required financial management for foreign investment.<br />
Neither for the environment nor for jobs in this country<br />
such a general actionism is of any help. Still, politics needs<br />
to turn to indeed difficult conflicting priorities of politics,<br />
economy and citizens (voters).<br />
And this by not only the dimension of “environment”<br />
but also equally several other dimensions such as business<br />
and economic aspects, social interests, jobs and responsibility<br />
for future generations.<br />
The politics has to be granted that of course it is difficult<br />
to nearly impossible to foresightfully valuate single<br />
measures especially for the job market. A reliable economic<br />
model for governmental intervention does not exist.<br />
Diverse models between centrally planned economy and<br />
market liberalism is subject of discussion of the savants<br />
and belongs to the catalogue of the politics.<br />
There is one thing experience and common sense show:<br />
Permanent measures guided by governmental intervention,<br />
be it directly for jobs or particular industries are not<br />
beneficial. They only lead to distortion of the national<br />
performance and will place growing strains on the national<br />
economy. This gradual loss of in the end social security is<br />
dramatic.<br />
In the past decades the peaceful use of nuclear energy<br />
has contributed a considerable share to jobs, social security<br />
and the environment. In the 1990s and the 2000s around<br />
40,000 people were working directly in nuclear energy. In<br />
France, as country with the highest ratio of nuclear energy<br />
it is said that 400,000 jobs are directly or induced related<br />
to nuclear energy. The greatest, but not directly perceptible<br />
benefit of nuclear power and the performance of its<br />
employees lies in their macro-economic contribution. The<br />
non-subsidized jobs contribute significantly to an economically<br />
stable and attractive investment and market environment<br />
through favourable electricity generation costs and<br />
are thus a basis for a secure and viable infrastructure.<br />
However, it is the politics themselves that is called upon<br />
for work and social welfare in an increasingly distorted<br />
energy policy: only a sustainable and fair framework without<br />
permanent money transfers for all technologies creates jobs<br />
and promotes social development before the reality of<br />
globalisation with all its negative consequences will catch up.<br />
Christopher Weßelmann<br />
– Editor in Chief –<br />
EDITORIAL<br />
Editorial<br />
Energy Economy: Under the Banner of Jobs
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
4<br />
EDITORIAL<br />
Christopher<br />
Weßelmann<br />
– Chefredakteur –<br />
Energiewirtschaft:<br />
Im Zeichen von Arbeitsplätzen<br />
Liebe Leserin, lieber Leser, es sind wahrlich keine guten Nachrichten, mit denen die zwei großen integrierten<br />
Technologieunternehmen General Electric (GE) und Siemens kurz vor Weihnachten und Jahreswechsel an ihre Belegschaft<br />
gingen: GE kündigte an weltweit rund 12.000 Arbeitsplätze in seiner Kraftwerkssparte zu streichen, 6.800 davon in Europa;<br />
Siemens plant, 6.900 Stellen zu streichen, 6.100 davon im Kraftwerksbereich. Damit setzen beide Konzerne ihren Arbeitsplatzabbau<br />
in diesem Bereich fort. GE hatte schon mit der Übernahme des Kraftwerksbereichs der französischen Alstom massiv in<br />
Europa Stellen gestrichen und bei Siemens ist Arbeitsplatzabbau im Energiesektor fast schon ein Dauerthema.<br />
Von beiden Unternehmen wird angeführt, dass dies letzt endlich<br />
eine Reaktion auf die veränderte weltweite Investitionslandschaft<br />
sei. Die Anzahl von Projekten für die vor allem diese<br />
Sparten tragenden konventionellen Gas- und Kohlekraft werke<br />
würden deutlich zurück gehen, sodass bei gleich bleibendem<br />
Angebot der ein weiterer Preisverfall unausweichlich sei und<br />
damit nicht mehr ausreichend Margen gegeben seien.<br />
Mit wenigen Ausnahmen waren pauschale Schuldzuweisungen,<br />
wie in der Vergangenheit üblich, mit dem Hinweis<br />
„man habe sich nicht frühzeitig genug auf andere Produkte,<br />
sprich erneuerbare Energien, gestützt,“ ausgeblieben. Vielleicht<br />
waren hier politischen Protagonisten doch einige Fakten aus<br />
dem Jahresverlauf präsent, die auch für die Erneuerbaren als<br />
vermeintlicher Arbeitsplatzmotor eher dämpfende Aussichten<br />
bescheinigen. Die Insolvenz der deutschen Solarworld AG, der<br />
Niedergang des chinesischen „Solar Valley“ und auch der nur<br />
wenige Wochen vorher seitens der Siemens Wind- Tochter<br />
„ Siemensgamesa“ angekündigte weltweite Stellen abbau im<br />
Umfang von voraussichtlich 6.000 Arbeitsplätzen – bei einer<br />
Gesamtbelegschaft von rund. 27.000 noch wesentlich eingreifender<br />
als beim konventionellen Geschäft – sind Einzelbeispiele.<br />
Zu betrachten ist also eher die Gesamtsituation, die sich<br />
gemäß dem aktuellen „World Energy Investment 2017“ Report<br />
der Internationalen Energie Agentur (International Energy<br />
Agency) deutlich in den vergangenen Jahren gewandelt hat.<br />
Von den erfassten weltweiten Gesamtinvestitionen im<br />
Energie sektor in Höhe von 1.700 Mrd. US-$ im Jahr 2016<br />
(dies entspricht rund 2,2 % des globalen Bruttosozialproduktes,<br />
bedeutet im Vorjahresvergleich mit 1.900 Mrd. US-$<br />
aber auch einen Rückgang um 12 %) entfällt erstmals seit dem<br />
Jahr 2000 der größte Anteil auf die Stromerzeugung, die<br />
damit die in den Vorjahren deutlich höheren Investitionen in<br />
den Öl & Gas Sektor ablöst. Die rund 724 Mrd. US-$ teilen sich<br />
auf in 277 Mrd. US-$ für Netze (39 %), 297 US-$ für Erneuerbare<br />
(41 %) und 143 Mrd. US-$ für konventionelle Erzeugung<br />
und sie liegen noch deutlich im oberen Bereich der vergangenen<br />
20 Jahre – Inflationsbereinigt lagen z.B. die Gesamtinvestitionen<br />
im Jahr 2000 bei rund 750 Mrd. US-$. Zu beachten<br />
ist, dass diese Zahlen nicht den tatsächlichen Beitrag für<br />
Stromversorgung und Stromversorgungssicherheit abbilden.<br />
In Erzeugungsleistung gemessen ergeben sich für das Jahr<br />
2016 rund 165.000 MW an neu installierten Anlagen im<br />
Bereich der Erneuerbaren und rund 95.000 MW an konventionellen<br />
Anlagen. Der schlussendlich tatsächliche, mögliche<br />
Beitrag für die Energieversorgung spiegelt dann noch ein ganz<br />
anderes Verhältnis wider, denn aufgrund der natürlichen<br />
Verfügbarkeit bei den Erneuerbaren und mit den technischen<br />
Verfügbarkeiten der Konventionellen würden die 165.000 MW<br />
in etwa 25.000 MW an konventioneller Leistung entsprechen.<br />
In Summe einer Betrachtung des Investitionskuchens<br />
sowie der erwähnten weiter hohen Investitionen verbleibt die<br />
Frage nach den Gründen für den Stellenabbau.<br />
Hier spielt dann doch einmal mein ungeliebter, weil meist<br />
ohne Inhalte gefüllter Begriff der Globalisierung die Rolle.<br />
Der Markt für alle Anlagen und Einrichtungen in der Stromerzeugung<br />
hat sich stark gewandelt. Zum einen, weil sich die<br />
Investitionen regional erheblich verschoben haben. Investiert<br />
wird heute außerhalb von Europa und Nordamerika, in Asien, in<br />
Afrika, in Südamerika. Und, was noch wesentlicher ist, auch die<br />
Herstellerlandschaft hat sich in Richtung Asien verschoben.<br />
Hinzu kommt eine erkennbare Verschlechterung des poli tischen<br />
Investitionsumfeld für die konventionelle Stromer zeugung in<br />
westlichen Ländern, auch wenn es um Exporte und damit<br />
eigene, heimische Arbeits plätze geht. Fehlende Kreditbürgschaften<br />
und eine z.B. von Politiken in Frankreich gebotenes<br />
Verbot für jegliche staat liche Unterstützung beim Export<br />
konventioneller Technologie verschärfen die Situation. Unter<br />
dem pauschalisierenden Begriff „Grüner Investitionen“ erhofft<br />
man sich Popularität. Ob die Akteure wissen, dass „neue“ Akteure<br />
aus Asien prompt eine solche sich auftuende Lücke schließen<br />
und nicht nur die Technologie mitbringen sondern auch das für<br />
Auslands investitionen erforderliche Finanzmanagement, sollte<br />
gefragt werden. Für die Umwelt bringt solcher pauschaler<br />
Aktio nismus jedenfalls nichts, und für Arbeitsplätze hierzulande<br />
auch nicht. Dennoch muss sich Politik im zugegeben<br />
schwie rigen Spannungsfeld von Politik, Wirtschaft und Bürger<br />
(Wähler) nicht nur der Dimension „Umwelt“ zuwenden,<br />
sondern weitere wie Volk- und Betriebswirtschaftliche Aspekte,<br />
soziale Interessen, Arbeitsplätze und Verantwortung für<br />
zukünftige Generationen gleichermaßen berücksichtigen.<br />
Dabei ist der Politik zugute zu halten, dass es natürlich<br />
schwierig bis unmöglich ist, Einzelmaßnahmen gerade für<br />
den Arbeitsmarkt vorausschauend zu bewerten. Ein ver lässliches<br />
Volkswirtschaftliches Modell für staatliche Interventionen<br />
gibt es nicht. Die verschiedensten Modelle zwischen<br />
Planwirtschaft und vollständigem Marktliberalismus sind seit<br />
jeher Diskussionsgegenstand der Gelehrten und gehören zum<br />
Katalog der Politik. Eines zeigen die einfache Erfahrung und<br />
der gesunde Menschenverstand: Dauerhaft durch staatliche<br />
Interventionen gelenkte Maßnahmen, sei es direkt für Arbeitsplätze<br />
oder einzelne Wirtschaftszweige, zeichnen sich<br />
nicht aus. Diese führen nur zu Verzerrungen der nationalen<br />
Leistung und setzen die Nationalökonomie im heutigen<br />
globalen Wettbewerb später unter Druck. Dieser schleichende<br />
Verlust von am Ende sozialer Sicherheit ist dramatisch.<br />
Die friedliche Nutzung der Kernenergie hat in den vergangenen<br />
Jahrzehnten ihren volkswirtschaftlich bedeutenden<br />
Beitrag für Arbeitsplätze, soziale Sicherung und Umwelt<br />
geleistet. In den 1990er und 2000er Jahren waren in Deutschland<br />
rund 40.000 Menschen direkt für die Kernenergie tätig. In<br />
Frankreich, dem Land mit dem weltweit höchsten Kernenergieanteil,<br />
wird von 400.000 Arbeitsplätzen gesprochen, die direkt<br />
oder induziert in Zusammenhang mit der Kernenergie stehen.<br />
Der weitaus größte aber nicht direkt fühlbar Nutzen der Kernenergie<br />
und der Leistung ihrer Beschäftigten liegt im volkswirtschaftlichen<br />
Beitrag. Die nicht subventionierten Arbeitsplätze<br />
tragen über günstige Stromerzeugungskosten wesentlich für<br />
ein ökonomisch stabiles und attraktives Investitions- und<br />
Marktumfeld bei und sind damit eine Grundlage für eine<br />
sichere und Erfolg versprechende Infrastruktur.<br />
Doch gefordert ist für Arbeit und Soziales in einer immer<br />
mehr und mehr verzerrten Energiepolitik dann doch die Politik<br />
selbst: Nur zukunftsfähige, auf Dauerhaftigkeit zielende und<br />
faire Rahmenbedingungen ohne dauerhafte Geldtransfers für<br />
alle Technologien schaffen Arbeitsplätze und fördern soziale<br />
Entwicklung, bevor einen die Realität der Globalisierung mit<br />
allen negativen Konsequenzen einholen wird.<br />
Editorial<br />
Energy Economy: Under the Banner of Jobs
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Atomrecht – Was Sie wissen müssen 30.01.<strong>2018</strong><br />
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Erfolgreicher Wissenstransfer in der Kerntechnik<br />
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Advancing Your Nuclear English<br />
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Stilllegung, Rückbau und Entsorgung – 24.09. - 25.09.<strong>2018</strong> Berlin<br />
Recht und Praxis<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
6<br />
Issue 1<br />
January<br />
CONTENTS<br />
13<br />
ETSON Strategic<br />
Orientations<br />
on Research Activities<br />
| | View of two of four reactors at the Ringhals nuclear power plant site in the Varberg Municipality approximately 65 km south<br />
of Gothenburg, Sweden. (Courtesy: Vattenfall AB)<br />
Editorial<br />
Energy Economy: Under the Banner of Jobs 3<br />
Energiewirtschaft:<br />
Im Zeichen von Arbeitsplätzens 4<br />
Abstracts | English 8<br />
Abstracts | German 9<br />
Energy Policy, Economy and Law<br />
ETSON Strategic Orientations<br />
on Research Activities.<br />
ETSON Research Group Activity 13<br />
J.P. Van Dorsselaere, M. Barrachin, D. Millington, M. Adorni,<br />
M. Hrehor, F. Mascari, A. Schaffrath, I. Tiselj, E. Uspuras,<br />
Y. Yamamoto, D. Gumenyuk, N. Fedotova, O. Cronvall and P. Liska<br />
Spotlight on Nuclear Law<br />
Council Regulation of the European Dual Use<br />
Regulation – A Never Ending Story? 19<br />
Die Novellierung der europäischen Dual-Use<br />
Verordnung – eine unendliche Geschichte? 19<br />
Ulrike Feldmann<br />
10<br />
DAtF Notes 20<br />
| | AP1000 new build in Haiyang, China.<br />
Inside Nuclear with NucNet<br />
UK Is Leading the Way<br />
With Clear Strategy for Nuclear 10<br />
NucNet<br />
Calendar 12<br />
21<br />
| | Nuclear Triple “S”.<br />
Contents
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
7<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards:<br />
An Application of an Integrated Approach 21<br />
Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy,<br />
Robert Rodger and Jonathan Scott<br />
Fuel<br />
Review of Fuel Safety Criteria in France 38<br />
Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne<br />
and Olivier Dubois<br />
CONTENTS<br />
Operation and New Build<br />
Clearance of Surface-contaminated Objects<br />
from the Controlled Area of a Nuclear Facility:<br />
Application of the SUDOQU Methodology 29<br />
F. Russo, C. Mommaert and T. van Dillen<br />
AMNT 2017<br />
Key Topic |<br />
Outstanding Know-How & Sustainable Innovations<br />
Technical Session:<br />
Reactor Physics, Thermo and Fluid Dynamics<br />
Neutron Flux Oscillations Phenomena 44<br />
Joachim Herb<br />
Key Topic |<br />
Enhanced Safety & Operation Excellence<br />
Focus Session:<br />
Radiation Protection 46<br />
29<br />
Erik Baumann and Angelika Bohnstedt<br />
| Variation of the total dose values in the analysed scenarios.<br />
AMNT <strong>2018</strong><br />
Preliminary Programme 47<br />
|38<br />
Decommissioning and Waste Management<br />
Carbon-14 Speciation During Anoxic Corrosion<br />
of Activated Steel in a Repository Environment 34<br />
KTG Inside 54<br />
E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat<br />
34<br />
54<br />
| | KTG Inside. Horst Kemmeter, speaking at a WiN meeting in Biblis.<br />
| | Sketch of the reactor.<br />
News 57<br />
Nuclear Today<br />
‘Newcomer’ Nuclear Nation<br />
Leads Way into New Nuclear Year 66<br />
John Shepherd<br />
Imprint 11<br />
| | Topics reviewed in the frame of French rulemaking<br />
on fuel safety criteria.<br />
AMNT <strong>2018</strong>: Registration Form . . . . . . . . . . . Insert<br />
Contents
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
8<br />
ABSTRACTS | ENGLISH<br />
UK Is Leading the Way With Clear Strategy<br />
for Nuclear<br />
NucNet | Page 10<br />
The UK is Europe’s most prominent leader in nuclear<br />
development because of the government’s clear<br />
strategy of supporting nuclear energy as part of its<br />
future energy mix, a senior official from US-based<br />
nuclear equipment manufacturer Westinghouse<br />
Electric Company said. Mr Kirst told that the UK<br />
government’s decision to support the financing of<br />
new energy projects, including nuclear, by way of a<br />
contract for difference scheme was a breakthrough.<br />
Additionally potential for nuclear development in<br />
other EU member states is possible in Poland and the<br />
Czech Republic where also new nuclear capacities<br />
are possible. Potential exists also in non-EU countries<br />
like Turkey and the Ukraine.<br />
ETSON Strategic Orientations on Research<br />
Activities. ETSON Research Group Activity<br />
J.P. Van Dorsselaere, M. Barrachin, D. Millington,<br />
M. Adorni, M. Hrehor, F. Mascari, A. Schaffrath,<br />
I. Tiselj, E. Uspuras, Y. Yamamoto, D. Gumenyuk,<br />
N. Fedotova, O. Cronvall and P. Liska | Page 13<br />
In 2011, ETSON published the “Position Paper of<br />
the Technical Safety Organizations: Research Needs<br />
in Nuclear Safety for Gen 2 and Gen 3 NPPs”. This<br />
paper, published only a few months after the<br />
Fukushima- Daiichi severe accidents, presented the<br />
priorities for R&D on the main pending safety<br />
issues. It was produced by the ETSON Research<br />
Group (ERG) that has the mandate of identifying<br />
and prioritizing safety research needs, sharing<br />
information on research projects in which ETSON<br />
members are involved, defining and launching new<br />
research projects and disseminating knowledge<br />
among ETSON members. Six years after this<br />
publication, many R&D international projects<br />
finished in diverse frames, and other ones have<br />
started. In particular a lot of work was done (and is<br />
going on…) on the analysis of the Fukushima-<br />
Daiichi severe accidents. Meanwhile a roadmap on<br />
research on Gen. 2 and 3 nuclear power plants<br />
(NPP), including safety aspects, was produced by<br />
the NUGENIA association, followed by a more<br />
detailed document as “NUGENIA global vision”. It<br />
was also demonstrated that the ETSON R&D<br />
priorities were consistent with the implementation<br />
of the 2014 Euratom Directive on safety of nuclear<br />
installations.<br />
Council Regulation of the European Dual<br />
Use Regulation – A Never Ending Story?<br />
Ulrike Feldmann | Page 19<br />
For the first time, the EC Council Regulation of<br />
19 December 1994 established a Community regime<br />
for the control of exports of dual-use items. In 2000,<br />
the first major revision of the dual-use regime came<br />
into force, subjecting not only sensitive material, i.<br />
e. plutonium and highly enriched uranium, but also<br />
the entire category 0 (nuclear material, installations,<br />
equipment) to a licensing requirement for intra-<br />
Community shipments. This revision was revised a<br />
few months later due to inappropriate content by<br />
removing a small proportion of nuclear goods. A<br />
further comprehensive new revision was published<br />
in 2009. However, the EU Commission’s current<br />
proposal to revise Annex IV of the regulation does<br />
not do justice to the objective of free trade of goods<br />
and the maintenance of the competitiveness of<br />
European industry from the point of view of the<br />
European nuclear industry, as well as from the point<br />
of view of the non-nuclear industry in the EU.<br />
Nuclear Safety, Security and Safeguards:<br />
An Application of an Integrated Approach<br />
Howard Chapman, Jeremy Edwards,<br />
Joshua Fitzpatrick, Colette Grundy,<br />
Robert Rodger and Jonathan Scott | Page 21<br />
National Nuclear Laboratory has recently produced<br />
a paper regarding the integrated approach of<br />
nuclear safety, security and safeguards. The paper<br />
considered the international acknowledgement of<br />
the inter-relationships and potential benefits to be<br />
gained through improved integration of the nuclear<br />
‘3S’; Safety, Security and Safeguards. It considered<br />
that combining capabilities into one synergistic<br />
team can provide improved performance and value.<br />
This approach to integration has been adopted, and<br />
benefits realised by the National Nuclear Laboratory<br />
through creation of a Safety, Security and<br />
Safeguards team. In some instances the interface is<br />
clear and established, as is the case between safety<br />
and security in the areas of Vital Area Identification.<br />
In others the interface is developing such as the<br />
utilisation of safeguards related techniques such as<br />
nuclear material accountancy and control to<br />
enhance the security of materials. This paper looks<br />
at a practical example of the progress to date in<br />
implementing Triple S by a duty holder.<br />
Clearance of Surface-contaminated Objects<br />
from the Controlled Area of a Nuclear<br />
Facility: Application of the SUDOQU<br />
Methodology<br />
F. Russo, C. Mommaert and T. van Dillen | Page 29<br />
The lack of clearly defined surface-clearance levels in<br />
the Belgian regulation led Bel V to start a collaboration<br />
with the Dutch National Institute for Public<br />
Health and the Environment (RIVM) to evaluate the<br />
applicability of the SUDOQU methodology for the<br />
derivation of nuclide-specific surface-clearance<br />
criteria for objects released from nuclear facilities.<br />
SUDOQU is a methodology for the dose assessment<br />
of exposure to a surface-contaminated object, with<br />
the innovative assumption of a time-dependent<br />
surface activity whose evolution is influenced by<br />
removal and deposition mechanisms. In this work,<br />
calculations were performed to evaluate the annual<br />
effective dose resulting from the use of a typical<br />
office item, e.g. a bookcase. Preliminary results allow<br />
understanding the interdependencies between the<br />
model’s underlying mechanisms, and show a strong<br />
sensitivity to the main input parameters. The results<br />
were benchmarked against those from a model described<br />
in Radiation Protection 101, to investigate<br />
the impact of the model’s main assumptions. Results<br />
of the two models were in good agreement.<br />
The SUDOQU methodology appears to be a flexible<br />
and powerful tool, suitable for the proposed application.<br />
Therefore, the project will be extended to<br />
more generic study cases, to eventually develop surface-clearance<br />
levels applicable to objects leaving<br />
nuclear facilities.<br />
Carbon-14 Speciation During Anoxic<br />
Corrosion of Activated Steel in a Repository<br />
Environment<br />
E. Wieland, B.Z. Cvetkovic, D. Kunz,<br />
G. Salazar and S. Szidat | Page 34<br />
Radioactive waste contains significant amounts<br />
of 14 C which has been identified a key radionuclide<br />
in safety assessments. In Switzerland, the 14 C inventory<br />
of a cement-based repository for low- and<br />
intermediate-level radioactive waste (L/ILW) is<br />
mainly associated with activated steel (~85 %). 14 C<br />
is produced by 14 N activation in steel parts exposed<br />
to thermal neutron flux in light water reactors.<br />
Release of 14 C occurs in the near field of a deep<br />
geological repository due to anoxic corrosion of<br />
activated steel. Although the 14 C inventory of the<br />
L/ILW repository and the sources of 14 C are well<br />
known, the formation of 14 C species during steel<br />
corrosion is only poorly understood. The aim of the<br />
present study was to identify and quantify the<br />
14 C-bearing carbon species formed during the<br />
anoxic corrosion of iron and steel and further to<br />
determine the 14C speciation in a corrosion experiment<br />
with activated steel. All experiments were<br />
conducted in conditions similar to those anticipated<br />
in the near field of a cement-based repository.<br />
Review of Fuel Safety Criteria in France<br />
Sandrine Boutin, Stephanie Graff,<br />
Aude Foucher-Taisne and Olivier Dubois | Page 38<br />
Fuel safety criteria for the first barrier, based on<br />
state-of-the-art at the time, were first defined in the<br />
1970s and came from the United States, when the<br />
French nuclear program was initiated. Since then,<br />
there has been continuous progress in knowledge<br />
and in collecting experimental results thanks to the<br />
experiments carried out by utilities and research<br />
institutes, to the operating experience, as well as to<br />
the generic R&D programs, which aim notably at<br />
improving computation methodologies, especially<br />
in Reactivity-Initiated accident and Loss-of-Coolant<br />
Accident conditions. In this context, the French<br />
utility EDF proposed new fuel safety criteria, or<br />
reviewed and completed existing safety demonstration<br />
covering the normal operating, incidental<br />
and accidental conditions of Pressurised Water<br />
Reactors. IRSN assessed EDF’s proposals and presented<br />
its conclusions to the Advisory Committee<br />
for Reactors Safety of the Nuclear Safety Authority<br />
in June 2017. This review focused on the relevance<br />
of historical limit values or parameters of fuel safety<br />
criteria and their adequacy with the state-of-the-art<br />
concerning fuel physical phenomena (e.g. Pellet-<br />
Cladding Mechanical Interaction in incidental conditions,<br />
clad embrittlement due to high temperature<br />
oxidation in accidental conditions, clad ballooning<br />
and burst during boiling crisis and fuel melting).<br />
AMNT 2017: Outstanding Know-How &<br />
Sustainable Innovations – Technical Session:<br />
Reactor Physics, Thermo and Fluid Dynamics<br />
Enhanced Safety & Operation Excellence –<br />
Focus Session: Radiation Protection<br />
Joachim Herb, Erik Baumann and<br />
Angelika Bohnstedt | Page 44<br />
Summary report on the Key Topics “Outstanding<br />
Know-How & Sustainable Innovations – Technical<br />
Session: Reactor Physics, Thermo and Fluid<br />
Dynamics” and “Enhanced Safety & Operation Excellence<br />
– Focus Session: Radiation Protection” of<br />
the 48 th Annual Meeting on Nuclear Technology<br />
(AMNT 2017) held in Berlin, 16 to 17 May 2017.<br />
‘Newcomer’ Nuclear Nation Leads Way Into<br />
New Nuclear Year<br />
John Shepherd | Page 66<br />
At the start of a new year, it is appropriate that a<br />
‘newcomer’ nuclear nation has launched work on<br />
building its first nuclear power plant. First nuclear<br />
safety-related concrete has been poured for the<br />
plant at Rooppur in Bangladesh – making the South<br />
Asia nation the first in 30 years to start building its<br />
first commercial reactor unit following the United<br />
Arab Emirates in 2012 and Belarus in 2013.<br />
Despite setbacks that nuclear has endured in recent<br />
years, there are nearly 60 reactors under construction<br />
around the world, mostly in Asia. Some<br />
447 commercial reactor units are in operation in<br />
30 countries.<br />
Abstracts | English
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Großbritannien ist führend mit<br />
seiner klares Strategie für die Kernenergie<br />
NucNet | Seite 10<br />
Großbritannien ist in Europa führend bei der<br />
zukünftigen Kernenergieentwicklung aufgrund der<br />
klaren Strategie der Regierung, die Kernenergie als<br />
Teil ihres zukünftigen Energiemixes zu unterstützen.<br />
Dies hob Michael Kirst voms US-Kern technik<br />
unternehmen West inghouse Electric Company<br />
hervor. Die Entscheidung der britischen Regierung,<br />
die Finanzierung neuer Energieprojekte, einschließlich<br />
der Kernenergie, im Wege eines<br />
Differenz vertrags zu unterstützen, sei ein Durchbruch<br />
gewesen. Darüber hinaus sind in anderen<br />
EU-Mitgliedsstaaten, wie Polen und Tschechien,<br />
Potenziale auch für neue Kernkraftwerke vorhanden.<br />
Potenziale bestehen auch in Nicht-EU-<br />
Ländern, so in der Türkei und der Ukraine.<br />
ETSON Strategische Ausrichtung<br />
für Forschungsaktivitäten.<br />
Aktivitäten der ETSON-Forschungsgruppe<br />
J.P. Van Dorsselaere, M. Barrachin, D. Millington,<br />
M. Adorni, M. Hrehor, F. Mascari, A. Schaffrath,<br />
I. Tiselj, E. Uspuras, Y. Yamamoto, D. Gumenyuk,<br />
N. Fedotova, O. Cronvall und P. Liska | Seite 13<br />
Im Jahr 2011 veröffentlichte ETSON das „Positionspapier<br />
der Technischen Sicherheitsorganisationen:<br />
Forschungsbedarf für die nukleare Sicherheit für<br />
die Kernkraftwerke der Generation 2 und 3“.<br />
Nur wenige Monate nach den schweren Unfällen<br />
von Fukushima-Daiichi wurden Prioritäten für<br />
Forschung und Entwicklung in Bezug auf wichtige<br />
noch offene Fragen zur Sicherheit vorgestellt. Das<br />
Positionspapier wurde von der ETSON Research<br />
Group (ERG) erstellt, die den Auftrag hat, den<br />
Bedarf an Sicherheitsforschung zu ermitteln und<br />
zu priorisieren, Informationen über Forschungs projekte<br />
von ETSON-Mitgliedern auszutauschen, neue<br />
Forschungsprojekte zu definieren und zu lancieren<br />
und den ETSON-Mitgliedern Informationen bereit<br />
zu stellen. Sechs Jahre nach dieser Veröffentlichung<br />
sind viele internationale F&E-Projekte abge schlossen,<br />
andere haben begonnen. Insbesondere an der<br />
Analyse der schweren Unfälle von Fukushima-<br />
Daiichi wurde gearbeitet. Zwischenzeitlich hat<br />
NUGENIA einen Fahrplan für die Sicherheitsforschung<br />
erstellt und das detaillierte Dokument<br />
„NUGENIA Global Vision“ veröffentlicht. Die F&E-<br />
Prioritäten von ETSON stehen zudem in Einklang<br />
mit der Umsetzung der Euratom-Richt linie 2014.<br />
Die Novellierung der europäischen<br />
Dual Use-Verordnung – eine unendliche<br />
Geschichte?<br />
Ulrike Feldmann | Seite 19<br />
Erstmalig wurde mit der Verordnung des Rates vom<br />
19.12.1994 eine Gemeinschaftsregelung für die<br />
Ausfuhrkontrolle von Gütern mit doppeltem Verwendungszweck<br />
geschaffen. Im Jahr 2000 fand die<br />
erste größere Revision der Dual-Use Regelungen<br />
statt, mit der für den Nuklearbereich nicht nur<br />
sensitives Material, d.h. Plutonium und hochangereichertes<br />
Uran sondern die gesamte Kategorie<br />
0 (Nuklearmaterial, Anlagen, Ausrüstung) auch<br />
einer Genehmigungspflicht für die innergemeinschaftliche<br />
Verbringung unterworfen wurde, die<br />
aufgrund nicht angebrachter Inhalte wenige<br />
Monate später revidiert wurde durch Herausnahme<br />
eines kleinen Teils von Nukleargütern. 2009<br />
erschien eine weitere umfassende neue Revision.<br />
Der aktuelle Revisionsvorschlag der EU-Kommission<br />
zum Annex IV der Verordnung wird dem<br />
Ziel des freien Warenverkehrs und dem Erhalt der<br />
Wettbewerbsfähigkeit der europäischen Industrie<br />
jedoch aus Sicht der europäischen Nuklearindustrie<br />
wie auch aus Sicht der nicht-nuklearen Industrie in<br />
der EU nicht gerecht.<br />
Nukleare Sicherheit, Gefahrenabwehr und<br />
Safeguards: Anwendung eines integrierten<br />
Ansatzes<br />
Howard Chapman, Jeremy Edwards,<br />
Joshua Fitzpatrick, Colette Grundy,<br />
Robert Rodger und Jonathan Scott | Seite 21<br />
Das National Nuclear Laboratory hat eine Studie<br />
über einen integrierten Ansatz zur nuklearen<br />
Sicherheit, sowie Gefahrenabwehr und Safeguards<br />
erstellt. Vorgestellt werden die Wechselbeziehungen<br />
und Vorteile, die durch eine bessere<br />
Integration der nuklearen“3S“ (Safety, Security and<br />
Safeguards) erzielt werden können. Ein integrierter<br />
Anssatz kann dabei potenzielle Synergien schöpfen<br />
und Vorteile erschließen. Dieser integrierte Ansatz<br />
wurde bei der Bildung eines Teams für Sicherheit,<br />
Gefahrenabwehr und Safeguards des NNL übernommen.<br />
In einigen Anwendungsfällen sind die<br />
Schnittstellen eindeutig in anderen müssen sie<br />
weiter entwickelt werden. Vorgestellt wird ein<br />
praktisches Beispiel für die bisherigen Fortschritte<br />
bei der Umsetzung von Triple S anhand eines<br />
Sicherheitsbeauftragen.<br />
Freigabe oberflächenkontaminierter<br />
Objekte aus dem Kontrollbereich<br />
eines Kernkraftwerkes<br />
Anwendung der SUDOQU-Methode<br />
F. Russo, C. Mommaert und T. van Dillen | Seite 29<br />
Das Fehlen definierter Grenzwerte für die Oberflächenkontamination<br />
in der betreffenden belgischen<br />
Verordnung veranlasste Bel V in Zusammenarbeit mit<br />
dem National Institute for Public Health and the<br />
Environment (Niederlande) die Anwendung der<br />
SUDOQU-Methode für die Ableitung nuklidspezifischer<br />
Oberflächendosiskriterien für Objekte zu<br />
evaluieren, die aus kerntechnischen Anlagen freigemessen<br />
werden sollen. SUDOQU ist eine Methode zur<br />
Dosisbewertung der Exposition eines oberflächenkontaminierten<br />
Objekts unter der Annahme einer<br />
zeitabhängigen Oberflächenaktivität, deren Entwicklung<br />
von Entfernungs- und Ablagerungsmechanismen<br />
beeinflusst wird. Berechnungen zur Ermittlung<br />
der effektiven Jahresdosis werden vorgestellt,<br />
die sich aus der Verwendung eines typischen Büroartikels<br />
ergibt. Vorläufige Ergebnisse erlauben es,<br />
die Wechselwirkungen zwischen den zugrunde<br />
liegenden Mechanismen des Modells zu verstehen<br />
und zeigen eine starke Sensitivität gegenüber den<br />
wichtigsten Eingangsparametern. Die Ergebnisse<br />
wurden mit denen eines weiteren beschriebenen<br />
Modells verglichen. Die Ergebnisse der beiden<br />
Modelle stimmten gut überein.<br />
Die SUDOQU-Methode scheint ein flexibles und<br />
leistungsfähiges Werkzeug zu sein, das für die<br />
vorgeschlagene Anwendung geeignet ist. Das<br />
Projekt wird auf allgemeinere Fälle ausgeweitet, um<br />
Oberflächenfreigabekriterien zu entwickeln, die für<br />
Objekte aus kerntechnischen Anlagen anwendbar<br />
sind.<br />
Kohlenstoff-14-Verhalten bei der<br />
anaerober Korrosion von aktiviertem Stahl<br />
in einer Endlagerumgebung<br />
E. Wieland, B.Z. Cvetkovic, D. Kunz,<br />
G. Salazar und S. Szidat | Seite 34<br />
Radioaktive Abfälle enthalten signifikante Mengen<br />
von 14 C, die in Sicherheitsbewertungen als ein<br />
Leitradionuklid identifiziert wurden. In der Schweiz<br />
wird das 14 C-Inventar eines Endlagers für mit Zement<br />
konditionierte schwach- und mittelradioaktive<br />
Abfälle hauptsächlich von aktiviertem Stahl (~85 %)<br />
dominiert. 14 C wird durch 14 N-Aktivierung in Stahlkomponenten<br />
gebildet, die dem thermischen Neutronenfluss<br />
in Leichtwasserreaktoren ausgesetzt<br />
sind. Die Freisetzung von 14 C erfolgt im Nahfeld eines<br />
geologischen Tiefenlagers durch anaerobe Korrosion<br />
des aktivierten Stahls. Obwohl das 14 C-Inventar des<br />
Endlagers und die Quellen von 14 C bekannt sind, ist<br />
zur Bildung von 14 C-Ver bindungen bei der Korrosion<br />
von Stahl nur wenig bekannt. Das Ziel der vorliegenden<br />
Studie war es, die 14 C-haltigen Kohlenstoffver<br />
bindungen, die während der anaeroben<br />
Korrosion von Eisen und Stahl gebildet werden, zu<br />
identifizieren und quantifizieren und die 14 C-Verbindungen<br />
in einem Korrosionsexperiment mit<br />
aktiviertem Stahl zu bestimmen. Alle Experimente<br />
wurden unter ähn lichen Bedingungen wie im<br />
Nahfeld eines Endlagers durchgeführt.<br />
Überprüfung der Kriterien für die Sicherheit<br />
von Kernbrennstoff in Frankreich<br />
Sandrine Boutin, Stephanie Graff,<br />
Aude Foucher-Taisne und Olivier Dubois | Seite 38<br />
Die Kriterien für die Sicherheit der ersten Barriere<br />
des Kernbrennstoff gegenüber Spalt produkt freisetzung<br />
wurden in den 1970er Jahren definiert als<br />
das französische Atomprogramm initiiert wurde.<br />
Seitdem haben sich Wissen und Erfahrungen<br />
dank der von den Kernkraftwerksbetreibern und<br />
Forschungsinstituten durchgeführten Experimente,<br />
Betriebserfahrungen sowie generischer F&E-<br />
Programme, die insbesondere auf die Verbesserung<br />
der Berechnungsmethoden abzielen, kontinuierlich<br />
weiterentwickelt. Der französische Energieversorger<br />
EDF schläg neue Kriterien für die Brennstoffsicherheit<br />
vor und überprüft und ergänzt<br />
be stehende Sicherheitskriterien, die sich auf<br />
die normalen Betriebs-, Ereignis- und Unfallbedingungen<br />
von Druckwasserreaktoren beziehen.<br />
IRSN hat die Vorschläge des EDF bewertet und seine<br />
Schlussfolgerungen im Juni 2017 dem Beratenden<br />
Ausschuss für Reaktorsicherheit der Französischen<br />
Behörde für nukleare Sicherheit vorgelegt.<br />
AMNT 2017: Outstanding Know-How &<br />
Sustainable Innovations – Technical Session:<br />
Reactor Physics, Thermo and Fluid Dynamics<br />
Enhanced Safety & Operation Excellence –<br />
Focus Session: Radiation Protection<br />
Joachim Herb, Erik Baumann und<br />
Angelika Bohnstedt | Seite 44<br />
Zusammenfassender Bericht zu den Sessions der<br />
Key Topics „Outstanding Know-How & Sustainable<br />
Innovations – Technical Session: Reactor Physics,<br />
Thermo and Fluid Dynamics“ und „Enhanced Safety<br />
& Operation Excellence – Focus Session: Radiation<br />
Protection“ des 48 th Annual Meeting on Nuclear<br />
Technology (AMNT 2017), Berlin, 16 bis 17 Mai<br />
2017.<br />
Ein Newcomer führt die Kernenergie<br />
in das Neue Jahr<br />
John Shepherd | Seite 66<br />
Zu Beginn des neuen Jahres weist ein “Newcomer“<br />
mit dem Bau des ersten Kernkraftwerks den Weg.<br />
Für das Fundament des Kernkraftwerks in Rooppur<br />
in Bangladesch wurde der erste Beton gegossen.<br />
Damit ist die südasiatische Nation eine weitere, die<br />
nach den Vereinigten Arabischen Emiraten 2012<br />
und Weißrussland 2013, mit dem Bau eines ersten<br />
kommerziellen Reaktors begonnen hat.<br />
Trotz der Rückschläge für die Kernenergie in den<br />
letzten Jahren, sind weltweit fast 60 Reaktoren in<br />
Bau, vor allem in Asien. 447 kommerzielle Reaktorblöcke<br />
sind in 30 Ländern in Betrieb.<br />
9<br />
ABSTRACTS | GERMAN<br />
Abstracts | German
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
10<br />
INSIDE NUCLEAR WITH NUCNET<br />
UK Is Leading the Way<br />
With Clear Strategy for Nuclear<br />
NucNet<br />
The UK is Europe’s most prominent leader in nuclear development because of the government’s clear<br />
strategy of supporting nuclear energy as part of its future energy mix, a senior official from US-based nuclear<br />
equipment manufacturer Westinghouse Electric Company said.<br />
Michael Kirst, Westinghouse’s vice-president of<br />
strategy for Europe, Middle East and Africa<br />
(EMEA), warned, however, that choices about nuclear<br />
development must be based on technology, and not on the<br />
type of financing package. “We now have a banking contest<br />
and not a technology contest and this is not healthy for the<br />
industry or the energy system,” he said.<br />
Mr Kirst told reporters in Brussels that the UK government’s<br />
decision to support the financing of new energy<br />
projects, including nuclear, by way of a contract for<br />
difference (CfD) scheme was a breakthrough.<br />
“The UK government made it clear they need these new<br />
nuclear capacities”, he said. The UK model provides a “fair<br />
foundation” where all low-carbon technologies were given<br />
exactly the same access to state support.<br />
Mr Kirst said Westinghouse, a privately owned company,<br />
does not have access to state support on demand, unlike its<br />
major competitors in the nuclear industry, which are<br />
“somehow state-owned or state-controlled”. A clear market<br />
signal for private investors in nuclear development is therefore<br />
essential because it allows choices based on technology,<br />
rather than on a financing package, Mr Kirst said.<br />
Speaking about NuGen’s planned three-unit Moorside<br />
nuclear project in Cumbria, northwest England, the<br />
company’s president for EMEA, Luc Van Hulle, said there<br />
are “a couple of options on the table” and Westinghouse’s<br />
AP1000 Generation III+ pressurised water reactor<br />
technology is still potentially one of these options.<br />
The future of the Moorside project to build three<br />
AP1000s has been overshadowed by Westinghouse’s filing<br />
for Chapter 11 bankruptcy protection in the US in March<br />
2017, along with Westinghouse owner Toshiba’s financial<br />
woes and its decision to no longer serve as a contractor of<br />
engineering, procurement and construction for overseas<br />
nuclear projects.<br />
Mr Van Hulle said the Moorside project became “more<br />
complicated” after Engie sold its 40 % stake in NuGen to<br />
Toshiba in April 2017, making the Japanese company the<br />
sole owner of the project. But he said Westinghouse is<br />
confident that the project will proceed “one way or<br />
another”. He said the fate of the project is in the hands of<br />
the UK government and NuGen’s owner Toshiba.<br />
Last month state media reported that China General<br />
Nuclear Power Corporation (CGN) is considering investing<br />
in Moorside, while in March 2017, South Korea’s Korea<br />
Electric Power Corporation (Kepco) expressed an interest in<br />
taking a stake in NuGen.<br />
Mr Van Hulle said that holding on to the AP1000 design<br />
will be the securest and fastest way to realise the Moorside<br />
project because the plant completed the UK’s generic<br />
design assessment (GDA) review by regulators in the UK in<br />
March 2017.<br />
If NuGen chooses another technology, the process of<br />
going through another GDA process could delay the project<br />
by four or five years, he said.<br />
“Clearly there will be a shift in the start date from 2025<br />
to later in the 2020s, but the plant could still be up and<br />
running before 2030,” NuGen’s chief executive officer Tom<br />
Samson told Reuters last week.<br />
Mr Samson said the timing will largely depend on the<br />
technology choice, because the new bidders may want to<br />
bring in their own designs. However, Mr Samson said:<br />
“We are not ruling out any technology at this stage.”<br />
In the US, the expected delay to the Vogtle nuclear<br />
project and the cancellation of the Summer project in<br />
South Carolina was not related to the AP1000 technology,<br />
Mr Van Hulle said.<br />
He said the AP1000 design is “safe and sound” and the<br />
AP1000 reactor units being built in China will prove this<br />
once they enter commercial operation.<br />
There are four AP1000 nuclear units under construction<br />
in China – two at Sanmen and two at Haiyang – all expected<br />
to become commercially operational in <strong>2018</strong>.<br />
| | AP1000 new build in Haiyang, China.<br />
South Carolina Electric and Santee Cooper, the two US<br />
utilities that co-own the Summer AP1000 project, decided<br />
to suspend its construction in July 2017 quoting cost<br />
overruns and schedule delays.<br />
Mr Van Hulle said the utilities’ decision to stop construction<br />
was “saddening” because of the advanced stage<br />
of development, with all nuclear steam supply systems<br />
having been installed. He said the Summer units will not be<br />
completed in the “foreseeable future”, but there is a<br />
possibility that a new owner could take over the project.<br />
In September 2017, the owners of the two-unit Vogtle<br />
AP1000 project in Georgia recommended completing<br />
construction, despite Westinghouse’s financial woes and<br />
increased costs.<br />
The two new reactors at Vogtle, units 3 and 4, under<br />
construction since 2013, represent the first US deployment<br />
of the AP1000 technology.<br />
According to Mr Van Hulle, despite its current difficulties<br />
in the US, Westinghouse has a “very sound base<br />
business” which will serve as the backbone of the<br />
company’s future.<br />
In August 2017, Westinghouse submitted a five-year<br />
business plan to the company’s debtor-in-possession (DIP)<br />
financing lenders and the unsecured creditors committee.<br />
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The company said at the time that this marked a critical<br />
milestone in the Chapter 11 bankruptcy process.<br />
The plan integrates Westinghouse’s initiatives into a<br />
five-year financial forecast and would result in projected<br />
savings of $20 5m (€ 174 m) expected to improve earnings<br />
before interest, taxes, depreciation and amortisation<br />
(EBITDA) over the five-year term.<br />
Westinghouse said the plan supports the operation of its<br />
core businesses and its new projects business. One component<br />
of the savings will be global staff reductions, starting<br />
with 7 % of staff being made redundant in fiscal year 2017.<br />
Since filing for Chapter 11 in March 2017, Westinghouse<br />
has obtained approval of an $ 800 m DIP financing package<br />
and has negotiated a long-term services agreement with<br />
Southern Nuclear Company for the two AP1000 plants<br />
under construction at Vogtle.<br />
“We are well on track with exiting the Chapter 11<br />
process”, Mr Van Hulle said.<br />
Asked to comment on the potential for nuclear<br />
development in other EU member states, Mr Van Hulle said<br />
Bulgaria, Hungary, Poland, and the Czech Republic could<br />
be expected to develop existing or new nuclear capacities.<br />
Potential exists also in non-EU countries like Switzerland,<br />
Turkey and particularly Ukraine, he said.<br />
According to Mr Kirst, Ukraine’s reactor fleet operates<br />
at an average load factor of about 70 % compared to 85 to<br />
90 % in the US and EU. “There is a lot of untapped energy<br />
that can come online at a very low cost and this is what<br />
we have been suggesting to the Ukrainian government”,<br />
Mr Kirst said.<br />
Mr Van Hulle said there is also an opportunity for<br />
Westinghouse to expand its business relationships in<br />
Ukraine in terms of fuel supplies and plant operation,<br />
availability and energy distribution.<br />
“With the amount of reactors they have they can be<br />
really influential in non-Russia based VVER technology”,<br />
he noted.<br />
Westinghouse has contracts to supply nuclear fuel for six<br />
VVER reactor units in Ukraine, as well as core monitoring<br />
systems for Zaporozhye-5, and a potential uprate project at<br />
South Ukraine-3.<br />
Ukraine operates a fleet of 15 commercial units, all of<br />
the VVER pressurised water reactor design and built<br />
during the Soviet Era.<br />
Mr Kirst said Ukraine is the only country which has<br />
significantly diversified its nuclear fuel supply away from<br />
Russia, while EU counties which use VVER reactors remain<br />
completely dependent on Russian supply.<br />
“There have not been significant efforts in Brussels to<br />
address that issue, which is interesting considering that<br />
they are talking about an energy union and the need for<br />
secure and diverse energy supplies”, he said.<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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12<br />
CALENDAR<br />
Calendar<br />
<strong>2018</strong><br />
30.01.-31.01.<strong>2018</strong><br />
NNBS Egypt <strong>2018</strong> — Nuclear New Build Summit<br />
Egypt <strong>2018</strong>. Cairo, Egypt, InforValue Consulting<br />
Company, nuclearegypt.com<br />
05.02.-07.02.<strong>2018</strong><br />
Components and Structures under Severe<br />
Accident Loading Cossal (COSSAL). Cologne,<br />
Germany. OECD/NEA, GRS,<br />
www.grs.de, www.oecd-nea-org<br />
07.02.-08.02.<strong>2018</strong><br />
8. Symposium Stilllegung und Abbau<br />
kerntechnischer Anlagen. Hanover, Germany.<br />
TÜV Nord, www.tuev.nord.de<br />
26.02.-01.03.<strong>2018</strong><br />
Nuclear and Emerging Technologies for Space<br />
<strong>2018</strong>. Las Vegas, NV, USA. American Nuclear Society<br />
(ANS), www.ans.org<br />
01.03.<strong>2018</strong><br />
7. Fachgespräch Endlagerbergbau. Essen,<br />
Germany, DMT, GNS, www.dmt-goup.com<br />
04.03.-09.03.<strong>2018</strong><br />
82. Jahrestagung der DPG. Erlangen, Germany,<br />
Deutsche Physikalische Gesellschaft (DPG),<br />
www.dpg-physik.de<br />
11.03.-17.03.<strong>2018</strong><br />
International Youth Nuclear Congress (IYNC).<br />
Bariloche, Argentina, IYNC and WiN Global,<br />
www.iync.org/category/iync<strong>2018</strong>/<br />
26.03.-27.03.<strong>2018</strong><br />
Fusion energy using tokamaks: can development<br />
be accelerated? London, United Kingdom,<br />
The Royal Society, royalsociety.org<br />
08.04.-11.04.<strong>2018</strong><br />
International Congress on Advances in Nuclear<br />
Power Plants – ICAPP 18. Charlotte, NC, USA,<br />
American Nuclear Society (ANS), www.ans.org<br />
08.04.-13.04.<strong>2018</strong><br />
11 th International Conference on Methods and<br />
Applications of Radioanalytical Chemistry –<br />
MARC XI. Kailua-Kona, HI, USA, American Nuclear<br />
Society (ANS), www.ans.org<br />
17.04.-19.04.<strong>2018</strong><br />
World Nuclear Fuel Cycle <strong>2018</strong>. Madrid, Spain,<br />
World Nuclear Association (WNA),<br />
www.world-nuclear.org<br />
22.04.-26.04.<strong>2018</strong><br />
Reactor Physics Paving the Way Towords More<br />
Efficient Systems – PHYSOR <strong>2018</strong>. Cancun, Mexico,<br />
www.physor<strong>2018</strong>.mx<br />
08.05.-10.05.<strong>2018</strong><br />
29 th Conference of the Nuclear Societies in Israel.<br />
Herzliya, Israel. Israel Nuclear Society and Israel<br />
Society for Radiation Protection, ins-conference.com<br />
13.05.-19.05.<strong>2018</strong><br />
BEPU-<strong>2018</strong> — ANS International Conference on<br />
Best-Estimate Plus Uncertainties Methods. Lucca,<br />
Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,<br />
ANS, IAEA, NEA, www.nineeng.com/bepu/<br />
13.05.-18.05.<strong>2018</strong><br />
RadChem <strong>2018</strong> — 18 th Radiochemical<br />
Conference. Marianske Lazne, Czech Republic,<br />
www.radchem.cz<br />
14.05.-16.05.<strong>2018</strong><br />
ATOMEXPO <strong>2018</strong>. Sochi, Russia, atomexpo.ru<br />
15.05.-17.05.<strong>2018</strong><br />
11 th International Conference on the Transport,<br />
Storage, and Disposal of Radioactive Materials.<br />
London, United Kingdom, Nuclear Institute,<br />
www.nuclearinst.com<br />
20.05.-23.05.<strong>2018</strong><br />
5 th Asian and Oceanic IRPA Regional Congress on<br />
Radiation Protection – AOCRP5. Melbourne,<br />
Australia, Australian Radiation Protection Society<br />
(ARPS) and International Radiation Protection<br />
Association (IRPA), www.aocrp-5.org<br />
29.05.-30.05.<strong>2018</strong><br />
49 th Annual Meeting on Nuclear Technology<br />
AMNT <strong>2018</strong> | 49. Jahrestagung Kerntechnik.<br />
Berlin, Germany, DAtF and KTG,<br />
www.nucleartech-meeting.com<br />
03.06.-07.06.<strong>2018</strong><br />
38 th CNS Annual Conference and 42nd CNS-CNA<br />
Student Conference. Saskotoon, SK, Canada,<br />
Candian Nuclear Society CNS, www.cns-snc.ca<br />
03.06.-06.06.<strong>2018</strong><br />
HND<strong>2018</strong> 12 th International Conference of the<br />
Croatian Nuclear Society. Zadar, Croatia, Croatian<br />
Nuclear Society, www.nuklearno-drustvo.hr<br />
04.06.-07.06.<strong>2018</strong><br />
10 th Symposium on CBRNE Threats. Rovaniemi,<br />
Finland, Finnish Nuclear Society, ats-fns.fi<br />
04.06.-08.06.<strong>2018</strong><br />
5 th European IRPA Congress – Encouraging<br />
Sustainability in Radiation Protection. The Hague,<br />
The Netherlands, Dutch Society for Radiation<br />
Protection (NVS), local organiser, irpa<strong>2018</strong>europe.com<br />
06.06.-08.06.<strong>2018</strong><br />
2 nd Workshop on Safety of Extended Dry Storage<br />
of Spent Nuclear Fuel. Garching near Munich,<br />
German, GRS, www.grs.de<br />
17.06.-21.06.<strong>2018</strong><br />
ANS Annual Meeting “Future of Nuclear in the<br />
Shifting Energy Landscape: Safety, Sustainability,<br />
and Flexibility”. Philadelphia, PA, USA, American<br />
Nuclear Society (ANS), www.ans.org<br />
25.06.-26.06.<strong>2018</strong><br />
index<strong>2018</strong> – International Nuclear Digital<br />
Experience. Paris, France, Société Française<br />
d’Energie Nucléaire, www.sfen.org,<br />
www.sfen-index<strong>2018</strong>.org<br />
27.06.-29.06.<strong>2018</strong><br />
EEM — <strong>2018</strong> 15 th International Conference<br />
on the European Energy Market. Lodz, Poland,<br />
Lodz University of Technology, Institute of Electrical<br />
Power Engineering, Association of Polish Electrical<br />
Engineers (SEP), www.eem18.eu<br />
29.07.-02.08.<strong>2018</strong><br />
International Nuclear Physics Conference 2019.<br />
Glasgow, United Kingdom, www.iop.org<br />
05.08.-08.08.<strong>2018</strong><br />
Utility Working Conference and Vendor<br />
Technology Expo. Amelia Island, FL, USA,<br />
American Nuclear Society (ANS), www.ans.org<br />
22.08.-31.08.<strong>2018</strong><br />
Frédéric Joliot/Otto Hahn (FJOH) Summer School<br />
FJOH-<strong>2018</strong> – Maximizing the Benefits of Experiments<br />
for the Simulation, Design and Analysis of<br />
Reactors. Aix-en-Provence, France, Nuclear Energy<br />
Division of Commissariat à l’énergie atomique et aux<br />
énergies alternatives (CEA) and Karlsruher Institut<br />
für Technologie (KIT), www.fjohss.eu<br />
28.08.-31.08.<strong>2018</strong><br />
TINCE <strong>2018</strong> – Technological Innovations in<br />
Nuclear Civil Engineering. Paris Saclay, France,<br />
Société Française d’Energie Nucléaire, www.sfen.org,<br />
www.sfen-tince<strong>2018</strong>.org<br />
05.09.-07.09.<strong>2018</strong><br />
World Nuclear Association Symposium <strong>2018</strong>.<br />
London, United Kingdom, World Nuclear Association<br />
(WNA), www.world-nuclear.org<br />
09.09.-14.09.<strong>2018</strong><br />
21 st International Conference on Water Chemistry<br />
in Nuclear Reactor Systems. EPRI – Electric Power<br />
Research Institute, San Francisco, CA, USA,<br />
www.epri.com<br />
09.09.-14.09.<strong>2018</strong><br />
Plutonium Futures – The Science <strong>2018</strong>. San Diego,<br />
United States, American Nuclear Society (ANS),<br />
www.ans.org<br />
10.09.-13.09.<strong>2018</strong><br />
Nuclear Energy in New Europe – NENE <strong>2018</strong>.<br />
Portoroz, Slovenia, Nuclear Society of Slovenia,<br />
www.nss.si/nene<strong>2018</strong>/<br />
17.09.-21.09.<strong>2018</strong><br />
62 nd IAEA General Conference. Vienna, Austria.<br />
International Atomic Energy Agency (IAEA),<br />
www.iaea.org<br />
17.09.-20.09.<strong>2018</strong><br />
FONTEVRAUD 9. Avignon, France, Société Française<br />
d’Energie Nucléaire (SFEN), www.sfen-fontevraud9.org<br />
17.09.-19.09.<strong>2018</strong><br />
4 th International Conference on Physics and<br />
Technology of Reactors and Applications –<br />
PHYTRA4. Marrakech, Morocco, Moroccan<br />
Association for Nuclear Engineering and Reactor<br />
Technology (GMTR), National Center for Energy,<br />
Sciences and Nuclear Techniques (CNESTEN) and<br />
Moroccan Agency for Nuclear and Radiological<br />
Safety and Security (AMSSNuR), phytra4.gmtr.ma<br />
30.09.-05.10.<strong>2018</strong><br />
Pacific Nuclear Basin Conferences – PBNC <strong>2018</strong>.<br />
San Francisco, CA, USA, American Nuclear Society<br />
(ANS), www.ans.org<br />
02.10.-04.10.<strong>2018</strong><br />
7 th EU Nuclear Power Plant Simulation ENPPS<br />
Forum. Birmingham, United Kingdom, Nuclear<br />
Training & Simulation Group, www.enpps.tech<br />
14.10.-18.10.<strong>2018</strong><br />
12 th International Topical Meeting on Nuclear<br />
Reactor Thermal-Hydraulics, Operation and<br />
Safety – NUTHOS-12. Qingdao, China, Elsevier,<br />
www.nuthos-12.org<br />
14.10.-18.10.<strong>2018</strong><br />
NuMat <strong>2018</strong>. Seattle, United States, www.elsevier.com<br />
16.10.-17.10.<strong>2018</strong><br />
4 th GIF Symposium 16-17 Oct. <strong>2018</strong> at the<br />
8 th edition of Atoms for the Future. Paris, France,<br />
www.gen-4.org<br />
22.10.-24.10.<strong>2018</strong><br />
DEM <strong>2018</strong> Dismantling Challenges: Industrial<br />
Reality, Prospects and Feedback Experience. Paris<br />
Saclay, France, Société Française d’Energie Nucléaire,<br />
www.sfen.org, www.sfen-dem<strong>2018</strong>.org<br />
22.10.-26.10.<strong>2018</strong><br />
NUWCEM <strong>2018</strong> Cement-based Materials<br />
for Nuclear Wates. Avignon, France, French<br />
Commission for Atomic and Alternative Energies<br />
and Société Française d’Energie Nucléaire,<br />
www.sfen-nuwcem<strong>2018</strong>.org<br />
24.10.-25.10.<strong>2018</strong><br />
Chemistry in Power Plant. Magdeburg, Germany,<br />
VGB PowerTech e.V., www.vgb.org<br />
11.11.-15.11.<strong>2018</strong><br />
ANS Winter Meeting. Orlando, FL, USA, American<br />
Nuclear Society (ANS), www.ans.org<br />
Calendar
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
ETSON Strategic Orientations on Research<br />
Activities. ETSON Research Group Activity<br />
J.P. Van Dorsselaere, M. Barrachin, D. Millington, M. Adorni, M. Hrehor, F. Mascari, A. Schaffrath, I.<br />
Tiselj, E. Uspuras, Y. Yamamoto, D. Gumenyuk, N. Fedotova, O. Cronvall and P. Liska<br />
1 Introduction In October 2011, ETSON published the “Position Paper of the Technical Safety Organizations:<br />
Research Needs in Nuclear Safety for Gen 2 and Gen 3 NPPs” [1]. This paper, published only a few months after the<br />
Fukushima-Daiichi severe accidents in Japan, presented the R&D priorities on the main pending safety issues. It was<br />
produced by the ETSON Research Group (ERG) that has the mandate of identifying and prioritizing safety research<br />
needs, sharing information on research projects in which ETSON members are involved, defining and launching new<br />
research projects and disseminating knowledge among ETSON members.<br />
Six years after the above publication, many R&D international<br />
projects in frames such as OECD/NEA/CSNI and<br />
Euratom have finished and others have started. In<br />
particular a lot of work was done (and is going on…) on<br />
the analysis of the Fukushima-Daiichi severe accidents.<br />
Meanwhile a roadmap on research on Gen.II and III<br />
nuclear power plants (NPP), including safety aspects,<br />
was elaborated by the NUGENIA association and published<br />
in 2013 [2], followed in April 2015 by a more detailed<br />
document as “NUGENIA global vision” [3].<br />
Thus in 2016-2017, the ERG judged it necessary to<br />
perform an update of the ETSON ranking of R&D priorities,<br />
accounting for recent outcomes of research projects (and,<br />
for severe accidents, knowledge gained on the Fukushima-<br />
Daiichi accidents) and for the NUGENIA R&D roadmaps.<br />
The main objective was to underline a possible convergence<br />
of topics for further R&D, but accounting for current<br />
international R&D projects to avoid duplication of efforts.<br />
2 Process of ranking of priorities<br />
Thirteen ETSON members participated to the exercise<br />
focusing on the safety aspects with the challenge to agree<br />
on a short list of high priority topics and avoid the topics<br />
where significant R&D is ongoing. A good example of<br />
the latter case is In-Vessel-Melt-Retention during a severe<br />
accident where many organizations from Europe (and<br />
beyond) participate in the IVMR H2020 project [4]. For<br />
the sake of simplification, the process was based on the<br />
list of R&D challenges and issues from the NUGENIA<br />
roadmap (each challenge includes several specific issues).<br />
The partners were asked to:<br />
• Select up to 10 highest-priority challenges: give<br />
the mark 1 for the most important,…, 10 for the less<br />
important,<br />
• Then, for each of them, select up to 3 issues: give<br />
the mark 1 for the most important..., 3 for the less<br />
important.<br />
The ranking process was based on the list of R&D highpriority<br />
issues (around 150) from the latest NUGENIA<br />
R&D roadmap. This list covers the 6 following topical<br />
areas: plant safety and risk assessment, severe accidents,<br />
improved reactor operation, integrity assessment of<br />
systems, structures and components, fuel development,<br />
waste and spent fuel management and decommissioning,<br />
innovative LWR design and technology.<br />
The results indicated a rather large scattering of votes<br />
on issues but also the possibility of identifying issues with<br />
a majority of votes. The average ranking was the sum of<br />
marks divided by number of votes. The combined ranking<br />
of challenges and issues was then obtained as “challenge<br />
average ranking” multiplied by the “issue average ranking”.<br />
The smallest figures have the highest priority.<br />
Eight issues, described in the Section 3, were selected<br />
as the highest priority (the order of presentation does not<br />
represent a decreasing order of priority, the issues are in<br />
the order of the NUGENIA roadmap). This Section<br />
summarizes the importance of the issue for safety, the<br />
state of knowledge and the remaining gaps, and the international<br />
context such as ongoing or starting R&D projects.<br />
3 High priority issues<br />
3.1 Improved thermal-hydraulics evaluation<br />
for the existing plants<br />
Most of the thermal-hydraulic phenomena during<br />
accidents in NPPs occur at the scale of NPP cooling<br />
systems (thermal-hydraulics in Spent Fuel Pools or SFP is<br />
addressed in § 3.5). The NPP response often represents a<br />
complex interplay of the processes and phenomena in the<br />
subsystems, which can be reproduced or analyzed only<br />
with an experimental facility with a similar complexity or<br />
with a simulation system code that contains models of all<br />
relevant subsystems. Large integral facilities and system<br />
codes thus represent a basis for NPP safety analyses. More<br />
or less an integral facility was built in the past (or is being<br />
built) to correspond to every major NPP type, and thus was<br />
(or is) used to examine the plant performance during<br />
safety relevant scenarios. Such review of integral facilities<br />
and experiments was prepared by OECD/NEA/CSNI [5].<br />
Some of these facilities have already been dismantled,<br />
some of them are maintained (PKL in Germany, as well as<br />
INKA for Gen.3+ BWR safety systems, and LSTF in Japan),<br />
while the countries with long term nuclear goals upgrade<br />
(MOTEL in Finland) or build entirely new (ACME in China)<br />
facilities. These experiments were and are still used for<br />
validation and verification of system codes (CATHARE,<br />
ATHLET, TRACE, RELAP ...) that represent indispensable<br />
tools for safety analyses.<br />
A complementary approach to the integral thermalhydraulics<br />
testing is the “bottom-up” approach, which<br />
actually means experimental and numerical studies of<br />
separate effects at larger scales under well-defined initial<br />
and boundary conditions. These test facilities are more<br />
accessible for academic institutions and can be roughly<br />
divided into problems of single-phase and two(multi)-<br />
phase flow phenomena. Single-phase experiments and<br />
computational fluid dynamics (CFD) can be considered a<br />
mature research field, where even blind predictions of<br />
rather complex flows with heat transfer (pressurized<br />
thermal shock, natural convection) and mixing of species<br />
<strong>atw</strong>-Special „Eurosafe<br />
2017“. In cooperation<br />
with the EUROSAFE<br />
2017 partners,<br />
Bel V (Belgium),<br />
CSN (Spain), CV REZ<br />
(Czech Republic),<br />
MTA EK (Hungary),<br />
GRS ( Germany), ANVS<br />
(The Netherlands),<br />
INRNE BAS (Bulgaria),<br />
IRSN (France), NRA<br />
(Japan), JSI (Slovenia),<br />
LEI (Lithuania),<br />
PSI (Switzerland),<br />
SSM (Sweden),<br />
SEC NRS (Russia),<br />
SSTC NRS (Ukraine),<br />
VTT (Finland),<br />
VUJE (Slovakia),<br />
Wood (United<br />
Kingdom).<br />
Revised version<br />
of a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and<br />
7 November 2017.<br />
13<br />
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ENERGY POLICY, ECONOMY AND LAW 14<br />
(boron dilution) will closely approach the measurements.<br />
Tackling the two(multi)-phase phenomena is much more<br />
difficult. Just like in 1D system codes, detailed 3D twophase<br />
flow models still rely on a number of (semi)empirical<br />
closure relations, which must be carefully considered<br />
for each particular geometry and phenomenon. Blind predictions<br />
are successful only in some simple configurations,<br />
while predictions of complex phenomena like critical heat<br />
flux with 3D CFD models are not much more accurate than<br />
with 1D sub-channel codes (NURESAFE project, Section<br />
2.4 of [1]).<br />
From the TSO point of view, one of the most important<br />
research directions is upgrading of system codes with<br />
(quasi)3D modules for 3D components, especially the<br />
reactor vessel [6]. These coarse grid models can be tuned<br />
with CFD results and high-resolution experiments. These<br />
activities aim at coupling with 3D neutronic models and<br />
more detailed description of the heat transfer and mixing<br />
in the core region. Rough 3D approximations are used<br />
and are applicable also in the simulations of SFP and<br />
containment thermal-hydraulics, and represent a basis for<br />
severe accident simulations.<br />
For TSOs, more attention should probably be focused<br />
on integral studies, which are typically much more<br />
expensive, and can be as such seen as a critical infrastructure<br />
[1]. Research in smaller test facilities on<br />
phenomena such as single bubble, smallest turbulent<br />
eddy,... will not disappear, as they are relevant for many<br />
non-nuclear problems, while the equipment, knowledge<br />
and experts in the field and the integral facilities are much<br />
more difficult to maintain.<br />
3.2 Impact of single or multiple external events<br />
Many ETSON members contributed to the EU FP7<br />
ASAMPSA_E project [7], which began in 2013 and<br />
concluded at the end of 2016; the project was led by IRSN<br />
with 28 partners in 18 European countries. The aim was to<br />
support the systematic extension of PSA to all potential<br />
natural or man-made external and internal hazards.<br />
Documents were developed to guide European stakeholders<br />
in conducting extended PSAs and ensuring that all<br />
dominant risks are identified and managed. The project<br />
identified areas for future development relating to external<br />
hazards; the majority of these also apply to deterministic<br />
methods, which with PSA form the key aspects of hazard<br />
analysis.<br />
For the external flooding hazard, work identified to<br />
address the following shortfalls of current methodologies<br />
included:<br />
• Limitations in modelling and forecasting the physical<br />
phenomena and conditions leading to external flooding<br />
hazard,<br />
• Uncertainties in estimation of the impact of climate<br />
change on external flooding events,<br />
• Lack of site-specific data and limitations of spatial<br />
modelling and downscaling methods,<br />
• Difficulties in quantification of uncertainties for<br />
common-cause failures,<br />
• Difficulties in integrated modelling of hazard internal<br />
and external impact assessment,<br />
• Modelling of water propagation on the site and inside<br />
the buildings.<br />
For meteorological hazards, the recommendations<br />
included:<br />
• The provision of a better understanding and means<br />
for quantifying the correlation mechanisms between<br />
extreme weather events,<br />
• An analysis of the time of the occurrence of extreme<br />
hazard events and simultaneous evaluation of the<br />
atmospheric states at the time of the hazard,<br />
• More accurate estimation of the impact of climate<br />
change on extreme meteorological events,<br />
• Development and validation of downscaling methods<br />
and tools for analysing and characterizing spatially<br />
distributed extreme data.<br />
For the seismic hazard, the recommendations included:<br />
• The reduction of aleatory and epistemic uncertainties<br />
in both the derivation of the seismic hazard and the<br />
methods used to derive fragility curves,<br />
• Improved methods for deriving conditional probabilities<br />
of seismically induced consequential events such as<br />
fire and flood.<br />
For many hazard types, the need for work on treatment of<br />
hazard combinations was also identified. There is a need<br />
for a formalised approach for assessing and screening<br />
hazards in which a primary external hazard would cause<br />
one or more secondary hazards, or in which multiple<br />
hazards occur together as a result of a common event or<br />
underlying cause. Combinations of external and internal<br />
hazards also need to be considered more rigorously and<br />
systematically.<br />
The need for integration of natural external hazards<br />
in the plant safety case and PSA was identified previously<br />
[1] which recommended that the identification of a<br />
comprehensive list of hazards, the site specific screening of<br />
hazards, and the definition of the design basis hazards and<br />
hazard combinations are required. It recommended that a<br />
methodology or procedure was needed to integrate these<br />
into the overall safety case and PSA. The ASAMPSA_E<br />
project went some way towards achieving this objective.<br />
The needs identified in ASAMPSA_E are partly covered<br />
by the NARSIS (New Approach to Reactor Safety Improvement)<br />
H2020 project [8], coordinated by CEA (France),<br />
which recently started with the contribution of ENEA, JSI,<br />
IRSN and VTT. In summary, it addresses improvements on<br />
characterization of natural external hazards (concomitant<br />
external events…), on the fragility of NPP Structures,<br />
Systems and Components (SSC), on a combination of<br />
risk integration with uncertainty quantification, and on<br />
integration of expert-based information within PSA<br />
methodology.<br />
3.3 Methodologies for beyond design basis<br />
assessments<br />
Until recently, the safety assessment in the design of NPPs<br />
was mainly focused on evaluation of postulated transients<br />
and design basis accidents (DBA) and demonstration that<br />
the systems of the plant can ensure that the prescribed<br />
limits for fuel damage and radiation consequences are<br />
not exceeded. Analysis of beyond design conditions was<br />
generally treated as a complementary one and was mainly<br />
used to evaluate the progression of the accident sequences<br />
accompanied by multiple failures of systems, equipment<br />
and components, for a more precise definition of accident<br />
end states in the framework of probabilistic risk assessment<br />
and for identification of operator actions for bringing<br />
the plant into the controlled state and/or mitigating the<br />
consequences.<br />
With update of the IAEA requirements ([9] in particular),<br />
the analysis of design extension conditions that<br />
include multiple failure events without nuclear fuel<br />
melting, as well as severe accidents, becomes an intrinsic<br />
part of the plant safety assessment, and appropriate safety<br />
features for preventing such conditions from arising, or,<br />
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if they do arise, for controlling them and mitigating their<br />
consequences, are required to be included in the NPP<br />
design.<br />
Individual aspects of the methodology for beyond design<br />
basis accidents (BDBA), with different levels of detail, are<br />
reflected in national regulations of the ETSON member<br />
states. Basic considerations on assessing the design extension<br />
conditions can also be found in IAEA documents [10,<br />
11]. However some questions still need to be addressed, e.g.<br />
how to ensure that all relevant scenarios are considered,<br />
what is the extent of failures to be considered, how the<br />
uncertainties shall be identified and accounted for? Therefore<br />
there is a need to provide TSOs with unified detailed<br />
guidelines that cover all BDBA stages, starting from the<br />
deterministic and probabilistic criteria for selection of<br />
corresponding scenarios, assumptions on systems/equipment<br />
operability (including non-safety graded systems),<br />
and ending with the evaluation of assessment results, and<br />
establishes interfaces with practical applications of assessment<br />
results, including identification and justification of the<br />
provisions which are incorporated in the plant design or to<br />
be implemented as safety upgrade measures for mitigating<br />
the consequences of such events, identification of operator<br />
preventive and mitigating actions, etc.<br />
The important aspects to be addressed in the guidelines<br />
are incorporation of up-to-date results of R&D in the area<br />
of phenomenology, validation of computer codes and<br />
models and procedure for treatment of inherent uncertainties<br />
associated with current knowledge.<br />
The development of such guidelines is a complex and<br />
rather immense task. Therefore, a possibility for a first phase<br />
could be to collect information on respective experience<br />
of the participants, systemize and critically analyse this<br />
information to identify the existing gaps and then to elaborate<br />
solutions for enhancing the BDBA methodology.<br />
3.4 Development and validation of severe<br />
accident integral codes<br />
Considering the complexity and different mutual interacting<br />
phenomena in severe accident (SA) progression and<br />
the possible source term release to the environment,<br />
research is fundamental in order to characterize the main<br />
phenomena determining the NPP transient evolution and<br />
to support severe accident management (SAM). With this<br />
in mind, a key role is given to the state-of-art SA integral<br />
codes (as ASTEC [12] and MELCOR [13] that are mostly<br />
used within TSOs and safety authorities, but also MAAP<br />
used mainly by the industry) that store all the knowledge<br />
developed in the last decades from the experimental<br />
activities.<br />
With the target of assessing SAM, some modelling<br />
uncertainties still present, sometimes closely linked to<br />
remaining uncertainties on the knowledge of phenomena<br />
itself, should be addressed. The latest status of SA research<br />
highest priorities issued from the SARNET European<br />
network is presented in [14, 15]. Among them, the<br />
modelling improvements must address in priority:<br />
• The coolability of the degraded core and the phenomena<br />
necessary to assess the In-Vessel Melt Retention<br />
strategy,<br />
• The coolability of corium during Molten Core Concrete<br />
Interaction in the NPP cavity after a possible vessel<br />
failure,<br />
• The mitigation of potential source term (mainly<br />
ruthenium and iodine), in particular the use of filtered<br />
containment venting systems (FCVS) and the related<br />
efficiency, including the accident long term situation.<br />
An essential field of applications of such codes in the<br />
next years concerns the need to improve SAM guidelines.<br />
In addition, for plant applications, uncertainty analysis<br />
should be systematically performed (e.g. by using tools<br />
such as DAKOTA, RAVEN, SUNSET, SUSA, etc). More and<br />
more code-to-code exercises called “crosswalk” activities<br />
(e.g. involving the teams of code developers and thus<br />
going much more deeply than classical benchmark<br />
exercises) should be continued (see examples in [16, 17])<br />
in order to identify the modelling differences affecting<br />
code prediction results.<br />
In order to reduce the code user-effect [18], considering<br />
the SA complexity, a high level understanding of the<br />
phenomena/processes and of the use of such codes is<br />
required from code users. It is important to continue three<br />
types of ongoing actions:<br />
• Users’ training programs, led by international<br />
recognized experts,<br />
• Well-defined international cooperation platform of<br />
research activities where exchange of opinions,<br />
methods, experimental/calculated data, ideas and<br />
possible interactions between code users and developers<br />
take place (e.g. SARNET network set up under the<br />
European Commission FP, ASTEC-User Club sponsored<br />
by IRSN, CSARP/MCAP organized by USNRC, OECD/<br />
NEA/CSNI ISP, IAEA ICSP and research and innovation<br />
through EU-FP). In this framework the code-to-code<br />
benchmark exercises (such as the exercise in [19]), as<br />
well as independent user crosswalk activities, will allow<br />
to characterize also the influence of user effect on the<br />
different code predictions.<br />
• Availability of user manual and guidelines to be<br />
provided to the user, in addition to the complete code<br />
documentation (models, numerics, assessment), as<br />
well as development of graphical-user-interfaces [20]<br />
to support the user in the input-deck preparation and to<br />
make the post-processing of the data easier (a good<br />
example of such tool is SNAP developed for USNRC for<br />
use with MELCOR and other codes).<br />
In relation to the extension of SA prediction capability,<br />
another useful action is coupling of SA integral codes with<br />
specialized codes designed to predict the impact of the<br />
source term in the surrounding environment (source term<br />
release, transport, dispersion). This permits a best estimate<br />
evaluation of the source term and a consequent detailed<br />
consequence analyses to support emergency preparedness<br />
and response. An example is the coupling between<br />
MELCOR and MACCS SNL tools developed for USNRC.<br />
A long term consideration could be related to the development<br />
of advanced software platforms where SA integral<br />
codes can be coupled with specific detailed codes (e.g.<br />
CFD) to get a more detailed characterization of SA<br />
progression, in terms of single specific phenomenon or/<br />
and 3D nature predictability.<br />
Finally, as an essential activity, validation of codes vs.<br />
experiments should obviously be performed in the future<br />
as a continous process, on current experiments often<br />
dedicated to mitigation aspects but also on the huge<br />
amount of SA experiments that were performed during<br />
more than 30 years. The codes application to the<br />
Fukushima-Daiichi accidents is also an important task<br />
planned in the next years.<br />
3.5 Spent fuel pool accident scenarios<br />
SFPs are large accident-hardened structures that are used<br />
to temporarily store irradiated nuclear fuel [20, 22]. Safety<br />
and security are continuously reassessed [23], e.g. after<br />
ENERGY POLICY, ECONOMY AND LAW 15<br />
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ENERGY POLICY, ECONOMY AND LAW 16<br />
the terrorist attacks in the USA on September 11, 2001 and<br />
the Fukushima Daiichi accident in March 2011 [24],<br />
although the SFPs and the fuel stored in the pools remained<br />
safe during the accident. Considering all possible initiating<br />
events from safety as well as security perspectives, and the<br />
assumption that the accident cannot be prevented or<br />
mitigated, some SFP scenarios could possibly lead to large<br />
radiological consequences on-site and off-site.<br />
The main knowledge gaps are identified thanks to a<br />
recently completed OECD/NEA/CSNI activity, led by IRSN<br />
with the participation (among the international panel<br />
of experts) of ETSON members Bel V, GRS, PSI and NRA,<br />
on applying a Phenomena Identification and Ranking<br />
Technique (PIRT) on SFPs under loss-of-cooling and<br />
loss-of-coolant accidents conditions [20]. The resulting<br />
phenomena of primary interest for further research can be<br />
summarized as follows:<br />
• Cladding chemical reactions with mixed steam-air<br />
environments for all type of fuel claddings present in<br />
SFPs and also the low temperature range,<br />
• Thermal-hydraulic and heat transfer phenomena for<br />
the coolability of partly or completely uncovered fuel<br />
assemblies,<br />
• Thermal-hydraulic behaviour and large-scale natural<br />
circulation flow pattern that evolves in the SFP with<br />
fuel assemblies covered with water,<br />
• Spray cooling of uncovered spent fuel assemblies in<br />
typical storage rack designs.<br />
Quite a few experiments, specifically targeted to SFP<br />
accidents, are underway or planned. Improvements of<br />
models and simulation codes are still necessary, and their<br />
validation will continue against the produced data.<br />
Regarding applicability of codes, sensitivity and uncertainty<br />
analyses should be considered an integral part of<br />
their applications for SFPs accidents conditions.<br />
National projects focusing on SFP issues are addressed<br />
by several ETSON members, e.g. in cooperation with<br />
universities and research institutes in case of Bel V [25], by<br />
launching experimental programs by IRSN [26], related<br />
to analysis of processes in SFP for LEI, sensitivity analysis<br />
of various modelling options on SFP accidents in SSTC<br />
NRS etc.<br />
3.6 Corium thermophysical and thermodynamic<br />
properties<br />
During a severe accident sequence in LWRs, thermodynamic<br />
models are required to predict the behaviour of<br />
the melts (so-called corium) formed from the degradation<br />
of the core materials, the fission product (FP) releases and<br />
the residual power within the corium different phases.<br />
Data such as the composition of the phases present in the<br />
corium and its physical-chemical properties (solidus and<br />
liquidus temperatures, heat capacities, enthalpies …) are<br />
key parameters for modelling, among other things, the<br />
corium flow properties, the FP distribution between the<br />
gas and the condensed phases and then for modelling of<br />
the progression of the accident.<br />
Since 1990’s, in the framework of projects in the frame<br />
of the EC (COLOSS, SARNET…), the International Science<br />
and Technology Center (CORPHAD and PRECOS) and the<br />
OECD (MASCA [27]), SA experts have been interested in<br />
the assessment of thermodynamic data for a number of<br />
compounds of reactor materials and fission products and<br />
more complex phases. The most common thermodynamic<br />
data assessment approach for the chemical species of<br />
interest is the CALPHAD method [28]. All properties are<br />
derived from the Gibbs energy expression for each phase.<br />
Based on physical models of the different phases, such<br />
expression depends on various parameters, the values<br />
of which are optimised in order to best fit available<br />
experimental data.<br />
Databases thus obtained are more than mere compilations<br />
of thermodynamic data from various sources.<br />
Their constitution and maintenance needs considerable<br />
work for self-consistency analysis, to ensure that all<br />
the available experimental information is satisfactorily<br />
reproduced. Updating and improving the database<br />
becomes then a regular task, tightly linked to the needs of<br />
end-users.<br />
IRSN is developing, with the SIMAP French Laboratory<br />
scientific support, two consistent thermodynamic<br />
data bases for use for the interpretation of SA experiments<br />
and modelling. NUCLEA [29] is mainly used in research<br />
related to the core degradation (in- and ex-vessel) while<br />
MEPHISTA addresses the fuel and FP behaviour in normal<br />
and off-normal conditions. Both databases are currently<br />
used by a large number of institutes, industrial partners,<br />
and universities, including a few ETSON partners (VTT,<br />
soon PSI), EDF, CEA, Areva, KAERI (South Korea), JAEA<br />
(Japan) and others. The OECD-NEA Thermodynamics of<br />
Advanced Fuels – International Database (TAF-ID) project<br />
[30] (2013-2016) made available a comprehensive,<br />
internationally recognized and quality-assured database<br />
of phase diagrams and thermodynamic properties of<br />
advanced nuclear fuels. Its main goal consists in providing<br />
a computational tool to perform thermodynamic calculations<br />
on both fuel and structural materials for SA in<br />
LWRs and for the design of advanced fuel materials (MOX,<br />
metallic, carbide, nitride fuels) for Generation IV reactors.<br />
The recently launched OECD/NEA Thermodynamic<br />
Characterisation of Fuel Debris and Fission Products<br />
(TCOFF) project (2017-2019), involving 16 partners, aims<br />
at improving the existing thermodynamic databases<br />
(e.g. NUCLEA and TAF-ID) for scenario analyses of SA<br />
progression, looking particularly at the Fukushima-Daiichi<br />
accident.<br />
To date, the main gaps of knowledge in databases are<br />
the following ones:<br />
• The interactions between molten U-Zr-O and iron (and<br />
steel) within the vessel since they impact the heat flux<br />
to the vessel in order to determine the conditions (in<br />
particular time and location) of an eventual rupture, in<br />
particular for a molten metal layer located on top of the<br />
oxide one. Some work has been done in the framework<br />
of the MASCA and MASCA2 projects but it would be<br />
necessary to extend it to MOX fuel.<br />
• The impact of the stainless steel oxide components on<br />
the thermochemistry of the corium-concrete mixtures<br />
which should be experimentally investigated.<br />
• The activity coefficients of the Ag-In-Cd control rod<br />
elements in the melts are a very important item to<br />
derive reliable expressions for vapor pressures of<br />
absorber elements. Vaporization of these elements<br />
during a SA is of prime interest for reactors with<br />
Ag-In-Cd control rods. They actually constitute the<br />
main contributors in terms of mass of the aerosol<br />
release into the reactor coolant system and overall,<br />
they greatly impact the aerosol deposition and<br />
the source term behaviours. In fact, silver and cadmium<br />
are very reactive with iodine which is known to<br />
be a major contributor to the gaseous source term<br />
to environment.<br />
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3.7 Ageing/degradation mechanisms,<br />
modelling and materials properties<br />
for metallic components<br />
Many operating NPPs are nearing or have exceeded<br />
their original design lifetime (often 40 years). To safely<br />
continue operation beyond that, i.e. to enter the long term<br />
operation (LTO) period, necessitates considerable technical<br />
preparations and permission from the domestic<br />
regulator. This needs to take thoroughly into account the<br />
ageing/degradation mechanisms, through e.g. knowledge<br />
of materials properties and computational modelling.<br />
There are several potential ageing/degradation<br />
mechanisms affecting metallic NPP components, mainly:<br />
irradiation and thermal embrittlement, fatigue, stress<br />
corrosion cracking (SCC), general and local corrosion,<br />
flow accelerated corrosion, creep and mechanical wear.<br />
The commonly used steel types include: ferritic steels,<br />
austenitic stainless steels and nickel-base alloys. The<br />
susceptibility to degradation mechanisms depends mainly<br />
on physical loads, material properties and process<br />
environment. Important material properties include yield<br />
& tensile strength, fracture toughness and carbon content.<br />
There are still knowledge gaps concerning understanding<br />
irradiation embrittlement, fatigue, SCC and mechanical<br />
wear as well as joint action of degradation mechanisms.<br />
It is necessary to computationally model the propagation<br />
of degradation in metallic NPP components. When considering<br />
LTO, this is called “time limited ageing analysis”<br />
(TLAA). Most TLAAs necessitate knowing the temperature<br />
and stress distributions across the components. These are<br />
computed with heat transfer and structural mechanics<br />
analyses, typically applying numerical finite element (FE)<br />
codes. These results are used as input data in the ensuing<br />
degradation propagation analyses. For local flaws, e.g.<br />
cracks, these analyses are carried out applying fracture<br />
mechanics and empirically derived crack growth correlations.<br />
There are still gaps concerning modelling of irradiation<br />
embrittlement, thermal fatigue, SCC and mechanical<br />
wear as well as joint action of degradation mechanisms.<br />
Current research in Europe is performed in the frame of<br />
Euratom projects: irradiation embrittlement in SOTERIA,<br />
joint action of corrosion and fatigue in INCEFA+, thermal<br />
fatigue and fracture mechanics based modelling of<br />
degradation mechanisms in ATLAS+. Despite this<br />
intensive activity in Europe, this issue selection underlines<br />
the very high importance given by TSOs on such issue.<br />
3.8 Small modular reactors<br />
Currently Small Modular Reactor (SMR) concepts are<br />
discussed as one main option for new builds worldwide.<br />
This revival in SMRs is driven by the potential for enhanced<br />
safety and security while reducing capital costs and thus<br />
investment risks, through design simplification. SMRs<br />
introduce flexibility on locations unable to accommodate<br />
larger NPPs and can be operated under onshore, offshore<br />
and subsea-based conditions. Improved technologies and<br />
methods will be implemented, thus contributing to the<br />
demand of higher safety and reliability without sacrificing<br />
the long lasting operation experience of LWR technology.<br />
The European nuclear industry has developed no<br />
near-term feasibly deployable SMR [31] and countries<br />
have just begun to build-up the necessary regulatory<br />
structures and capacities. SMR based on LWR technology<br />
offer advantages due to the experience of the nuclear<br />
stakeholders (especially of the regulators) with LWR<br />
technology collected in the last decades. Therefore for<br />
ETSON, the priority concerns are LWR-type SMRs, and the<br />
basis for further success is the edge in knowledge, which<br />
also includes validated simulation tools.<br />
Several international activities were initiated concerning<br />
the identification and closure of open SMR issues.<br />
Several workshops and studies took place in the IAEA and<br />
OECD/NEA frame [32, 33, 34, 35]. In the UK a feasibility<br />
Study on SMR was published in 2014 to identify inter alia<br />
the best value for the UK. Several European TSOs deal with<br />
this issue, whereby the GRS study [36] is recognized as<br />
one of the most extensive works on this topic. The aims of<br />
the latter were to set-up a sound overview on current SMR,<br />
to identify essential issues of reactor safety research and<br />
future R&D projects, and to identify needs for adaption of<br />
system codes used in reactor safety research as well as<br />
approval and supervisory procedures. This overview<br />
consists of the description of 69 SMR diverse concepts (32<br />
LWR, 22 liquid metal cooled reactors, 2 heavy water cooled<br />
reactors, 9 gas cooled reactors and 4 molten salt reactors).<br />
It provides information e.g. about the core, the cooling<br />
circuits and the safety systems. The safety relevant issues<br />
of the selected SMR concepts were identified on the basis<br />
of the defense-in-depth concept, which is one core issue of<br />
the new Euratom Safety Directive 2014 (see the ETSON<br />
paper [37]). Further on, it was evaluated whether these<br />
safety systems and measures can already be simulated<br />
with the existing nuclear simulation chains and where<br />
further code development and validation are necessary.<br />
In general the existing codes are a good basis for the<br />
simulation of SMR. However the safety-related im provements<br />
of these advanced reactors, in general, still require a<br />
considerable effort for further development and validation.<br />
Both require new experiments with advanced (two-phase<br />
flow) measuring techniques. In addition to component tests,<br />
in which the start-up and operating behaviour has to be investigated<br />
under defined and idealized initial and boundary<br />
conditions, integral tests are required for the investigation of<br />
the mutual interaction of different passive safety systems or<br />
different trains of one passive safety system required for<br />
( severe) accident control. For such investigations, already<br />
existing large European experimental facilities for the investigation<br />
of passive safety systems (such as INKA in AREVA<br />
GmbH, PANDA in PSI or SPES in SIET-ENEA) can be applied.<br />
Main topics for improvements are e.g. advanced fuel<br />
patterns, innovative fuel and cladding design, increase of<br />
enrichment and burn-up, longer fuel cycles, boron-free<br />
cores, (new) working fluids with extended scopes, passive<br />
safety systems and their mutual interactions, natural<br />
circulation and flow instabilities, innovative heat<br />
exchanger designs (such as plate and helically coiled heat<br />
exchangers, heat pipes), 2D/3D models for simulation of<br />
temperature and velocity fields in large water pools.<br />
4 Conclusion<br />
The R&D highest priority needs that are described in this<br />
paper correspond mostly to the objectives of the new<br />
2014/87 Euratom Directive on the safety of nuclear<br />
installations, as shown in the ETSON EUROSAFE-2015<br />
paper [37]. In particular they aim at preventing accidents<br />
through defence in depth and at avoiding radioactive<br />
releases outside a nuclear installation. They were also<br />
already identified in the ETSON 2011 position paper [1].<br />
This ranking will first serve as basis for new potential<br />
research projects, either to be performed by ETSON<br />
partners only or as a kernel to be proposed in a larger<br />
frame such as NUGENIA or H2020. The ranking may<br />
also serve as the ETSON input to future roadmaps or to<br />
inter national R&D projects.<br />
ENERGY POLICY, ECONOMY AND LAW 17<br />
Energy Policy, Economy and Law<br />
ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity<br />
J.P. Van Dorsselaere, M. Barrachin, D. Millington, M. Adorni, M. Hrehor, F. Mascari, A. Schaffrath, I. Tiselj, E. Uspuras, Y. Yamamoto, D. Gumenyuk, N. Fedotova, O. Cronvall and P. Liska
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
ENERGY POLICY, ECONOMY AND LAW 18<br />
References<br />
1. ETSON/2011-001, Position paper of the Technical Safety<br />
Organisations: Research needs in nuclear safety for GEN 2 and<br />
GEN 3 NPPs, October 2011.<br />
2. NUGENIA roadmap 2013, see www.nugenia.org.<br />
3. NUGENIA Global Vision – Revision 1.1 – April 2015.<br />
4. F. Fichot et al., Status of the IVMR project: First steps towards a<br />
new methodology to assess In-Vessel Retention Strategy for<br />
high-power reactors, Proceedings of ERMSAR 2017, Warsaw,<br />
Poland, May 16-18, (2017).<br />
5. OECD/NEA/CSNI/R(2016)14, A state-of-the-art report on scaling<br />
in system thermal-hydraulics applications to nuclear reactor<br />
safety and design, 2017 (pg. 395).<br />
6. SNETP Deployment Strategy, www.snetp.eu, Dec. 2015.<br />
7. E. Raimond et al., Main findings and perspectives for research<br />
activities of the European project ASAMPSA_E, Forum EUROSAFE,<br />
Paris, 6-7 Nov.2017.<br />
8. E. Foerster et al., NARSIS – New Appproach to Reactor Safety<br />
Improvements, Proceedings of NUGENIA annual Forum 2017,<br />
Amsterdam (The Netherlands), 28-30 March 2017.<br />
9. SSR-2.1 (Rev.1), Safety of Nuclear Power Plants: Design, Specific<br />
Safety Requirements. IAEA, Vienna, 2016.<br />
10. IAEA-TECDOC-1791, Considerations on the Application of the<br />
IAEA Safety Requirements for the Design of Nuclear Power Plants.<br />
IAEA, Vienna, 2016.<br />
11. SSG-2, Deterministic Safety Analysis for Nuclear Power Plants,<br />
Specific Safety Guide. IAEA, Vienna, 2009.<br />
12. P. Chatelard et al., Main Modelling features of ASTEC V2.1 major<br />
version, Annals of Nuclear Energy, Vol 93, pp. 83-93, July 2016.<br />
13. MELCOR Computer Code Manuals, Vol.1: Primer and Users’<br />
Guide, SAND 2015-6691 R; Vol.2: Reference Manual,<br />
SAND 2015-6692 R; Vol.3: MELCOR Assessment Problems,<br />
SAND 2015-6693 R, Sandia National Laboratories, USA (2015).<br />
14. W. Klein-Heßling et al, Conclusions on severe accident research<br />
priorities, Annals of Nuclear Energy 74 (2014) 4–11.<br />
15. J.-P. Van Dorsselaere et al., Recent severe accident research<br />
synthesis of the major outcomes from the SARNET network,<br />
Nuclear Engineering and Design 291 (2015) 19–34.<br />
16. Modular Accident Analysis Program (MAAP) – MELCOR<br />
Crosswalk, Phase 1 Study, 3002004449, Technical Update,<br />
November 2014, EPRI.<br />
17. S. Belon et al., Insight of Core Degradation Simulation in Integral<br />
Codes Throughout ASTEC/MELCOR Crosswalk Comparisons and<br />
ASTEC Sensitivity Studies, Proceedings of ERMSAR 2017, Warsaw,<br />
Poland, May 16-18, (2017).<br />
18. Approaches and Tools for Severe Accident Analysis for Nuclear<br />
Power Plants, IAEA Safety Reports Series No. 56, IAEA, Vienna,<br />
2008.<br />
19. F. Mascari et al., Analyses of an Unmitigated Station Blackout<br />
Transient With ASTEC, MAAP And MELCOR Code, 9 th Meeting of<br />
the European MELCOR User Group, Madrid (Spain), April 6-7,<br />
2017.<br />
20. NEA/CSNI/R(98)22: Good Practices for User Effect Reduction,<br />
Status Report, November 1998.<br />
21. Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss<br />
of-Coolant Accident Conditions, NEA/CSNI/R(2015)2, May 2015.<br />
22. NEA/CSNI/R(2017) Phenomena Identification and Ranking Table<br />
(PIRT) on Spent Fuel Pools under Loss-of-Cooling and Loss-of-<br />
Coolant Accident Conditions (under publication).<br />
23. Safety and security of commercial spent nuclear fuel storage,<br />
2006, Public report ISBN 0-309-09647-2, The National<br />
Academies Press, Washington, DC, USA.<br />
24. The Fukushima Daiichi Nuclear Accident: Final report of the AESJ<br />
investigation committee. 2015, Tokyo, Japan: Springer.<br />
25. Bousbia Salah, A. and J. Vlassenbroeck. Survey of some safety<br />
issues related to some specific phenomena under natural<br />
circulation flow conditions. In: EUROSAFE 2012, November 5-6,<br />
2012 Brussels, Belgium.<br />
26. Mutelle, H., et al. A new research program on accidents in spent<br />
fuel pools: The DENOPI project. In: 2014 Water Reactor Fuel<br />
Performance Meeting (WRFPM-2014), September 14-17, 2014,<br />
Sendai, Japan.<br />
27. V.G. Asmolov et al., Atomic Energy 104(4) (2008) 273.<br />
28. R. Schmid-Fetzer et al., Assessment techniques, database design<br />
and software facilities for thermodynamics and diffusion,<br />
Calphad, 2007, 31, pp.38-52.<br />
29. S. Bakardjieva et al., Improvement of the European thermodynamic<br />
database NUCLEA, Progress in Nuclear Energy, volume 52,<br />
2010, pp.84-96.<br />
30. C. Gueneau et al., FUELBASE, TAF-ID databases and OC software:<br />
advanced computational tools to perform thermodynamic<br />
calculations on nuclear fuel materials, Proceedings of ERMSAR<br />
2015, Marseille (France), 24-26 March 2015.<br />
31. H. Subki, Advances in development and Deployment of Small<br />
Modular Reactor Design and Technology, ANNuR – IAEA – U.S.<br />
NRC Workshop, SMR Safety and Licensing, Jan. 12-15, 2016.<br />
32. IAEA workshop on Safety and Licensing Requirements for SMR,<br />
Vienna (Austria), January 2016.<br />
33. IAEA International Topical Issues in Nuclear Installations, June<br />
2017.<br />
34. OECD/NEA SMR: Nuclear Energy Market - Potential for Near-term<br />
Deployment, 2016.<br />
35. IAEA TECDOC 1733: Evaluation of Advanced Thermohydraulic<br />
System Codes for Design and Safety Analysis of Integral Type<br />
Reactors, Vienna, 2014.<br />
36. S. Buchholz, A. Krüssenberg, A. Schaffrath, Study of safety and<br />
international development of small modular reactors (SMR),<br />
Proceedings of 16th International Topical Meeting on Nuclear<br />
Reactor Thermalhydraulics (NURETH-16), Chicago, IL, USA<br />
(2015).<br />
37. J. P. Van Dorsselaere, J. Mustoe, S. Power, M. Adorni,<br />
A Schaffrath, A. Nieminen, ETSON views on R&D priorities for<br />
implementation of the 2014 Euratom Directive on safety of<br />
nuclear installations, Kerntechnik: Vol. 81, No. 5, pp. 527-534.<br />
Authors<br />
J.P. Van Dorsselaere (Contact author)<br />
M. Barrachin<br />
IRSN, Centre de Cadarache, BP3,<br />
13115 Saint Paul les Durance Cedex, France<br />
D. Millington<br />
Wood RSD, 305 Bridgewater Place, Birchwood Park,<br />
Warrington WA3 6XF, UK<br />
M. Adorni<br />
BelV, 148 Walcourtstraat, B-1070 Brussels, Belgium<br />
M. Hrehor<br />
CV REZ, Centrum Vyzkumu Rez, Husinec – Rez 130,<br />
250 68 Rez, Czech Republic<br />
F. Mascari<br />
ENEA, Via Martiri di Monte Sole, 4, 40129 Bologna, Italy<br />
A. Schaffrath<br />
Gesellschaft für Anlagen- und Reaktorsicherheit (GRS)<br />
gGmbH, Forschungszentrum, Boltzmannstraße 14,<br />
85748 Garching bei München, Germany<br />
I. Tiselj<br />
JSI, Jozef Stefan Institute, Jamova cesta 39,<br />
SI-1000 Ljubljana, Slovenia<br />
E. Uspuras<br />
LEI, Lithuanian Energy Institute, Breslaujos 3,<br />
LT-44403 Kaunas, Lituania<br />
Y. Yamamoto<br />
NRA, Nuclear Regulation Authority, Toranomon Towers<br />
Office, 4-1-28 Toranomon Minato-ku, Tokyo 105-0001, Japan<br />
D. Gumenyuk<br />
SSTC-NRS, State Scientific and Technical Center,<br />
35-37 Radhospna Str., 03142 Kiev, Ukraine<br />
N. Fedotova<br />
SEC-NRS, Scientific and Engineering Center for Nuclear and<br />
Radiation Safety, Malaya Krasnoselskaya st. 2/8, building 5,<br />
Moscow, 107140, Russia<br />
O. Cronvall<br />
VTT Technical Research Centre of Finland Ltd, Vuorimiehentie<br />
5, P.O.Box 1000, FI-02044, Finland<br />
P. Liska<br />
VUJE, Okruzna 5, 91864 Trnava, Slovakia<br />
Energy Policy, Economy and Law<br />
ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity<br />
J.P. Van Dorsselaere, M. Barrachin, D. Millington, M. Adorni, M. Hrehor, F. Mascari, A. Schaffrath, I. Tiselj, E. Uspuras, Y. Yamamoto, D. Gumenyuk, N. Fedotova, O. Cronvall and P. Liska
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Die Novellierung der europäischen Dual-Use Verordnung –<br />
eine unendliche Geschichte?<br />
19<br />
Ulrike Feldmann<br />
Entwicklung der europäischen Dual-Use Verordnung Erstmalig wurde mit der Verordnung (EG) Nr. 3381/94<br />
des Rates vom 19. 12.1994 (ABl. Nr. L 367 vom 31.12.1994, S. 1) eine Gemeinschaftsregelung für die Ausfuhrkontrolle<br />
von Gütern mit doppeltem Verwendungszweck geschaffen. Mit der Verordnung (EG) Nr. 1334/2000 vom 22.06.2000<br />
(ABl. Nr. L 159 vom 30.06.2000, S. 1) fand die erste größere Revision der Dual-Use Regelungen statt, mit der für den<br />
Nuklearbereich nicht – wie bis dato – nur sensitives Material, d.h. Plutonium und hochangereichertes Uran, sondern die<br />
gesamte Kategorie 0 (Nuklearmaterial, Anlagen, Ausrüstung) auch einer Genehmigungspflicht für die innergemeinschaftliche<br />
Verbringung unterworfen wurde. Außerdem wurde mit der Verordnung 1334/2000 in Art. 21 Abs. 1<br />
bestimmt, dass die Nukleargüter der Kategorie 0 nicht Gegenstand einer Allgemeingenehmigung sein können. Die<br />
EU-Kommission erkannte dann schnell, dass das „Kind mit dem Bade ausgeschüttet“ und mit der rigorosen Erfassung<br />
aller Nukleargüter der Kategorie 0 der innergemeinschaftliche Handel unnötig behindert wurde und nahm wenige<br />
Monate später mit der Verordnung (EG) Nr. 2889/2000 vom 22.12.2000 einen kleinen Teil von Nukleargütern aus der<br />
innergemeinschaftlichen Verbringungsgenehmigungspflicht wieder aus.<br />
Ab 2006 arbeitete die Kommission an einer weiteren<br />
umfassenden neuen Revision, um u.a. die UN Resolution<br />
1540 vom 28.04.2004 zur Nichtverbreitung von chemischen,<br />
biologischen, nuklearen Waffen und ihrer Trägersysteme<br />
durch Verschärfung der Exportkontrolle umzusetzen<br />
(z.B. durch Ausweitung des Geltungsbereichs auch<br />
auf Vermittlungsdienstleistungen und Einbeziehung des<br />
Technologietransfers, d.h. Bereitstellen von Software und<br />
Technologie), aber auch um das Genehmigungsverfahren<br />
zu beschleunigen und zu ver einfachen (z.B. durch die Einführung<br />
neuer Allgemeingenehmi gungen der Gemeinschaft<br />
für nicht-nukleare Dual-Use Güter). Nachdem die EU-<br />
Kommission aufgrund massiver Kritik aus den Mitgliedstaaten<br />
wie auch von Seiten der Industrie einen Teil<br />
ihrer –praxisuntauglichen – Novellierungsvorschläge zurück<br />
gezogen hatte, konnte die Revision verabschiedet<br />
werden und erschien im Amtsblatt der EU als Verordnung<br />
(EG) 428/2009 (ABL. Nr. L 134 vom 29.05.2009).<br />
Novellierung der Dual-Use-Verordnung<br />
428/2009/EG<br />
Bereits vor der Verabschiedung der Verordnung 428/2009<br />
hatte die EU-Kommission angekündigt, in einem weiteren<br />
Schritt den Annex IV der Verordnung zu novellieren.<br />
Sicherlich auch bedingt durch den Wechsel in der EU-<br />
Kommission legte die derzeit amtierende EU-Kommission<br />
erst im Herbst 2016 einen Revisionsentwurf vor, der jedoch<br />
über eine bloße Überarbeitung des Annex IV weit hinaus<br />
geht. Angedacht war von der Vorgänger-Kommission,<br />
mit der Novellierung die gestiegenen Sicherheitsanforderungen<br />
mit dem Grundsatz des freien Warenverkehrs<br />
und dem Erhalt der Wettbewerbsfähigkeit der europäischen<br />
Industrie zu einem besseren Ausgleich zu bringen<br />
als bisher. Der Revisionsvorschlag der jetzigen EU-<br />
Kommission wird diesem Ziel jedoch aus Sicht der<br />
europäischen Nuklearindustrie wie auch aus Sicht der<br />
nicht-nuklearen Industrie in der EU nicht gerecht.<br />
Schutz von Menschenrechten und Cyber-Überwachungstechnologien<br />
Im Vordergrund der Kritik steht sowohl der Vorschlag, in<br />
die Dual-Use Verordnung den Schutz von Menschenrechten<br />
aufzunehmen als auch der Vorschlag, Cyber-Überwachungstechnologien<br />
als neuen Typus eines Dual-Use<br />
Gutes in die Verordnung zu integrieren. Der Export von<br />
Technologien soll stärker kontrolliert werden, wenn das<br />
Risiko besteht, dass diese Technologien zur Überwachung<br />
von Menschen genutzt werden können. Zweifellos ist der<br />
Schutz von Menschenrechten ein hohes Gut. Angesichts<br />
der weitreichenden und rasanten geopolitischen Veränderungen<br />
wie auch angesichts ständig sich erweiternder<br />
Möglichkeiten zur digitalen Überwachung muss die<br />
Exportpolitik dieser Entwicklung zweifellos Rechnung<br />
tragen. Dies sollte allerdings auf gesicherter gesetzlicher<br />
Grundlage erfolgen. Zudem sollten verschärfte Kontrollregelungen<br />
praktikabel und sinnvoll sein und mit Augenmaß<br />
festgelegt werden. Zu bedenken ist dabei, dass heutzutage<br />
Überwachungstechnologie in vielen Produkten enthalten<br />
ist und zahlreiche Unternehmen ihre Waren weltweit vermarkten.<br />
Des weiteren sollten verschärfte Kontrollregelungen<br />
nicht dazu führen, dass Verbringung und Export von<br />
Nukleargütern grundlos strengeren Kontrollen unterworfen<br />
werden als andere Dual-Use- Güter.<br />
Bedenken gegen den Kommissionsvorschlag<br />
Jedoch bestehen zum einen Zweifel an der Mandatierung<br />
der EU-Kommission. Zum anderen fehlt es an einer klaren<br />
Definition der Menschenrechte im Kommissionsentwurf<br />
selber. Außerdem divergieren die Definitionen im Katalog<br />
der Menschenrechte in der Europäischen Menschenrechtskonvention<br />
und in der UN-Menschenrechtscharta. Hinzu<br />
kommt, dass der Kommissionsentwurf dem Exporteur, dem<br />
Broker und/oder demjenigen, der technische Überwachung<br />
zur Verfügung stellt, eine Prüfungs- und Informationspflicht<br />
auferlegt, deren Erfüllung jedenfalls ohne nähere Erläuterung<br />
(z.B. durch einen ent sprechenden Leitfaden) in vielen<br />
Fällen nicht leistbar ist. Insbesondere kleinere Unternehmen<br />
werden fachlich, zeitlich und personell nicht in der<br />
Lage sein zu beurteilen, ob das zu exportierende Gut in<br />
dem Empfängerland z.B. im Zusammen hang mit einem<br />
bewaffneten Konflikt oder einem terroristischen Akt oder<br />
von einem Dritten dazu benutzt werden soll, schwerwiegende<br />
Menschenrechts verletzungen zu begehen. Mit<br />
einem noch so guten „ Internal Compliance Programme“<br />
(ICP) werden sich diese Fragen oftmals nur unzureichend<br />
lösen lassen. Der Schutz von Menschenrechten ist im Inund<br />
Ausland im Übrigen zuvörderst eine Staatsaufgabe.<br />
Die Rechts unsicherheit auf Seiten der Unternehmen dürfte<br />
– auch nach Einschätzung der deutschen Behörden – dazu<br />
führen, dass sich die Unternehmer vermehrt ratsuchend an<br />
die zuständige Genehmigungsbehörde wenden werden, so<br />
dass deren Fallzahlen und damit die Wahrscheinlichkeit für<br />
längere Genehmigungsverfahren steigen werden. Ähnliche<br />
Bedenken bestehen gegen die Einführung einer „Catch-All“<br />
Regelung, nach der alle Internet–Überwachungstechnologien<br />
prinzipiell einer Exportgenehmigung bedürfen.<br />
SPOTLIGHT ON NUCLEAR LAW<br />
Spotlight on Nuclear Law<br />
Council Regulation of the European Dual Use Regulation – A Never Ending Story? ı Ulrike Feldmann
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
20<br />
DATF NOTES<br />
Die EU-Kommission hat im Laufe 2017 zwar einige<br />
Änderungen an ihrem Entwurf konzediert, darunter auch<br />
den Vorschlag für eine Verlängerung der – zunächst im<br />
Kommissionsentwurf auf ein Jahr festgelegten – Genehmigungsdauer<br />
sowie die Einführung einer Allgemeingenehmigung<br />
für Großprojekte aufgegriffen, ist aber z.B. auf<br />
Vorschläge für einen mehr risikobasierten Ansatz bei<br />
Nukleargütern oder für die Einführung von EU-Allgemeingenehmigungen<br />
soweit ersichtlich bisher nicht eingegangen.<br />
Allerdings beabsichtigt die Kommission, in der<br />
zweiten Dezemberhälfte wieder ein Exportkontrolle-<br />
Forum unter Beteiligung der Industrie zu veranstalten. In<br />
Fachkreisen wird es jedoch für wenig wahrscheinlich<br />
gehalten, dass die Kommission ihre Position aufgrund des<br />
Exportkontrollforums noch wesentlich ändern wird.<br />
Haltung des Europäischen Rates und<br />
des Parlaments<br />
Während der Europäische Rat sich bisher eher abwartend<br />
verhalten hat, hat sich das Europäische Parlament (EP)<br />
intensiv mit dem Vorschlag der EU-Kommission befasst.<br />
Zum Berichterstatter für die Revision der Dual-Use-<br />
Verordnung hatte das EP in 2017 MdEP Prof. Dr. Klaus Buchner<br />
bestimmt. Buchner ist u.a. Mitglied im EP-Ausschuss für<br />
auswärtige Angelegenheiten sowie in den EP-Unterausschüssen<br />
für Menschenrechte, Sicherheit und Verteidigung. Außerdem<br />
ist er stellvertretendes Mitglied im EP-Ausschuss für<br />
internationalen Handel (INTA), der federführend für die<br />
Revision der Dual-Use-Verordnung ist. Der Ausschuss INTA<br />
hat in seinem Berichtsentwurf zu dem Kommissionsentwurf<br />
424 Änderungsvorschläge gemacht (z.B. Ausdehnung<br />
des Schutzes der Menschenrechte, Veröffentlichung der<br />
Abwägungskriterien für die Exportkontrolle von Dual-Use-<br />
Gütern, Klarstellung des Begriffs des Exporteurs sowie Ablehnung<br />
einer Allgemeingenehmigung für Großprojekte als<br />
zu nuklearbezogen). Am 23. November 2017 hat der Ausschuss<br />
INTA in erster und einziger Lesung mit der ganz<br />
überwiegenden Mehrheit der Stimmen dafür gestimmt, die<br />
Exportkontrollen von Überwachungstechnologien deutlich<br />
auszuweiten und die Menschenrechte zum zentralen Bestandteil<br />
der Exportkontrolle zu machen. Berichterstatter<br />
Buchner befürchtet jedoch, wie sich seiner Presseerklärung<br />
vom 23.11.2017 zu der Beschlussfassung im INTA-Ausschuss<br />
vom selben Tag entnehmen lässt, „dass die Industrie,<br />
die um ihre Geschäfte bangt, massiven Druck ausübt, und<br />
mithilfe ihrer Lobbyisten die Verabschiedung der<br />
Ver ordnung bremst und versucht abzuschwächen.“ Es<br />
besteht die Gefahr so Buchner, „dass die gute, heute vom<br />
INTA-Ausschuss beschlossene Reform von der Industrie<br />
mit der willfährigen Unterstützung konservativer Abgeordneter<br />
im Plenum verwässert wird“.<br />
Wer solcherart versucht, einen stärkeren Schutz von<br />
Menschenrechten einzufordern, dürfte damit allerdings<br />
sich und seiner Sache einen Bärendienst erweisen.<br />
Ausblick<br />
Sollte das Plenum, wie terminiert am 16. Januar <strong>2018</strong> einen<br />
Beschluss zum Novellierungsentwurf fassen (was nicht<br />
sicher ist), dürfte sich der Rat vermutlich ab Februar/März<br />
<strong>2018</strong> intensiver mit der Thematik befassen. Die Geschichte<br />
um die Novellierung der Dual-Use Verordnung geht also<br />
zumindest noch ein Weilchen weiter.<br />
Autorin<br />
Ulrike Feldmann<br />
Berlin, Germany<br />
Notes<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 />
New Explanatory Video:<br />
Dismantling – 60 Seconds<br />
The essentials of decommissioning and dismantling a nuclear<br />
power plant in 60 seconds:<br />
• Who is responsible?<br />
• Who supervises it?<br />
• What happens with the material?<br />
You get brief answers on these and more questions<br />
in this explanatory video from DAtF (in German).<br />
3 The complete video can be watched at www.kernenergie.de<br />
or at the DAtF YouTube channel.<br />
3 A more comprehensive explanatory video, a brochure of DAtF<br />
on Decommissioning of NPPs and additional Information<br />
(all in German) are available on www.kernenergie.de.<br />
New Edition of the Brochure<br />
on the Final Disposal<br />
of High Radioactive Waste<br />
The brochure “Endlagerung hochradiaoktiver Abfälle” (in German)<br />
gives you a comprehensive overview on:<br />
• The history of final disposal in Germany<br />
and current waste management<br />
• How the new site selection process will run<br />
and what are the safety criteria<br />
• Who will run the process, who will be involved<br />
and how it is paid for<br />
3 These and other issues surrounding the management<br />
of highly active waste in Germany are addressed<br />
in the brochure available online and in print.<br />
DAtF Notes
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Nuclear Safety, Security and Safeguards:<br />
An Application of an Integrated Approach<br />
Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger<br />
and Jonathan Scott<br />
1 Introduction At the 34 th G8 1 Summit in Japan in 2008 the assembled leaders acknowledged the role of<br />
nuclear power in reducing CO 2 emissions. Part of the final communique stated their commitment to the highest possible<br />
standards on “nuclear non-proliferation, safeguards, safety and security” [2]. They recognised that synergies exist<br />
between the 3Ss, (nuclear safety, nuclear security, and nuclear safeguards) and considered it was important that the<br />
separate disciplines are integrated, and that the 3S infrastructure is strengthened through international cooperation<br />
and assistance.<br />
In order to identify the synergies<br />
between the individual specialisms,<br />
international legislation and regulatory<br />
regimes are reviewed before<br />
considering the methods and assessment<br />
techniques used. We then<br />
consider which approaches can contribute<br />
most to improving the integration<br />
of the nuclear 3S, and recount<br />
practical experience of implementing<br />
the Triple S approach.<br />
The aims for the individual<br />
specialisms are:<br />
• Safety is aimed at protecting<br />
workers and the public from the<br />
harmful effects of radiation (or<br />
chemicals or other hazards);<br />
• Security is aimed at preventing<br />
malicious acts that might harm a<br />
nuclear facility (sabotage) or result<br />
in the loss (theft) of nuclear<br />
materials; and<br />
• Safeguards are aimed at preventing<br />
the diversion of nuclear materials<br />
from a civil nuclear programme<br />
to nuclear weapons purposes.<br />
The 3Ss share the same overall objectives<br />
of protecting the public and the<br />
environment from potential hazards.<br />
They use similar principles to achieve<br />
protection; multiple barriers, defence<br />
in depth, decision analysis and consequence<br />
assessment. The regulatory<br />
regimes for all 3Ss use, in the main,<br />
the same processes; assessment, permissioning,<br />
inspection, enforcement<br />
and influence [3].<br />
3.1 Safety<br />
The International Atomic Energy<br />
Agency (IAEA) Fundamental Safety<br />
Principles document [4] states “The<br />
fundamental nuclear safety objective<br />
is to protect people and the environment<br />
from the harmful effects of<br />
ionising radiation”<br />
“To ensure that facilities are operated<br />
and activities conducted so as to<br />
achieve the highest standards of safety<br />
that can reasonably be achieved,<br />
measures have to be taken:<br />
a) To control the radiation exposure<br />
of people and to prevent the release<br />
of radioactive material to the<br />
environment;<br />
b) To restrict the likelihood of events<br />
that might lead to a loss of control<br />
over a nuclear reactor core, nuclear<br />
chain reaction, radioactive source<br />
or any other source of radiation;<br />
and<br />
c) To mitigate the consequences of<br />
such events if they were to occur”.<br />
3.2 Security<br />
Nuclear security focuses on the prevention,<br />
detection and response to<br />
malicious acts involving or directed at<br />
nuclear material, other radioactive<br />
material, associated facilities, or<br />
associated activities [5]. The objectives<br />
of a State’s Physical Protection<br />
Regime [6] should be:<br />
a) To protect against unauthorised<br />
removal;<br />
b) To locate and recover missing<br />
nuclear material;<br />
c) To protect against sabotage; and<br />
d) To mitigate or minimize effects of<br />
sabotage.<br />
3.3 Safeguards<br />
The objective of Safeguards is to prevent<br />
the diversion of nuclear material<br />
from peaceful use to nuclear weapons<br />
or other nuclear explosive devices<br />
(Article III.1 of the Non-Proliferation<br />
Treaty (NPT)).<br />
4 Approaches<br />
4.1 Safety<br />
The concept of defence in depth is<br />
fundamental to nuclear safety to<br />
prevent accidents and if prevention<br />
fails, to limit potential consequences.<br />
Nuclear Safety Assessment has a<br />
number of complementary analysis<br />
Safety Security Safeguards<br />
Convention on Nuclear Safety<br />
Convention on Assistance<br />
in the Case of a Nuclear Accident<br />
Convention on the Physical Protection<br />
of Nuclear Materials (CPPNM)<br />
United Nations (UN) International<br />
Convention for the Suppression<br />
of Acts of Nuclear Terrorism<br />
IAEA Statute<br />
<strong>atw</strong>-Special „Eurosafe<br />
2017“. In cooperation<br />
with the EUROSAFE<br />
2017 partners,<br />
Bel V (Belgium),<br />
CSN (Spain), CV REZ<br />
(Czech Republic),<br />
MTA EK (Hungary),<br />
GRS (Germany), ANVS<br />
(The Netherlands),<br />
INRNE BAS (Bulgaria),<br />
IRSN (France),<br />
NRA (Japan),<br />
JSI (Slovenia),<br />
LEI (Lithuania),<br />
PSI (Switzerland),<br />
SSM (Sweden),<br />
SEC NRS (Russia),<br />
SSTC NRS (Ukraine),<br />
VTT (Finland),<br />
VUJE (Slovakia),<br />
Wood (United<br />
Kingdom).<br />
Revised version<br />
of a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and<br />
7 November 2017.<br />
1) Canada, France,<br />
Germany, Italy,<br />
Japan, Russia,<br />
United Kingdom,<br />
United States<br />
and European<br />
Commission<br />
Non Proliferation Treaty<br />
(NPT)<br />
21<br />
ENVIRONMENT AND SAFETY<br />
2 International statues and<br />
agreements<br />
Some of the main international<br />
statutes (written law passed by a<br />
legislative body) and agreements for<br />
the 3Ss is presented in Table 1.<br />
3 Nuclear 3S objectives<br />
By considering the objectives of each<br />
of the 3Ss it becomes clear that they<br />
share the same broad aim and desired<br />
outcomes.<br />
Convention on the Early<br />
Notification of a Nuclear Accident<br />
or Radiological Emergency<br />
Threats to International Peace and<br />
Security caused by Terrorist Acts –<br />
UN Resolution 1373<br />
Code of Conduct on the Safety and Security of Radioactive Sources<br />
Joint Convention on the Safety<br />
of Spent Fuel Management and<br />
on the Safety of Radioactive Waste<br />
Management<br />
Code of Conduct on the Safety<br />
of Research Reactors<br />
| | Tab. 1.<br />
International Legislation and Agreements.<br />
Safeguards Agreements<br />
Additional Protocols<br />
Non-proliferation of Weapons<br />
of Mass Destruction –<br />
United Nations Security Council<br />
(UNSC) Resolution 1540<br />
Comprehensive Test Ban Treaty<br />
(CTBT)<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards: An Application of an Integrated Approach ı Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger and Jonathan Scott
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
ENVIRONMENT AND SAFETY 22<br />
| | Fig. 1.<br />
Schematic showing the general ranges of applicability of the 3 methods of Fault Analysis 2,3 .<br />
attitudes in organizations and individuals<br />
which establishes that, as an<br />
overriding priority, protection and<br />
safety issues receive the attention<br />
warranted by their significance”<br />
[9]. The development of a good<br />
safety culture requires a transparent<br />
approach to information sharing and<br />
dissemination. This helps ensure that<br />
incident reoccurrences can be prevented,<br />
and others who may be using<br />
the same or similar equipment,<br />
techniques or procedures can review<br />
their arrangements to prevent a<br />
similar incident.<br />
“The existence of a good safety<br />
culture is a prerequisite for the<br />
implementation of a good safety case.<br />
The converse is also true” [10]. This<br />
enables a good safety case to be<br />
translated into beneficial changes<br />
in behaviour associated with the<br />
existing safety culture and arrangements<br />
for the management of safety.<br />
Practicing a graded approach<br />
to safety ensures that the effort<br />
expanded is proportionate to the<br />
possible consequences. Figure 1 is<br />
from the Office for Nuclear Regulation<br />
Safety Assessment Principles [11]<br />
and shows the applicability for<br />
the methods of fault analysis; PSA,<br />
DBA and SAA. Thus more assessment<br />
effort is expended on those higher<br />
consequence and higher frequency<br />
events.<br />
2) Office for Nuclear<br />
Regulation [11].<br />
3) Target 4 (BSL):<br />
‘ Target 4 is<br />
intended to provide<br />
a broad indication<br />
of where DBA might<br />
be expected to be<br />
applied’ [11]. BSL –<br />
Basic Safety Level<br />
4) Based upon a<br />
Sandia National<br />
Laboratories<br />
diagram<br />
| | Fig. 2.<br />
Design and Evaluation Process Outline 4 .<br />
techniques to demonstrate the<br />
effectiveness of defence in depth,<br />
such as:<br />
• Design Basis Analysis (DBA): to<br />
ensure that the design is robust,<br />
fault tolerant and has effective<br />
safety measures;<br />
• Probabilistic Safety Analysis (PSA):<br />
to ensure risks are acceptable,<br />
understand inter-dependencies<br />
and to evaluate failures; and<br />
• Severe Accident Analysis (SAA): to<br />
determine further practicable<br />
measures to improve defence in<br />
depth.<br />
The hierarchical view deviations,<br />
incidents and accidents for nuclear<br />
facilities is compared against five<br />
levels of defence in depth [7] for<br />
safety:<br />
• Preventing deviations from normal<br />
operations;<br />
• Controlling deviations from operational<br />
states;<br />
• Controlling accidents within the<br />
design basis;<br />
• Mitigating accidents and ensuring<br />
confinement of radioactive materials;<br />
and<br />
• Mitigating the radiological consequences<br />
of radioactive releases.<br />
This hierarchical view allows<br />
designers, operators and others to<br />
identify where they can most effectively<br />
contribute to maintaining safety.<br />
The Safety Case is a well-documented<br />
approach normally used by<br />
regulators for proportionally assessing<br />
the safety submissions against<br />
the radiological hazards presented.<br />
Safety cases are typically defined as a<br />
“ structured argument, supported by a<br />
body of evidence that provides a<br />
compelling, comprehensible and valid<br />
case that a system is safe for a given<br />
application in a given operating<br />
environment” [8].<br />
For the safe operation of a nuclear<br />
site, facility or activity an effective<br />
safety culture needs to be in-place and<br />
fostered. Safety culture is defined as<br />
“The assembly of characteristics and<br />
4.2 Security<br />
A number of methodologies are used<br />
in security to increase the likelihood<br />
of creating and maintaining secure<br />
operations. An example holistic<br />
approach is the Design and Evaluation<br />
Process Outline (DEPO) (Figure 2)<br />
[12]. The physical protection system<br />
(PPS) is developed from determining<br />
the targets to be protected from the<br />
postulated malicious capabilities, and<br />
then designing for delay, detection,<br />
assessment and response. Vulnerability<br />
assessment is undertaken to<br />
ensure that the PPS is likely to be<br />
effective and depending on the outcome<br />
the design will be refined or<br />
implemented.<br />
However, a number of assessment<br />
techniques need to be deployed<br />
and the associated performance<br />
measures calculated and considered<br />
for operational acceptance. For<br />
example, a sensitive detector with a<br />
high probability of detection may<br />
detect all intrusions but have a high<br />
false alarm rate such that responders<br />
ignore the alarms being received.<br />
Defence in depth for security<br />
[7] comprise layers of physical and<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards: An Application of an Integrated Approach ı Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger and Jonathan Scott
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
| | Fig. 3.<br />
Schematic showing the general ranges of applicability for 3 security assessment methods.<br />
technical measures, along with<br />
operational and procedural protection<br />
that have to be overcome or circumvented<br />
by an adversary. The defence<br />
in depth approach should be applied<br />
in the following order:<br />
• Detecting a potential malicious act;<br />
• Delaying the adversary to allow an<br />
appropriate response; and<br />
• Responding to or neutralising the<br />
attack.<br />
Risk management techniques also<br />
need to be used to balance investment<br />
required to prevent high consequence<br />
low probability events and low consequence<br />
high probability events without<br />
putting an unnecessary burden on<br />
operational processes.<br />
A State’s Design Basis Threat<br />
(DBT) [13] details the capability and<br />
capacity of the malicious actors that<br />
any PPS should counter to maintain<br />
the security of nuclear material<br />
and other radioactive materials.<br />
Using that as a basis for the threats,<br />
scenarios should be developed that<br />
will be used in vulnerability assessments<br />
to determine how effective<br />
the security arrangements are likely to<br />
be in practice. Any weaknesses should<br />
be identified so that compensatory or<br />
enhancements in the arrangements<br />
can be implemented. The type and<br />
quality of assessment techniques that<br />
should be undertaken is suggested in<br />
Figure 3.<br />
The Nuclear Security Case 5 , part of<br />
the Nuclear Site Security Plan, is a<br />
more recent development than the<br />
Safety Case. Like the safety case the<br />
security case “should justify the<br />
claims, arguments and rationale for<br />
the ‘duty holders’ security regime by<br />
substantiating the security arrangements<br />
for a site, plant, activity, operation<br />
or modification. It should provide<br />
written evidence that the relevant<br />
security standards have been or are<br />
going to be met. It should also demonstrate<br />
that the risk posed by malicious<br />
activity has been reduced as far as<br />
could be reasonably expected” [14].<br />
As with the safety case effort should<br />
be expended to reviewing security<br />
as a system rather than as individual<br />
components.<br />
Risk management techniques<br />
are used to manage any variations<br />
between the optimal arrangements<br />
and what is in currently in place,<br />
particularly when a possible vulnerability<br />
is identified.<br />
Vulnerability assessment techniques<br />
to determine the performance<br />
of security arrangements involve<br />
many aspects of the system performance<br />
including the probability of<br />
detection, probability of interruption,<br />
probability of neutralisation and<br />
probability of effectiveness. The<br />
IAEA Nuclear Security Assessment<br />
Methodology (NUSAM) Co-ordinated<br />
Research Programme (CRP) has been<br />
establishing a risk-informed, performance-based<br />
methodological framework<br />
for nuclear security assessment<br />
at sites, facilities and activities so that<br />
practitioners will be better informed<br />
of the approaches and techniques that<br />
can be used, and those that provide<br />
the most effective assessment and value<br />
for the different facility type. The<br />
CRP also allows the different methods<br />
to be compared and helps to identify<br />
the comparative strengths, weaknesses<br />
and limitations of the alternative<br />
approaches. This should ensure<br />
a consistency of approach in security<br />
assessment, and therefore by implication<br />
a baseline standard for international<br />
approaches.<br />
As in the case of safety for the<br />
secure operation of a nuclear site,<br />
facility or activity, an effective security<br />
culture needs to be in-place and<br />
fostered. Security culture is defined<br />
as “The assembly of characteristics,<br />
attitudes and behaviour of individuals,<br />
organizations and institutions<br />
which serves as a means to support<br />
and enhance nuclear security” [15].<br />
Security and safety culture are<br />
both based upon the principles of<br />
adopting a questioning attitude, rigorous<br />
and prudent approaches, and<br />
effective communication.<br />
4.3 Safeguards<br />
Underpinning and implementing<br />
the principles within the Non-Proliferation<br />
Treaty (NPT) the main<br />
approaches used by the safeguards<br />
community for the protection of civil<br />
nuclear material preventing it from<br />
being redirected into weapons activities<br />
is ‘Safeguards by Design’ [16]<br />
and nuclear materials accountancy<br />
and control (NMAC). The physical<br />
arrangements including Tamper<br />
Indicator Devices, multiple barriers,<br />
NMAC and facility arrangements such<br />
as Material Balance Areas provide<br />
additional measures for defence<br />
in depth aiding the inspection of<br />
material and the ability to detect<br />
potential diversion.<br />
Inspection and material characterisation<br />
activities are used in decision<br />
analysis to determine whether<br />
the plant or facility is operating to<br />
specification and agreement.<br />
Safety, security and safeguards<br />
broadly follow the same principles to<br />
achieve protection.<br />
5 3S synergies<br />
The synergies and major considerations<br />
in the nuclear 3S are shown in<br />
Figure 4 [17]. This identifies the main<br />
issues and considerations within<br />
the 3S and where they intersect and<br />
overlap, irrespective of the type of<br />
regulatory regime.<br />
5.1 Triple S<br />
Moving into the practice of 3S,<br />
through the applications of methods<br />
and techniques we use the term Triple<br />
S. Thus, when Figure 4 is revised<br />
with a selection of typical, but not<br />
exhaustive, activities and assessments<br />
5) Part of NNL’s<br />
approach to<br />
demonstrating<br />
compliance with<br />
ONR’s Nuclear<br />
Security Assessment<br />
Principles (SyAPs).<br />
ENVIRONMENT AND SAFETY 23<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards: An Application of an Integrated Approach ı Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger and Jonathan Scott
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
ENVIRONMENT AND SAFETY 24<br />
| | Fig. 4.<br />
Nuclear 3S 6 .<br />
| | Fig. 5.<br />
Nuclear Triple S.<br />
6) Based upon ‘An<br />
integrated approach<br />
to nuclear safety<br />
and security: in the<br />
context of 3S’,<br />
Jor-Shan Choi,<br />
Tokyo, Japan,<br />
9 December 2011<br />
7) An area inside a<br />
protected area containing<br />
equipment,<br />
systems or devices,<br />
or nuclear material,<br />
the sabotage of<br />
which could directly<br />
or indirectly lead to<br />
high radiological<br />
consequences [18].<br />
that are undertaken the resulting<br />
synergies, overlaps and interaction<br />
are presented in Figure 5. For<br />
example, Vital Area Identification<br />
(VAI) is a process to identify potential<br />
high consequence targets so that<br />
protection can be provided to prevent<br />
or reduce the likelihood of sabotage<br />
[18]. Although VAI is primarily driven<br />
by a security need the contribution<br />
from safety specialists’ considering<br />
the potential consequences and<br />
operational limitations is important,<br />
and is therefore shown in the intersection<br />
of safety and security. There<br />
are some activities that all specialisms<br />
contribute to, such as new nuclear<br />
build and stakeholder engagement,<br />
and is shown across all three sections<br />
of the diagram requiring input from<br />
all three. It is where activities fall into<br />
more than one area that deliberate<br />
and positive interactions between all<br />
the specialisms will provide added<br />
value and potential conflicts are<br />
averted or minimised. Each specialist<br />
develops a clearer understanding of<br />
the needs, intentions and priorities of<br />
the other specialists, resulting in an<br />
integrated approach to Triple S. Thus<br />
time, effort and cost are minimised as<br />
plant workarounds, reworks or design<br />
changes are prevented, and operational<br />
arrangements can be considered<br />
earlier in the project.<br />
Exploring this in further detail,<br />
safety and security, followed by<br />
security and safeguards, is where the<br />
largest interaction, potential synergies<br />
and similar approaches are to be<br />
found.<br />
5.2 Safety and security<br />
Security requires extensive safety<br />
input for the identification of Vital<br />
Areas 7 . The safety assessments including<br />
radiological consequence<br />
modelling, radiological hazard analysis,<br />
PSA, SAA, internal and external<br />
hazards, and layout design all contribute<br />
to identifying potential Vital<br />
Areas.<br />
The design basis accidents and<br />
design basis threats (DBT) approaches<br />
in both specialisms guide designers,<br />
practitioners and assessors to adequately<br />
consider those threats that<br />
may need to be countered.<br />
Safety and security both use a<br />
graded approach. The relative importance<br />
of accident prevention and<br />
mitigation measures is expressed in<br />
terms of the adverse consequences for<br />
public and worker health. Likewise<br />
the relative importance of security<br />
measures is directed towards preventing<br />
and limiting what are considered<br />
high and low consequence<br />
events.<br />
Prevention, Response, Control<br />
and Management effort to counter<br />
malicious attack for security, or accidents<br />
in safety require considerations<br />
on the speed of progress of an incident,<br />
the potential consequences of<br />
those responses and management<br />
actions and how to minimise the<br />
impact on the plant, people, public<br />
and environment.<br />
Approaches and methods used in<br />
the minimisation of impact for radiological<br />
consequence through ‘As Low<br />
As Reasonably Practicable’ (ALARP)<br />
practices in safety are commensurate<br />
with those used by security not to<br />
create ‘As Secure As Reasonable<br />
Practicable’ (ASARP) but rather the<br />
introduction of risk management<br />
practices to manage potential vulnerabilities<br />
identified through PPS<br />
evaluation activities.<br />
Safety and security both encourage<br />
and embrace Advisory Missions and<br />
inspections; from the World Association<br />
of Nuclear Operators (WANO),<br />
Integrated Regulatory Review Service<br />
(IRRS) and Operational Safety Review<br />
Team (OSART) for safety; from the<br />
International Physical Protection<br />
Advisory Service (IPPAS) for security,<br />
and which is understood to potentially<br />
be expanded to include a module on<br />
NMAC.<br />
Safety and security both attempt to<br />
foster positive cultures that identify<br />
and report problems and issues. However<br />
the transparent and open communications<br />
of safety may conflict with<br />
the ‘need to know’ principles employed<br />
in security. Appropriate implementation<br />
of ‘need to know’ principle where<br />
consideration is given to what is<br />
‘ needed to be known’ can d irect appropriate<br />
filtering and redaction so that<br />
appropriate inter actions can occur<br />
without compro mising security of<br />
materials or information. For example,<br />
consequence assessors do not need to<br />
know the locations or means that<br />
material can be acquired by a perpetrator<br />
to undertake the assessment.<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards: An Application of an Integrated Approach ı Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger and Jonathan Scott
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Thus, by fostering an approach<br />
that integrates both safety and<br />
security in a mutually supporting<br />
manner through peer-to-peer and<br />
other challenges of behaviours creates<br />
the opportunity for reinforcement of<br />
positive behaviours.<br />
5.3 Safety and safeguards<br />
Safety and safeguards particularly<br />
interact during the design activities<br />
around ‘Safeguards by Design’ [19],<br />
and then through the construction<br />
phase. One example of this interaction<br />
is preventing diversion through layout<br />
design, aiding inventory control<br />
(Nuclear Material Accounting and<br />
Control – NMAC) through criticality<br />
control accountancy measures.<br />
Inspections are carried out as part<br />
of an integrated safeguards regime<br />
and undertaken by inspectors from<br />
the IAEA, Euratom or the Brazilian-<br />
Argentine Agency for Accounting and<br />
Control of Nuclear Materials (ABACC).<br />
5.4 Security and safeguards<br />
Security and safeguards should interact<br />
during the design activities around<br />
‘Safeguards by Design’ [19], and then<br />
through the operational phase of the<br />
facility or plant. Both safeguards and<br />
security [20] are aimed at deterring<br />
and detecting unauthorised removal<br />
of nuclear material, providing assurance<br />
that all nuclear material is secure<br />
and timely detection of any material<br />
loss.<br />
There are areas where security and<br />
safeguards can interact to improve<br />
effectiveness and efficiency in achieving<br />
their objectives such as research<br />
and development of Non-destructive<br />
Assay equipment and surveillance<br />
system, analysis capability (i.e.<br />
Nuclear Forensics, Destructive Analysis)<br />
and, Security- and Safeguardsby-<br />
design. Further, enhancing nuclear<br />
security may be achieved through the<br />
use of nuclear material accounting<br />
and control systems [21]. This is an<br />
approach being advocated by the IAEA<br />
and clearly demonstrates how existing<br />
accountancy measures can be utilised<br />
to provide a potential additional<br />
means through which the theft of<br />
material by an insider can be detected.<br />
6 Duty holder application<br />
of triple S integration<br />
The integration of Nuclear Safety,<br />
Nuclear Security and Nuclear Safeguards<br />
can be beneficial for nuclear<br />
site duty holders, operators and<br />
tenant organisations. More broadly,<br />
the early integration and interaction<br />
of safety and security in critical<br />
national infrastructure (CNI) and other<br />
projects that require a security input<br />
is of value. A duty holder can begin<br />
the integration of Safety, Security<br />
and Safeguards (SSS) by the formation<br />
of a SSS team, bringing together<br />
Safety, Security, Safeguards and the<br />
broader safety disciplines. The following<br />
application of Triple S integration<br />
shows how a security technique can be<br />
applied to a recent change in nuclear<br />
registration within the UK and how<br />
this technique can be bolstered by the<br />
Safety and Safeguards disciplines.<br />
6.1 NNL application and<br />
experience<br />
Returning to the UK nuclear industry,<br />
the Office for Nuclear Regulation<br />
(ONR), recently replaced its security<br />
guidance to support the regulations<br />
by introducing the Security Assessment<br />
Principles (SyAPs) as a replacement<br />
for the National Objectives,<br />
Requirements and Model Standards<br />
(NORMS). NORMS was considered by<br />
some to be a prescriptive approach to<br />
nuclear security regulation. It set out<br />
security objectives that dutyholders<br />
were expected to meet. However,<br />
some in the industry viewed the<br />
suggested Model Standards, that were<br />
presented as what may allow a facility<br />
or site to meet regulatory compliance<br />
was provided as guidance. The introduction<br />
of SyAPs is a move to an outcomes-based<br />
regulatory regime and a<br />
non-prescriptive approach to nuclear<br />
security, giving duty holders more<br />
freedom and therefore more space for<br />
Triple S integration. Importantly in<br />
the context of 3S principles, SyAPs are<br />
more in line with the Safety Assessment<br />
Principles (SAPs), reinforcing<br />
the benefits of adopting an integrated<br />
approach to safety and security, and<br />
working together, learning for each<br />
other, and adapting methodologies to<br />
meet similar regulatory expectations.<br />
The integration of Triple S has been<br />
recognised by the ONR as an efficient<br />
way of thinking, this is reflected in the<br />
formation of the Security Informed<br />
Nuclear Safety (SINS) team within<br />
ONR.<br />
However, with the introduction of<br />
SyAPs, duty holders across the UK<br />
must review their current security<br />
arrangements so that the requirements<br />
of SyAPs can be met. Reviewing<br />
nuclear site security measures in line<br />
with SyAPs using a team that includes<br />
specialists from the three disciplines;<br />
Safety, Security and where appropriate<br />
Safeguards; will allow duty<br />
holders to better address the principles<br />
and gain organisational value.<br />
6.2 Operational requirements<br />
The Centre for the Protection of<br />
National Infrastructure (CPNI) is the<br />
government authority for protective<br />
security advice to the UK national<br />
infrastructure. [22]. CPNI provides<br />
tools to help CNI companies and<br />
organisations undertake an improved<br />
security assessment of their sites, and<br />
their methods are often considered<br />
‘best relevant practice’ and serve as a<br />
logical approach to an outcomes-based<br />
regulatory regime such as SyAPs.<br />
One such method promulgated by<br />
CPNI is the Operational Requirements<br />
(ORs) process. [23] The OR process<br />
identifies, develops and aids justification<br />
of actions to be taken and<br />
investments to be made to protect<br />
assets. [24] The OR process consists of<br />
two levels; Level 1 OR seeks to:<br />
• Identify assets and critical infrastructure<br />
• Identify threats and vulnerabilities<br />
• Assess possible risks<br />
• Identify risk mitigation options<br />
and develop a Strategic Security<br />
Plan (SSP) Review organisational<br />
readiness to deliver the developed<br />
SSP.<br />
Level 2 OR is a continuation of the<br />
Level 1 OR. It is concerned with<br />
in-depth analysis of requirements<br />
suggested as a result of the security<br />
posture formed from the Level 1 OR<br />
process. An example application of<br />
the OR process with regards to SyAPs<br />
can be seen below (Figure 6).<br />
The OR process provides a useful<br />
vehicle for the integration of Safety,<br />
Security and Safeguards, from the<br />
perspective of ‘Safety and Safeguards<br />
informed Security’ (SSIS 8 ). SyAPs<br />
requires the categorisation of nuclear<br />
sites and facilities, and nuclear<br />
material (NM) and other radioactive<br />
materials (ORM) for both theft and<br />
sabotage; it follows logically to<br />
integrate safety specialisms when<br />
considering potential consequences<br />
(categorisation) and the malicious<br />
actions that may be undertaken to<br />
achieve such consequences, as well as<br />
considering the implications of operational<br />
‘flow’ of material around a<br />
proposed facility from an NMAC<br />
perspective. Such perspectives may<br />
further inform the design process at a<br />
high level (remembering the purpose<br />
of OR1).<br />
Using the principles behind the OR<br />
process and SSIS in conjunction,<br />
assets, vulnerabilities, risks and<br />
mitigations are found, resulting in<br />
a security posture for the site that<br />
takes due account of safety and safeguards.<br />
Triple S integration allows<br />
8) Coined herein to<br />
describe the<br />
intermediate stage<br />
between individual,<br />
isolated Safety,<br />
Security and<br />
Safeguards<br />
functions and the<br />
notion of fully<br />
integrated ‘SSS’.<br />
ENVIRONMENT AND SAFETY 25<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
ENVIRONMENT AND SAFETY 26<br />
| | Fig. 6.<br />
Example of the integration of Triple S into phases of an OR process.<br />
| | Fig. 7.<br />
Level 1 OR, Categorisation.<br />
reinforcement of decisions, shown<br />
above as the SAPs supports the regulatory<br />
guidance for security categorisation.<br />
Additionally, the use of nuclear<br />
and radiological safety consequence<br />
analysis supports the security categorisation<br />
for sabotage. The Level 1<br />
OR process can be used for SyAPs<br />
reviews as shown in Figure 7.<br />
Once a duty holder has categorised<br />
their nuclear site, the Level 2 OR<br />
process can assess the individual<br />
security requirements of site and<br />
facility areas (Figure 8).<br />
The Level 2 OR process assesses<br />
the site in terms of its security<br />
capabilities. The outcomes of the<br />
Level 2 OR process are a set of<br />
performance requirements against<br />
the defined functions that the Physical<br />
Protection System (PPS) must meet in<br />
order to be compliant with the<br />
standards held by SyAPs; those being<br />
Delay, Detect, Assess, Control of<br />
Access and Insider Mitigation.<br />
SAPs feeds into the regulatory<br />
guidance that underlies the security<br />
assessment. Initial attempts at applying<br />
the safety methodology of HAZOPs<br />
(using keywords to explore potential<br />
issues in the design and test for ‘compliance’)<br />
resulted in a level of success.<br />
However, this experiment highlighted<br />
the fundamental differences between<br />
safety and security, in that safety<br />
can be probabilistically assessed and<br />
security cannot. Said differently, the<br />
laws of physics and attributes of<br />
systems/components determine what<br />
is and is not possible in the world of<br />
safety. In the world of security, outcomes<br />
are more strongly determined<br />
by malicious capabilities (knowledge<br />
and resources) and their imagination;<br />
as such security scenarios cannot be<br />
conceptualised deterministically and<br />
calculated probabilistically.<br />
The outcomes of the Level 2 OR<br />
process are defined and communicated<br />
in a Performance Specification.<br />
The Performance Specification relays<br />
the PPS specifications that the duty<br />
holder requires to the design process.<br />
The PPS design process is carried out<br />
using aspects of Safety, Security and<br />
Safeguarding to update the nuclear<br />
facility and maintain high standards<br />
in all 3 fields (Figure 9).<br />
SSS (or SSIS) can influence specific<br />
design aspects, such as turnstile<br />
requirements (linking access control<br />
and emergency egress), material store<br />
access and surveillance features<br />
( security and safeguards) and material<br />
handling limits in specified areas<br />
(radio logical protection and counterdiversion/insider<br />
threat mitigation).<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards: An Application of an Integrated Approach ı Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger and Jonathan Scott
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Duty Holders can validate the<br />
results of the OR process through<br />
further vulnerability assessments of<br />
the derived PPS. This will require<br />
further Triple S integration as each<br />
field will require an analysis of<br />
updated facilities to ensure that high<br />
standards continue (Figure 10).<br />
Integrating Triple S in the early<br />
stages of a design project (similarly to<br />
the application above) can prevent<br />
future costs and save time. Our<br />
experience has shown a greater<br />
engagement and interaction of previously<br />
disparate disciplines, whose<br />
assessment needed to be rationalised<br />
and integrated, often leading to<br />
re-assessment/re-work to ensure<br />
consistency of assessment boundaries,<br />
assumptions, etc.<br />
If integration is not considered,<br />
then designs implemented by one<br />
discipline can interfere with designs<br />
implemented by another and the<br />
earlier examples of beneficial integration<br />
may be pre-empted by the<br />
need for avoidance of conflicts or<br />
issues. A basic example of issues raised<br />
by the lack of integration can be<br />
shown by the installation of security<br />
fences interfering with on-site fire<br />
safety (evacuation routes), forcing an<br />
expensive retrofit on the security<br />
fence.<br />
7 Conclusion<br />
This paper has covered NNL’s progress<br />
to date in Triple S integration ( referred<br />
to as SSIS rather than SSS) and its<br />
implementation in a new concept<br />
design project.<br />
Whilst the potential benefits of an<br />
integrated Triple S approach are<br />
abundantly clear, it is somewhat more<br />
difficult to realise these conceptual<br />
benefits practically. The National<br />
Nuclear Laboratory (NNL) has made<br />
significant progress in its own<br />
approach to aligning the three<br />
disciplines, though the approach could<br />
still be described more as ‘Safety and<br />
Safeguards Informed Security’ (SSIS).<br />
Experience thus far has identified that<br />
specialists in Triple S disciplines need<br />
to become more aware of the priorities,<br />
approaches, methods and drivers<br />
of other specialists delivering their<br />
respective objectives to develop and<br />
promote an integrated approach.<br />
NNL has observed more effective<br />
cross-specialism communication and<br />
interactions and much heightened<br />
awareness and interaction between<br />
the broader organisation and Triple S<br />
functions. Triple S can lead to increasing<br />
professionalism as methods<br />
and techniques used by one group of<br />
| | Fig. 8.<br />
Level 2 OR, specific requirements of PPS.<br />
| | Fig. 9.<br />
Performance Specification and Design Stages of OR.<br />
| | Fig. 10.<br />
Vulnerability Assessment of new Facility Design.<br />
specialists are adapted and used by<br />
others through sharing of knowledge<br />
and learning from experience. Interaction<br />
with the other specialists can<br />
lead individuals to reconsider how to<br />
undertake work and what information<br />
is important such that safety, security<br />
and safeguards are integrated in a<br />
holistic manner.<br />
Further, integration of 3S is more<br />
likely to be achieved and be effective<br />
in the early design and construction<br />
phases of a project, with the positive<br />
effects being realised as cost and<br />
efficiency benefits throughout operation.<br />
Early interaction reduces the<br />
potential for conflict by identifying<br />
where negative interactions might<br />
ENVIRONMENT AND SAFETY 27<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
ENVIRONMENT AND SAFETY 28<br />
occur, thus, potentially expensive<br />
rework or compromises are removed.<br />
The application of the OR process<br />
to a SyAPs review shows that safety,<br />
security and safeguards can bolster<br />
the effectiveness of new design projects.<br />
It shows the importance of<br />
integration and its cost and time<br />
saving potential, and that the<br />
legitimacy of the Triple S approach<br />
spans beyond the conceptual stage.<br />
Abbreviations<br />
3S Safety, Security and Safeguards<br />
ABACC Brazilian-Argentine Agency for<br />
Accounting and Control of Nuclear<br />
Materials<br />
ALARP As Low As Reasonably Practicable<br />
ASARP As Secure As Reasonably Practicable<br />
CPPNM Convention on the Physical<br />
Protection of Nuclear Materials<br />
CRP<br />
DBA<br />
DBT<br />
Co-ordinated Research Programme<br />
Design Basis Analysis<br />
Design Basis Threat<br />
DEPO Design and Evaluation Process<br />
Outline<br />
IAEA International Atomic Energy Agency<br />
IPPAS International Physical Protection<br />
Advisory Service<br />
IRRS Integrated Regulatory Review Service<br />
NMAC Nuclear Material Accountancy and<br />
Control<br />
NPT Non Proliferation Treaty<br />
NUSAM Nuclear Security Assessment<br />
Methodologies<br />
OSART Operational Safety Review Team<br />
PPS Physical Protection System<br />
PSA Probabilistic Safety Analysis<br />
SAA Severe Accident Analysis<br />
Triple S Safety, Security and Safeguards<br />
UN United Nations<br />
UNSC United Nations Security Council<br />
VAI Vital Area Identification<br />
WANO World Association of Nuclear<br />
Operators<br />
References<br />
[5] International Atomic Energy Agency,<br />
Objectives and Essential Elements of a<br />
State's Nuclear Security Regime, Vienna:<br />
International Atomic Energy Agency,<br />
2013.<br />
[6] International Atomic Energy Agency,<br />
Nuclear Security Recommendations on<br />
Physical Protection of Nuclear Material<br />
and Nuclear Facilities, Vienna: International<br />
Atomic Energy Agency,<br />
January 2011.<br />
[7] International Atomic Energy Agency,<br />
Management of the Interface between<br />
Nuclear Safety and Security for Research<br />
Reactors, International Atomic Energy<br />
Agency, Vienna, 2016.<br />
[8] J. Inge, The Safety Case, Its Development<br />
and Use in the United Kingdom 2 (n.d.),<br />
Ministry of Defence.<br />
[9] International Atomic Energy Agency,<br />
IAEA Safety Glossary, International<br />
Atomic Energy Agency, Vienna, 2007.<br />
[10] R. J. Cullen, Safety Culture: Cornerstone<br />
of the Nuclear Safety Case, in Hazards<br />
XXI: Process Safety and Environmental<br />
Protection in a Changing World,<br />
Manchester, 2009.<br />
[11] Office for Nuclear Regulation, Safety<br />
Assessment Principles for Nuclear<br />
Facilities, Office for Nuclear Regulation,<br />
Bootle, 2014.<br />
[12] Sandia National Laboratory and Japan<br />
Atomic Energy Agency, Security by<br />
Design Handbook, 2013.<br />
[13] International Atomic Energy Agency,<br />
Development, Use and Maintenance of<br />
the Design Basis Threat: Implementing<br />
Guide, Vienna: International Atomic<br />
Energy Agency, 2009.<br />
[14] Office for Nuclear Regulation, Guidance<br />
on the Purpose, Scope and Quality of a<br />
Nuclear Site Security Plan, Office for<br />
Nuclear Regulation, Bootle, 2016.<br />
[15] International Atomic Energy Agency,<br />
Nuclear Security Culture, International<br />
Atomic Energy Agency, Vienna, 2008.<br />
[16] R. S. Bean, T. A. Bjornard und<br />
D. J. Hebditch, Safeguards-by-Design:<br />
An Element of 3S Integration, in IAEA<br />
Symposium on Nuclear Safety, April<br />
2009.<br />
[17] J.-S. Choi, An integrated approach to<br />
nuclear safety and security: in the<br />
context of 3S, in JAEA International<br />
Forum on Peaceful Use of Nuclear<br />
Energy and Nuclear Security, Tokyo,<br />
2011.<br />
[22] CPNI, About CPNI, 2017. [Online].<br />
Available:<br />
https://www.cpni.gov.uk/about-cpni.<br />
[Zugriff am 10 October 2017].<br />
[23] CPNI, Operational Requirements, 2017.<br />
[Online]. Available:<br />
https://www.cpni.gov.uk/operationalrequirements.<br />
[Zugriff am 10 October 2017].<br />
[24] CPNI, Guide to Producing Operational<br />
Requirements for Security Measures,<br />
2016.<br />
Authors<br />
Howard Chapman<br />
Jeremy Edwards<br />
Joshua Fitzpatrick<br />
Colette Grundy<br />
Robert Rodger<br />
Jonathan Scott<br />
National Nuclear Laboratory<br />
Fifth Floor, Chadwick House<br />
Warrington Road, Birchwood Park,<br />
Warrington, WA3 6AE,<br />
United Kingdom<br />
[1] NNL, Nuclear Safety, Security and<br />
Safeguards: An Integrated Approach,<br />
2017.<br />
[2] Ministry of Foreign Affairs of Japan,<br />
International Initiative on 3S-Based<br />
Nuclear Energy Infrastracture, G8<br />
Hokkaido Toyako, Institute of Oriental<br />
Culture, University of Tokyo, Hokkaido<br />
Toyako, 2008.<br />
[3] M. Weightman, Leadership and<br />
Organisational Aspects, Bootle: Office<br />
for Nuclear Regulation, 2011.<br />
[4] International Atomic Energy Agency,<br />
Fundamental Safety Principles, Vienna:<br />
International Atomic Energy Agency,<br />
November 2006.<br />
[18] International Atomic Energy Agency,<br />
Identification of Vital Areas at Nuclear<br />
Facilities, International Atomic Energy<br />
Agency, Vienna, 2012.<br />
[19] R. S. Bean, J. W. Hockert und<br />
D. J. Hebditch, Integrating Safeguards<br />
and Security with Safety into Design,<br />
in 19 th Annual EFCOG Safety Analysis<br />
Workshop, 2009.<br />
[20] K. Murakami, Nuclear Safeguards<br />
Concepts, Requirements, and Principles<br />
applicable to Nuclear Security, July 2012.<br />
[21] International Atomic Energy Agency,<br />
Use of Nuclear Material Accounting and<br />
Control for Nuclear Security Purposes at<br />
Facilities, International Atomic Energy<br />
Agency, Vienna, 2015.<br />
Environment and Safety<br />
Nuclear Safety, Security and Safeguards: An Application of an Integrated Approach ı Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger and Jonathan Scott
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Clearance of Surface-contaminated<br />
Objects from the Controlled Area<br />
of a Nuclear Facility: Application of the<br />
SUDOQU Methodology<br />
F. Russo, C. Mommaert and T. van Dillen<br />
1 Introduction During and after the Fukushima nuclear accident, the possibility existed that surface-contaminated<br />
consumer goods, freight containers and conveyances would be imported from Japan, which revealed the need<br />
for proper criteria and screening levels for surface contamination of these items, to insure protection of the public.<br />
In this framework, it was concluded<br />
that the then existing dose-calculation<br />
models mostly addressed exposure<br />
scenarios for occupationally exposed<br />
workers, which were generally not<br />
aimed at properly evaluating the<br />
effective dose incurred by members of<br />
the public exposed to surface-contaminated<br />
objects. The main difference<br />
between occupational and public<br />
exposure scenarios is that, while<br />
workers may frequently be exposed<br />
to freshly contaminated objects,<br />
members of the public are likely to<br />
come in contact with only one (same)<br />
object during a prolonged period of<br />
time. Therefore, while the hypothesis<br />
of a constant contamination level may<br />
suffice for occupationally exposed<br />
workers, it is less realistic for objects<br />
handled by members of the public,<br />
where the initial contamination<br />
present on the object will be affected<br />
by several removal mechanisms,<br />
which need to be considered when<br />
evaluating the annual effective dose.<br />
Based on these findings, the Dutch<br />
National Institute for Public Health<br />
and the Environment (RIVM) developed<br />
the SUDOQU (SUrface DOse<br />
QUantification) methodology [1] for<br />
the evaluation of the annual effective<br />
dose for members of the public<br />
resulting from exposure to surfacecontaminated<br />
objects. It assumes<br />
time-dependent surface- and air- contamination<br />
levels, whose evolution is<br />
governed by a system of coupled<br />
differential equations, describing the<br />
mass balance imposed by the involved<br />
mechanisms. The surface-activity concentration<br />
(Bq/cm 2 ) is considered to<br />
decrease by radioactive decay,<br />
resuspension and wipe-off (transfer<br />
of activity to the hands). The resuspended<br />
activity contributes to the<br />
(increase in) air-activity concentration<br />
(Bq/m 3 ) and can, in turn, partly re-deposit<br />
onto the object surface. The air<br />
activity concentration is further<br />
affected by radioactive decay and<br />
ventilation. Different exposure pathways<br />
are considered: external-gammaradiation<br />
exposure, inhalation, indirect<br />
ingestion and skin contamination<br />
through wipe-off. The effective dose<br />
can then be calculated as the sum of<br />
the contributions of the exposure<br />
pathways. Based on these intrinsic<br />
properties, the SUDOQU methodology<br />
is particularly attractive for clearance<br />
and exemption calculations, especially<br />
when considering public reuse<br />
scenarios, because they often involve<br />
the prolonged use of the same object.<br />
Therefore, in 2016, a collaboration<br />
was started between Bel V and RIVM,<br />
to extend the scope of the SUDOQU<br />
model, and to test its suitability for the<br />
derivation of surface-clearance levels<br />
for objects released from the controlled<br />
area of a nuclear facility.<br />
2 Objectives and<br />
methodology<br />
The results presented in this paper<br />
were obtained in the framework of a<br />
pilot project, having as main objective<br />
to investigate the applicability of the<br />
SUDOQU methodology for clearance<br />
calculations, and to gain a better<br />
understanding of the interplay among<br />
the involved mechanisms and how<br />
this affects the resulting total effective<br />
dose. This was achieved by performing<br />
deterministic calculations<br />
of the annual effective dose resulting<br />
from exposure to a typical office<br />
item, i.e. a bookcase, considering<br />
different scenarios of use and different<br />
nuclides.<br />
2.1 Reference scenario<br />
In the reference scenario (scenario 1),<br />
a bookcase is considered that leaves<br />
the controlled area of a nuclear facility<br />
with a homogeneous surface contamination<br />
of 1 Bq/cm 2 (different<br />
radionuclides are considered, as<br />
explained further in this Section).<br />
Next, the bookcase is placed in an<br />
office with a 50-m 2 area and a 2.5-m<br />
height and is used by an “average”<br />
office worker, who will be exposed to<br />
the contaminated surface. During<br />
working hours (i.e. 8 h/d, 5 d/w, and<br />
52 w/y, resulting in 2080 h/y, thus<br />
in a duty factor f exp =0.24 [1]) the<br />
worker is in the office at a distance of<br />
3 m from the contaminated bookcase,<br />
by which he incurs a certain exposure<br />
by external (gamma) radiation. The<br />
bookcase is assumed to be contaminated<br />
only on its front panel,<br />
characterised by a 6-m 2 surface.<br />
For the calculation of the externalradiation<br />
dose contribution, the<br />
conversion factor from ambient<br />
dose equivalent to effective dose<br />
(E/H*(10)) is set equal to one, which<br />
is conservative for any irradiation<br />
geometry in the photon energy range<br />
of the considered nuclides. During<br />
office hours, the worker is assumed<br />
to occasionally touch the bookcase,<br />
thereby wiping off some activity from<br />
its surface, with a frequency of<br />
approximately once every three<br />
hours (ϕ = 0.31 h -1 during use) and<br />
an efficiency of 20% (f oth = 0.2, corresponding<br />
to the ratio of the contamination<br />
level of the hands after a<br />
wipe-off event and that of the bookcase).<br />
Activity is also transferred<br />
indirectly to the face after contact<br />
with the hands. This transfer is<br />
modelled by an efficiency of f htf =0.2<br />
(ratio of contamination levels of face<br />
and hands). The individual will thus<br />
incur a skin equivalent dose following<br />
contamination of the skin area of<br />
the hands (A hands =400 cm 2 ) and<br />
of the face (A face =100 cm 2 ), which<br />
eventually also contributes to the<br />
effective dose. Furthermore, part of<br />
the activity on the hands will be<br />
transferred to the mouth (indirect<br />
ingestion): this is assumed to occur<br />
with a frequency equal to that of<br />
wipe-off (0.31 h -1 ). The activity transferred<br />
from the hands to the mouth<br />
per ingestion event is set equal to<br />
100 % (f htm =1) of the activity present<br />
<strong>atw</strong>-Special „Eurosafe<br />
2017“. In cooperation<br />
with the EUROSAFE<br />
2017 partners,<br />
Bel V (Belgium),<br />
CSN (Spain), CV REZ<br />
(Czech Republic),<br />
MTA EK (Hungary),<br />
GRS (Germany), ANVS<br />
(The Netherlands),<br />
INRNE BAS (Bulgaria),<br />
IRSN (France),<br />
NRA (Japan),<br />
JSI (Slovenia),<br />
LEI (Lithuania),<br />
PSI (Switzerland),<br />
SSM (Sweden),<br />
SEC NRS (Russia),<br />
SSTC NRS (Ukraine),<br />
VTT (Finland),<br />
VUJE (Slovakia),<br />
Wood (United<br />
Kingdom).<br />
Revised version<br />
of a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and<br />
7 November 2017.<br />
29<br />
OPERATION AND NEW BUILD<br />
Operation and New Build<br />
Clearance of Surface-contaminated Objects from the Controlled Area of a Nuclear Facility: Application of the SUDOQU Methodology ı F. Russo, C. Mommaert and T. van Dillen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
OPERATION AND NEW BUILD 30<br />
on the ingested area, but it is assumed<br />
that, per ingestion event, indirect<br />
ingestion occurs only from a limited<br />
fraction of the surface of the hands,<br />
i.e. f ing A hands , with f ing =0.01. Activity<br />
on the hands is assumed not to be<br />
affected by removal through indirect<br />
ingestion or transfer to the face<br />
( conservative approach). Moreover, it<br />
is assumed that a certain fraction of<br />
activity is re-suspended from the<br />
object surface and becomes airborne,<br />
therefore producing an effective dose<br />
contribution through inhalation. The<br />
dose conversion factors for the inhalation<br />
and ingestion pathways are<br />
those indicated in ICRP Publication 72<br />
[3] for an adult member of the public.<br />
More specifically regarding the inhalation<br />
dose, when dose conversion<br />
coefficients for different lung absorption<br />
types are available, the most<br />
conservative value is chosen.<br />
To study the role of the characteristics<br />
(type and energy) of the<br />
emitted radiation, the dose calculation<br />
for this scenario was performed<br />
for several radionuclides (βγ- or pure<br />
β-emitters) and for a radioisotopic<br />
composition typical of a nuclear<br />
power plant, as indicated in the first<br />
row of Table 1. The latter is labelled<br />
“NPP” in Table 1 and Figure 1, and<br />
corresponds to the nuclide vector of<br />
the whole site of the nuclear power<br />
plant in Doel, Belgium, in the year<br />
2015-2016. Radioactive progeny is<br />
here considered to contribute to the<br />
dose only if equilibrium can be<br />
reached within the time-integration<br />
period of one year and for sufficiently<br />
large branching ratios. For the list of<br />
considered nuclides, this is the case<br />
for Cs-137 (including Ba-137m) and<br />
Sr-90 (including Y-90).<br />
2.2 Alternative scenarios<br />
Starting from the reference scenario<br />
described in Sect. 2.1, several alternative<br />
scenarios were developed, by<br />
varying one parameter at a time. This<br />
was done to analyse the effect of<br />
separate parameter variations on<br />
the effective dose and therefore to<br />
identify the most relevant parameters,<br />
to which the results are most sensitive.<br />
This study then serves as basis for a<br />
more detailed sensitivity analysis.<br />
In this study, five alternative<br />
scenarios were developed from the<br />
reference scenario. In scenario 2, the<br />
distance of the worker to the contaminated<br />
bookcase is increased from<br />
3 m to 4.5 m. In scenario 3, wipe-off<br />
events are assumed to occur with an<br />
increased frequency of once per hour<br />
(during use), instead of once every<br />
three hours (the ingestion frequency,<br />
instead, remains unvaried with<br />
respect to the reference scenario). In<br />
scenarios 4 and 5, the transfer<br />
efficiency f oth is decreased from 0.2 to<br />
0.1 and 0.05, respectively. In scenario<br />
6, the worker benefits from six weeks<br />
of holiday, thus is only exposed during<br />
46 weeks per year. As a result, the<br />
duty factor decreases from 0.24 to<br />
0.22.<br />
3 Preliminary results<br />
The obtained results are summarised<br />
in Table 1, reporting the total annual<br />
effective dose in the six scenarios for<br />
all considered nuclides and for the<br />
NPP nuclide vector.<br />
It can be noticed from Table 1 that<br />
the dose values for the considered<br />
nuclides range from about 10 -1 µSv/y<br />
for isotopes as Ni-63 and Co-57, to<br />
values as high as 10² µSv/y for Pu-241.<br />
The dose values resulting from exposure<br />
to the NPP isotopic vector are<br />
similar to those of its most abundant<br />
radionuclide, i.e. Co-60.<br />
The (large) differences among<br />
the considered nuclides are related<br />
to the characteristics of the emitted<br />
radiation (type and energy of emitted<br />
particles), the half-life of the nuclides<br />
and the metabolic behaviour of these<br />
elements when ingested or inhaled.<br />
Note that, in general, results of a dose<br />
evaluation will also strongly depend<br />
on the type of object (geometry, surface<br />
area, distance) and how exactly it<br />
is used or handled. Effective doses<br />
presented in Table 1 for the bookcase<br />
may thus differ significantly from<br />
those for other objects released from a<br />
nuclear facility, because the relevant<br />
exposure pathways may contribute<br />
differently to the effective dose, in<br />
absolute and relative sense. Variations<br />
between nuclides may then also be<br />
different from those observed in<br />
Table 1, depending on their dominant<br />
exposure pathways. The comparison<br />
of results for several objects is currently<br />
under investigation.<br />
Furthermore, Figure 1 illustrates,<br />
for each nuclide, the relative dose,<br />
defined as the ratio of the dose in a<br />
specific scenario and the dose in the<br />
reference scenario. Elimination of the<br />
absolute differences by such normalisation<br />
enables a way to compare the<br />
relative impact of parameter changes<br />
for the considered nuclides, thus a<br />
comparison of parameter sensitivity<br />
between nuclides. It can be observed<br />
that, in most of the alternative<br />
scenarios, the variation of the dose<br />
with respect to the reference scenario<br />
is rather heterogeneous for the considered<br />
radionuclides. For example,<br />
considering scenario 2, in which the<br />
distance to the object is increased,<br />
the total dose for βγ-emitters decreases<br />
as a consequence of the reduction<br />
of the external-gamma-radiation<br />
dose, which is here the only contribution<br />
affected by (a change in)<br />
distance. The relative decrease, however,<br />
is not the same for all nuclides,<br />
as it depends on the relative contribution<br />
of the external-gamma-radiation<br />
pathway to the total dose, which<br />
differs per nuclide. Accordingly,<br />
reducing the distance with respect to<br />
the object would lead to an increase<br />
of the total dose, which is more pronounced<br />
when the external-gammaradiation<br />
exposure term is more<br />
dominant: it can be shown that,<br />
for the βγ-emitters considered here,<br />
the total dose increases by a factor<br />
between two and four when the<br />
distance is reduced to 1 m. For purebeta<br />
emitters, in which the externalgamma-radiation<br />
component is absent,<br />
the dose is not affected by a<br />
variation of the distance. A certain<br />
dose contribution could result from<br />
external-beta radiation, but is not<br />
considered here. In scenario 3, in<br />
Scen. Na-22 Mn-54 Co-56 Co-57 Co-58 Co-60 Zn-65 Cs-134 Cs-137 Eu-152 Ni-63 Sr-90 Pu-241 NPP<br />
1 3.86 0.97 1.87 0.21 0.56 5.18 1.38 8.05 6.91 3.65 0.10 17.04 87.64 4.20<br />
2 2.63 0.58 1.11 0.14 0.34 3.77 1.14 7.15 6.53 2.91 0.10 17.04 87.64 3.30<br />
3 2.07 0.55 1.31 0.12 0.40 2.73 0.81 4.35 3.59 1.90 0.05 8.85 45.66 2.21<br />
4 3.71 1.03 1.87 0.21 0.54 5.38 1.10 6.21 5.48 3.98 0.09 13.94 104.59 4.11<br />
5 3.58 1.06 1.87 0.21 0.53 5.46 0.92 4.96 4.49 4.16 0.08 11.76 114.91 4.01<br />
6 3.58 0.90 1.70 0.19 0.50 4.82 1.27 7.47 6.43 3.40 0.09 15.86 81.55 3.90<br />
| | Tab. 1.<br />
Total annual effective dose [µSv/y] for all the considered nuclides in the six scenarios (see Sect. 2.1 and 2.2) for the contaminated bookcase.<br />
Operation and New Build<br />
Clearance of Surface-contaminated Objects from the Controlled Area of a Nuclear Facility: Application of the SUDOQU Methodology ı F. Russo, C. Mommaert and T. van Dillen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
| | Fig. 1.<br />
Variation of the total dose values in the six analysed scenarios for the bookcase with respect to the reference scenario (i.e. scenario 1).<br />
which the wipe-off frequency is increased,<br />
a decrease of the total dose is<br />
observed for all nuclides. This can be<br />
attributed to a more rapid removal<br />
from the surface, which leads to a<br />
reduction of the time-integrated<br />
surface- and air-contamination levels,<br />
and thus to a decrease of all dose<br />
contributions. This decrease is of<br />
course larger when wipe-off is a more<br />
dominant mechanism for removal of<br />
surface activity. As a result, in the case<br />
of long-lived radionuclides, for which<br />
radioactive decay does not constitute<br />
a competing removal mechanism, the<br />
wipe-off process will have a larger<br />
relative contribution, and the final<br />
result will be more sensitive to a variation<br />
in this mechanism: such radionuclides<br />
will, therefore, show a larger<br />
decrease than shorter-lived nuclides<br />
as Co-56 and Co-58. In scenarios 4<br />
and 5, a decrease in the transfer<br />
efficiency has two opposite effects. On<br />
the one hand, activity residing on the<br />
object surface will be removed at a<br />
slower rate, leading to an increase of<br />
the time-integrated surface-contamination<br />
level (TISC). As a result, more<br />
activity is available for resuspension,<br />
thus the time-integrated air-contamination<br />
level (TIAC) also increases.<br />
Since the external-gamma-radiation<br />
dose is proportional to TISC and the<br />
committed effective dose from inhalation<br />
is proportional to TIAC, both dose<br />
contributions increase with respect to<br />
the reference scenario. On the other<br />
hand, the effective-dose contributions<br />
from indirect ingestion and skin contamination<br />
are both proportional to<br />
the product f oth TISC (f oth decreases,<br />
TISC increases). For the assumptions<br />
made here, the product f oth TISC<br />
decreases, thus the latter dose contributions<br />
decrease. Altogether, the<br />
total annual effective dose is the result<br />
of the balance between the opposite<br />
trends of these considered dose<br />
contributions. For some nuclides (e.g.<br />
Co-60, Mn-54, Pu-241, and Eu-152)<br />
the total dose increases as a result of<br />
the increase of the external-radiation<br />
exposure or inhalation contribution<br />
(or a combination of both). For other<br />
nuclides (e.g. Cs-137, Cs-134, Zn-65,<br />
Sr-90) the total dose follows the<br />
decreasing trend of its leading contribution,<br />
i.e. ingestion. In other cases<br />
(Co-56 and Co-57), the total dose<br />
marginally changes, due to the fact<br />
that the opposite effects approximately<br />
cancel each other out. Finally,<br />
in scenario 6, a decrease of the exposure<br />
duration leads to an (approximately)<br />
identical decrease in the total<br />
dose for all nuclides (the relative<br />
values in this scenario range between<br />
0.90 and 0.95).<br />
3.1 Benchmarking study<br />
The results obtained with SUDOQU<br />
were compared to the results obtained<br />
with the model described in RP101<br />
[2]. A graphical illustration of this<br />
comparison is provided in Figure 3.2.<br />
The RP101-model was chosen for the<br />
benchmarking study because one of<br />
the scenarios studied in RP101 considers<br />
a surface-contaminated tool<br />
cabinet, which is comparable to the<br />
bookcase considered in this paper.<br />
Moreover, like SUDOQU, the RP101-<br />
model assumes a non-constant surface<br />
activity. However, a fundamental<br />
difference between the two models<br />
is that the RP101-model only considers<br />
radioactive decay as a removal<br />
mechanism, whereas the SUDOQU<br />
model considers other processes<br />
affecting the evolution of the contamination<br />
level (Sect. 1). Another<br />
important difference concerns the<br />
removability of surface contamination:<br />
in SUDOQU, 100 % of the surface<br />
activity is assumed to be remov able,<br />
with a transfer efficiency of 20 %; in<br />
RP101, only 10 % of the total surface<br />
activity is removable, and the transfer<br />
efficiency is equal to 10 %. These<br />
differences lead to dissimilar (relative)<br />
contributions of the exposure<br />
pathways in the two models.<br />
In this study, parameter values<br />
defining the exposure geometry and<br />
duration in SUDOQU were harmonised<br />
with those in RP101. In this way,<br />
differences in dose results between<br />
the two models are only related to<br />
differences in model construction<br />
and the (remaining) underlying<br />
assumptions.<br />
As a first step of the benchmarking<br />
study, values of the remaining parameters<br />
were left unvaried in SUDOQU<br />
(i.e. values from Sect. 2.1), with the<br />
aim of comparing the two models<br />
based on their main, default assumptions<br />
and to investigate their impact<br />
on the results. The assumption in<br />
RP101 that only 10 % of the total<br />
surface activity is removable enhances<br />
the dose contribution from externalgamma-radiation<br />
exposure, as the<br />
remaining 90 % of the surface activity<br />
contributes exclusively to this pathway,<br />
while only being modified by<br />
radioactive decay. On the other hand,<br />
the contribution of the other exposure<br />
pathways, related to activity removal<br />
from the surface (resuspension and<br />
wipe-off), will be reduced in RP101<br />
with respect to those in SUDOQU, for<br />
which 100 % of the surface activity is<br />
removable and may thus contribute to<br />
these pathways (inhalation, ingestion<br />
and skin contamination). Again, the<br />
net outcome depends on the balance<br />
OPERATION AND NEW BUILD 31<br />
Operation and New Build<br />
Clearance of Surface-contaminated Objects from the Controlled Area of a Nuclear Facility: Application of the SUDOQU Methodology ı F. Russo, C. Mommaert and T. van Dillen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
OPERATION AND NEW BUILD 32<br />
| | Fig. 2.<br />
Comparison of the total annual effective dose obtained for several radionuclides with the model in RP101 [2] and with SUDOQU. Values labelled as SUDOQU*<br />
are obtained by applying the same removable fraction and wipe-off efficiency as those used in RP101.<br />
between these opposite effects. For<br />
Co-60 and Na-22, the smaller value<br />
of the external-radiation dose in<br />
SUDOQU (with respect to that in<br />
RP101) is not fully compensated by<br />
the larger values of the other dose<br />
contributions, leading to a slightly<br />
smaller total dose in SUDOQU. For<br />
the other considered nuclides ( Cs-137,<br />
Sr-90 and Pu-241), the opposite<br />
occurs, leading to more conservative<br />
results in SUDOQU.<br />
An additional comparison was<br />
made by implementing in SUDOQU<br />
the same assumptions as in RP101<br />
concerning the removable fraction<br />
and the transfer efficiency. These<br />
results are shown in Figure 2 as well,<br />
indicated by the label SUDOQU*.<br />
Due to these assumptions, the<br />
external-gamma-radiation exposure<br />
in SUDOQU* now increases to values<br />
larger than those in RP101, while the<br />
other dose contributions decrease,<br />
although still being larger than the<br />
values obtained in RP101. As a result,<br />
the annual dose values obtained with<br />
SUDOQU* are more conservative for<br />
all considered nuclides, but in good<br />
agreement with the RP101-results.<br />
4 Conclusions<br />
The SUDOQU model [1] enables dose<br />
evaluations for exposure to a surfacecontaminated<br />
object. It is characterised<br />
by the innovative and distinctive<br />
assumption of time-dependent<br />
surface- and indoor air-contamination<br />
levels governed by mass-balance<br />
equations based on the following<br />
mechanisms: radioactive decay,<br />
resuspension, wipe-off, deposition<br />
and ventilation. These features make<br />
the SUDOQU methodology a suitable<br />
candidate for performing clearance<br />
calculations based on reuse scenarios,<br />
where the individual is likely to be<br />
exposed to the same object throughout<br />
the year, and for which the<br />
assumption of constant contamination<br />
levels would be unrealistically<br />
conservative. In this work, a surfacecontaminated<br />
bookcase released from<br />
the controlled area of a nuclear facility<br />
is studied, with the aim of assessing<br />
the applicability of SUDOQU for<br />
the development of surface-clearance<br />
criteria for nuclear facilities. Deterministic<br />
calculations of the annual<br />
effective dose were thus conducted for<br />
several nuclides in different scenarios<br />
of use. First, the results in this paper<br />
reveal a strong nuclide dependency:<br />
even within the same category of<br />
emitters there can be pronounced<br />
differences in absolute dose values,<br />
depending on the radiological characteristics<br />
of the nuclides and their metabolic<br />
behaviour and radiobiological<br />
impact on the human body. Moreover,<br />
the consideration of a mass balance<br />
describing the time evolution of the<br />
contamination levels causes the total<br />
annual dose to be the result of<br />
a delicate interplay of the involved<br />
elements. In this way, a variation of a<br />
certain input parameter may lead to<br />
opposite effects on the various dose<br />
contributions, and thus to a total dose<br />
that either decreases, increases or<br />
remains constant. The net outcome<br />
again depends on the characteristics<br />
of the nuclide and on the specifics of<br />
the exposure scenario. The results<br />
obtained with SUDOQU were benchmarked<br />
against the results reported in<br />
RP101 [2] for the reuse scenario of a<br />
tool cabinet, and the two models<br />
proved to be in good agreement.<br />
The results presented in this paper<br />
not only demonstrate the suitability of<br />
SUDOQU for dose assessments related<br />
to clearance of objects from nuclear<br />
facilities, but they are also a good<br />
starting point to better understand<br />
the intricate interplay among the<br />
involved mechanisms. Their interaction<br />
also disclosed the importance and<br />
difficulty of a detailed sensitivity<br />
analysis. Future work will focus on the<br />
development of surface clearance<br />
levels based on probabilistic and<br />
realistically conservative dose assessments.<br />
References<br />
[1] T. van Dillen, SUDOQU: a new dose<br />
model to derive criteria for surface<br />
contamination of non-food (consumer)<br />
goods, containers and conveyances,<br />
Radiation Protection Dosimetry,<br />
164(1-2) (2015), pp. 160-164.<br />
[2] Radiation Protection 101: Basis for the<br />
definition of surface contamination<br />
clearance levels for the recycling or<br />
reuse of metals arising from<br />
dismantling of nuclear installations,<br />
European Commission, 1998.<br />
[3] ICRP, Age-dependent Doses to the<br />
Members of the Public from Intake of<br />
Radionuclides - Part 5 Compilation of<br />
Ingestion and Inhalation Coefficients,<br />
ICRP Publication 72. Ann. ICRP 26 (1),<br />
1995.<br />
Authors<br />
F. Russo<br />
C. Mommaert<br />
Bel V<br />
Rue Walcourt, 148<br />
1070 Brussels,<br />
Belgium<br />
T. van Dillen<br />
National Institute for Public Health<br />
and the Environment (RIVM)<br />
P.O. Box 1<br />
3720 BA Bilthoven,<br />
The Netherlands<br />
Operation and New Build<br />
Clearance of Surface-contaminated Objects from the Controlled Area of a Nuclear Facility: Application of the SUDOQU Methodology ı F. Russo, C. Mommaert and T. van Dillen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
34<br />
DECOMMISSIONING AND WASTE MANAGEMENT<br />
<strong>atw</strong>-Special „Eurosafe<br />
2017“. In cooperation<br />
with the EUROSAFE<br />
2017 partners,<br />
Bel V (Belgium),<br />
CSN (Spain), CV REZ<br />
(Czech Republic),<br />
MTA EK (Hungary),<br />
GRS (Germany), ANVS<br />
(The Netherlands),<br />
INRNE BAS (Bulgaria),<br />
IRSN (France),<br />
NRA (Japan),<br />
JSI (Slovenia),<br />
LEI (Lithuania),<br />
PSI (Switzerland),<br />
SSM (Sweden),<br />
SEC NRS (Russia),<br />
SSTC NRS (Ukraine),<br />
VTT (Finland),<br />
VUJE (Slovakia),<br />
Wood (United<br />
Kingdom).<br />
Revised version<br />
of a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and<br />
7 November 2017.<br />
Carbon-14 Speciation During<br />
Anoxic Corrosion of Activated Steel<br />
in a Repository Environment<br />
E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat<br />
1 Introduction Carbon-14 is an important radionuclide in the inventory of radioactive waste [1,2] and, due its<br />
long half-life (5730 y), it has been identified a key radionuclide in safety assessments [3,4]. 14 C is of specific concern due<br />
to its potential presence as either dissolved or gaseous species in the disposal facility and the host rock, the high mobility<br />
of dissolved carbon compounds in the geosphere caused by weak interaction with mineral surfaces in near neutral<br />
conditions, and eventually because it can be incorporated in the human food chain. Current safety assessments are<br />
based on specific assumptions regarding the rate of 14 C release from potential sources, the 14 C speciation upon release<br />
and the mobility of the different chemical forms of 14 C in the cementitious near field and the host rock [1].<br />
The main source of 14 C in L/ILW in<br />
Switzerland are activated metallic<br />
nuclear fuel components and reactor<br />
core components as well as spent<br />
filters and ion exchange resins used in<br />
light water reactors (LWR) for the<br />
removal of radioactive contaminants<br />
in a number of liquid processes and<br />
waste streams. Compilations of the<br />
activity inventories revealed that<br />
in the already existing and future<br />
arisings of radioactive waste in<br />
Switzerland, the 14 C inventory is<br />
mainly associated with activated<br />
(or irradiated, respectively) steel<br />
(~85 %) while the 14 C inventories<br />
associated with nuclear fuel components<br />
(e.g. Zircaloy) and wastes<br />
from the treatment of reactor coolants<br />
(e.g. spent ion exchange resins)<br />
are much less. 14 C in activated steel<br />
results mainly from 14 N activation<br />
( 14 N(n,p) 14 C) [2]. Release of 14 C<br />
occurs during anoxic corrosion of<br />
activated steel in the cementitious<br />
near field of the L/ILW repository.<br />
Recent reviews of corrosion rates<br />
suggest that steel corrosion in these<br />
conditions is a very slow process [5,6].<br />
Carbon-14 can be released in a<br />
variety of organic and inorganic<br />
chemical forms. 14 C will decay within<br />
a disposal facility if the 14 C-bearing<br />
compounds are retained by interaction<br />
with the materials of the<br />
engineered barrier. For example,<br />
inorganic carbon, i.e. 14 CO 2 and its<br />
bases, is expected to precipitate as<br />
calcium carbonate within a cementbased<br />
repository or undergo 14 CO 3<br />
2-<br />
isotopic exchange with carbonate<br />
minerals. For this reason inorganic 14 C<br />
has only a negligible impact on the<br />
14 C-based dose release. By contrast,<br />
gaseous species containing 14 C, such<br />
as 14 CH 4 , 14 CO etc., could form and<br />
migrate with bulk gas from the near<br />
field into the host rock. It is indicated<br />
from previous studies that a limited<br />
number of small organic molecules<br />
| | Fig. 1.<br />
Schematic presentation of the design of the corrosion experiment. Reactor set-up for the corrosion experiment with activated steel<br />
(top); analytical procedures for the detection of 14C-bearing dissolved organic compounds (bottom) and gaseous species (right).<br />
are likely to be formed in the course<br />
of the anoxic corrosion of activated<br />
steel in alkaline conditions, in particular<br />
reduced hydrocarbons, such<br />
as methane, ethane etc., and oxidized<br />
hydrocarbons, such as alcohols,<br />
aldehydes and carboxylic acids [7].<br />
It is to be noted that both oxidized<br />
and reduced hydrocarbons have been<br />
observed in anoxic iron-water systems<br />
in anoxic (near neutral to alkaline)<br />
conditions which seems to be inconsistent<br />
with a view to the negative<br />
redox potential associated with the<br />
systems [8].<br />
Although the 14 C inventory associated<br />
with activated steel is well<br />
known, our understanding of the<br />
chemical form of the 14 C-bearing<br />
compounds produced in the course of<br />
the anoxic corrosion of activated steel<br />
is limited. The present study is aimed<br />
to fill this knowledge gap.<br />
2 Corrosion study<br />
with activated steel<br />
The schematic presentation of the<br />
experimental design is displayed in<br />
Figure 1 which includes a reactor<br />
system to perform the corrosion<br />
experiment with activated steel and<br />
analytical methods for the identification<br />
and quantification of the<br />
14 C-bearing compounds in the liquid<br />
and gas phases. The corrosion study<br />
was supposed to be carried out using<br />
steel components exposed to neutron<br />
flux in a Swiss nuclear power plant<br />
(NPP). To this end five irradiated steel<br />
guide-tube nuts were retrieved from<br />
the storage pool of NPP Gösgen during<br />
the annual maintenance work in<br />
2012 and transferred to the PSI<br />
hotlaboratory. The nuts had been<br />
positioned at the bottom end of<br />
fuel rods and exposed to a thermal<br />
neutron flux for ~2 years. Each nut<br />
weighed ~5 g and had a contact dose<br />
Decommissioning and Waste Management<br />
Carbon-14 Speciation During Anoxic Corrosion of Activated Steel in a Repository Environment<br />
ı E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
rate ~150 mSv/h (predominantly<br />
caused by 60 Co). When planning<br />
the corrosion experiment several<br />
constraints had to be taken into<br />
consideration: The low 14 C inventory<br />
of the activated steel samples<br />
(~18 kBq/g) [10] in combination<br />
with the fact that only a small amount<br />
of activated steel could be used in a<br />
corrosion experiment outside a hot<br />
cell due to the high dose rate of the<br />
material and the very slow corrosion<br />
of stainless steel in alkaline conditions<br />
(typically
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
DECOMMISSIONING AND WASTE MANAGEMENT 36<br />
A B C<br />
| | Fig. 3.<br />
A) Sketch of the reactor, B) picture of the lead shielding with door and<br />
C) the sampling system for liquid and gaseous samples placed outside the lead shielding.<br />
a naturally occurring radionuclide<br />
produced in the upper atmosphere<br />
and present in the chemical form<br />
14 CO 2 (activity of 1 m 3 air ~53 mBq).<br />
Furthermore, alkaline solutions are<br />
commonly known as a sink for CO 2<br />
and therefore for 14 CO 2 . Hence, the<br />
14 C background concentration accumulated<br />
in the course of the corrosion<br />
experiment with activated steel could<br />
be affected by an undesirable uptake<br />
of 14 C from the atmosphere in any<br />
stage of sample preparation and<br />
handling. The average 14 C background<br />
was determined to be F 14 C =<br />
0.06 ± 0.02 (F 14 C = fraction modern)<br />
in samples collected after high performance<br />
ion exchange chromatography<br />
(HPIEC) and using pre-cleaned plastic<br />
vials for injection and collection. This<br />
value is about an order of a magnitude<br />
higher than background values<br />
achieved in radio carbon dating.<br />
Sample preparation for compoundspecific<br />
14 C AMS method involves<br />
va rious dilution processes during<br />
chromatographic separation of single<br />
compounds that had to be considered<br />
adequately in order to reach the target<br />
dynamic range of the AMS (Figure 2).<br />
The analytical protocol required<br />
that dilution of the samples by a<br />
factor 1:25 and 1:50 occurred in the<br />
course of the separation by HPIEC.<br />
Tests measurements carried out at<br />
increasing concentrations of 14 C-<br />
labelled carboxylic acids standards<br />
allowed the dynamic range of the<br />
AMS-based analytical method to be<br />
determined (~0.06 - ~50 F 14 C).<br />
Recovery of the compound-specific<br />
14 C AMS method was determined<br />
using four different 14 C-labelled<br />
carboxylic acids ( 14 C-acetic acid,<br />
14 C-formic acid, 14 C-malonic acid and<br />
14 C-oxalic acid) dissolved in either<br />
deionized, decarbonated water (ultrapure<br />
water generated by Millipore<br />
Gradient A10 water purification<br />
system) or in ACW (pH 12.5). The<br />
samples were sequentially injected<br />
into the HPIEC system as single compounds.<br />
The corresponding fractions<br />
of the 14 C-labelled carboxylic acids<br />
were collected and analyzed by AMS<br />
[11, 12]. Recoveries (%) were determined<br />
using single compounds and<br />
mixtures of the compounds. In all<br />
cases recovery was found to be close<br />
to 100 % (97 ±17 %) [12].<br />
Corrosion studies with unirradiated<br />
iron powders revealed that<br />
volatile organic compounds, such as<br />
alkanes, alkenes, alcohols, aldehydes,<br />
are also formed during iron corrosion<br />
[9] which requires the development of<br />
a compound-specific 14 C AMS<br />
analytical method for 14 C-bearing<br />
v olatile species. The analytical<br />
approach is currently being developed<br />
in a way similar to that previously<br />
elaborated for dissolved organic<br />
compounds and is based on gas<br />
chromatographic (GC) separation of<br />
single compounds in combination<br />
with 14 C detection by AMS. To this<br />
end, the GC system has to be coupled<br />
directly to a combustion reactor and a<br />
fraction sampling system for 14 CO 2<br />
(Figure 1). Coupling of the three<br />
devices, i.e. GC, com-bustion reactor<br />
and fraction collector, is still under<br />
development.<br />
2.4 Development of the<br />
corrosion reactor<br />
The experimental set-up for the longterm<br />
corrosion experiment with the<br />
activated steel nut specimens consists<br />
of a custom-made gas-tight over pressure<br />
reactor placed within a 10 cm<br />
thick lead shielding (Figure 3). For the<br />
experiments two activated steel nut<br />
segments of ~1 g each were immersed<br />
in 300 mL ACW (pH 12.5) under a N 2<br />
atmosphere (200 mL). The reactor is<br />
equipped with a digital pressure transmitter,<br />
a temperature sensor and a<br />
sensor to detect dissolved oxygen<br />
( Visiferm DO Arc, Hamilton, USA).<br />
The overpressure reactor is designed in<br />
such a way that all mani pulations<br />
necessary for regular sampling can be<br />
carried out outside the lead shielding<br />
to minimize exposure of the experimentalist<br />
to radiation. Leak tests<br />
confirmed gas-tightness of the reactor.<br />
2.5 Start of the corrosion<br />
experiment<br />
The corrosion experiment with the<br />
activated steel nut segments was<br />
started in May 2016. Results from the<br />
first few samplings are exemplarily<br />
listed in Table 1. They show an<br />
increase in the activity of total organic<br />
14 C (TO 14 C) with time, thus indicating<br />
progressing corrosion. At present,<br />
however, identification and quantification<br />
of the individual 14 C-bearing<br />
organic compounds by compoundspecific<br />
14 C AMS is not yet possible<br />
because their concentration is still<br />
below the detection limit of the<br />
compound-specific 14 C AMS method.<br />
As a consequence, the analytical<br />
methodology is currently further<br />
improved by developing a procedure<br />
that allows pre-concentration of<br />
the liquid samples collected by the<br />
fraction collector.<br />
Time TO 14 C TOC Hydrocarbons [µM] Carboxylic acids [µM]<br />
[d] [F 14 C] [Bq/L] [ppm] Methane Ethane Ethene Foramte Acetate Oxalate Gycolate Lactate<br />
0 0.00 0.00 - - - < 5 n.d. < 0.1 n.d. n.d.<br />
1 0.10 0.04 - n.d. n.d. n.d. 7 n.d. 0.3 0.4 n.d.<br />
15 0.99 0.45 2.44 n.d. n.d. n.d. 8 n.d. 0.5 1.3 1.6<br />
29 1.56 0.70 2.60 n.d. n.d. n.d. 7 n.d. 0.5 1.4 1.2<br />
93 3.53 1.60 4.67 0.42 n.d. n.d. 13 n.d. 0.7 1.7 2.8<br />
| | Tab. 1.<br />
Compilation of the first results from the corrosion study with activated steel.<br />
Decommissioning and Waste Management<br />
Carbon-14 Speciation During Anoxic Corrosion of Activated Steel in a Repository Environment<br />
ı E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
The total concentration of organic<br />
carbon ( 12 C + 14 C), i.e. TOC, also<br />
tends to increase with time. Note that<br />
TOC accounts for the total concentration<br />
of organic compounds, that<br />
is, 12 C-bearing and 14 C-bearing compounds.<br />
However, the concentration<br />
of the 14 C-bearing compounds is<br />
orders of magnitudes lower than that<br />
of the corresponding 12 C-bearing<br />
counter parts. The concentration of<br />
the analysed 12 C-bearing individual<br />
compounds, i.e. hydrocarbons and<br />
carboxylic acids (Table 1), is still<br />
below (n.d.) or close to the detection<br />
limits of the analytical techniques<br />
(GC-MS and HPIEC-MS, respectively).<br />
Note that the latter analytical techniques,<br />
i.e. GC-MS and HPIEC-MS,<br />
can be used to detect separately<br />
both 12 C-bearing and 14C-bearing<br />
species of the same kind based on<br />
their differences in the mass of carbon.<br />
Again, the concentration of the 14 C-<br />
bearing compounds is orders of<br />
magnitudes lower than that of the<br />
corresponding<br />
12 C-bearing counterparts.<br />
Thus, the concentrations of the<br />
hydrocarbons and carboxylic acids<br />
listed in Table 1 correspond to those<br />
of the respective 12 C-bearing organic<br />
compounds. The first results clearly<br />
support the need of a very sensitive<br />
AMS-based analytical method for<br />
the detection of both volatile and<br />
dissolved 14 C-bearing carbon species,<br />
i.e. compound-specific 14 C AMS.<br />
3 Summary<br />
Our current understanding of the type<br />
of 14 C-bearing species produced<br />
during anoxic corrosion of activated<br />
metals is very limited. This information,<br />
however, is required in conjunction<br />
with safety assessment of<br />
nuclear waste repositories containing<br />
activated metals (e.g. activated steel,<br />
Zircaloy) as waste materials. A unique<br />
corrosion experiment with activated<br />
steel from NPP Gösgen, Switzerland,<br />
is currently being carried out with the<br />
aim of identifying and quantifying the<br />
14 C-bearing carbon species produced<br />
in the course of the corrosion process<br />
under hyper-alkaline, anoxic conditions.<br />
A specific analytical technique<br />
was developed by combining chro matographic<br />
separation of 14 C-bearing<br />
individual compounds with 14 C detection<br />
by AMS (compound-specific 14 C<br />
AMS). This approach was chosen<br />
because the concentrations of these<br />
compounds was expected to be extremly<br />
low due to low amount of<br />
activated steel that could be used in<br />
the experiment, the low corrosion rate<br />
of steel in hyper-alkaline conditions<br />
and the low 14 C inventory determined<br />
for activated steel. The compoundspecific<br />
14 C AMS method is characterized<br />
by a low 14 C detection limit<br />
and a large dynamic range (~3 orders<br />
of a magnitude) and therefore it is<br />
well suited for application in the corrosion<br />
experiment with activated<br />
steel. The method was developed for<br />
selected, potentially 14 C-bearing compounds<br />
of interest as previous studies<br />
with unirradiated iron have shown<br />
that only a limited number of carbon<br />
species are formed during corrosion.<br />
The specific set-up developed for<br />
the corrosion experiment with activated<br />
steel allows continuous monitoring<br />
of important physico-chemical<br />
parameters (pressure, temperature,<br />
dissolved oxygen) and further allows<br />
sampling of liquid and gas phase from<br />
the reactor to be conducted outside<br />
the lead shielding. Analysis of the<br />
liquid and gas phases from the first<br />
sampling campaigns show that the<br />
concentrations of the individual<br />
organic compounds ( 12 C- and 14 C-<br />
bearing) are still very low, i.e. below<br />
or close to the detection limit of the<br />
analytical methods used in this study.<br />
Nevertheless, the total organic 14 C<br />
content increases with time, indicating<br />
progressing corrosion. This<br />
increase in TO 14 C is slow in line with<br />
the very slow corrosion of steel in<br />
alkaline media. The analytical method<br />
will be developed further to identify<br />
and quantify the 14 C-bearing single<br />
compounds in future samplings.<br />
Acknowledgement<br />
We thank NPP Gösgen for providing<br />
the irradiated steel nuts and<br />
Ines Günther- Leopold (PSI), Matthias<br />
Martin (PSI) and Robin Grabherr (PSI)<br />
for sample preparation. Partial<br />
funding for this project was provided<br />
by swissnuclear and the National<br />
Cooperative for the Disposal of<br />
Radioactive Waste (Nagra), Switzerland.<br />
The project has received funding<br />
from the European Union's European<br />
Atomic Energy Community's ( Euratom)<br />
Seventh Framework Programme FP7/<br />
2007-2013 under grant agreement<br />
no. 604779, the CAST project.<br />
References<br />
[1] L. Johnson and B. Schwyn, 2008.<br />
Proceedings of a Nagra/RWMC workshop<br />
on the release and transport of<br />
C-14 in repository environments, Nagra<br />
Working Report NAB 08-22, Nagra,<br />
Wettingen, Switzerland.<br />
[2] M.-S. Yim and F. Caron, 2006. Life cycle<br />
and management of carbon-14 from<br />
nuclear power generation, Prog. Nucl.<br />
Energ. 48, 2-36.<br />
[3] Nagra, 2002. Project Opalinus Clay:<br />
Safety Report. Demonstration of<br />
Disposal Feasibility for Spent fuel,<br />
Vitrified High-level Waste and Longlived<br />
Intermediate-level Waste<br />
(Entsorgungsnachweis), Nagra<br />
Technical Report NTB 02-05, Nagra,<br />
Wettingen, Switzerland.<br />
[4] Nuclear Decommissioning Authority,<br />
2012. Geological Disposal. Carbon-14<br />
Project - Phase 1 Report,<br />
NDA/RWMD/092, United Kingdom.<br />
[5] N.R. Smart et al., 2004. The Anaerobic<br />
Corrosion of Carbon and Stainless Steel<br />
in Simulated Cementitious Repository<br />
Environments: A Summary Review of<br />
Nirex Research. AEAT/ERRA-0313, AEA<br />
Technology, Harwell, United Kingdom.<br />
[6] N. Diomidis, 2014. Scientific Basis for<br />
the Production of Gas due to Corrosion<br />
in a Deep Geological Repository, Nagra<br />
Working Report NAB 14-21, Nagra,<br />
Wettingen, Switzerland.<br />
[7] E. Wieland and W. Hummel, 2015.<br />
Formation and stability of carbon-14<br />
containing organic compounds in<br />
alkaline iron-water-systems: Preliminary<br />
assessment based on a literature survey<br />
and thermodynamic modelling,<br />
Mineral. Mag. 79, 1275-1286.<br />
[8] D. B. Vance, 1996. Redox reactions in<br />
remediation, Environ. Technol. 6, 24-25.<br />
[9] B. Cvetković et al., 2017. Formation of<br />
low molecular weight organic<br />
compounds during anoxic corrosion of<br />
zero-valent iron in alkaline conditions.<br />
Environm. Eng. Sci. (accepted).<br />
[10] D. Schumann et al., 2014.<br />
Determination of the 14 C content in<br />
activated steel components from a<br />
neutron spallation source and a nuclear<br />
power plant. Anal. Chem. 86,<br />
5448-5454.<br />
[11] S. Szidat et al., 2014. 14 C analysis and<br />
sample preparation at the new Bern<br />
laboratory for the analysis of<br />
radiocarbon with AMS (LARA).<br />
Radiocarbon 56, 561-566.<br />
[12] B. Cvetković et al., 2017. Analysis of<br />
carbon-14 containing corrosion<br />
products released from activated steel<br />
by accelerator mass spectrometry.<br />
Analyst (in prep.).<br />
Authors<br />
E. Wieland<br />
B.Z. Cvetković<br />
D. Kunz<br />
Paul Scherrer Institut<br />
Laboratory for Waste Management<br />
5232 Villigen PSI, Switzerland<br />
G. Salazar<br />
S. Szidat<br />
University of Bern<br />
Department of Chemistry and<br />
Biochemistry & Oeschger Centre<br />
for Climate Change Research<br />
3012 Bern, Switzerland<br />
DECOMMISSIONING AND WASTE MANAGEMENT 37<br />
Decommissioning and Waste Management<br />
Carbon-14 Speciation During Anoxic Corrosion of Activated Steel in a Repository Environment ı E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
38<br />
FUEL<br />
<strong>atw</strong>-Special „Eurosafe<br />
2017“. In cooperation<br />
with the EUROSAFE<br />
2017 partners,<br />
Bel V (Belgium),<br />
CSN (Spain), CV REZ<br />
(Czech Republic),<br />
MTA EK (Hungary),<br />
GRS (Germany), ANVS<br />
(The Netherlands),<br />
INRNE BAS (Bulgaria),<br />
IRSN (France),<br />
NRA (Japan),<br />
JSI (Slovenia),<br />
LEI (Lithuania),<br />
PSI (Switzerland),<br />
SSM (Sweden),<br />
SEC NRS (Russia),<br />
SSTC NRS (Ukraine),<br />
VTT (Finland),<br />
VUJE (Slovakia),<br />
Wood (United<br />
Kingdom).<br />
Revised version<br />
of a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and<br />
7 November 2017.<br />
1) Reactivity control<br />
is ensured notably<br />
by the motion of<br />
rod cluster control<br />
assemblies<br />
requiring not to<br />
exceed a limited<br />
fuel assembly<br />
deformation.<br />
2) Core coolability<br />
requires not to<br />
exceed a limited<br />
deformation of the<br />
fuel rods geometry.<br />
3) Fission products<br />
containment is<br />
primarily ensured<br />
by the first barrier<br />
integrity.<br />
4) M5 is the reference<br />
alloy designed by<br />
AREVA while ZIRLO<br />
and Optimized<br />
ZIRLO are Westinghouse’s<br />
alloys (the<br />
historical Zircaloy-4<br />
cladding is no<br />
longer loaded in<br />
EDF’s reactors since<br />
the end of 2016).<br />
5) Zr + 2H 2 O → ZrO 2 +<br />
2H 2<br />
Review of Fuel Safety Criteria in France<br />
Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois<br />
1 Background Depending on design basis condition of Pressurized Water Reactors (PWRs), the safety<br />
objective is either preventing or mitigating the release of fission products and other contaminants to the environment.<br />
Fuel is involved in each of the three reactor safety functions: reactivity control 1 , core coolability 2 and fission products<br />
containment 3 . A main issue in the safety demonstration for the French PWRs is to respect the objectives related to the<br />
barriers behavior, depending on Plant Condition Category (PCC) divided into four categories: normal operation<br />
( PCC-1), incident transients (PCC-2), moderate frequency accident transients (PCC-3) and hypothetical accident<br />
transients (PCC-4).<br />
The objectives associated with the<br />
first barrier are the following:<br />
• for PCC-1 and PCC-2, the fuel rods<br />
must remain intact;<br />
• for PCC-3 and PCC-4, although<br />
fuel rod integrity may be lost, the<br />
number of damaged fuel rods<br />
must be limited (in PCC-4, a more<br />
extensive number of damaged<br />
fuel rods is allowed than in PCC-3)<br />
and the geometrical structure of<br />
the core must not be damaged in<br />
order to ensure an adequate core<br />
coolability.<br />
For each category and transients type,<br />
these objectives are then expressed as<br />
requirements associated to the limitative<br />
physical phenomena occurring<br />
during PCC. Afterwards, the requirements<br />
are supported by fuel safety<br />
criteria that are limit values on computable<br />
metrics representative of the<br />
relevant physical phenomena. These<br />
limit values are determined by experiments<br />
intended to be representative of<br />
situations encountered in PCC.<br />
In France, the fuel safety criteria<br />
(and notably their limit values) came<br />
in the 1970s from Westinghouse’s<br />
license. At that time state-of-the-art<br />
and computing capacities lead to<br />
establish decoupling criteria enabling<br />
to implement simplified and robust<br />
approaches to analyze the more complex<br />
and severe accidental conditions.<br />
For instance, to maintain core coolability,<br />
requirements may be based on<br />
either fuel rod cladding integrity or<br />
the absence of fuel dispersal in<br />
the primary coolant. Indeed, such<br />
requirements avoid notably studying<br />
the impact of hot or melt fuel interaction<br />
with water on core coolability.<br />
Since the French nuclear program<br />
was initiated, both operating experience,<br />
experiments carried out by<br />
operators and research institutes as<br />
well as international R&D programs,<br />
which aim at improving computation<br />
methodologies, have allowed continuous<br />
progress in knowledge and<br />
in collecting experimental results,<br />
especially in RIA (Reactivity-Initiated<br />
accident) and LOCA (Loss-of-Coolant<br />
Accident) conditions. Moreover, new<br />
cladding alloys characterized by<br />
enhanced performances, especially<br />
regarding cladding corrosion during<br />
operating conditions, have been<br />
introduced in French PWRs (such as<br />
M5, ZIRLO and Optimized ZIRLO 4 ).<br />
Besides, although some operating<br />
conditions have changed, notably<br />
with strech-out operating conditions<br />
and with the increase of maximum<br />
allowed fuel burn-up, most of fuel<br />
safety criteria have not been reviewed<br />
since EDF’s Nuclear Power Plants<br />
(NPPs) were designed, except those<br />
concerning LOCA, which have<br />
changed as a result of rulemaking<br />
occured between 2008 and 2016 (see<br />
Eurosafe 2016) and those concerning<br />
Pellet-Cladding Interaction assisted<br />
by Stress Corrosion Cracking (PCI-<br />
SCC) in PCC-2 which have been introduced<br />
since the 90’s.<br />
In this context, the fuel safety<br />
criteria were reviewed from 2011 to<br />
2017 in order to assess, on the one<br />
hand the sufficiency and validity of<br />
current requirements and fuel safety<br />
criteria relating to all fuel degradation<br />
modes in the light of state-of-the-art<br />
and operating conditions. The consistency<br />
of the fuel rod behavior under<br />
the reference PCCs with the assumptions<br />
used in radiological consequences<br />
studies was also assessed.<br />
Thus, the review concerned the<br />
following limitative physical phenomena:<br />
• cladding embrittlement due to<br />
corrosion. In PWRs, fuel rod<br />
cladding in Zirconium alloy is<br />
oxidized by the primary coolant 5 ,<br />
which leads to the development of<br />
an oxide layer at the clad outer<br />
surface and to the absorption of a<br />
portion of the hydrogen in the<br />
cladding, leading to precipitated<br />
hydrides. As a consequence, cladding<br />
strength decreases [2, 1]. The<br />
kinetics of oxidation depends on<br />
clad temperature, which is about<br />
350 °C in normal operations. If a<br />
PCC-2 may lead to a rise in clad<br />
temperature to a value in the range<br />
of 450 to 480 °C, clad temperature<br />
under PCC-3 and PCC-4 is higher<br />
(> 700 °C) due to boiling crisis;<br />
• clad failure due to Pellet-Cladding<br />
Mechanical Interaction (PCMI)<br />
and PCI-SCC. During transients<br />
characterized by an increase of the<br />
reactor power, the heating of fuel<br />
pellets induces their thermal<br />
expansion and potentially fission-<br />
gas-induced fuel swelling, resulting<br />
in a thermomechanical loading<br />
(stress and strain) on the cladding<br />
and potentially to clad failure.<br />
Depending on power increase<br />
during the transient and on<br />
the level of clad embrittlement,<br />
two clad failures types may be<br />
observed. On the one hand, the<br />
clad loading may be purely<br />
mechanical (PCMI) under the<br />
effect of the stress exerted by<br />
pellets on clad. Hydride precipitation,<br />
in particular in high burnup<br />
fuel rods, plays an important role<br />
in the incipient cracking initiated<br />
at the cladding outer surface which<br />
penetrates inwards, resulting in<br />
though-wall cracking (with the<br />
risk of fuel dispersal in the primary<br />
coolant) [3, 4]. This phenomenon<br />
is associated with power pulses<br />
characterized by a rapid power<br />
increase. On the other hand, in<br />
conjunction with some corrosive<br />
fission products, such as iodine,<br />
expelled from pellets, the clad<br />
loading may be assisted by SCC,<br />
clad failure may be initiated at the<br />
cladding inner surface leading to<br />
clad perforation (without the risk<br />
of fuel dispersal in the primary<br />
coolant) [5] (see PCI workshop at<br />
Luca in 2016). This phenomenon<br />
is associated with power ramps<br />
characterized by a lower power<br />
rate than for pulses and followed<br />
by an holding time at the ramp<br />
terminal level;<br />
• consequences of Departure from<br />
Nucleate Boiling (DNB). Due to<br />
boiling crisis occurrence, clad temperature<br />
can increase suddenly,<br />
reaching also high value (>700 °C)<br />
Fuel<br />
Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
for several seconds. This may lead<br />
either to clad ballooning up to<br />
burst if the rod internal pressure<br />
due to fission gas releases (during<br />
normal operation and transient)<br />
is higher than the external one<br />
( especially for medium burnup<br />
fuel rods), or to clad collapse on<br />
the fuel pellets in the opposite<br />
case. Moreover, the overheated<br />
clads, embrittled by high temperature<br />
oxidation, may lead to their<br />
failure due to the application<br />
of thermal stress during the<br />
rewetting phase [6, 7, 8];<br />
• consequences of fuel melting. In the<br />
extreme case of an excessive temperature<br />
rise of fuel rods due to a<br />
major reactivity insertion or boiling<br />
crisis, fuel rods may melt at least<br />
partially (especially for fresh or<br />
very low burn-up fuel). Indeed,<br />
since the fissile content becomes<br />
low at high burnup, the possibility<br />
of pellet melting is very weak even<br />
taking into account the reduction of<br />
the melting point due to burn-up.<br />
Fuel pellets melting generally leads<br />
to clad fragmentation and clad<br />
failure mode depends on fuel type<br />
(UO 2 versus MOX) [6, 9, 10, 11].<br />
Moreover, the current EDF’s NPPs<br />
operating conditions which are<br />
allowed must be taken into account<br />
in the safety demonstration. Two<br />
phenomena need to be dealt with:<br />
• fuel assemblies may undergo bow<br />
in PWRs due to hydraulic loads<br />
exerted by the water, mechanical<br />
loads applied by the top nozzle,<br />
irradiation and temperature. The<br />
design of fuel assemblies, particularly<br />
the thickness and material of<br />
the guide thimble, their position in<br />
the core, as well as the duration of<br />
their irra diation, also play a role in<br />
the assembly bow. The magnitude<br />
of the bow measured during refueling<br />
outages for some PWRs 6 is<br />
in the order of a few millimetres<br />
and can be as much as 20 mm in<br />
case of excessive assembly bow<br />
[12, 13]. This potentially has an<br />
impact on the in-core power distribution<br />
(at the pin scale) and on<br />
the safety analyses supporting the<br />
plant operations which rely on the<br />
hypothesis of a uniform water gap<br />
between fuel assemblies ;<br />
• leaking fuel rods [14]. Even if<br />
it is an infrequent event, in EDF’s<br />
reactors, some fuel rods may lose<br />
their integrity, for example as the<br />
result of cladding wear due to the<br />
vibration of a loose part 7 stuck in a<br />
grid cell or due to design or manufacture<br />
defects. The presence of a<br />
| | Fig. 1.<br />
Topics reviewed in the frame of French rulemaking on fuel safety criteria.<br />
primary defect (original loss of fuel<br />
rod integrity) allows water to enter<br />
into rods, which frequently leads to<br />
a fairly well explained physicochemical<br />
mechanism linked to<br />
steam oxidation at the inside cladding<br />
surface, and to the occurrence<br />
of a secondary defect. In this area,<br />
which is typically located at about<br />
two or three meters from the original<br />
defect, the cladding becomes<br />
very brittle and can fail inducing<br />
a fuel dissemination in the reactor<br />
coolant system, even in normal<br />
operating conditions [15, 16]. The<br />
impact of this dissemination is<br />
taken into account by the radiochemical<br />
specifications in the<br />
Operating Technical Specifications.<br />
Due to some leaking fuel<br />
rods in reactor, Rod Ejection Accident<br />
(REA) may lead to sudden<br />
fuel rods failures near the ejected<br />
control rod and to the dispersal<br />
of fuel pellets fragments in the<br />
primary coolant, and thus to a violent<br />
thermal interaction between<br />
fuel pellets fragments and the<br />
coolant. This interaction would<br />
lead to a strong primary coolant<br />
pressure increase and to a production<br />
of a steam zone, which could<br />
dry out the neighbouring rods<br />
(near the ejected control rod) up to<br />
their failure. In addition, the<br />
primary coolant pressure would<br />
propagate to neighbouring rods<br />
and to the reactor vessel, potentially<br />
damaging them.<br />
In the French regulatory framework,<br />
new fuel safety criteria are suggested<br />
by the French utility EDF on request<br />
of the French Nuclear Safety Autho rity<br />
(ASN) and submitted to it for approval.<br />
The safety assessment of EDF’s proposals<br />
(based on test results, studies,<br />
operating experience feedback, examinations<br />
of irradiated fuel rods…)<br />
is made by Institute of Radiological<br />
Protection and Nuclear Safety (IRSN).<br />
Based on IRSN’s technical assessment,<br />
the Advisory Committee for<br />
Reactors Safety of the Nuclear Safety<br />
Authority (ASN) meeting about the<br />
French rulemaking on fuel safety<br />
criteria related to PCC-1, PCC-2,<br />
PCC-3 and PCC-4 (except for LOCA)<br />
was held in June 2017. The new<br />
criteria are then assumed to be applied<br />
for EDF’s French PWR (except for EPR)<br />
and for claddings loaded in these<br />
reactors (except for Zircaloy-4 which<br />
is not used anymore in fresh fuel).<br />
In this way, the paper describes the<br />
main conclusions of IRSN’s assessment<br />
about the evolutions of fuel<br />
safety criteria for each PCC and each<br />
limitative physical phenomena. The<br />
following Figure 1 gives an overwiew<br />
of French rulemaking.<br />
2 Fuel safety criteria<br />
before the french<br />
rulemaking<br />
2.1 In PCC-1 and PCC-2<br />
At the reactor design stage, two<br />
requirements associated with physical<br />
phenomena likely to affect the fuel<br />
rod integrity were used to design<br />
reactor protection systems: the<br />
absence of DNB and the absence of<br />
6) The fuel assembly<br />
bow is not measurable<br />
in core but out<br />
of core during<br />
refueling outages<br />
for some EDF PWRs.<br />
7) A loose part is a<br />
fragment, usually<br />
metal and very<br />
small (less than<br />
three millimetres),<br />
which has generally<br />
come off a larger<br />
part during operating,<br />
e.g. when<br />
fuel assemblies are<br />
being handled.<br />
FUEL 39<br />
Fuel<br />
Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
FUEL 40<br />
8) The CFHR criterion<br />
adopted in the<br />
French safety<br />
demonstration<br />
results from the<br />
interpretation of<br />
critical flux tests<br />
performed for a<br />
given fuel assembly.<br />
For this reason,<br />
the CHFR criterion<br />
is likely to undergo<br />
changes in case of<br />
modification to the<br />
fuel materials and<br />
design.<br />
9) A RIA is caused by<br />
a control REA,<br />
which is defined as<br />
the mechanical<br />
failure of a Rod<br />
Cluster Control Assembly<br />
(RCCA)<br />
drive mechanism<br />
casing, located on<br />
top of the reactor<br />
pressure vessel<br />
which is ejected<br />
vertically from the<br />
reactor core due to<br />
the high coolant<br />
pressure. Such a<br />
RIA is characterized<br />
by a very rapid increase<br />
of reactivity<br />
and power in some<br />
rods of the reactor.<br />
10) EDF’s safety<br />
domain for REA:<br />
Oxide thickness,<br />
enthalpy variation,<br />
pulse width, clad<br />
temperature.<br />
11) ECR : Equivalent<br />
Cladding Reacted.<br />
12) Expansion due to<br />
compression using<br />
various PWR cladding<br />
alloys and<br />
performed at<br />
350°C and 10 -4 s -1 .<br />
13) Uni-axial tensile<br />
tests using<br />
transverse samples<br />
and carried out<br />
from 280°C to<br />
400°C at 10 -2 s -1 .<br />
fuel melting. In the 1990s, the absence<br />
of clad failure due to PCI-SCC was<br />
added. Criteria were thus defined:<br />
• in order to avoid DNB, the Critical<br />
Heat Flux Ratio (CHFR) must<br />
remain above a critical value<br />
d epending on the fuel assembly 8 ;<br />
• in order to avoid fuel melting,<br />
the maximum Linear Power<br />
Density (LPD) must remain below<br />
590 W/cm;<br />
• in order to avoid clad failure due to<br />
PCI-SCC, some thermo- mechanical<br />
limits must be verified.<br />
In addition, fuel rod design criteria<br />
were used to check that fuel rods<br />
behave correctly during transients as<br />
regards to:<br />
• cladding corrosion. In PCC-1, oxide<br />
thickness shall not exceed 100 µm.<br />
In PCC-2, clad temperature at the<br />
interface between the metal and<br />
the oxide shall not exceed 425 °C;<br />
• PCMI. In PCC-1 and PCC-2, the<br />
circumferential clad strain shall<br />
not exceed 1 %.<br />
2.2 In PCC-3 and PCC-4<br />
In France, at the start of the industrial<br />
exploitation of NPPs, specific requirements<br />
and empirical criteria were<br />
defined to demonstrate core coolability,<br />
especially for Rod Ejection<br />
Accident (REA) 9 :<br />
• to ensure that there is no hot or<br />
molten fuel dispersal in the<br />
primary coolant during REA, the<br />
maximum fuel enthalpy is limited<br />
to 200 cal/g, the limit coming from<br />
Westinghouse’s extrapolation of<br />
fuel behavior established on the<br />
basis of RIA full-scale SPERT-CDC<br />
tests carried out at zero-power on<br />
fresh and very low irradiated UO 2<br />
fuel. This criterion is applicable for<br />
mean fuel assembly burn-up up to<br />
33 GWd/tU;<br />
• regarding PCMI, the progressive<br />
increase of fuel assembly discharge<br />
burn-up led ASN to ask EDF to<br />
demonstrate that the previous<br />
criteria were still applicable for<br />
REA. Thus, some full-scale tests<br />
carried out in the French CABRI<br />
test reactor and in the Japanese<br />
NSRR test reactor using high<br />
burn-up fuel rods led to fuel<br />
dispersal in the primary coolant for<br />
fuel enthalpy far below 200 cal/g.<br />
These tests clearly showed that this<br />
criterion was no longer relevant.<br />
Based on the results of full-scale<br />
tests, EDF established an empirical<br />
safety domain defined by four<br />
parameters 10<br />
which intends precluding<br />
PCMI clad failure and<br />
burst during boiling crisis for high<br />
mean fuel assembly burn-up<br />
(> 47 GWd/tU);<br />
• the maximum peak clad temperature<br />
must remain below 1,482 °C<br />
(2,700 °F). This limit was taken<br />
from fuel failure boundary for<br />
LOCA conditions. The rational for<br />
retaining a higher temperature<br />
limit for non-LOCA transients,<br />
such as REA, is that film boiling<br />
occurs briefly during those<br />
transients, so that fuel rods could<br />
withstand this brief dry-out<br />
without suffering serious damage.<br />
In addition, the number of fuel rod<br />
failures must be calculated so that the<br />
radiological doses to the public can be<br />
estimated. A requirement is defined to<br />
limit the number of fuel rods affected<br />
by DNB. The conservative assumption<br />
is that all fuel rods entering into<br />
boiling crisis are assumed to fail.<br />
Thus, the percentage of fuel rods<br />
likely to suffer DNB is limited to 5 % in<br />
PCC-3 and to 10 % in PCC-4. Besides,<br />
all fuel rods that experience fuel<br />
melting, especially for REA, are<br />
assumed to be failed for radiological<br />
doses calculations. Nevertheless, only<br />
a limited amount of fuel melting is<br />
accepted, less than 10 % of pellet<br />
volume.<br />
3 Evolution of fuel safety<br />
criteria<br />
3.1 Clad embrittlement<br />
due to corrosion<br />
During operating conditions, it is no<br />
longer necessary, for cladding alloys<br />
loaded in EDF’s reactors (M5, and<br />
Optimized ZIRLO), to verify the oxide<br />
thickness criterion limited to 100 µm<br />
because of their improved corrosion<br />
resistance. However, as in-reactor<br />
hydrogen content has a major impact<br />
on clad behavior under PCMI during<br />
incidental and accidental conditions,<br />
the validity of the various criteria<br />
ensuring clad non-failure under PCMI<br />
conditions relies on compliance with<br />
limits of hydrogen content (see<br />
§ 4.2.1).<br />
During incidental conditions, the<br />
absence of corrosion acceleration is<br />
not likely to occur for cladding alloys<br />
loaded in EDF’s reactors because of<br />
their corrosion resistance and the<br />
temperatures likely to be reached<br />
during PCC-2. Verification that the<br />
clad temperature at the interface<br />
between the metal and the oxide<br />
remains below 425°C is therefore no<br />
longer necessary.<br />
In accidental conditions, the<br />
current clad temperature criterion<br />
limited to 1482°C does not take into<br />
account the time spent at high<br />
temperature during boiling crisis,<br />
even though cladding oxidation rate<br />
is dependent on this. By analysing<br />
experimental results available in the<br />
literature, EDF plans to complete this<br />
criterion by defining a new oxidation<br />
rate (ECR 11 ) limit, which is expressed<br />
as a function of maximum clad<br />
temperature and based on DNB tests<br />
carried out in PBF reactor [7]. IRSN<br />
considered that, although this<br />
approach is acceptable, EDF hasn’t<br />
taken into account all physical<br />
phenomena that are likely to induce<br />
clad embrittlement nor measurement<br />
uncertainties to define the ECR limit.<br />
EDF will complete its approach and<br />
review this new criterion.<br />
3.2 Clad failure due to PCMI<br />
and PCI-SCC<br />
3.2.1 PCMI clad failure<br />
For PCC-2 power ramps likely to<br />
induce PCMI clad failure, the clad<br />
strain limit of 2 % is raised instead of<br />
1 % until the in-reactor hydrogen<br />
content is below 250 ppm, based on<br />
representative analytical tests 12 . In<br />
addition, the uncontrolled with drawal<br />
of control rod assembly bank(s) at<br />
zero power is a particular PCC-2<br />
transient leading to a rapid power<br />
excursion, which may also induce<br />
PCMI clad failure. Up to now, no<br />
criterion was established for this<br />
transient. That is why, a specific limit<br />
of 1 % of plastic clad strain has been<br />
defined to ensure clad non-failure until<br />
the in-reactor hydrogen content is<br />
below 805 ppm. This criterion is based<br />
on appropriate analytical tests 13 . IRSN<br />
concludes that these evolutions, based<br />
on a cautious interpretation of tests<br />
results, are acceptable.<br />
No requirement and fuel safety<br />
criterion ensuring core coolability<br />
were defined for mean fuel assembly<br />
burn-up between 33 and 47 GWd/tU<br />
in REA transients. Moreover, SPERT,<br />
CABRI and NSRR tests were carried<br />
out at zero-power while French safety<br />
demonstration requires REA studies<br />
for all initial power levels. That is why,<br />
EDF has revised existing criteria and<br />
completed the safety demonstration<br />
for fuel assembly burn-up higher than<br />
33 GWd/tU. The new acceptance<br />
criteria, expressed by enthalpy rise<br />
and pulse width, aim at precluding<br />
PCMI clad failure. Their limits depend<br />
on cladding corrosion performances,<br />
more specifically on in-reactor hydrogen<br />
content which is of interest to<br />
cope with PCMI behavior. More precisely,<br />
EDF’s approach to define the<br />
Fuel<br />
Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
new REA criteria depends on the fuel<br />
rods types:<br />
• for UO 2 fuel rods with ZIRLO, Optimized<br />
ZIRLO and M5 claddings,<br />
the approach has been based on<br />
the interpretation with SCANAIR<br />
code [17] of some full-scale RIA<br />
tests carried out in CABRI and<br />
NSRR reactors and associated with<br />
PCMI issue. But, the threshold<br />
values of enthalpy rise and pulse<br />
width are different for M5 than for<br />
ZIRLO and Optimized ZIRLO due<br />
to specific cladding corrosion performances.<br />
Regarding M5, IRSN<br />
considers acceptable the 150 cal/g<br />
of enthalpy rise criterion (the pulse<br />
width limit definition being in<br />
progress and the hydrogen content<br />
limit is 160 ppm). However, concerning<br />
ZIRLO and Optimized<br />
ZIRLO, IRSN identifies that no<br />
uncertainty about experimental<br />
data has been taken into account<br />
by EDF to calculate the enthalpy<br />
rise limit from the restrictive test,<br />
CABRI CIP0-1 14 , which will lead<br />
EDF to review the definition of the<br />
associated criterion;<br />
• for MOX fuel rods with M5 cladding,<br />
EDF has used SCANAIR code<br />
to reproduce PCMI behavior for<br />
MOX fuel based on a specific RIA<br />
test carried out on UO 2 fuel and<br />
related to ballooning. IRSN considers<br />
that the approach is complicated<br />
and unsupported. Eventually,<br />
EDF plans to define fuel<br />
safety criteria for MOX fuel rods<br />
with M5 on the basis of the analysis<br />
of specific integral RIA tests devoted<br />
to MOX, as it has been done<br />
for UO 2 fuel rods.<br />
For REA initiated at non-zero power<br />
levels, EDF has developed an approach<br />
which aims at demonstrating that the<br />
REA initiated at zero-power is the<br />
most limiting compared to transients<br />
initiated at higher power levels. IRSN<br />
estimates that EDF’s approach, based<br />
on the comparison of thermo- mechanical<br />
parameters calculated with<br />
SCANAIR code for the PCMI behavior,<br />
is acceptable. EDF will apply this<br />
approach for each NPPs series.<br />
As in-reactor hydrogen content<br />
plays an important role in the definition<br />
of criteria related to PCMI,<br />
IRSN will assess EDF’s correlations<br />
giving hydrogen content as a function<br />
of oxide thickness.<br />
3.2.2 PCI-SCC clad failure<br />
The risk of PCI-SCC clad failure is<br />
currently taken into account in PCC-2<br />
studies for which fuel rods integrity<br />
must be demonstrated. However,<br />
some PCC-3 or PCC-4 transients lead<br />
to PCI-SCC. If the corresponding clad<br />
failure mode is not likely to lead to a<br />
loss of core coolability, the risk still<br />
needs to be assessed for PCC-3 and<br />
PCC-4 transients in order to ensure<br />
that the radiological consequences of<br />
the concerned accidents are conservatively<br />
assessed. Thus, EDF has<br />
developed an approach to verify the<br />
absence of any risk of clad failure in<br />
case of Uncontrolled Control Rod<br />
Withdrawal accident at non-zero<br />
power level (PCC-3). IRSN considers<br />
this approach to be acceptable.<br />
Another transient, the Steam Line<br />
Break accident initiated at non-zero<br />
power level (PCC-4) is also likely to<br />
lead to PCI-SCC clad failure. EDF<br />
has provided justification concerning<br />
some reactors concluding that the<br />
PCI-SCC clad failure risk is no greater<br />
than for PCC-2 transients. For IRSN,<br />
the justification still needs to be confirmed<br />
and extended to all reactors.<br />
3.3 Consequences of DNB<br />
In order to demonstrate the absence<br />
of fuel dispersal in the primary coolant<br />
after clads ballooning and burst<br />
during boiling crisis, EDF has proposed<br />
two approaches depending on<br />
transients:<br />
• for REA, the approach is based<br />
on the comparison between the<br />
restrictive PCMI criterion and<br />
results of various full-scale tests<br />
associated with ballooning and<br />
burst (IGR, BIGR, NSRR, PBF – [18,<br />
19, 20]). In the available experimental<br />
database, no fuel dispersal<br />
is observed up to EDF fuel rods<br />
burn-up discharge limit (57 GWd/<br />
tU) and up to the enthalpy rise<br />
limit of 150 cal/g (see § 4.2.1);<br />
• for Uncontrolled Control Rod<br />
With drawal at non-zero power<br />
level (PCC-3) and Locked Rotor<br />
(PCC-4) accidents, EDF has compared<br />
the maximum fuel rod<br />
burn-up calculated beyond which<br />
boiling crisis is avoided and the<br />
current non-dispersal threshold 15 .<br />
However, as the absence of fuel<br />
dispersal has been demonstrated<br />
with a very small margin, IRSN<br />
considers that EDF will have to<br />
update its safety demonstration for<br />
each ten-yearly outage review or in<br />
case of modifications deemed to<br />
impact this conclusion.<br />
Besides, questionning the conservative<br />
assumption is that all fuel<br />
rods entering into boiling crisis are<br />
assumed to fail, EDF foresees to limit<br />
(up to 5 % for PCC-3 or 10 % for<br />
PCC-4) the number of broken rods<br />
due to ballooning during boiling<br />
crisis. From EDF’s point of view, the<br />
current criterion related to radiological<br />
doses calculations is based on a<br />
very conservative assumption considering<br />
that all fuel rod entering into<br />
boiling crisis is supposed to be failed<br />
[21]. By applying a fuel rod burn-up<br />
threshold calculated with SCANAIR<br />
code [17] depending on fuel rod<br />
design and irradiation, some fraction<br />
of fuel rods can be excluded from the<br />
counting of failed rods. IRSN considers<br />
acceptable this method. However,<br />
in case of plant operating conditions<br />
modifications (for the future), EDF’s<br />
evolution could lead to increase<br />
radiological consequences, which is<br />
not acceptable for IRSN.<br />
Finally, regarding on-going RIA<br />
investigations and research programs,<br />
IRSN considers namely that Cabri<br />
International Project (CIP 16 ) tests<br />
planned in the CABRI-water loop<br />
facility may be used to analyse clad<br />
behavior during boiling crisis notably<br />
for high fuel burn-up and will improve<br />
knowledge on the MOX fuel behavior.<br />
3.4 Consequences<br />
of fuel melting<br />
In the current safety demonstration,<br />
no requirement associated with fuel<br />
safety criterion was defined concerning<br />
fuel melting risk during PCC-3. In<br />
order to adress this gap, EDF plans to<br />
verify the limit of 10 % molten fuel at<br />
the pellet centre for the Uncontrolled<br />
Control Rod Withdrawal accident<br />
initiated at non-zero power level. For<br />
IRSN, this evolution is acceptable, but<br />
the radiological doses calculations<br />
related to this transient will have to be<br />
assessed consistently with the new<br />
criterion.<br />
Moreover, like the NRC’s requirement,<br />
a limited amount of fuel melting<br />
is acceptable provided it is restricted<br />
to the fuel centerline region and is<br />
less than 10% of pellet volume [22].<br />
Indeed, during REA (PCC-4), due to<br />
the effects of edge peaked power and<br />
lower solidus temperature, fuel rods<br />
may undergo fuel melting in the pellet<br />
periphery. Thus, fuel melting outside<br />
the centerline region is precluded to<br />
avoid molten fuel coolant interaction.<br />
Therefore, EDF will demonstrate that<br />
this requirement is satisfied based on<br />
appropriate analysis rules.<br />
Besides, with regard to the<br />
200 cal/g of maximum fuel enthalpy<br />
criterion for REA (applied to fuel<br />
assemblies with burn-ups up to<br />
33 GWd/tU), EDF confirmed its validity<br />
for MOX fuel on the basis of the<br />
CABRI REP-Na9 test 17 . However, IRSN<br />
14) For CIP0-1, the<br />
measured<br />
hydrogen content<br />
is 1000 ppm.<br />
15) Established at<br />
55,2 GWd/tU in<br />
mean fuel rod<br />
burn-up, based on<br />
Halden and<br />
Studsvik LOCA<br />
tests.<br />
16) CABRI CIP: Tests<br />
with water coolant<br />
loop plan to start<br />
in <strong>2018</strong>.<br />
17) CABRI REP-Na9<br />
was carried out on<br />
MOX fuel with a<br />
low clad corrosion<br />
and a fuel rod<br />
burn-up of 28<br />
GWd/tU. The<br />
tested fuel rod was<br />
not failed for a<br />
maximum fuel<br />
enthalpy of<br />
200 cal/g.<br />
FUEL 41<br />
Fuel<br />
Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
FUEL 42<br />
underlines that further tests on MOX<br />
fuel would improve knowledge about<br />
its sensibility as regard to fuel melting,<br />
especially for high burn-up and high<br />
plutonium levels characteristics of<br />
MOX fuel loaded in EDF’s reactors.<br />
4 Taking into account<br />
rod failures and<br />
assembly bow<br />
4.1 Impact of fuel assembly<br />
bowing on safety<br />
demonstration<br />
Fuel assemblies distort in-core and the<br />
gap between fuel assemblies can vary<br />
away from the design value. Using<br />
several ex-core fuel assembly bow<br />
measurements (from different reactors<br />
and cycles), EDF has developed<br />
a mechanical model to estimate interassembly<br />
gap size distributions in<br />
cores. IRSN assessed assumptions and<br />
considered that the predictive model<br />
is consistent with the current state- ofthe-art.<br />
In addition to slower drop<br />
times of RCCA due to friction in guide<br />
tubes, fuel assembly distortion potentially<br />
leads to neutronic, thermohydraulic<br />
and mechanical impacts on<br />
safety demonstration:<br />
• the presence of larger inter-assembly<br />
gaps causes a local variation in<br />
the fuel-to-moderator ratio and<br />
hence the local neutron moderation.<br />
Basically, as fuel assemblies<br />
move apart, the concentration of<br />
thermal neutrons in the gap increases<br />
and so does the power in<br />
peripheral pins. EDF has developed<br />
a new methodology for quantifying<br />
and taking into account this effect<br />
in the safety demonstration. IRSN<br />
estimates this methodology satisfactory;<br />
• for the same reasons, the Critical<br />
Heat Flux Ratio (CHFR) decreases<br />
at periphery, but also the hydraulic<br />
diameter of the corresponding flow<br />
channel increases, so does the<br />
CHFR. Because of these antagonistic<br />
effects, the flow channel in<br />
which the minimum CHFR is<br />
reached could become a peripheral<br />
one (instead of a channel within<br />
the fuel assembly in the current<br />
safety demonstration). In such<br />
flow channel, grid straps do not<br />
have mixing vanes, which significantly<br />
reduces CHFR. For EDF,<br />
the minimum CHFR remains in the<br />
center of the fuel assembly. However,<br />
to evaluate the global effect,<br />
EDF realized sensitivity studies<br />
notably using unappropriate CHF<br />
correlations. Because of the large<br />
number of justifications still to be<br />
provided, IRSN can’t conclude on<br />
EDF evaluation;<br />
• the presence of smaller water gaps,<br />
and particularly the existence of<br />
contact between grids, is likely to<br />
increase the maximum impact<br />
forces on fuel assembly grids under<br />
seismic and LOCA loads. EDF not<br />
yet assessed the effect of variable<br />
inter-assembly gaps, repre sentative<br />
of the in-reactor situation, on the<br />
assembly grids buckling risk. In<br />
addition, for IRSN, the validation<br />
of EDF’s model to calculate the<br />
impact force on grids during<br />
accidentel conditions needs to be<br />
completed, particularly because<br />
it doesn’t include a comparison<br />
with sufficiently representative<br />
tests results. Thus, the safety<br />
demonstration will be updated.<br />
4.2 Leaking fuel rods<br />
during normal operating<br />
conditions<br />
The behavior of defective fuel rods,<br />
especially under REA, is an important<br />
aspect of safe reactor operation, since<br />
some EDF’s reactors (7 out of the 58<br />
operating reactors currently) contain<br />
a very small percentage of leaking fuel<br />
rods (only 0,11% leaking fuel assemblies).<br />
This issue has been assessed for<br />
several years. The complexity of the<br />
physical phenomena to be taken into<br />
account and the lack of available<br />
experimental data on waterlogged<br />
fuel rods under this transient explain<br />
the difficulty to conclude on the<br />
potential unwanted effects: surrounding<br />
fuel rods failures due to<br />
mechanical and thermal effects or<br />
even potential vessel damage [23,<br />
24]. IRSN considered that EDF’s<br />
demonstration takes into account<br />
satisfactorily the state-of-the-art.<br />
Finally, the large pressure pulse does<br />
not lead to additional fuel rods failures<br />
nor to vessel damage. However, for<br />
IRSN, EDF should still justify that the<br />
models used for assessing thermal<br />
interaction and its consequences are<br />
appropriate.<br />
Considering other PCC-2 and<br />
PCC-4 transients, IRSN estimates that<br />
it is likely that in many cases, application<br />
of stress would lead to the fuel<br />
rods failure in the secondary defect<br />
area and to fuel dispersal in the<br />
primary coolant. However, these<br />
phenomena are unlikely to affect the<br />
core coolability or to have any significant<br />
impact on the the radiological<br />
doses calculations, except for steam<br />
generator tube rupture accidents.<br />
Indeed, these transients are characterized<br />
by a break in the second<br />
barrier, containment bypass and the<br />
possibility that some contaminated<br />
reactor coolant will be released into<br />
the environment. EDF will study<br />
the potential consequences of this<br />
scenario.<br />
References<br />
1. A.M. Garde et al., Hydrogen Pick-Up<br />
Fraction for ZIRLO Cladding Corrosion<br />
and Resulting Impact on the Cladding<br />
Integrity, Proceedings of Top Fuel 2009<br />
Paris, France, September 6-10 (2009)<br />
2. S. K. Yagnik, R-C Kuo, Y.R. Rashid et al.,<br />
Effect of hydrides on the mechanical<br />
properties of Zircaloy-4, Proceedings of<br />
the 2004 International Meeting on<br />
LWR Fuel Performance, Orlando,<br />
Florida, September (2004)<br />
3. R.L. Yang, R.O. Montgomery,<br />
N. Waeckel, EPRI TR #1002865, Topical<br />
report on reactivity-initiated accident:<br />
bases for RIA fuel and core coolability<br />
criteria (2002)<br />
4. T. Sugiyama, High burnup fuel behavior<br />
under high temperature RIA conditions,<br />
FSRM 2010, Tokai, Japan, May (2010)<br />
5. B. Julien et al., Performance of<br />
advanced fuel product under PCI<br />
conditions, Proceedings of the 2004<br />
International Meeting on LWR Fuel<br />
Performance, Orlando, Florida,<br />
September 19-22 (2004)<br />
6. P.E. Macdonald, W.J. Quapp et al.,<br />
Response of unirratiated and irratiated<br />
PWR fuel rods tested under Powercooling<br />
mismatch conditions, Nuclear<br />
Safety, vol.19, n°4, (1978)<br />
7. F. M. Haggag, Zircaloy-cladding<br />
embrittlement criteria : comparison of<br />
in-pile and out-of-pile results, NUREG/<br />
CR-2757 (1982)<br />
8. T. Fuketa, Transient response of LWR<br />
fuels (RIA), Compr. Nucl. Mater.<br />
579-593 (2012)<br />
9. W.G. Lussie, The response of mixed<br />
oxide fuel rods to power bursts,<br />
IN-ITR-114, Idaho Nuclear Corporation<br />
(1970)<br />
10. W.G. Lussie, The response of UO2 fuel<br />
rods to power bursts, IN-ITR-112, Idaho<br />
Nuclear Corporation (1970)<br />
11. M.D. Freshley, Behavior of discret<br />
plutonium dioxide particles in mixedoxide<br />
fuel during rapid power transient,<br />
Nuclear technology, Vol.15 (1972)<br />
12. N. Waeckel, Fuel Assembly distortion in<br />
EDF NPPs, Oral communication on<br />
OECD WGFS, Paris (2014)<br />
13. C. Durand, Fuel bowing performances,<br />
EDF oral communication at OECD NEA<br />
Workshop Advanced fuel modelling for<br />
safety and performance enhancement<br />
(2017)<br />
14. Report OECD NEA/CSNI/R(2014)10,<br />
Leaking Fuel Impacts and Practices<br />
(2014)<br />
15. Y. KIM, S. KIM, Kinetic studies on<br />
massive hydriding of commercial<br />
zirconium alloy tubing, Journal of<br />
nuclear materials, 270, pp. 147-153<br />
(1999)<br />
Fuel<br />
Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
44<br />
AMNT 2017<br />
16. D. H. Locke, The behavior of defective<br />
reactor fuel, Nuclear Engineering and<br />
Design (1972)<br />
17. A. Moal, V. Georgenthum,<br />
O. Marchand, SCANAIR: A transient fuel<br />
performance code Part One: General<br />
modelling description, Nuclear<br />
Engineering and Design, Vol. 280,<br />
pp. 150-171 (2014)<br />
18. NUREG/IA-0213, Experimental study of<br />
narrow pulse effects on the behavior of<br />
high burn-up fuel rods with Zr 1 % Nb<br />
cladding and UO2 fuel (VVER type)<br />
under reactivity-initiated accident<br />
conditions: program approach and<br />
analysis of results, (2006)<br />
19. M. Ishikawa, A study of fuel behavior<br />
under reactivity initiated accident<br />
conditions – Review, Journal of nuclear<br />
materials, Vol. 95, pp. 1-30 (1980)<br />
20. OCDE NEA Report N°6847, Nuclear fuel<br />
behavior under Reactivity-initiated<br />
accident (RIA) conditions (2010)<br />
21. C. Bernaudat, J. Guion, N. Waeckel,<br />
IAEA Technical meeting on fuel behaviour<br />
and modelling under severe transient<br />
and LOCA conditions, Mito<br />
(Japon) (2011)<br />
22. Draft regulatory guide DG-1327,<br />
Pressurized water reactor control rod<br />
ejection and boiling water reactor<br />
control rod drop accidents, U.S, Nuclear<br />
Regulatory Commission (NRC),<br />
Washington DC (2016)<br />
23. S. Tanzawa and T. Fujishiro, Effects of<br />
waterlogged fuel rod rupture on<br />
adjacent fuel rods and channel box<br />
under RIA conditions, Nucl. Sci. and<br />
Tech., 24(1):23-32 (1987)<br />
24. T. Sugiyama and T. Fuketa, Mechanical<br />
energy generation during high burnup<br />
fuel failure under reactivity initiated<br />
accident conditions, Nucl. Sci. and Tech.,<br />
37(10):877-886 (2000)<br />
Authors<br />
Sandrine Boutin<br />
Stephanie Graff<br />
Aude Foucher-Taisne<br />
Olivier Dubois<br />
Institut de radioprotection<br />
et de sûreté nucléaire<br />
B.P. 17<br />
92262 Fontenay-aux-Roses,<br />
France<br />
48 th Annual Meeting on Nuclear Technology (AMNT 2017)<br />
Key Topic | Outstanding Know-How &<br />
Sustainable Innovations<br />
Technical Session: Reactor Physics,<br />
Thermo and Fluid Dynamics<br />
Neutron Flux Oscillations Phenomena<br />
Joachim Herb<br />
The Technical Session about Neutron Flux Oscillation Phenomena was chaired by Joachim Herb (Gesellschaft für<br />
Anlagen und Reaktorsicherheit (GRS) GmbH) and well attended by approx. 50 listeners. It comprised of three keynotes<br />
and two technical presentations. The main topics were the significant changes of the neutron flux noise levels in<br />
different German and foreign pressurized water reactors (PWRs). For about ten years an increase in neutron noise<br />
levels has been observed in German PWRs. During the following five years the noise levels have been decreasing again.<br />
In principle, a correlation of the neutron noise levels to the use of certain fuel element types was observed and the<br />
phenomenon of neutron flux oscillations had been known since decades. Nevertheless, no self-consistent physical<br />
theory exists so far, which can explain the observed changes and the absolute levels of the observed neutron flux noise<br />
levels. Therefore, safety authorities, technical support organizations (TSO), utilities as well as research organizations<br />
showed increased interest in this topic during the last years. The results of the corresponding work as well as an outlook<br />
into soon-starting research projects were given in this session.<br />
The first keynote of the session about<br />
Neutron Flux Oscillations in PWR:<br />
Safety Relevance was presented<br />
by Kai-Martin Haendel (TÜV Nord<br />
EnSys GmbH & Co. KG, Germany).<br />
Mr. Haendel reported that the source<br />
of the low frequency neutron flux<br />
noise (< 2 Hz) had unexpectedly<br />
changed which led to sporadic erroneous<br />
activations of surveillance<br />
signals (rod drop, reactor power<br />
limitation) in the reactor limitation<br />
system despite the existing filtering<br />
of the neutron flux signal. A review of<br />
the limitation and protection systems<br />
was necessary to demonstrate that<br />
safety functions were not compromised<br />
by the higher levels of neutron<br />
noise and that the actions of the<br />
limitation system comply with the<br />
given safety criteria, i.e. the safetyrelated<br />
parameters adhere to all safety<br />
limits under all design accidental<br />
conditions. For the purpose of the<br />
rod drop detection and the short-time<br />
corrected thermal reactor power it<br />
was shown that, as long as the delay<br />
time of the filters stayed below certain<br />
limits, all safety key parameters were<br />
met. A reduction of the reactor power<br />
results also in a decrease of the<br />
neutron noise level and hence in the<br />
absence of any erroneous activation of<br />
the rod drop signal and a strongly<br />
reduced occurrence of erroneous activations<br />
of the reactor power signal.<br />
Marcus Seidl (PreussenElektra<br />
GmbH, Germany) presented the second<br />
keynote with the title Neutron Flux<br />
Oscillations in PWR: Operational<br />
Experience. While neutron noise so<br />
far has mainly been explained empirically<br />
the existing theoretical frameworks<br />
are unable to describe all its<br />
observed properties in Konvoi and<br />
Vor-Konvoi reactors in a consistent<br />
manner. This is likely due to the fact<br />
that a suitable (and not jet existing)<br />
AMNT 2017<br />
Technical Session: Neutron Flux Oscillations Phenomena ı Joachim Herb
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
theory needs to couple the neutronics,<br />
thermal-hydraulics and mechanical<br />
properties of the core. The goal is<br />
difficult to achieve because on the one<br />
hand almost no suitable coupling<br />
schemes exist at the moment for this<br />
purpose. It is also difficult on the other<br />
hand, because neutron noise mostly<br />
has been treated as an unwanted signal<br />
(besides being used for some recurrent<br />
component oscillation checks) in commercial<br />
power reactors in the sense<br />
that is has been suitably filtered or<br />
reduced by means of power reductions.<br />
As other fields of science have<br />
shown in the past the analysis of noisy<br />
signals can often lead to better instruments<br />
and in turn to the detection<br />
of hitherto unrecognized phenomena.<br />
So the main motivation to continue<br />
current efforts to consistently explain<br />
the neutron noise signals is to get a<br />
better understanding of the mechanical<br />
and thermal-hydraulic behaviour<br />
of fuel assemblies under operating<br />
conditions. For this purpose, it might<br />
be necessary to corroborate upcoming<br />
theoretical explanations by means of<br />
better in-core temperature and mechanical<br />
oscillation measurements. In<br />
practice this can pay off in an improvement<br />
of reactor performance by<br />
leading to better fuel assembly designs<br />
and an improved thermal margin<br />
determination.<br />
The third keynote about Neutron<br />
Flux Oscillations in PWR: Clarification<br />
of Possible Causes was<br />
presented by Christophe Demazière<br />
(Chalmers University of Technology,<br />
Sweden). He gave a brief account of<br />
the capabilities of core monitoring<br />
using noise analysis, including a<br />
historic overview starting 1949 with<br />
early development in noise analysis at<br />
the Clinton Pile at ORNL, USA, later<br />
works on the detection of excessive<br />
vibrations of control rods, core-barrel<br />
vibrations, estimations of in-core<br />
coolant velocities, detector tube<br />
impacting and the analyses of BWR<br />
instabilities. To generalize the use of<br />
noise analysis, it is necessary to invert<br />
the reactor transfer function, which<br />
describes the effect of local disturbances<br />
on the measured neutron flux<br />
noise. Then Christophe Demazière<br />
introduced the Horizon 2020 EUproject<br />
CORTEX (CORe monitoring<br />
Techniques and EXperimental validation<br />
and demonstration) which was<br />
expected to start on September 1 st ,<br />
2017. The project aims are the<br />
development of high fidelity tools<br />
for simulating stationary fluctuations,<br />
the validation of those tools against<br />
experiments to be performed at<br />
research reactors, the development<br />
of advanced signal processing techniques<br />
(to be combined with the simulation<br />
tools), the demonstration of the<br />
proposed methods for both on-line<br />
and off-line core diagnostics and<br />
monitoring and the dissemination of<br />
the knowledge gathered from within<br />
the project to stakeholders in the<br />
nuclear sector. The project will be led<br />
and coordinated by Chalmers University<br />
of Technology. 17 European<br />
organizations (from eight countries)<br />
and two non-European organizations<br />
will be involved in the project.<br />
Additionally, there will be an Advisory<br />
End-User Group for the project.<br />
Gaëtan Girardin (Kernkraftwerk<br />
Gösgen-Däniken AG, Switzerland)<br />
sum marized the recent investigation<br />
on Neutron Flux Oscillation Phenomena<br />
at Kernkraftwerk Gösgen<br />
(KKG), which is a 3-loop pre-KONVOI<br />
type PWR. It was observed that the<br />
global amplitudes of the power oscillations<br />
had slowly and monotonously<br />
increased during the last seven operating<br />
cycles. Moreover, no modification<br />
of importance had been done on<br />
the primary circuit and the reactor<br />
core over the last years that could<br />
possibly explain the amplitude increase<br />
of the neutron noise. In order<br />
to determine the possible reason<br />
of the neutron noise increase, the<br />
already existing neutron flux measurements<br />
were completed during the last<br />
cycle by two extensive measurement<br />
campaigns: one mid of cycle and the<br />
second one end of cycle. Based on<br />
these new measurements, it was<br />
obtained and confirmed that the<br />
largest noise amplitudes are located in<br />
one quadrant of the core between<br />
Loop 1 and 3, and the simultaneous<br />
measurements revealed that the noise<br />
signals at two opposite sides of the<br />
core had strong negative correlations.<br />
Moreover, no time shifts were found<br />
in the axial measurements between<br />
the top and bottom neutron signals. It<br />
was also found that the highest amplitudes<br />
had not increased over last cycle<br />
compared to previous increase in the<br />
previous cycles. The observed saturation<br />
of the noise amplitudes at quite<br />
high amplitudes were correlated to a<br />
core fully loaded with HTP design<br />
fuel assemblies. The ex-core filters<br />
were calibrated in a way so that few<br />
activations of the power limitation<br />
system were observed. It was also<br />
observed that there existed a relationship<br />
between fuel assembly bowing<br />
and noise amplitudes. Based on the<br />
analyses a stabilization of neutron<br />
noise amplitudes was expected.<br />
The final presentation was given<br />
by Joachim Herb (Gesellschaft für<br />
Anlagen- und Reaktorsicherheit, (GRS)<br />
gGmbH, Germany) about the Analyses<br />
of Possible Explanations for the<br />
Neutron Flux Fluctuations in<br />
German PWR. He reported, that no<br />
comprehensive theory existed yet<br />
which could explain the observed<br />
neutron flux fluctuation levels based<br />
on first physical principles. Therefore,<br />
GRS has started investigations on<br />
which combination of thermal hy draulics,<br />
structural mechanics and neutron<br />
physics models were able to explain<br />
the observed neutron flux fluctuation<br />
and the change in the observed levels.<br />
The analyses based on the evaluation<br />
of measurements in German PWRs.<br />
Using simple models, parts of the<br />
observations could be explained: A<br />
basic coupled thermal hydraulics/<br />
point neutron kinetics model could<br />
reproduce the shape of the neutron<br />
flux noise spectrum as well as the<br />
linear dependency between the noise<br />
level and the moderator temperature<br />
coefficient, but it could not explain<br />
the spatial correlations between the<br />
signals of different detectors. A point<br />
source model for the neutron flux was<br />
used to consistently explain the observations<br />
at the different neutron flux<br />
detector locations, but it could not<br />
explain the shape of the noise spectrum.<br />
A model based on the modification<br />
of the cross sections of the<br />
neutron reflector was able to produce<br />
flux changes of about 10 %, but it had<br />
to be shown what could cause the<br />
assumed changes of the cross sections.<br />
Also, different mechanical explanations<br />
were discussed based on the assumption<br />
of core-wide motions of fuel<br />
assemblies and further core internals.<br />
These motions might be produced by<br />
excitations at the natural frequency,<br />
forced excitations and/or self-excitation<br />
due to fluid-structure interaction<br />
with the coolant. Overall, it was concluded<br />
that the phenomena is very<br />
likely caused by a combination of<br />
different physical effects which<br />
requires further work on the combination<br />
of different physical models<br />
and coupled simulations.<br />
Author<br />
Joachim Herb<br />
Gesellschaft für Anlagen- und<br />
Reaktorsicherheit (GRS) gGmbH<br />
Abteilung Kühlkreislauf /<br />
Cooling Circuit Department<br />
Bereich Reaktorsicherheitsforschung<br />
/ Reactor Safety<br />
Research Division<br />
Forschungszentrum<br />
Boltzmannstraße 14<br />
85748 Garching b. München,<br />
Germany<br />
45<br />
AMNT 2017<br />
AMNT 2017<br />
Technical Session: Neutron Flux Oscillations Phenomena ı Joachim Herb
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
46<br />
AMNT 2017<br />
Key Topic | Enhanced Safety & Operation<br />
Excellence<br />
Focus Session: Radiation Protection<br />
Erik Baumann and Angelika Bohnstedt<br />
The objectives of radiation protection are to minimize the negative health effects due to radiation. Over many past<br />
decades, the regulatory environment, i.e. the various international and national codes and standards but also<br />
recommendations issued by IAEA and IRCP, was always subject to continuous development reflecting up to date<br />
knowledge and experience. Currently, discussions focus on “new” areas of human activities like decommissioning and<br />
on “new radioactive substances” and potential threats associated with their handling (handling and treatment of<br />
substances containing naturally occurring radioactive materials). Latter already became subject to regulations issued<br />
by EURATOM. For topics like decommissioning, statement can be found doubting that the existing regulations address<br />
radiation protection in a sufficient manner.<br />
Authors<br />
Erik Bauman<br />
New NP GmbH<br />
Paul-Gossen-Str. 100<br />
91052 Erlangen,<br />
Germany<br />
Dr. Angelika<br />
Bohnstedt<br />
Karlsruhe Institute<br />
of Technology (KIT)<br />
Programm Nukleare<br />
Entsorgung, Sicherheit<br />
und Strahlenforschung<br />
(NUSAFE)<br />
Hermann-von-<br />
Helmholtz-Platz 1<br />
76344 Eggenstein-<br />
Leopoldshafen,<br />
Germany<br />
In the Focus Session Radiation<br />
Protection – What about the basic<br />
principles and objectives in the<br />
current regulatory environment?<br />
three presentations (a forth one had<br />
to be cancelled on short term for<br />
personal reasons) directed the view<br />
on different aspects. This gave the<br />
occasion to the 25 to 30 participants<br />
for interesting questions and a fruitful<br />
exchange of opinion not only with the<br />
lecturer but also among each other.<br />
Especially the presentation dealing<br />
in a somewhat provocative way<br />
with the subject ‘Hormesis’ led to a<br />
motivated discussion in the audience<br />
about different point of views.<br />
In the first presentation Hormesis<br />
– a Miracle in reality? Discussion Required<br />
Jan-Christian Lewitz (LTZ-<br />
Consulting GmbH) started with from<br />
literature compiled controversial conclusions<br />
about the amount of harmed<br />
people by the Chernobyl accident. This<br />
was followed by a provocative statement<br />
about the hormesis principle in<br />
the way “when unhealthy things<br />
become healthy” and “it is just<br />
depending on the right dose”. He quoted<br />
the explanation for hormesis as “biopositive<br />
reaction of biological systems”<br />
but also restricted that there are<br />
“no general mechanism known for the<br />
different hormetic effects” and indicated<br />
that hormesis is not con sidered<br />
for risk assessment. Then Mr. Lewitz<br />
showed curves about the dose/effect<br />
relation and the LNT model and<br />
remarked that little scientific evidence<br />
of any measurable adverse health<br />
effects at radiation doses below about<br />
100 mSv is at the moment available.<br />
As a discussible example for another<br />
effect he shortly gave an overview of an<br />
incident in Taiwan in the eighties of<br />
the last century where buildings, used<br />
by about 10,000 people, were constructed<br />
with Co-60 contaminated<br />
steel. Higher-than-normal radiation<br />
levels were discovered after 9 years and<br />
therefor surveys for cancer and birth<br />
defects in this group of persons, some<br />
lived up to 20 years in the building,<br />
where executed. Mr. Lewitz presented<br />
the result of the survey with a lower<br />
mortality in the examined group than<br />
in the normal average public.<br />
He ended his presentation with the<br />
questions “What should be looked<br />
after and be obeyed?” and “Is Optimization<br />
below 100 mSv/y justified in<br />
regard to limited resources?” and<br />
encouraged the audience to discuss<br />
with him his challenging point of view.<br />
The second lecture Radiation<br />
Instrumentation and Measurement<br />
Technologies for High Radiation<br />
Fields was given by Dr. Marina Sokcic-<br />
Kostic (NUKEM Technologies Engineering<br />
Services GmbH) who talked about<br />
the possibility to monitor radioactive<br />
materials in high dose-rate environments<br />
where common types of gamma<br />
detectors reach their limits. The first<br />
instrument she presented was a<br />
Geiger- Mueller-Counter where by<br />
switching on and off the counting<br />
tube dead-times can be avoided.<br />
Next Ms. Sokcic-Kostic remarked that<br />
measurement of particle radiation in<br />
the presence of high gamma fields is<br />
quite challenging. She explained a<br />
fission chamber, operable for gamma<br />
radiation up to 50 to 100 Sv/h, where<br />
ionization efficiency is set very low, so<br />
that mainly the fission products<br />
are measured and additionally by<br />
adjusting the pulse heights neutrons<br />
can be separated from gammas. Afterwards<br />
she presented some applications<br />
of this chamber. One device<br />
with several chambers is used to<br />
characterize irradiated fuel assemblies<br />
in a storage pond by passive neutron<br />
and passive gamma counting.<br />
Another one she explained where the<br />
chamber is combined with other<br />
measurement instruments works with<br />
active neutron detection monitors<br />
using external neutron or ion sources.<br />
Ms. Sokcic-Kostic conclude her pre sentation<br />
with a Gamma camera which is<br />
able to localize hot spots in waste<br />
packages and some information about<br />
Cherenkov detectors.<br />
With the final talk Predictions of<br />
Expected Dose Rates by validated<br />
Activation Calculations as Input for<br />
a step-wise Decommissioning and<br />
Dismantling of a Nuclear Power<br />
Plant Dr. L. Schlömer (together with<br />
Dr. S. Tittelbach and Prof. P.-W.<br />
Phlippen; all WTI Wissenschaftlich-<br />
Tech nische Ingenieurberatung GmbH)<br />
changed the subject to Monte-Carlo<br />
modelling. He listed the specific<br />
requirements for decommissioning<br />
like licensing, planning of packaging,<br />
probing and of course cost estimation.<br />
Then he showed that Monte-Carlo<br />
code coupled with modern variance<br />
reduced techniques (ADVANTG) is a<br />
good solution for radiological characterization<br />
while reducing number<br />
of samples and related costs. Mr.<br />
Schlömer commented that even more<br />
detailed calculations are able with<br />
an activation and decay module<br />
(ORIGEN-S). With an example for a<br />
BWR (same for a PWR) he explained<br />
the steps which have to be performed<br />
for the calculation procedure to get<br />
from a technical drawing of a reactor<br />
to a detailed MCNP-model. To validate<br />
the method a comparison of<br />
measured and calculated dose rates is<br />
necessary. Therefore, Mr. Schlömer<br />
continued, dose rate measurements<br />
have to be executed on defined places<br />
between RPV and biological shield. He<br />
concluded his presentation with the<br />
outcome that the methods of validation<br />
show good results for the BWR<br />
and the PWR.<br />
AMNT 2017<br />
Focus Session: Radiation Protection ı Erik Baumann and Angelika Bohnstedt
The International Expert Conference on Nuclear Technology<br />
Estrel Convention<br />
Center Berlin<br />
29 – 30 May<br />
<strong>2018</strong><br />
Germany<br />
AMNT <strong>2018</strong><br />
Key Topics<br />
Outstanding Know-How &<br />
Sustainable Innovations<br />
Enhanced Safety &<br />
Operation Excellence<br />
Decommissioning Experience &<br />
Waste Management Solutions<br />
Preliminary Programme<br />
December 15, 2017<br />
Subject to change.<br />
www.nucleartech-meeting.com
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
48<br />
Plenary Session<br />
Tuesday ı May 29 th <strong>2018</strong><br />
All contributions translated simultaneously<br />
in English/German.<br />
Key Topic<br />
Outstanding Know-How &<br />
Sustainable Innovations<br />
AMNT <strong>2018</strong><br />
Welcome and Opening Address<br />
| | Dr. Ralf Güldner, President of DAtF, Germany<br />
33<br />
Policy<br />
Continuity or Disruption, What Future<br />
for EU-UK Nuclear Partnership<br />
| | Greg Clark MP, Secretary of State for Business,<br />
Energy and Industrial Strategy, United Kingdom<br />
(TBC)<br />
Decommissiong and Interim Storage<br />
after Assignment of Responsibilities<br />
Rückbau und Zwischenlagerung<br />
nach der Neuordnung<br />
| | Dr. Dr. Jan Backmann, Head of Reactor Safety and<br />
Radiation Protection, Ministry of Energy,<br />
Agriculture, the Environment and Digitalization<br />
of Schleswig-Holstein, Germany<br />
33<br />
Economy<br />
The NDA's Current Strategy and<br />
its Long term Objectives<br />
| | David Peattie, CEO, Nuclear Decommissioning<br />
Authority (NDA), United Kingdom<br />
Nuclear Power under Current Market Conditions<br />
in Switzerland<br />
| | Dr. Willibald Kohlpaintner, Head of Nuclear Energy<br />
Division, Axpo Holding AG, Switzerland<br />
33<br />
Competence<br />
How Does Nuclear Phase-Out Affect<br />
the International Business of German<br />
Technical and Scientific Support<br />
Organisations?<br />
| | Dr. Dirk Stenkamp, CEO, TÜV Nord Group,<br />
Germany<br />
Phase-Out in Germany –<br />
We Are International!<br />
| | Carsten Haferkamp, Managing Director,<br />
New NP GmbH<br />
33<br />
Communications<br />
Trust Building by Participation – National<br />
Societal Advisory Committee's Challenging<br />
Objective<br />
Vertrauen schaffen durch Partizipation –<br />
Die große Aufgabe des Nationalen Begleitgremiums<br />
bei der Endlagersuche in Deutschland<br />
| | Prof. Dr. Klaus Töpfer, Former Federal Minister,<br />
Member of the National Societal Advisory<br />
Committee, Germany (TBC)<br />
33<br />
Waste Management<br />
Site Selection in Practice:<br />
Challenges at the Start of the Process<br />
Standortauswahl in der Praxis: Herausforderungen<br />
am Neubeginn des Verfahrens<br />
Panel<br />
| | Ursula Heinen-Esser, Managing Director,<br />
Bundesgesellschaft für Endlagerung (BGE),<br />
Germany<br />
| | N.N.<br />
| | N.N.<br />
| | N.N.<br />
Moderator<br />
| | Johannes Pennekamp,<br />
Frankfurter Allgemeine Zeitung, Germany<br />
Award Ceremony<br />
Award of the Honorary Membership of KTG<br />
| | Presented by Frank Apel, Chairperson of KTG,<br />
Germany<br />
Outside the Box<br />
Black Holes, Multidimensionality and Entropy<br />
– Limits of Reality<br />
| | Dr. Maria J. Rodriguez, Research Group Leader,<br />
Gravitational and Black Hole Theory, Max Planck<br />
Institute for Gravitational Physics, Germany<br />
Focus Sessions<br />
Tuesday ı 29 th May <strong>2018</strong><br />
International Regulation | Radiation<br />
Protection: The Implementation of the<br />
EU Basic Safety Standards Directive<br />
2013/59 and the Release of Radioactive<br />
Material from Regulatory Control<br />
Coordinator:<br />
| | Dr. Christian Raetzke, CONLAR Consulting on<br />
Nuclear Law, Licensing and Regulation, Germany<br />
The EU Basic Safety Standards Directive has to be<br />
implemented in national law by 6 February <strong>2018</strong>. In<br />
Germany a new Act on Radiation Protection has<br />
been created. The changes present many challenges<br />
to regulators and industry alike in EU countries. The<br />
session will particularly focus on the release of radioactive<br />
material from regu latory control and will put it<br />
in the context of the new Directive.<br />
The Implementation of the New EU BSS<br />
in France<br />
| | Sidonie Royer-Maucotel, Commissariat<br />
á l'Énergie Atomique et aux Énergies Alternatives<br />
(CEA), France (TBC)<br />
The Implementation of the New EU BSS<br />
in Germany<br />
| | Dr. Goli-Schabnam Akbarian, Federal Ministry for<br />
the Environment, Nature Conservation, Building<br />
and Nuclear Safety (BMUB), Germany<br />
Comparative Overview of Regulations<br />
for Clearance in NEA Member States<br />
| | Edward Lazo, OECD Nuclear Energy Agency (NEA),<br />
France (TBC)<br />
Necessary Modifications on Clearance<br />
Regulations in Germany and Switzerland –<br />
Comparative Analysis<br />
| | Dr. Jörg Feinhals, DMT GmbH & Co. KG, Secretary<br />
of the Working Group Disposal/Directory of Fachverband<br />
für Strahlenschutz e. V. (Radiation<br />
Protection Association)<br />
Safety of Advanced<br />
Nuclear Power Plants<br />
Tuesday ı 29 th May <strong>2018</strong><br />
Plenary Closing Remarks<br />
| | Frank Apel, Chairperson of KTG, Germany<br />
Coordinators:<br />
| | Dr. Andreas Schaffrath, Gesellschaft<br />
für Anlagen- und Reaktorsicherheit (GRS) gGmbH,<br />
Dr. Thomas Mull, New NP GmbH<br />
Social Evening<br />
DAtF-Reception and<br />
Meet-and-Greet in the Exhibiton Area<br />
New Builds in UK<br />
| | N.N.<br />
AMNT <strong>2018</strong><br />
Preliminary Programme
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Current Developments in China<br />
| | Prof. Xu Cheng, Karlsruhe Institute of Technology<br />
(KIT), Germany<br />
Status on Thermal-Hydraulic Passive Safety<br />
Systems Design and Safety Assessment<br />
| | Prof. Francesco D'Auria, University of PISA, Italy<br />
Reactor Safety Research in Germany<br />
| | Dr. Thomas Nunnemann, Federal Ministry for<br />
Economic Affairs and Energy (BMWi), Germany<br />
Einzeleffekt- und Integralexperimente<br />
zur Untersuchung des Anlauf- und<br />
Betriebsverhalten passiver Systeme<br />
| | Dr. Thomas Mull, New NP GmbH, Germany;<br />
Prof. S. Leyer, Université du Luxembourg,<br />
Luxembourg; Prof. Uwe Hampel,<br />
Dr. Christoph Schuster, Tech nische Universität<br />
Dresden (TUD), Germany<br />
Modellierung passiver Systeme mit<br />
der nuklearen Rechenkette der GRS<br />
| | Dr. Andreas Schaffrath, S. Buchholz,<br />
Dr. A. Krüsssenberg, Gesellschaft für Anlagenund<br />
Reaktorsicherheit (GRS) gGmbH, Germany<br />
Technical Sessions<br />
Wednesday ı 30 th May <strong>2018</strong><br />
Outstanding Know-How &<br />
Sustainable Innovations<br />
Chair & Keynote Coordinator:<br />
| | Dr. Matthias Lamm<br />
Know-How, New Build and Innovations<br />
Keynote<br />
Can Nuclear Energy Thrive in a Carbon-<br />
Constrained World? – Findings From a<br />
New MIT Study<br />
| | Jacopo Buongiorno, TEPCO Professor and<br />
Associate Department Head, Director, Center for<br />
Advanced Nuclear Energy Systems (CANES),<br />
Massachusetts Institute of Technology (MIT), USA<br />
Keynote<br />
AP1000 –<br />
On the Way to Commercial Operation<br />
| | Tba<br />
Operational Readiness<br />
of the Barakah Nuclear Power Plant<br />
| | Dr. Rolf Janke, Nawah Energy Company, Licensing<br />
& Regulatory Affairs, United Arab Emirates<br />
Russian Reactor Technologies:<br />
Basic Development Trend and “Waiting List”<br />
| | Dr. Andrey Gagarinskiy, NRC Kurchatov Institute,<br />
Russia<br />
Advanced Load Following Control<br />
with Predictive Reactivity Management<br />
(ALFC-PREDICTOR)<br />
| | Andreas Kuhn, New NP GmbH, Germany<br />
Improving Knowledge Transfer Through<br />
Interactive Learning Strategies<br />
| | Jeanne Bargsten, TÜV SÜD Energietechnik GmbH<br />
BW, Germany<br />
Digital Transformation in Nuclear Industry –<br />
Focus: Backoffice Applications<br />
| | Dr. Jan Leilich, New NP GmbH, Germany<br />
Keynote<br />
Co-Generation – A Game Changer<br />
in Polands New Build Plans?<br />
| | Prof. Dr. hab. Grzegorz Wrochna, National Centre<br />
for Nuclear Research, Poland<br />
The Future of Nuclear Power<br />
Chair:<br />
| | Dr. Thomas Mull & Fabian Weyermann<br />
Keynote<br />
DEMO – The Remaining Crucial Step T owards<br />
the Exploitation of Fusion Power After ITER<br />
| | Dr. Gianfranco Federici, EUROfusion, Spain<br />
Application of Variance Reduction Techniques<br />
in Neutronics Shielding Calculations of the<br />
Stellarator Power Reactor HELIAS<br />
| | André Häußler, Karlsruhe Institute of Technology<br />
(KIT), Germany<br />
Synergistic Effect of H and He on W Grain<br />
Boundaries: A First-Principles Study<br />
| | Litong Yang, Forschungszentrum Jülich GmbH,<br />
Germany<br />
Neutronics Analyses on the IFMIF- DONES Test<br />
Cell Bio-Shield and Liner<br />
| | Dr. Yuefeng Qiu, Karlsruhe Institute of Technology<br />
(KIT), Germany<br />
CFD Analysis of Centrifugal Liquid Metal<br />
Pumps<br />
| | Moritz Schenk, Karlsruhe Institute of Technology<br />
(KIT), Germany<br />
New Products, New Services<br />
Chair:<br />
| | Prof. Andreas Class and Ralf Schneider- Eickhoff<br />
Steam Generator Segmentation Innovation<br />
Project<br />
| | Niklas Bergh, Westinghouse Electric Germany<br />
GmbH, Germany<br />
ASME Nuclear Certification and<br />
Other Certification Programs<br />
| | Dr. Dirk Kölbl, CIS GmbH Consulting Inspection<br />
Services, TÜV Thüringen Group, Managing<br />
Director, Germany<br />
Equipment Qualification for Nuclear Power<br />
Plants – Ensuring the Compliance<br />
of Safety-Critical Nuclear Equipment<br />
| | Dr. Ailine Trometer, TÜV SÜD Energietechnik<br />
GmbH, Germany<br />
SISTec: Mathematical Calibration<br />
of Large Clearance Monitors<br />
| | Tim Thomas, Safetec Entsorgungs- und Sicherheitstechnik<br />
GmbH, Germany<br />
Perimeter Security System Peri-D-Fence-L1<br />
| | Steffen Christmann, Westinghouse Electric<br />
Germany GmbH, Germany<br />
A Multipurpose Inertial Electrostatic<br />
Confinement Fusion for Medical Isotopes<br />
Production<br />
| | Dr. Yasser Shaban, Southern Medical University,<br />
School of Biomedical Engineering, Expert<br />
Committee Member, China<br />
Neutronic Analysis of a Nuclear- Chicago NH3<br />
Neutron Howitzer<br />
| | Ahmet Ilker Topuz, Istanbul Technical University,<br />
Turkey<br />
Reactor Physics, Thermo and<br />
Fluid Dynamics<br />
Chair:<br />
| | Dr. Andreas Schaffrath<br />
Keynote Coordinator:<br />
| | Dr. Tatiana Salnikova<br />
Investigation of the Operation Mode<br />
of Passive Safety System 1<br />
PANAS: Experimental and Theoretical<br />
Investigation of Generic Thermal Hydraulic<br />
Issues of Passive Safety Systems<br />
| | Dr. Christoph Schuster, Technische Universität<br />
Dresden (TUD), Germany<br />
EASY – Evidence of Design Basis Accidents<br />
Mitigation Solely with Passive Safety Systems<br />
| | Sebastian Buchholz, Gesellschaft für Anlagenund<br />
Reaktorsicherheit (GRS) gGmbH, Germany<br />
Modelling of Condensation Inside<br />
an Inclined Pipe<br />
| | Amirhosein Moonesi Shabestary, Helmholtz-<br />
Zentrum Dresden-Rossendorf, Germany<br />
Performance of the Passive Flooding System<br />
in the Integral Tests of the Easy Project<br />
| | Nadine Kaczmarkiewicz, Deggendorf Institute of<br />
Technology, Mechanical Engineering, Germany<br />
Investigation of the Operation Mode of<br />
Passive Safety System 2<br />
Chair:<br />
| | Dr. Thomas Mull<br />
Investigation of Thermal Coupling Model for<br />
Evaporation Process in a Slightly Inclined Tube<br />
and Tube Bundles<br />
| | Yu Zhang, Technische Hochschule Deggendorf,<br />
Germany<br />
49<br />
AMNT <strong>2018</strong><br />
AMNT <strong>2018</strong><br />
Preliminary Programme
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
50<br />
AMNT <strong>2018</strong><br />
Experimental and Theoretical Investigation<br />
of Boiling in the Slightly Inclined Tubes of the<br />
Containment Cooling Condenser<br />
| | Frances Viereckl, TU Dresden, Chair of Hydrogen<br />
and Nuclear Energy, Germany<br />
Model Order Reduction of Low Pressure<br />
Natural Circulation System<br />
| | René Manthey, TU Dresden, Institute of Power<br />
Engineering, Germany<br />
Model Order Reduction of a High Pressure<br />
Natural Circulation System<br />
| | Alexander Knospe, TU Dresden, Institut für<br />
Energietechnik, Germany<br />
New Neutron Kinetic Developments<br />
and Findings<br />
Chair:<br />
| | Dr. Tatiana Salnikova<br />
Keynote<br />
Insights of End-of-Life Core Design<br />
from Utility Point of View<br />
| | Dr. Marcus Seidl, PreussenElektra GmbH,<br />
Germany<br />
Frequency-Domain Investigation<br />
on the Neutron Noise in KWU PWRs<br />
| | Marco Viebach, TU Dresden, Institut für<br />
Energietechnik, Germany<br />
Nuclear Energy Campus<br />
The Nuclear Energy CAMPUS leads through the<br />
world of radioactivity, nuclear technology and<br />
radiation protection with individual stations. There<br />
will be contact persons available at all of the themed<br />
stands to offer information in form of short talks,<br />
movies, demonstrations or experiments. Besides,<br />
information on study options and career perspectives<br />
within nuclear industry are provided. The CAMPUS<br />
language will be German..<br />
Welcome and Introduction<br />
| | Florian Gremme, Young Generation Network,<br />
KTG, Germany<br />
Post-Test Analysis of the RPV Lower Head Leak<br />
Experiment at the INKA Test Facility Using<br />
ATHLET<br />
| | Michael Sporn, TU Dresden, Institute of Power<br />
Engineering, Germany<br />
New Thermal Hydraulic Development<br />
and Findings<br />
Chair:<br />
| | Dr. Sanjeev Gupta<br />
Keynote<br />
International Cooperation in the Experimental<br />
Field of Nuclear Thermohydraulics: Primary<br />
Coolant Loop Test Facility (PKL)<br />
| | N.N., OECD, France<br />
Keynote<br />
International Cooperation<br />
on Pool Scrubbing Research:<br />
Examples of NUGENIA/IPRESCA Project<br />
| | Dr. Sanjeev Gupta, Becker Technologies GmbH,<br />
Germany<br />
Application of an Eulerian/Eulerian<br />
CFD Approach to Simulate the<br />
Thermohydraulics of Rod Bundles<br />
| | Dr. Wei Ding, Helmholtz Zentrum Dresden<br />
Rossendorf, Germany<br />
Analysis of the Fatigue of the Bolts in the<br />
Flange of a Reactor Pressure Vessel<br />
| | Fabian Gottlieb, Kraftanlagen Heidelberg GmbH,<br />
Technical Analysis, Germany<br />
Preliminary Results of Water Hammer<br />
Simulation in Two-Phase Flow Regimes<br />
Using the Code ATHLET 3.1A<br />
| | Christoph Bratfisch, Ruhr-Universität Bochum,<br />
Germany<br />
Design of Simplified and Optimized Heavy<br />
Liquid Metal Loop for Future Applications<br />
| | Dr.-Ing. Nader Ben Said, Westinghouse Electric<br />
Germany GmbH, Germany<br />
Investigation on Variation of Nodelized<br />
Macroscopic Cross Sections Driven by<br />
Deflection of Fuel assemblies with Serpent<br />
| | Nico Bernt, Technische Universität Dresden (TUD),<br />
Germany<br />
PWR Cycle Analysis With the GRS Core<br />
Simulator KMACS<br />
| | Dr. Matías Zilly, Gesellschaft für Anlagen- und<br />
Reaktorsicherheit (GRS) gGmbH, Germany<br />
Application of a Full-Core Statistical Approach<br />
in LB-LOCA Analysis<br />
| | Dr. Andreas Wensauer, PreussenElektra GmbH,<br />
Germany<br />
Nuclear Data Uncertainty Analyses<br />
with XSUSA and MCNP<br />
| | Dr. Winfried Zwermann, Gesellschaft für Anlagenund<br />
Reaktorsicherheit (GRS) gGmbH, Germany<br />
Workshop<br />
Young Scientists' Workshop<br />
Tuesday ı 29 th May <strong>2018</strong><br />
Wednesday ı 30 th May <strong>2018</strong><br />
Coordinator:<br />
| | Prof. Dr.-Ing. Jörg Starflinger,<br />
University of Stuttgart, Germany<br />
Jury:<br />
| | Prof. Dr. Marco K. Koch,<br />
Ruhr-Universität Bochum<br />
| | Prof. Dr. Jörg Starflinger, University of Stuttgart,<br />
Institut für Kernenergetik und Energiesysteme<br />
(IKE)<br />
| | Dr. Wolfgang Steinwarz<br />
| | Dr. Katharina Stummeyer, Gesellschaft<br />
für Anlagen- und Reaktorsicherheit (GRS) gGmbH<br />
Prize awarded by:<br />
| | GNS Gesellschaft für Nuklear-Service mbH and<br />
Forschungsinstitut für Kerntechnik und Energiewandlung<br />
e. V.<br />
Introducing of the<br />
Young Generation Network<br />
| | Yvonne Schmidt-Wohlfarth, Young Generation<br />
Network, KTG, Germany<br />
Nuclear Technology in and<br />
Beyond our Daily Lifes<br />
| | N.N.<br />
Working in NPPs<br />
| | Sebastian Hahn, Young Generation Network, KTG,<br />
Germany<br />
Radioactivity and Radiation Protection<br />
| | Sven Jansen, VKTA – Strahlenschutz, Analytik &<br />
Entsorgung Rossendorf e. V., Germany<br />
Final Disposal of Radioactive Waste<br />
| | Dr. Thilo von Berlepsch (BGE), Germany<br />
Nuclear Fusion<br />
| | André Häußler, Elena Nunnemann, Karlsruhe<br />
Institute of Technology (KIT), Germany<br />
Stations of Nuclear Campus<br />
1 NPPs & Decommissioning<br />
2 Electricity Market – Composition<br />
of the Electricity Price<br />
3 Packaging, Casks & Conditioning of Waste<br />
4 Nuclear Medicine Applications<br />
Modeling of Post-Dryout Heat Transfer<br />
| | Dali Yu, Karlsruhe Institute of Technology (KIT),<br />
Germany<br />
Detailed session programme to be announced.<br />
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Key Topic<br />
Enhanced Safety &<br />
Operation Excellence<br />
Focus Sessions<br />
Tuesday ı 29 th May <strong>2018</strong><br />
Radiation Protection during<br />
decommissioning – are there special<br />
needs?<br />
Coordinators:<br />
| | Eric Baumann, New NP GmbH, Germany<br />
Dr. Angelika Bohnstedt, Karlsruhe Institute of<br />
Technology (KIT), Germany<br />
The first area of the session addresses protection of<br />
personnel engaged in decommissioning activities.<br />
Decommissioning activities are associated with<br />
large changes in the NPP configuration. The overall<br />
remaining radioactive inventory will shift. Systems<br />
are taken out of service. Originally confined radioactive<br />
sources become open sources due to the fact<br />
that systems are decommissioned. Demolition of<br />
civil structures and other unit elements require<br />
additional technical systems to cope with air<br />
contamination.<br />
The second area of discussion deals with clearance of<br />
radioactive material and public acceptance. A large<br />
amount of medium and low level waste has to be<br />
examined and removed from the site. This topic<br />
also addresses the question of “conditional” and<br />
“ unconditional” release of components and material.<br />
It is not only a question about clearance levels but<br />
also about public acceptance of receiving cleared, i.e.<br />
non-nuclear, waste at normal landfill sites.<br />
To all decommissioning activities, the ALARA<br />
approach applies. German rules and regulations,<br />
e.g. Radiation Protection Ordinance, various KTA<br />
rules, and international rules (BSS) and recommendations<br />
issued e.g. by IAEA or EU provide an appropriate<br />
framework for workers protection. Is there a<br />
need for specific “German decommissioning rules”?<br />
The final closure of a site requires the removal of all<br />
material. The largest amount originating from the<br />
demolition of buildings is non-nuclear waste. Some<br />
amount of waste has gone through the clearance.<br />
Some amount of waste was never subject to nuclear<br />
regulatory surveillance because it originates from<br />
office buildings, cooling towers, turbine buildings<br />
(in PWR plants), pumping station structures and<br />
others. Beside construction waste, valuable raw<br />
materials are extracted – e.g. copper from electrical<br />
cables. How is the public acceptance of “evil stuff”<br />
from a nuclear power plant?<br />
This session tries addressing some of these questions<br />
and tries providing some answers. Some of the<br />
presentation will give an interesting introduction<br />
and the answer might be gained during lively<br />
discussions between the session participants.<br />
Detailed session programme to be announced.<br />
International Operational Experience<br />
Coordinator:<br />
| | Dr.-Ing. L. Mohrbach, VGB PowerTech e.V.,<br />
Germany<br />
The operation of nuclear power plants involves a<br />
wide scope of specialized areas of expertise, from<br />
materials to human factors. Beyond daily business,<br />
some background information from different fields<br />
of operational activities might not only be regarded<br />
as personally worthwhile but may also be well suited<br />
to complement the general knowledge base for<br />
nuclear.<br />
This session addresses some of these questions<br />
and tries providing some answers. Some of the<br />
presen tations will give an introduction and produce<br />
questions. The answer might be gained during lively<br />
discussions between the session participants.<br />
Summary of the QUENCH LOCA<br />
Experimental Program<br />
| | Dr. Andreas Wensauer, PreussenElektra GmbH,<br />
Germany<br />
Practical Safeguards in Nuclear Power Plants<br />
| | Dr. Irmgard Niemeyer, Dipl.-Ing. Katharina Aymanns,<br />
Forschungszentrum Jülich GmbH, Germany (TBC)<br />
Comparison of Employment Effects<br />
of Low-Carbon Generation Technologies<br />
| | Geoffrey Rothwell, OECD Nuclear Energy Agency<br />
(NEA), France<br />
Application of Lubricants and<br />
other Consumables in Nuclear Power Plants<br />
| | Dr. Fred Böttcher, EnBW Kernkraft GmbH;<br />
Dr. Dittmar Rutschow, VGB PowerTech e. V.,<br />
Germany<br />
New Developments in Radiation Protection<br />
| | N.N.<br />
Benefits of Simulator Training<br />
| | N.N., KSG Kraftwerks-Simulator- Gesellschaft mbH,<br />
GfS Gesellschaft für Simulatorforschung mbH,<br />
Germany<br />
Technical Sessions<br />
Wednesday ı 30 th May <strong>2018</strong><br />
Operation and Safety<br />
of Nuclear Installations<br />
Chair:<br />
| | Dr. Thorsten Hollands<br />
Keynote Coordinator:<br />
| | Dr. Erwin Fischer<br />
Chair:<br />
| | Dr. Thorsten Hollands<br />
Keynote<br />
Safe to the Last Day – A Challenge for Operators<br />
| | Christoph Heil, EnBW Kernkraft GmbH, Executive<br />
Director, Germany<br />
Keynote<br />
Is Safety Culture Perceptible and Measurable?<br />
| | Uwe Stoll, Gesellschaft für Anlagen- und<br />
Reaktorsicherheit (GRS) gGmbH, Scientific and<br />
Technical Director, Germany<br />
Keynote<br />
Preserving and Ensuring Competence<br />
and Motivation<br />
| | Dr. Frank Sommer, PreussenElektra GmbH,<br />
Head of CoC Operations, Germany<br />
Digital Transformation in Nuclear Industry –<br />
Focus: Site Applications<br />
| | Dr. Jan Leilich, New NP, IBGM Product Management,<br />
Germany<br />
Save to the Last Day – How to Manage the<br />
Complexity in a Multi-Year End of Life Process<br />
| | Prof. Dr. Rüdiger von Der Weth, Hochschule für<br />
Wirtschaft und Technik Dresden, Faculty of<br />
Business Administration, Germany<br />
Loca Scenario-Related Zinc Borate<br />
Precipitation Studies at Lab Scale<br />
| | Dr. Ulrich Harm, Technische Universität Dresden<br />
(TUD), Germany<br />
Simulation of Asymmetric Severe Accidents<br />
Using the Code System AC2<br />
| | Liviusz Lovasz, Gesellschaft für Anlagen- und<br />
Reaktorsicherheit (GRS) gGmbH, Germany<br />
Simulation of the Bundle Test QUENCH-07<br />
with the Severe Accident Analysis Codes<br />
ASTEC V2.1 and AC^2 – ATHLET CD<br />
| | Florian Gremme, Ruhr-Universität Bochum, Chair<br />
of Energy Systems and Energy Economics,<br />
Germany<br />
Sensitivity and Uncertainty Analysis<br />
of the MCCI Model Results in AC2/COCOSYS<br />
for the OECD-CCI3 Experiment<br />
| | Dr. Claus Spengler, Gesellschaft für Anlagen- und<br />
Reaktorsicherheit (GRS) gGmbH, Germany<br />
Dry Filter Method DFM 2.0 – The Newest<br />
Generation of Filtered Containment Venting<br />
System<br />
| | Dr. Peter Hausch, Caverion Deutschland GmbH,<br />
Business Unit Krantz, Germany<br />
3D Surface Radiation Dosimetry of<br />
a Nuclear-Chicago NH3 Neutron Howitzer<br />
| | Ahmet Ilker Topuz, Istanbul Technical University,<br />
Nuclear Energy, Turkey<br />
Chair:<br />
Dr. Jürgen Sydow<br />
TESPA-ROD Code Prediction of the Fuel Rod<br />
Behaviour During Long-Term Storage<br />
| | Dr. Heinz-Günther Sonnenburg, Gesellschaft für<br />
Anlagen- und Reaktorsicherheit (GRS), Germany<br />
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AMNT <strong>2018</strong><br />
Summary of Experimental Investigations<br />
at the ALADIN Test Facility<br />
for the Thermal Hydraulic Analysis<br />
of Accident Scenarios in Spent Fuel Pools<br />
| | Christine Partmann, Technische Universität<br />
Dresden (TUD), Germany<br />
Considerations for Multi Unit Effects<br />
in Probabilistic Risk Assessment<br />
| | Dr. Felix Philipp Sassen, Westinghouse Electric<br />
Germany GmbH, Germany<br />
Model-Based Vulnerability Analysis<br />
of Complex Infrastructures<br />
| | Mathias Lange, Hochschule Magdeburg- Stendal,<br />
Germany<br />
Canadian Nuclear Fire PRA<br />
| | Hossam Shalabi, Canadian Nuclear Safety<br />
Commission, Canada<br />
Key Topic<br />
Decommissioning<br />
Experience & Waste<br />
Management Solutions<br />
Focus Sessions<br />
Tuesday ı 29 th May <strong>2018</strong><br />
Post-operation and Decommissioning<br />
in Germany<br />
Coordinator:<br />
| | Dr. Erich Gerhards, PreussenElektra GmbH,<br />
Germany<br />
Preparing for Decommissioning – Meeting the<br />
Changing Requirements for Decommissioning<br />
| | N.N., OECD Nuclear Energy Agency (NEA)<br />
The Paradigm Shift in Nuclear Waste<br />
Management in Germany<br />
Coordinator:<br />
| | Michael Köbl, GNS Gesellschaft<br />
für Nuklear-Service mbH, Germany<br />
In summer 2017 the responsibilities for nuclear<br />
waste management in Germany have been fundamentally<br />
reorganized. While the operators remain<br />
responsible for the decommissioning and dismantling<br />
of their NPPs as well as the packaging of the<br />
nuclear waste, the German government assumes<br />
responsibility not only for final disposal, but additionally<br />
already for interim storage. This means that<br />
the waste pro ducers, who used to be obliged to store<br />
their HLW/ILW until the future availability of the federal<br />
repository “Konrad”, from now on can directly<br />
hand over their suitably packaged waste to the state<br />
owned interim storage facilities. This essentially new<br />
procedure poses huge challenges to the waste producers<br />
as well as to the authorities. It is the aim of<br />
this Focus Session to outline the new regulations and<br />
discuss the consequences for all the parties involved.<br />
TBA<br />
| | Responsible Authorities and Federal Corporations:<br />
BMUB, BfE, BGE, BGZ<br />
TBA<br />
| | Independent Experts<br />
TBA<br />
| | Waste Producers<br />
TBA<br />
| | Suppliers/Vendors<br />
Panel Discussion<br />
| | All Participants<br />
This session will be held in German<br />
with simultaneous English translation.<br />
Keynote<br />
Decommissioning and Waste Management of<br />
Obsolete Nuclear Research Facilities<br />
| | Dr. Vincenzo V. Rondinella, Joint Research Center<br />
(JRC) of the European Commission, Germany<br />
Keynote<br />
Global Status of Decommissioning<br />
| | Patrick J. O’Sullivan, International Atomic Energy<br />
Agency (IAEA), Austria<br />
Ventilation Concepts for Different Phases<br />
During Decommissioning of Nuclear Facilities<br />
| | Dirk Thybussek, Caverion Deutschland GmbH,<br />
Business Unit Krantz, Germany<br />
Bladecutter: A Novel Technology<br />
for Removing Nuclear Sludge<br />
| | Shuai Wang, The University of Manchester, School<br />
of Electrical and Electronic Engineering, United<br />
Kingdom<br />
Untersuchungen Zum Abtrag Asbesthaltiger<br />
Spachtelmasse Mittels Feuchtsandstrahlen<br />
| | Simone Müller, KIT – Rückbau konventioneller &<br />
kerntechnischer Bauwerke, Germany<br />
Das Ausschreibungsverfahren für<br />
den Abbau des Reaktordruckgefäßes und der<br />
RDG-Einbauten im Kernkraftwerk Lingen<br />
| | Stefan Lindemann, RWE Power AG, Germany<br />
How to Improve Decommissioning<br />
by Virtual Engineering Tools<br />
| | Prof. Dr. Ulrich W. Scherer, Hochschule Mannheim,<br />
Faculty Chemical Process Technology, Germany<br />
Characterizing the Radioactivity<br />
of the Concrete Shielding During<br />
Decommissioning of the LFR<br />
| | Perry Young, NRG, Research & Innovation – CP4S<br />
–, Netherlands<br />
Requirements on Operation and Decommissioning<br />
| | Dr. Heinz Drotleff, German Waste Management<br />
Commission, Germany<br />
The Role of Service Operation for Decommissioning<br />
– A Practitioner’s Experience<br />
| | Dr. Thomas Volmar, RWE Power AG, Germany<br />
Competencies and Ressources Required<br />
to Assure Safe Service Operation and<br />
Decommissioning<br />
| | A. Dinter, PreussenElektra GmbH, Germany<br />
Decommissioning and Service Operation<br />
in Sweden<br />
| | M. Bächler, UNIPER Technology, Germany<br />
Full Scope Approach – Hand over of<br />
Operations, Decommissioning, Dismantling<br />
and Waste Management<br />
| | Robert Bonner, AECOM, United Kingdom<br />
This session will be held in German<br />
with simultaneous English translation.<br />
Detailed session programme to be announced.<br />
Technical Sessions<br />
Wednesday ı 30 th May <strong>2018</strong><br />
Decommissioning Experience &<br />
Waste Management Solutions<br />
Chair:<br />
| | Martin Brandauer<br />
Keynote Coordinator:<br />
| | Thomas Seipolt<br />
Keynote<br />
Evaluation of Approaches to Automate<br />
Reactor Internals Segmentation/Evaluation<br />
of New or Enhanced Techniques<br />
for Concrete Decontamination<br />
| | PhD Richard Reid/Richard McGrath, The Electric<br />
Power Research Institute (EPRI), USA<br />
Application of the System FREMES<br />
to Characterize and Sort Soil During<br />
the Remediation of FBFCi Dessel<br />
Fuel Element Factory<br />
| | Felix Langer, NUKEM Technologies Engineering<br />
Services, O-P, Germany<br />
Design 3D, Laser Scanning and Radiological<br />
Data Visualization<br />
| | Sergi Milà, Westinghouse Electric Spain, Spain<br />
Full System Decontamination Project at Bohunice<br />
| | Randall Duncan, Westinghouse Electric Company,<br />
USA<br />
Under Water Cutting Technologies<br />
| | John Hubball, Westinghouse Electric Company,<br />
DDR&WM, USA<br />
Vorstellung eines Magnetfiltersystems<br />
zur Behandlung von Sekundärabfällen der<br />
Wasser-Abrasiv-Suspensions-Schneidtechnik<br />
| | Carla Krauß, Karlsruhe Institute of Technology<br />
(KIT), Germany<br />
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Decommissioning Characterisation Through<br />
Compressive Gamma-Ray Imaging<br />
| | Dr. David Boardman, ANSTO,<br />
Nuclear Stewardship, Australia<br />
AuDeKa: A BMBF Funded Project to Develop<br />
an Automated Deconta mination Cabin with<br />
Documentation Based on Industry 4.0 Features<br />
| | Franz Borrmann, Institut für Umwelt technologien<br />
& Strahlenschutz GmbH, Germany<br />
Lean Documentation Approach<br />
in Decommissioning<br />
| | Franz Borrmann, Institut für Umwelt technologien<br />
& Strahlenschutz GmbH, Germany<br />
Activation Analysis, Validation and Component-<br />
Wise Packaging Concept for the Decommission<br />
Planning of the Gundremmingen NPP<br />
| | Dr. Ben Volmert, Nagra, Inventory & Logistics,<br />
Switzerland<br />
Primary Circuit Decontamination<br />
at Biblis Unit A NPP<br />
| | Markus Thoma, Siempelkamp NIS<br />
Ingenieurgesellschaft mbH, Germany<br />
Rückbau und Entsorgung der Reaktordruckbehälter-Einbauten<br />
und der RDBs<br />
der Kernkraftwerke Philippsburg 1 (KKP 1)<br />
und Neckarwestheim I (GKN I)<br />
| | Dr. Bernhard Wiechers, Westinghouse Electric<br />
Germany GmbH, Decommis sioning, Dismantlling<br />
& Remediation, Germany<br />
This session can be held in German/English<br />
with simultaneous translation.<br />
Radioactive Waste Management,<br />
Storage and Disposal<br />
Chair:<br />
| | Dr. Alexander Zulauf<br />
Keynote Coordinator:<br />
| | Iris Graffunder<br />
Keynote<br />
Challenges in the Management of Concrete<br />
Waste from the Dismantling of Nuclear<br />
Facilities – Case Study Rheinsberg NPP<br />
| | Jörg Möller, EWN Entsorgungswerk<br />
für Nuklearanlagen GmbH, Germany<br />
Keynote<br />
Managing Waste at the Remote- handled<br />
Dismantling of Activated Concrete and Steel<br />
Structures of the Biological Shield of KNK<br />
| | Johannes Rausch, KTE Kerntechnische Entsorgung<br />
Karlsruhe GmbH, Germany<br />
Keynote<br />
Clearance Measurement of Demolition Waste:<br />
Measurement Process with High Operational<br />
Throughput<br />
| | Stefan Thierfeldt, Brenk Systemplanung GmbH,<br />
Germany<br />
Nuclear Energy and Society<br />
Engaging with Society – Past, Present and<br />
Future. Results From the Collabo rative<br />
Interdisciplinary Project HoNESt – History of<br />
Nuclear Energy and Society<br />
| | Dr. Jan-Henrik Meyer, University of Copenhagen,<br />
Saxo Institute, Denmark<br />
Waste Treatment<br />
Fortum NURES®-BORES Concept of Treating<br />
Liquid Radioactive Waste Containing Boron<br />
| | Dr. Jussi-Matti Mäki, Fortum Power and Heat Oy,<br />
Nuclear Services, Finland<br />
Investigations of Process Parameters Using<br />
Microwave Technology for the Treatment of<br />
Radioactive Waste<br />
| | Prof. Dr. Ulrich W. Scherer, Hochschule Mannheim,<br />
Faculty Chemical Process Technology, Germany<br />
Characterization<br />
Advantages and Limits of Spectroscopic<br />
Measurement for the Classification<br />
of Radioactive Wastes<br />
| | Dr. Marina Sokcic-Kostic, NUKEM Technologies<br />
Engineering Services, Engineering, Germany<br />
Waste Management<br />
Development of a Calculation Tool<br />
for Optimal Holistic Disposal Planning<br />
| | Dr. Anton Anthofer, VPC GmbH, Germany<br />
Endlagerdokumentation Neu Gedacht<br />
| | Dr. Anton Anthofer, VPC GmbH, Germany<br />
Development of a Monitoring Concept<br />
for Transport and Storage Containers<br />
for Spent Fuel and Heat-Generating<br />
High-Level Radioactive Waste on Prolonged<br />
Intermediate Storage<br />
| | Daniel Fiß, Hochschule Zittau/Görlitz, Germany<br />
Use of a Statistical Toolset for Risk Aware<br />
Package Planning of Activated Core Internals<br />
| | Dr. Maarten Becker, Institut für Umwelttechnologien<br />
& Strahlenschutz GmbH, Germany<br />
Use of Flexible Packaging and Real Time Assay<br />
Techniques to Divert Low Activity Waste LLW<br />
from the UK LLWR Facility<br />
| | Ian Wigginton, Nuvia Ltd, Waste & Environment,<br />
United Kingdom<br />
Packaging<br />
MOSAIK Casks for Transport, Storage and Final<br />
Disposal of All Kinds of Intermediate Level<br />
Waste – A Success Story Spanning More Than<br />
Three Decades<br />
| | Dr. Jörn Becker, GNS Gesellschaft für<br />
Nuklear-Service mbH, Technik, Germany<br />
One Cask Fits All – The New MOSAIK® II-S<br />
for All Kinds of Intermediate Level Waste<br />
| | David Bergandt, GNS Gesellschaft<br />
für Nuklear-Service mbH, TP2 Project<br />
Management, Germany<br />
GNS SBoX® A New Family of Robust,<br />
Self-Shielded Containers<br />
| | Martin Beverungen, GNS Gesellschaft<br />
für Nuklear-Service mbH, Germany<br />
Automated Ultrasonic Testing of CASTOR®<br />
Cask Bodiesin Serial Production –<br />
A Progress Report<br />
| | Jörg Frank, GNS Gesellschaft für Nuklear-Service<br />
mbH, Cask Manufacturing (Orders), Germany<br />
Quivers for Non Standard Fuel Rods –<br />
Advances and First Utilizations<br />
| | Olga Di Paola, GNS Gesellschaft für Nuklear-<br />
Service mbH, Germany<br />
Experiences in the Assessment<br />
of a Dual Purpose Transport Cask Loaded<br />
with Damaged Spent Nuclear Fuel<br />
| | Dr. Thorsten Schönfelder, Bundesanstalt für<br />
Materialforschung und -prüfung (BAM), Germany<br />
Preliminary Experimental Study on Reduction<br />
of Hydrogen Concentration in a Small- Scale<br />
Radioactive Waste Long-Term Storage<br />
Container with Catalysts<br />
| | Prof. Dr. Kazuyuki Takase, Nagaoka University<br />
of Technology, Japan<br />
Simulation-Based Investigation<br />
of Suitability of Thermography and Muon Flux<br />
Measurements for Non-Invasive Monitoring<br />
of Transport and Storage Containers<br />
for Spent Fuel<br />
| | Michael Wagner, Technische Universität Dresden<br />
(TUD), Germany<br />
Repository<br />
Entsorgung von Wärme Entwickelnden<br />
Radioaktiven Abfällen – Herausforderungen<br />
und Lösungsansätze<br />
| | Matthias Bode, Leibniz Universität Hannover,<br />
Germany<br />
This session can be held in German/English<br />
with simultaneous translation.<br />
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KTG INSIDE<br />
Fachgruppe Reaktorsicherheit:<br />
Vorstand neu aufgestellt<br />
Inside<br />
Dr. Tatiana Salnikova hat den Vorsitz der KTG Fachgruppe<br />
„Reaktorsicherheit“ von Uwe Stoll erfolgreich übernommen.<br />
Am Moskauer Energetischen Institut studierte Dr. Salnikova<br />
zunächst Umwelttechnik. Im Anschluss daran wechselte sie<br />
zum Kerntechnikstudium an die Hochschule Zittau/Görlitz.<br />
Ihre Promotion im Bereich der thermohydraulischen<br />
Modellierung von Brennelementen mithilfe numerischer<br />
Methoden erfolgte in Kooperation zwischen der TU Dresden<br />
und AREVA NP. Im Jahr 2007 startete Tatiana Salnikova als<br />
Projektleiterin bei der AREVA GmbH. Ihre Arbeitsschwerpunkte<br />
liegen heute auf dem Gebiet der nuklearen Sicherheit.<br />
Dazu gehören die Erstellung von Sicherheitsanalysen<br />
für KKW, die Mitarbeit in nationalen und internationalen<br />
Gremien wie Reaktor-Sicherheitskommission (RSK), International<br />
Atomic Energy Agency (IAEA), und Electric Power<br />
Research Institute (EPRI). Derzeit beschäftigt sie sich unter<br />
anderem mit Fragestellungen zur Lastwechselfahrweise<br />
von KKW. Ebenfalls seit diesem Jahr hat es bei der Position<br />
des Kassenwartes einen Wechsel von Dr. Walter Tromm zu<br />
Dr. Frank Sommer gegeben. Dr. Frank Sommer ist seit<br />
2013 Bereichsleiter für das Kompetenzcenter Betrieb der<br />
PreussenElektra GmbH in Hannover. Er studierte Maschinenbau<br />
an der Ruhr-Universität in Bochum und promovierte<br />
dort im Anschluss am Lehrstuhl für Strömungstechnik. Seit<br />
1992 ist Frank Sommer in verschiedenen Funktionen<br />
bei PreussenElektra bzw. ihren Vorgängerunternehmen beschäftigt.<br />
Für die geleistete Arbeit bedanken wir uns herzlich bei<br />
den Amtsvorgängern.<br />
Dr. Angelika Bohnstedt (KIT) als stellvertretende Fachgruppensprecherin<br />
gewählt. Herzlichen Glückwunsch.<br />
Erik Baumann<br />
Sprecher der Fachgruppe Strahlenschutz<br />
6. Bilaterales Treffen WiN Schweden<br />
und WiN Germany<br />
26./27. Oktober 2017 – Informationszentrum<br />
Kernkraftwerk Biblis<br />
Bereits zum sechsten Mal trafen sich schwedische und<br />
deutsche Women in Nuclear (WiN) zum Erfahrungsaustausch.<br />
Nach Oskarshamn im April 2016 lud Deutschland<br />
am 26./27.Oktober 2017 nach Biblis ein – das Kernkraftwerk<br />
Biblis war neben dem bilateralen Treffen auch Gastgeber<br />
für die Mitgliederversammlung von WiN Germany<br />
2017.<br />
Dr. Tatiana Salnikova<br />
(Sprecherin der KTG Fachgruppe Reaktorsicherheit)<br />
und Dr. Frank Sommer<br />
(Kassenwart der KTG Fachgruppe Reaktorsicherheit)<br />
| | „Insgesamt sind wir gut aufgestellt, um das Rückbauprojekt erfolgreich<br />
durchzuführen – es ist gut, dass der Rückbau jetzt begonnen hat“,<br />
resümiert Kemmeter am Ende seines Vortrags.<br />
Fachgruppe Strahlenschutz:<br />
Jahresrückblick 2017<br />
Der Schwerpunkt der Tätigkeit der KTG Fachgruppe<br />
Strahlenschutz lag 2017 in der Vorbereitung und Durchführung<br />
der Focus Session Radiation Protection im Rahmen<br />
des gemeinsam von der KTG e.V: und dem DAtF e.V. veranstalteten<br />
48. Annual Meeting on Nuclear Technology<br />
(AMNT <strong>2018</strong>, Jahrestagung Kerntechnik).<br />
Die Focus Session ist seit 2015 fester Bestandteil im Programm<br />
der AMNT. Durch die gemeinsame Anstrengung<br />
der Mitglieder der Fachgruppe gelang es auch 2017 eine<br />
interessante Focus Session mit dem Thema „Radiation<br />
Protection – What about the basic principles and objectives<br />
in the current regulatory environment?“ zu gestalten. Die<br />
Berichterstattung dazu findet sich in der Ausgabe 1 (<strong>2018</strong>)<br />
der <strong>atw</strong>.<br />
Am Rande der Jahrestagung Kerntechnik fand eine<br />
Versammlung der Fachgruppe Strahlenschutz statt, zu der<br />
alle Mitglieder vorab per E-Mail eingeladen worden waren.<br />
Ein wesentlicher Tagesordnungspunkt bestand in der Wahl<br />
eines neuen Stellvertreters, da der bisherige Stellvertreter,<br />
Herr Sinisa Simic nicht mehr zur Verfügung steht. Einstimmig<br />
wurde von den anwesenden Mitgliedern<br />
Der große Dank für die Einladung und finanzielle<br />
Unterstützung wurde seitens der WiNner persönlich dem<br />
Gastgeber Horst Kemmeter – Leiter des Kernkraftwerkes<br />
Biblis – überbracht, der in seinem Einführungsvortrag den<br />
Stand der Rückbauaktivitäten des KKW Biblis vorstellte<br />
und das Motto des WiN-Treffens The long way to green field<br />
durchaus passend für den Standort Biblis fand.<br />
Nach Besichtigung des Standortzwischenlager (SZL), in<br />
dem Castor®- und Mosaik-Behälter lagern, sowie der Baustelle<br />
des neu entstehenden LAW-II-Lagers (Low Active<br />
Waste-Lager) fasste Martina Etzmuß (Preussen Elektra) im<br />
Rahmen des offiziellen Vortragsprogrammes die politische<br />
Situation in Deutschland insbeson dere nach dem Erd beben<br />
und Tsunami in Japan und der sofortigen Still legung von 8<br />
Kraftwerksblöcken zusammen.<br />
Maria Taranger (Barsebäck AB) stellt die politischen<br />
Rahmenbe dingungen in Schweden vor: Das National<br />
Energy Agreement vom Juni 2015 hat zumindest für eine<br />
mittelfristige Sicherheit gesorgt, denn eine Stilllegung von<br />
KKWs aus politischen Gründen ist danach nicht mehr<br />
möglich.<br />
Das schützt jedoch nicht vor wirtschaftlichen Entscheidungen,<br />
so wie sie in Ringhals 1 und 2 von Vattenfall<br />
im letzten Jahr mit vorzeitiger Abschaltung getroffen<br />
wurden. Anna Collin (Ringhals AB) berichtete vom Projekt<br />
KTG Inside
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
STURE, mit dem die Stilllegung der beiden Ringhals-<br />
Blöcke geregelt ist. Gleichzeitig sollen die Blöcke 3 und 4<br />
sicher bis 2045 weiter betrieben werden. Dies bedeute ein<br />
starker Fokus auf den sogenannten Human Factor, wie<br />
Mitarbeiterqualifikationen und Flexibilität, so Collin.<br />
Katarina Andersson (OKG) und Maria Taranger (BKAB)<br />
stellten in einer gemeinsamen Präsentation die Rückbauaktivitäten<br />
von Barsebäck 1 und 2 sowie Rückbauplanungen<br />
von Oskarsham 1 und 2 vor. An vielen Stellen<br />
profitiert man von der guten Zusammenarbeit, trotzdem<br />
gäbe es standortspezifische Anforderungen.<br />
Strategische Aspekte des Abfallmanagements wurden<br />
von Sofia Eliasson (OKG) vorgestellt.<br />
Katrin Hertkorn-Kiefer (RWE) trug Einzelheiten zu den<br />
Rückbauprojekten der RWE vor und stellte fest, dass das<br />
Konzept des sicheren Einschlusses für Biblis keine Option<br />
gewesen wäre, es sei der Öffentlichkeit nicht mehr zu<br />
vermitteln.<br />
In die abschließende Diskussionsrunde Are we well<br />
prepared for dismantling? führte Martina Sturek (SKB) mit<br />
ihrem Vortrag zum schwedischen Entsorgungskonzept<br />
ein. In Schweden sind die Betreiber der Kernkraftwerke<br />
für die Entsorgung und Endlagerung verantwortlich.<br />
Sie haben hierfür die gemeinsame Gesellschaft Svensk<br />
Kärnbränslehantering AB (SKB) gegründet, die auch für<br />
Transporte und Zwischenlagerung zuständig ist. Der<br />
hochradioaktive Abfall soll im Wirtsgestein Granit im<br />
Endlager Forsmark gelagert werden.<br />
Dr. Christiane Vieh (BGE) vermittelte Eindrücke von der<br />
Verantwortung der BGE für die Endlager in Deutschland.<br />
Mit der Neugründung von zwei bundeseigenen Gesellschaften,<br />
der Bundesgesellschaft für Endlagerung (BGE)<br />
und der Bundesgesellschaft für Zwischenlagerung (BGZ)<br />
übernimmt die Bundesrepublik Deutschland die Verantwortung<br />
für die Zwischen- und Endlagerung von<br />
radioaktiven Abfällen. Hingegen sind die Betreiber der<br />
Kernkraftwerke weiterhin für den Rückbau Ihrer Anlagen<br />
nach der Stilllegung zuständig.<br />
In Deutschland ist die Suche nach einem Endlager für<br />
wärmeentwickelnde radioaktive Abfälle in vollem Gange<br />
und die Entscheidung für einen Standort wird im Jahr<br />
2031 erwartet. Der Schacht Konrad, ein stillgelegtes<br />
Eisenerz-Bergwerk, wird derzeit zum Endlager für radioaktive<br />
Abfälle mit vernachlässigbarer Wärmeentwicklung<br />
umgerüstet.<br />
Gabi Voigt – ja, die aktuelle WiN Global Präsidentin ist<br />
Mitglied von WiN Germany – berichtete unter anderem<br />
stolz, dass WiN-Global seit dem 20. August 2017 als NGO<br />
registriert wurde. Dies war eine notwendige Formalie für<br />
| | Martina Sturek, WiN Präsidentin von WiN Schweden – im Bild links mit<br />
WiN Germany Präsidentin Jutta Jené rechts) sowie WiN Global Präsidentin<br />
Gabi Voigt (Mitte) – hat ein Treffen voraussichtlich im November <strong>2018</strong> im<br />
Kernkraftwerk Ringhals angekündigt.<br />
| | Gabi Voigt, aktuelle WiN Global Präsidentin<br />
die WiN-Global Konferenz Ende August 2017 in China und<br />
wurde mit hohem Aufwand noch fristgerecht erreicht.<br />
Gabi blickt auf viele Aktivitäten im Jahr 2017 zurück und<br />
stellte fest, dass es sehr viel mehr Arbeit gewesen sei, als sie<br />
erwartet habe. Ziele für das kommende Jahr sind u.a. eine<br />
stärkere Präsenz in den sozialen Medien und dass verschiedene<br />
Konzepte der Zusammenarbeit (Memorandum<br />
of Understanding) mit anderen Organisationen wie WNA,<br />
ICRP, IRPA oder INYG mit Leben gefüllt würden.<br />
Verleihung des WiN Germany Preises 2017<br />
im Rahmen des bilateralen Treffens<br />
Zum ersten Mal in der Geschichte von WiN Germany e.V.<br />
fand die Präsentation der eingereichten wissenschaft lichen<br />
Arbeit für den WiN Germany Preis im Rahmen des bilateralen<br />
Treffens statt. Larissa Klaß, die zurzeit ihre Doktorarbeit<br />
am Forschungszentrum Jülich schreibt, trug aus ihrer<br />
Masterarbeit zum Thema Modified diglycolamides for a<br />
selective separation of Am (III): complexation, structural investigations<br />
and possible application vor. Das Fachwissen und<br />
die Eloquenz von Larissa beim Vortrag einschließlich ihrer<br />
55<br />
KTG INSIDE<br />
| | 19 WiNners aus Schweden und 24 WiNners aus Deutschland trafen sich beim 6. bilateralen Treffen der<br />
beiden WiN-Chapter am Standort Biblis<br />
| | WiN-Präsidentin Jutta Jené gratuliert Larissa Klaß – eine würdige<br />
WiN-Preisträgerin, die sich über die für sie einmalige Gelegenheit freute,<br />
vor einem ausschließlich weiblichen Publikum vortragen zu können.<br />
KTG Inside
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
56<br />
KTG INSIDE<br />
KTG Inside<br />
Verantwortlich<br />
für den Inhalt:<br />
Die Autoren.<br />
Lektorat:<br />
Sibille Wingens,<br />
Kerntechnische<br />
Gesellschaft e. V.<br />
(KTG)<br />
Robert-Koch-Platz 4<br />
10115 Berlin<br />
T: +49 30 498555-50<br />
F: +49 30 498555-51<br />
E-Mail: s.wingens@<br />
ktg.org<br />
www.ktg.org<br />
souveränen Antworten auf fachliche Detailnachfragen<br />
überzeugten das gesamte deutsch-schwedische Auditorium.<br />
Kurz vor Ende der diesjährigen Veranstaltung traf eine<br />
äußerst erfreuliche Nachricht von URENCO Deutschland<br />
GmbH ein: URENCO sponsort den WiN Germany Award mit<br />
1.500 Euro, da dem Unternehmen die Förderung von Frauen<br />
in der Kerntechnik am Herzen liegt und Bestandteil<br />
Herzlichen<br />
Glückwunsch<br />
Januar <strong>2018</strong><br />
91 Jahre wird<br />
1. Prof. Dr. Werner Oldekop,<br />
Braunschweig<br />
89 Jahre wird<br />
20. Dr. Devana Lavrencic-Cannata, Rom/I<br />
88 Jahre wird<br />
10. Dipl.-Ing. Hans-Peter Schmidt,<br />
Weinheim<br />
87 Jahre wird<br />
12. Dr. Rolf Hueper, Karlsruhe<br />
86 Jahre wird<br />
3. Dipl.-Ing. Fritz Kohlhaas, Kahl/Main<br />
85 Jahre wird<br />
9. Prof. Dr. Hellmut Wagner, Karlsruhe<br />
83 Jahre werden<br />
10. Dipl.-Ing. Walter Diefenbacher,<br />
Karlsruhe<br />
17. Dipl.-Ing. Helge Dyroff, Alzenau<br />
24. Theodor Himmel, Bad Honnef<br />
82 Jahre werden<br />
5. Obering. Peter Vetterlein, Oberursel<br />
23. Prof. Dr. Hartmut Schmoock,<br />
Norderstedt<br />
30. Dipl.-Phys. Wolfgang Borkowetz,<br />
Rüsselsheim<br />
30. Dipl.-Ing. Friedrich Morgenstern,<br />
Essen<br />
81 Jahre werden<br />
7. Dipl.-Ing. Albrecht Müller,<br />
Niederrodenbach<br />
9. Dipl.-Ing. Werner Rossbach,<br />
Bergisch Gladbach<br />
25. Dipl.-Ing. (FH) Heinz Wolf,<br />
Philippsburg<br />
80 Jahre werden<br />
7. Dipl.-Ing. Manfred Schirra, Stutensee<br />
8. Dipl.-Ing. Wolfgang Repke, Waldshut<br />
10. Dr. Dieter Türck, Dieburg<br />
12. Dipl.-Ing. Hans Dieter Adami, Rösrath<br />
18. Dr. Werner Katscher, Jülich<br />
22. Dr. Frank Müller, Erlangen<br />
79 Jahre werden<br />
11. Dipl.-Ing. Gerwin H. Rasche, Hasloch<br />
13. Dr. Udo Wehmann, Hildesheim<br />
16. Dr. Wolfgang Kersting, Blieskastel<br />
21. Prof. Dr. Detlef Filges, Langerwehe<br />
28. Dr. Sigwart Hiller, Lauf<br />
78 Jahre wird<br />
4. Dipl.-Ing. Wolfgang Semenau,<br />
Laudenbach<br />
77 Jahre werden<br />
3. Dipl.-Ing. Ferdinand Wind,<br />
Tettnang-Burgermoos<br />
12. Dr. Hand-G. Bogensberger,<br />
Anthem/USA<br />
15. Dipl.-Ing. Ulf Rösser,<br />
Heiligkreuzsteinach<br />
26. Dr. Heinrich Pierer von Esch, Erlangen<br />
76 Jahre werden<br />
6. Dipl.-Ing. Günter Höfer, Mainhausen<br />
31. Dipl.-Phys. Werner Scholtyssek,<br />
Stutensee<br />
75 Jahre werden<br />
19. Dr. Gerd Habedank,<br />
Seeheim-Jugenheim<br />
24. Dr. Günter Bäro Weinheim<br />
70 Jahre wird<br />
20. Dipl.-Ing. Edgar Bogusch, Erlangen<br />
60 Jahre werden<br />
7. Rüdiger König, Essen<br />
19. Dipl.-Ing. Erwin Neukäter, Sugiez/CH<br />
50 Jahre werden<br />
12. Dipl.-Phys. Karl Froschauer,<br />
Freigericht-Somborn<br />
19. Dipl.-Ing. Sönke Holländer, Essen<br />
21. Dipl.-Ing. Torsten Fricke, Hohnstorf<br />
Februar <strong>2018</strong><br />
90 Jahre wird<br />
10. Dipl.-Ing. Hans-Peter Schabert,<br />
Erlangen<br />
89 Jahre wird<br />
20. Dr. Helmut Hübel, Bensberg<br />
88 Jahre wird<br />
5. Dr. Eberhard Teuchert,<br />
Leverkusen<br />
ihres Nachhaltigkeitsprogrammes ist. Damit ist nicht nur<br />
der diesjährige Preis finanziert, sondern auch die Vergabe<br />
des WiN-Preises <strong>2018</strong> ist gesichert. Entsprechend groß viel<br />
der Beifall aus. WiN Germany sagt herzlichen Dank an<br />
URENCO Deutschland für die großzügige Spende und hofft<br />
auf Nachahmer!<br />
87 Jahre wird<br />
14. Dipl.-Ing. Heinrich Kahlow,<br />
Rheinsberg<br />
85 Jahre wird<br />
11. Dr. Rudolf Büchner, Dresden<br />
Yvonne Broy<br />
84 Jahre werden<br />
9. Dr. Horst Keese, Rodenbach<br />
12. Dipl.-Ing-. Horst Krause, Radebeul<br />
23. Prof. Dr. Dr.-Ing. E.h. Adolf Birkhofer,<br />
Grünwald<br />
82 Jahre werden<br />
6. Dr. Ashu-T. Bhattacharyya, Erkelenz<br />
17. Dr. Helfrid Lahr, Wedemark<br />
81 Jahre werden<br />
5. Prof. Dr. Arnulf Hübner, Berlin<br />
6. Dipl.-Ing. Heinrich Moers,<br />
Maitland/USA<br />
11. Dr. Günter Keil, Sankt Augustin<br />
18. Dipl.-Ing. Hans Wölfel, Heidelberg<br />
21. Dipl.-Ing. Hubert Andrae, Rösrath<br />
80 Jahre werden<br />
15. Dr. Hans-Heinrich Krug, Saarbrücken<br />
27. Dr. Klaus Wolfert, Ottobrunn<br />
79 Jahre werden<br />
3. Dr. Roland Bieselt, Kürten<br />
8. Dr. Joachim Madel, St. Ingbert<br />
8. Dr. Herbert Spierling, Dietzenbach<br />
22. Dr. Manfred Schwarz, Dresden<br />
<br />
28. November 2017<br />
Dipl.-Phys.<br />
Erich Neuburger<br />
Karlsruhe<br />
Die KTG verliert in ihm ein langjähriges<br />
aktives Mitglied, dem sie ein<br />
ehrendes Andenken bewahren wird.<br />
Ihren Familien gilt unsere Anteilnahme.<br />
KTG Inside
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
78 Jahre werden<br />
9. Dr. Gerhard Preusche,<br />
Herzogenaurach<br />
13. Dr. Hans-Ulrich Fabian, Gehrden<br />
14. Dipl.-Ing. Kurt Ebbinghaus,<br />
Bergisch Gladbach<br />
21. Dr. Jürgen Langeheine, Gauting<br />
23. Dr. Gerhard Heusener, Bruchsal<br />
25. Prof. Dr. Sigmar Wittig, Karlsruhe<br />
77 Jahre wird<br />
16. Dr. Jürgen Lockau, Erlangen<br />
76 Jahre werden<br />
6. Dr. Michael Schneeberger, Linz/A<br />
22. Cornelis Broeders, Linkenheim<br />
75 Jahre werden<br />
5. Dr. Joachim Banck, Heusenstamm<br />
9. Dr. Friedrich-Karl Boese, Leonberg<br />
13. Dr. Ingo-Armin Brestrich, Plankstadt<br />
20. Ing. Leonhard Irion, Rückersdorf<br />
28. Dr. Klaus Tägder, Sankt Augustin<br />
70 Jahre werden<br />
7. Dr. Hans-Hermann, Remagen<br />
8. Dr. Max Hillerbrand, Erlangen<br />
14. Reinhold Rothenbücher, Erlangen<br />
23. Dr. Rudolf Görtz, Salzgitter<br />
29. Dr. Anton von Gunten,<br />
Oberdiessbach<br />
65 Jahre werden<br />
3. Dr. Reinhard Knappik, Dresen<br />
20. Dipl.-Ing. Berthold Racky, Nidderau<br />
60 Jahre werden<br />
3. Prof. Dr. Sabine Prys, Offenburg<br />
3. Dipl.-Ing. Siegfried Wegerer,<br />
Tiefenbach<br />
10. Dipl.-Ing. (FH) Anton Hums,<br />
Essenbach<br />
50 Jahre werden<br />
5. Dr. Volker Wunder, Ottensoos<br />
20. Dr. Josef Engering, Jülich<br />
22. Toralf Wolf, Plauen<br />
28. Dipl.-Ing. Jörg Schneider, Radebeul<br />
Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag<br />
und wünscht ihnen weiterhin alles Gute!<br />
Wenn Sie keine<br />
Erwähnung Ihres<br />
Geburtstages in<br />
der <strong>atw</strong> wünschen,<br />
teilen Sie dies bitte<br />
rechtzeitig der KTG-<br />
Geschäftsstelle mit.<br />
57<br />
NEWS<br />
Top<br />
Back to the Future –<br />
MIT resurrects 1940s-era<br />
nuclear experiment<br />
(nei) On 2 December 2017, the Massachusetts<br />
Institute of Techno logy’s<br />
Nuclear Science and Engineering Department<br />
restarted a “subcritical<br />
graphite exponential pile” dating back<br />
to the earliest years of the atomic age.<br />
The pile is similar in design to the<br />
famous Chicago Pile-1 (CP-1) built by<br />
Enrico Fermi under the bleachers of<br />
the University of Chicago football<br />
stadium, which in 1942 initiated the<br />
world’s first man-made, self-sustaining<br />
fission chain reaction.<br />
“Those were the good old days,<br />
when scientists were more important<br />
than university football coaches…,”<br />
mused MIT Nuclear Reactor Laboratory<br />
Director David Moncton at a<br />
Dec. 2 ceremony attended by nearly<br />
50 faculty, students and guests marking<br />
the restart of the pile at 4:25 p.m.<br />
EST, 75 years to the minute since<br />
CP-1’s first criticality – and 60 years<br />
since MIT’s pile was first assembled by<br />
the university’s students.<br />
The university intends to use the<br />
unique assembly to teach students<br />
about “real world” reactor physics<br />
measurements and to conduct new<br />
experiments – including some relevant<br />
to advanced reactor designs, says<br />
MIT’s Professor of the Practice of<br />
Nuclear Science and Engineering<br />
Kord Smith.<br />
“This is an extremely important<br />
facility for teaching students about<br />
measuring reactor parameters,” Smith<br />
says. “This will give our students the<br />
rare opportunity to handle and load<br />
uranium fuel themselves.”<br />
The MIT graphite reactor consists<br />
of a 2.5-meter cubical pile of 30 metric<br />
tons of stacked graphite rectangular<br />
bars, with holes drilled at regular<br />
intervals to allow 2.5 metric tons of<br />
natural uranium metal fuel rods to be<br />
inserted. (Fermi chose the term “pile”<br />
from the word “pila,” which means<br />
stack in Italian.) With no moving<br />
parts, the only other components of<br />
the pile will be a plutonium-beryllium<br />
or californium-252 neutron “source”<br />
to drive the subcritical flux distribution,<br />
a neutron-absorbing cadmium<br />
rod to adjust subcritical reactivity, and<br />
indium activation foils to measure the<br />
spatial distribution of neutrons within<br />
the pile.<br />
Smith says Fermi’s original subcritical<br />
experiments were built to<br />
verify early nuclear physics theories<br />
about the size and spacing of fuel<br />
rods and the neutron slowing-down<br />
or “moderating” material needed to<br />
allow a neutron chain reaction to<br />
become self-sustaining.<br />
He explained that Fermi’s design<br />
was “brilliantly simple,” allowing the<br />
measurement of a single parameter –<br />
the axial profile of neutrons in the pile<br />
– to return information about how<br />
close the assembly was to a selfsustaining<br />
chain reaction and what<br />
scale-up of pile dimensions was<br />
needed in order for CP-1 to become<br />
an actual critical reactor.<br />
The simplicity of Fermi’s design<br />
allowed MIT’s pile, like many others at<br />
universities and laboratories around<br />
the country, to be built in a month’s<br />
time in 1957. However, with the<br />
advent of more powerful water-cooled<br />
reactors soon after, these teaching<br />
tools soon fell into disuse and were<br />
gradually forgotten. In fact, MIT’s<br />
| | MIT Nuclear Reactor Laboratory Director David Moncton (L) with Associate<br />
Department Head Jacopo Buongiorno, Professor of the Practice Kord Smith<br />
and Professor Emeritus Neil Todreas in front of the graphite exponential<br />
pile. (Photo: NEI, 4572)<br />
graphite pile was “rediscovered” last<br />
year, more or less hidden for decades<br />
under its aluminum metal covers.<br />
“We couldn’t believe the pile was<br />
still here,” Smith says.<br />
With the help of Moncton, departmental<br />
colleagues and staff, Smith<br />
restored the facility to working order<br />
in time for the Dec. 2 restart.<br />
Moncton, who operates the MIT<br />
Reactor – the second-largest universitybased<br />
research reactor in the country<br />
– says both the U.S. Nuclear Regulatory<br />
Commission and the U.S. Department<br />
of Energy have been very helpful and<br />
accommodating of MIT’s plans to<br />
restart the subcritical graphite pile.<br />
The university is awaiting an NRC<br />
operating license, which hopefully will<br />
be issued by the end of this year.<br />
Once that happens, Smith expects<br />
to use the pile for undergraduate and<br />
graduate courses in the fundamentals<br />
of reactor physics starting next year.<br />
Among the activities in which the<br />
students will be involved include<br />
“testing of physics kernels of neutron<br />
interactions within reactor-grade<br />
graphite,” he says.<br />
“Modeling and simulation are often<br />
oversold by those who have never<br />
done reactor measurements, and<br />
students are beginning to believe that<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
58<br />
NEWS<br />
*)<br />
Net-based values<br />
(Czech and Swiss<br />
nuclear power<br />
plants gross-based)<br />
1)<br />
Refueling<br />
2)<br />
Inspection<br />
3)<br />
Repair<br />
4)<br />
Stretch-out-operation<br />
5)<br />
Stretch-in-operation<br />
6)<br />
Hereof traction supply<br />
7)<br />
Incl. steam supply<br />
8)<br />
New nominal<br />
capacity since<br />
January 2016<br />
9)<br />
Data for the Leibstadt<br />
(CH) NPP will<br />
be published in a<br />
further issue of <strong>atw</strong><br />
BWR: Boiling<br />
Water Reactor<br />
PWR: Pressurised<br />
Water Reactor<br />
Source: VGB<br />
everything can be computed accurately,”<br />
Smith explains. “How ever,<br />
calculations are no better than the<br />
[underlying] physics models. Graphite<br />
piles are a great place to study the<br />
physics of how neutrons interact in a<br />
graphite-moderated system.”<br />
Another application will be to<br />
model neutron fields in solid media<br />
with large voids, with possible research<br />
applications for graphite- moderated<br />
advanced reactors and in test reactors<br />
like Idaho National Laboratory’s Transient<br />
Reactor Test Facility, also known<br />
as the TREAT reactor – which was restarted<br />
Nov. 15 after a 23-year operational<br />
hiatus. The TREAT reactor in<br />
turn will be used for tests that will support<br />
the development of advanced<br />
accident tolerant fuels for the U.S.<br />
commercial reactor fleet.<br />
In another echo of the past, Associate<br />
Department Head Jacopo<br />
Buongiorno sent away to Italy for a<br />
bottle of Chianti, which was duly<br />
signed by the 49 attendees at the<br />
ceremony – just as Fermi and his<br />
49 colleagues did 75 years ago.<br />
| | www.nei.org, 4572<br />
ONR (United Kingdom):<br />
Regulators approve new<br />
nuclear power station design<br />
(onr) The UK Advanced Boiling Water<br />
Reactor (UK ABWR), designed by<br />
Hitachi-GE link to external website, is<br />
suitable for construction in the UK,<br />
the regulators confirmed following<br />
completion of an in-depth assessment<br />
of the nuclear reactor design.<br />
The Office for Nuclear Regulation<br />
(ONR), the Environment Agency link<br />
to external website and Natural Resources<br />
Wales link to external website,<br />
the regulators who undertake the Generic<br />
Design Assessment of new reactor<br />
designs, are satisfied that this reactor<br />
meets regulatory expectations on<br />
safety, security and environmental<br />
protection at this stage of the regulatory<br />
process.<br />
ONR has issued a Design Acceptance<br />
Confirmation (DAC) and the<br />
environment agencies have issued a<br />
Statement of Design Acceptability<br />
(SoDA) to Hitachi-GE.<br />
Horizon Nuclear Power link to<br />
external website is proposing to build<br />
and operate two of these reactors in<br />
Wylfa Newydd on Anglesey and<br />
Oldbury- on-Severn near Thornbury<br />
in South Gloucestershire.<br />
Operating Results September 2017<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 910 880 720 649 662 5 631 282 252 863 137 100.00 96.19 99.80 94.83 99.16 94.46<br />
OL2 Olkiluoto BWR FI 910 880 720 659 969 4 443 996 242 261 136 100.00 75.15 99.87 74.01 100.73 74.55<br />
KCB Borssele PWR NL 512 484 720 361 342 2 300 987 157 105 428 100.00 69.12 100.00 69.73 98.08 66.95<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 1,2,7) PWR CH 380 365 76 24 852 2 087 110 130 319 266 10.56 84.34 9.09 83.76 8.56 82.96<br />
KKG Gösgen 7) PWR CH 1060 1010 720 758 046 6 231 443 302 842 078 100.00 90.67 99.99 90.20 99.33 89.74<br />
KKM Mühleberg 2) BWR CH 390 373 552 205 130 2 273 440 123 485 685 76.67 90.51 73.99 89.75 73.05 88.98<br />
CNT-I Trillo PWR ES 1066 1003 720 764 466 6 184 466 236 678 183 100.00 89.43 100.00 89.08 98.95 88.07<br />
Dukovany B1 PWR CZ 500 473 720 357 912 1 722 369 107 532 743 100.00 54.51 100.00 54.07 99.42 52.58<br />
Dukovany B2 PWR CZ 500 473 720 353 466 2 222 184 103 544 812 100.00 69.70 99.45 69.04 98.19 67.84<br />
Dukovany B3 PWR CZ 500 473 0 0 2 309 273 101 934 129 0 82.69 0 71.14 0 70.50<br />
Dukovany B4 PWR CZ 500 473 532 254 950 1 826 863 102 355 014 73.89 67.71 70.70 55.91 70.82 55.77<br />
Temelin B1 PWR CZ 1080 1030 720 778 027 7 081 092 104 709 251 100.00 100.00 99.99 99.95 100.05 100.08<br />
Temelin B2 PWR CZ 1080 1030 720 780 336 5 223 180 99 087 502 100.00 73.50 100.00 73.09 100.35 73.83<br />
Doel 1 PWR BE 454 433 720 324 229 2 613 885 133 226 857 100.00 88.70 99.87 88.12 98.81 87.70<br />
Doel 2 PWR BE 454 433 720 326 245 2 599 836 131 253 485 100.00 89.48 99.97 89.06 99.25 86.89<br />
Doel 3 PWR BE 1056 1006 524 556 750 6 732 621 251 169 221 72.72 96.62 72.46 96.41 72.84 96.81<br />
Doel 4 PWR BE 1084 1033 720 780 648 5 469 234 252 141 684 100.00 79.29 100.00 78.59 98.88 76.39<br />
Tihange 1 PWR BE 1009 962 282 277 896 2 690 977 289 954 051 39.13 42.34 39.02 41.87 38.26 40.70<br />
Tihange 2 PWR BE 1055 1008 720 758 970 5 084 166 246 603 234 100.00 78.51 100.00 73.85 100.47 73.83<br />
Tihange 3 PWR BE 1089 1038 720 774 234 7 050 423 266 531 120 100.00 100.00 99.97 99.98 98.60 98.72<br />
Operating Results October 2017<br />
Plant name<br />
Type<br />
Nominal<br />
capacity<br />
gross<br />
[MW]<br />
net<br />
[MW]<br />
Operating<br />
time<br />
generator<br />
[h]<br />
Energy utilisation<br />
Energy generated, gross Time availability Energy availability<br />
[MWh]<br />
[%]<br />
[%] *) [%] *)<br />
Month Year Since Month Year Month Year Month Year<br />
commissioning<br />
KBR Brokdorf DWR 1480 1410 745 1 011 300 8 549 400 318 131 734 100.00 86.71 99.50 86.46 97.13 83.84<br />
KKE Emsland 4) DWR 1406 1335 745 937 223 3 903 011 338 316 924 100.00 41.98 93.94 39.11 84.59 35.98<br />
KWG Grohnde DWR 1430 1360 745 1 004 762 9 304 398 333 303 977 100.00 91.93 99.93 91.77 95.81 90.70<br />
KRB B Gundremmingen SWR 1344 1284 745 970 799 8 126 396 365 069 095 100.00 87.01 94.85 83.35 90.42 77.21<br />
KRB C Gundremmingen 4) SWR 1344 1288 745 778 570 8 351 414 330 004 358 100.00 91.83 100.00 90.98 76.78 84.52<br />
KKI-2 Isar DWR 1485 1410 745 968 428 7 990 831 318 640 904 100.00 85.41 99.83 83.30 96.32 81.02<br />
KKP-2 Philippsburg DWR 1468 1402 745 1 073 129 9 378 353 339 453 163 100.00 89.84 99.71 89.37 96.66 86.22<br />
GKN-II Neckarwestheim DWR 1400 1310 745 1 046 248 5 745 846 353 059 535 100.00 55.80 99.92 55.72 94.15 52.80<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
Mark Foy, ONR’s Chief Nuclear<br />
Inspector said: “The completion of the<br />
generic design assessment of the UK<br />
ABWR is a significant step in our<br />
regulation of the overall process to<br />
construct this type of reactor in the<br />
UK, ensuring that the generic design<br />
meets the highest standards of safety<br />
that we expect in this country. We’re<br />
already working on our assessment<br />
of Horizon’s site licence application<br />
and on the development of the site<br />
specific safety case to progress, in<br />
due course, the construction and<br />
operation of these reactors at Wylfa<br />
Newydd.”<br />
Dr Jo Nettleton, Deputy Director<br />
for Radioactive Substances and Installations<br />
Regulation at the Environment<br />
Agency said: “We’ve concluded that<br />
the generic design of the UK ABWR<br />
should be capable of meeting the high<br />
standards of environment protection<br />
and waste management that we<br />
require in the UK. We only came<br />
to this conclusion after carefully reviewing<br />
the submissions provided by<br />
Hitachi- GE and their responses to the<br />
questions and issues we raised. We’ve<br />
also carefully considered all the comments<br />
we received from people during<br />
our public consultation and we’re<br />
grateful for all who took part for<br />
taking time to respond.”<br />
Tim Jones, Natural Resources<br />
Wales’s Executive Director for North<br />
and Mid Wales, said: “It is our job to<br />
ensure that any new nuclear power<br />
station will meet high standards of environmental<br />
protection and waste<br />
management, ensuring that our communities<br />
and environment are kept<br />
safe.<br />
“Following a public consultation<br />
on our initial findings, we have<br />
concluded that the UK ABWR design is<br />
acceptable. We will now work on the<br />
detailed assessments of the permits,<br />
licences and consents that Horizon<br />
Nuclear Power will need to have in<br />
place to build Wylfa Newydd.”<br />
The regulators have documented<br />
progress of each stage of their assessment<br />
through a series of reports on its<br />
joint website.<br />
| | www.onr.org.uk, 8874<br />
IAEA conference says more<br />
nuclear power needed to meet<br />
global goals on climate change<br />
(iaea) Nuclear power remains an<br />
important option for many countries<br />
to strengthen energy security and mitigate<br />
the effects of global warming and<br />
air pollution, but substantial growth in<br />
its use is needed for the world to meet<br />
its climate goals, according to an IAEA<br />
international conference that concluded<br />
in the United Arab Emirates.<br />
The some 700 participants from 67<br />
IAEA Member States and five international<br />
organizations who attended<br />
the event in Abu Dhabi this week<br />
enjoyed a wide convergence of views,<br />
Ambassador Hamad Alkaabi, president<br />
of the International Ministerial Conference<br />
on Nuclear Power in the 21 st Century,<br />
said in his concluding statement.<br />
“While respecting the right of each<br />
State to define its national energy<br />
policy, the Conference recognized that<br />
nuclear power remains an important<br />
option for many countries to improve<br />
energy security, reduce the impact of<br />
volatile fossil fuel prices and mitigate<br />
the effects of climate change and air<br />
pollution, including by backing up<br />
intermittent energy sources,” Alkaabi,<br />
the UAE’s Permanent Representative<br />
to the IAEA in Vienna, said at the conference’s<br />
closing session, attended by<br />
IAEA Director General Yukiya Amano.<br />
The three-day conference provided<br />
a forum for high-level dialogue on<br />
the role of nuclear power in the coming<br />
decades. Nuclear power emits<br />
virtually no greenhouse gases during<br />
operation. It produces 11 percent of<br />
the world’s electricity, which amounts<br />
to one-third of all electricity generated<br />
from low-carbon sources. Participants<br />
noted that some 6.5 million<br />
deaths a year are linked to air pollution,<br />
with that number set to increase<br />
significantly in the coming decades<br />
in the absence of greater action to<br />
curb emissions and expand access to<br />
low-carbon energy.<br />
To meet targets set out in the<br />
Paris Agreement on climate change,<br />
“substantial growth in nuclear<br />
electricity generation by 2050 will be<br />
required,” Alkaabi said, citing the<br />
International Energy Agency.<br />
While nuclear power will play a key<br />
role for many countries in achieving<br />
the Sustainable Development goals<br />
and reducing greenhouse-gas emissions,<br />
“nuclear is not currently attracting<br />
the necessary global investment” to<br />
limit the average global temperature<br />
increase to 2° C as required by the Paris<br />
Agreement, he said. “In addition, a<br />
number of plants are being shut down<br />
in some countries before the end of<br />
their safe operational lifetimes for both<br />
political and economic reasons.”<br />
The conference was the fourth<br />
such ministerial event following previous<br />
gatherings in Paris in 2005,<br />
Beijing in 2009 and St. Petersburg in<br />
2013. Organized in cooperation with<br />
the Nuclear Energy Agency (NEA) of<br />
the Organisation for Economic<br />
| | Panellists at the International Ministerial Conference on Nuclear Power<br />
in the 21 st Century, with the conference president, Ambassador Hamad<br />
Alkaabi of the UAE, second from right. (Photo: D. Calma/IAEA, 8345)<br />
Co- operation and Development, the<br />
conference was hosted by the UAE<br />
Government through the Ministry of<br />
Energy and the Federal Authority for<br />
Nuclear Regulation.<br />
Ministers and senior officials from<br />
IAEA Member States engaged in<br />
discussions on issues including their<br />
countries’ energy strategy and vision<br />
for the role of nuclear power and challenges<br />
to its introduction, continued<br />
operation and expansion. In addition,<br />
four panel sessions with selected<br />
speakers from diverse backgrounds<br />
discussed nuclear power and sustainable<br />
development; challenges to<br />
nuclear-power infrastructure development;<br />
nuclear safety and reliability;<br />
and innovations and advanced<br />
nuclear technologies.<br />
Alkaabi said participants widely<br />
agreed on other key areas, including<br />
the need to create an enabling environment<br />
to facilitate the introduction<br />
of nuclear power and ensure its safety<br />
and sustainability; that nuclear power<br />
is a safe, reliable and clean energy<br />
option; and that “innovations in technology<br />
design – including reactor size<br />
– as well as in investment and ownership<br />
models could facilitate the introduction<br />
of nuclear power in more<br />
countries.”<br />
Small modular reactors currently<br />
under development “may allow for<br />
expanded use of nuclear power – including<br />
on smaller grids and in remote<br />
settings, as well as for non-electrical<br />
applications – and improve access to<br />
nuclear energy,” the ambassador said.<br />
The conference repeatedly highlighted<br />
the importance of public<br />
confidence for the future of nuclear<br />
power. “Open and transparent decision<br />
making involving all stakeholders<br />
can improve the public perception of<br />
nuclear power and lead to broader<br />
public acceptance,” Alkaabi said.<br />
In conclusion, participants recognized<br />
the IAEA’s leading role in<br />
promoting peaceful uses of nuclear<br />
energy and supporting efforts to<br />
strengthen global nuclear safety,<br />
nuclear security and safeguards.<br />
| | www.iaea.org, 8345<br />
59<br />
NEWS<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
60<br />
NEWS<br />
World<br />
France postpones plans to<br />
reduce nuclear share after<br />
warning of shortages<br />
(nucnet) The French government has<br />
postponed a target to reduce the share<br />
of nuclear energy in the country’s<br />
energy mix after grid operator RTE<br />
warned it risked supply shortages after<br />
2020 and could miss a goal to lower<br />
carbon emissions. In 2015 the previous<br />
government of Francois Hollande<br />
established an energy transition law<br />
which set a target of reducing the<br />
share of nuclear in the energy mix to<br />
50% by 2025 from the current 75%.<br />
But environment minister Nicolas<br />
Hulot said on 8 November 2017 this<br />
would not be realistic. He said reducing<br />
the nuclear share in a hurry<br />
would increase France’s CO 2 emissions,<br />
endanger the security of power<br />
supply and put jobs at risk. Mr Hulot<br />
said president Emmanuel Macron’s<br />
government remains committed to<br />
reducing nuclear energy and ordered<br />
his ministry to produce a new timetable.<br />
He later said in a television<br />
interview that the government would<br />
be working towards a 2030 to 2035<br />
timeframe. RTE said in its 2017-2035<br />
Electricity Outlook that if France went<br />
ahead with plans to simultaneously<br />
shut down four 40-year-old nuclear<br />
reactors and all its coal-fired plants,<br />
there could be risks of power supply<br />
shortages. State-controlled utility<br />
EDF, which operates France’s 58<br />
commercial nuclear power plants, has<br />
argued instead to extend the operation<br />
of its nuclear fleet from 40 to at least<br />
50 years. France is the second largest<br />
generator of nuclear electricity behind<br />
the US. According to the International<br />
Atomic Energy Agency, France’s<br />
nuclear fleet produced almost 28% of<br />
the country’s electricity in 2016.<br />
| | www.gouvernement.fr, 7763<br />
Bill Gates’ TerraPower forms<br />
new company with China to<br />
develop twr technology<br />
(nucnet) TerraPower, the company<br />
founded in 2008 to develop advanced<br />
nuclear technology and backed by<br />
Microsoft founder Bill Gates, has<br />
signed a joint venture with China<br />
National Nuclear Corporation (CNNC)<br />
to form a company that will work to<br />
complete the Travelling Wave Reactor<br />
(TWR) design and commercialise TWR<br />
technology. TerraPower said on its<br />
website that the formation of the new<br />
company, Global Innovation Nuclear<br />
Energy Technology Company Ltd, was<br />
made possible under policies and<br />
agreements signed by the governments<br />
of the US and China. Terra Power said<br />
the collaboration with CNNC aims to<br />
pioneer new options in civilian nuclear<br />
energy that address safety, environmental<br />
and cost concerns. Unlike traditional<br />
nuclear reactors, TWR technology<br />
will be capable of using fuel made<br />
from depleted uranium, which is currently<br />
a waste byproduct of the<br />
uranium enrichment process. Its<br />
unique design gradually converts the<br />
fuel through a nuclear reaction without<br />
removing it from the reactor’s core,<br />
eliminating the need for reprocessing.<br />
This means the reactor can generate<br />
heat and produce electricity over a<br />
much longer period of continuous<br />
operation. Additionally, eliminating<br />
reprocessing reduces proliferation<br />
concerns, lowers the overall cost of the<br />
nuclear energy process, and helps to<br />
protect the environment by making use<br />
of a waste by-product and reducing the<br />
production of greenhouse gases. On<br />
3 November 2017 in Beijing, Mr Gates<br />
met the premier of China’s state council,<br />
Li Keqiang, to discuss increased<br />
cooperation between China and the<br />
US in the development of the next<br />
generation of reactor technologies.<br />
| | terrapower.com, 8832<br />
Barakah project brought $ 3.3 bn<br />
of economic benefit to UAE<br />
(nucnet) More than 1,400 local companies<br />
have been contracted in the<br />
development of the United Arab<br />
Emirates’ first nuclear power station<br />
project at Barakah, Mohamed Al-<br />
Hammadi, chief executive officer of<br />
the Emirates Nuclear Energy Corporation<br />
(Enec), told an International<br />
Atomic Energy Agency conference in<br />
Abu Dhabi. Mr Al-Hammadi told the<br />
International Ministerial Conference<br />
on Nuclear Power in the 21st Century<br />
that the construction of Barakah<br />
brought over $3.3bn (€2.8bn) worth of<br />
contracts to UAE-based companies,<br />
| | Barakah project brought $ 3.3 bn of economic<br />
benefit to UAE. View of the Barakah<br />
construction site in September 2017.<br />
(Courtesy: ENEC, 8877)<br />
providing economic benefits to the<br />
Gulf country. Enec signed a contract<br />
with Korea Electric Power Corporation<br />
in 2009 for building four APR-1400<br />
units at the Barakah station. Construction<br />
of the units began in 2012. Enec<br />
said yesterday that Unit 1 at Barakah is<br />
now more than 96% complete, Unit 2<br />
more than 87%, Unit 3 more than 78%<br />
and Unit 4 more than 58%. Overall,<br />
construction of the four units is more<br />
than 84% complete.<br />
| | www.enec.gov.ae, 8877<br />
Dominion to apply for second<br />
life extension at North Anna<br />
Nuclear Station – 80 operation<br />
years advised<br />
(nucnet) Dominion Energy Virginia has<br />
notified the US Nuclear Regulatory<br />
Commission that it intends to apply for<br />
a second 20-year life extension for the<br />
twin-reactor North Anna nuclear<br />
power station in Virginia. The company<br />
said it would file a licence renewal application<br />
with the NRC in 2020, following<br />
a similar application to extend the<br />
operating lifetime of two reactors at<br />
the Surry nuclear station, also in<br />
Virginia, to 80 years. Dominion said it<br />
expects to invest up to $4bn (€3.3bn)<br />
in upgrades to the two North Anna<br />
units and the two Surry units as<br />
part of the relicensing process. The<br />
Washington-based Nuclear Energy<br />
Institute said that of the 99 commercial<br />
nuclear power reactors operating in<br />
the US, 84 have had their original<br />
40-year operating licences extended to<br />
60 years. Three others that were issued<br />
licence renewals have since shut down.<br />
Another seven applications are under<br />
NRC review, and the remaining four<br />
are expected to apply between 2020<br />
and 2022. By 2040, half of the nation’s<br />
nuclear plants will have been operating<br />
for 60 years. Under its second<br />
licence renewal programme, the<br />
industry is planning for a second round<br />
of licence renewals to allow operation<br />
out to 80 years.<br />
| | www.dominion.com, 3882<br />
Household energy prices<br />
in the EU down compared<br />
with 2016<br />
(eurostat) In the European Union<br />
(EU), household electricity prices<br />
slightly decreased (-0.5%) on average<br />
between the first half of 2016 and the<br />
first half of 2017 to stand at €20.4 per<br />
100 kWh. Across the EU Member<br />
States, household electricity prices in<br />
the first half of 2017 ranged from<br />
below €10 per 100 kWh in Bulgaria to<br />
more than €30 per 100 kWh in<br />
Denmark and Germany.<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
| | First concrete poured for unit 1 at Bangladesh’s<br />
Rooppur. Artist’s view of the site<br />
with two reactors. (Courtesy: Rosatom, 7745)<br />
Highest increases in electricity<br />
prices in Cyprus, Greece and<br />
Belgium, largest falls in Italy,<br />
Croatia and Lithuania<br />
Across the EU Member States, the<br />
highest increase in household electricity<br />
prices in national currency<br />
between the first half of 2016 and<br />
the first half of 2017 was registered<br />
by far in Cyprus (+22.0%), followed<br />
by Greece (+12.8%), Belgium<br />
(+10.0%), Poland (+6.9%), Sweden<br />
(+5.5%) and Spain (+5.1%). In contrast,<br />
the most noticeable decreases<br />
were observed in Italy (-11.2%),<br />
Croatia (-10.2%) and Lithuania<br />
(-9.3%), well ahead of Luxembourg<br />
(-4.9%), Austria (-4.1%), Romania<br />
(-4.0%) and the Netherlands (-3.6%).<br />
Expressed in euro, average household<br />
electricity prices in the first half<br />
of 2017 were lowest in Bulgaria (€9.6<br />
per 100 kWh), Lithuania (€11.2) and<br />
Hungary (€11.3) and highest in<br />
Denmark and Germany (both €30.5)<br />
followed by Belgium (€28.0). The<br />
average electricity price in the EU<br />
was €20.4 per 100 kWh.<br />
When expressed in purchasing<br />
power standards (PPS), an artificial<br />
common reference currency that<br />
eliminates general price level differences<br />
between countries, it can be seen<br />
that, relative to the cost of other goods<br />
and services, the lowest household<br />
electricity prices were found in Finland<br />
(12.8 PPS per 100 kWh), Luxembourg<br />
(13.5) and the Netherlands (14.2),<br />
and the highest in Germany (28.7),<br />
Portugal (28.6), Poland (25.9),<br />
Belgium (25.6) and Spain (25.4).<br />
Half or more of the electricity<br />
price is made up of taxes and<br />
levies in Denmark, Germany and<br />
Portugal<br />
The share of taxes and levies in total<br />
household electricity prices varied<br />
significantly between Member States,<br />
ranging from two-thirds in Denmark<br />
(67% of household electricity price is<br />
made up of taxes and levies) and over<br />
half in Germany (54%) and Portugal<br />
(52%) to 5% in Malta in the first half<br />
of 2017. On average in the EU, taxes<br />
and levies accounted for more than a<br />
third (37%) of household electricity<br />
prices.<br />
| | ec.europa.eu, 8921<br />
Reactors<br />
Argentina to start construction<br />
of two new reactors<br />
(nucnet) Argentina plans to start construction<br />
of two new nuclear reactor<br />
units in the second half of <strong>2018</strong>,<br />
Argentina’s undersecretary for nuclear<br />
energy Julian Gadano told Reuters.<br />
Mr Gadano said Argentina is in the process<br />
of finalising negotiation of the<br />
commercial and financial contracts to<br />
build the two plants. In May 2017,<br />
Argentina signed a $12.5bn (€10.7bn)<br />
agreement with China for the construction<br />
and financing of two nuclear power<br />
units. According to the agreement,<br />
China’s National Nuclear Corporation<br />
and Nucleoeléctrica Argentina will begin<br />
construction of Atucha-3, a 700-<br />
MW Candu-6 pressurised heavy water<br />
reactor (PHWR), in <strong>2018</strong> and will start<br />
building a 1,000-MW Hualong One, or<br />
HPR1000, pressurised- water reactor<br />
unit in 2020. Argentina has three operating<br />
commercial power reactors – a<br />
Candu unit at the Embalse nuclear<br />
station and two PHWRs at Atucha.<br />
Under the May 2017 contract, China<br />
agreed to provide a long term-loan for<br />
85% of the required financing, which<br />
will be repaid when the plants begin<br />
generating electricity, according to<br />
comments at the time by Mr Gadano.<br />
| | www.na-sa.com.ar, 3345<br />
First concrete poured for unit<br />
1 at Bangladesh’s Rooppur<br />
(nucnet) First concrete was poured on<br />
30 November 2017 for the nuclear<br />
island basemat of Unit 1 at the planned<br />
Rooppur nuclear power station in<br />
Bangladesh, Russian state-owned<br />
nuclear corporation Rosatom said<br />
in a statement. The ceremony was attended<br />
by Rosatom’s director- general<br />
Alexey Likhachev and the prime minister<br />
of Bangladesh Sheikh Hasina, the<br />
statement said. In October 2013, Russia<br />
signed an agreement with Bangladesh<br />
for design work on Rooppur, on<br />
the banks of the Ganges river about<br />
160 km from the Bangladeshi capital<br />
Dhaka. In 2014, Rosatom said the<br />
Rooppur units – the first nuclear power<br />
reactors in Bangladesh – would both<br />
be 1,200-MW V-392M pressurised water<br />
reactors. According to Rosatom,<br />
the first unit at Rooppur is scheduled<br />
to begin commercial operation in 2023<br />
with the second unit following in<br />
2024. In July 2017, Russia agreed to<br />
release a state loan to finance the construction<br />
of the bulk of the Rooppur<br />
project. No mention was made of the<br />
amount of the loan, but earlier media<br />
reports put it at $12.6bn (€10.6bn).<br />
According to earlier reports, first concrete<br />
for Unit 1 at Rooppur was expected<br />
to be laid in December 2017.<br />
| | www.rosatom.ru, www.baec.gov.bd,<br />
7745<br />
Bulgaria extends Kozloduy-5<br />
operating licence by 10 years<br />
(nucnet) The operating licence for<br />
Unit 5 at the Kozloduy nuclear power<br />
station in Bulgaria has been extended<br />
by 10 years until 2027, the country’s<br />
energy ministry said. The 963-MW<br />
VVER V-320 unit, which began commercial<br />
operation in December 1988,<br />
could operate until 2047, the ministry<br />
said, but a 10-year extension is the<br />
longest allowed under Bulgarian law.<br />
Its existing operating licence was due<br />
to expire this month. Bulgaria has two<br />
nuclear units in commercial operation,<br />
Kozloduy-5 and Kozloduy-6.<br />
They are both Russian-designed<br />
VVERs and produce about 33% of the<br />
country’s electricity. The operating<br />
licence for Kozloduy-6 expires in<br />
August 2019. Extending the life of the<br />
two units is a priority for Bulgaria’s<br />
government, energy minister Temenuzhka<br />
Petkova said. Lachezar Kostov,<br />
the head of the Bulgarian Nuclear<br />
Regulatory Agency, said last year that<br />
the main tasks for Bulgaria’s nuclear<br />
energy sector are lifetime extensions<br />
at Kozloduy-5 and -6, modernisation<br />
of the two units by increasing<br />
their capacity, construction of a new<br />
unit at Kozloduy, and development of<br />
a national repository for low- and<br />
medium-level radioactive waste.<br />
| | www.kznpp.org, 8834<br />
Excavation of foundation pit<br />
begins at Iran’s Bushehr-2<br />
(nucnet) Excavation of the foundation<br />
pit for Iran’s Bushehr-2 nuclear power<br />
plant began on 31 October 2017,<br />
Russian state-owned nuclear corporation<br />
Rosatom said in a statement.<br />
The start of work was given in a<br />
ground-breaking ceremony attended<br />
by Rosatom’s director-general Alexey<br />
Likhachev and Ali Akbar Salehi, head<br />
of the Atomic Energy Organisation of<br />
Iran, the statement said. In March<br />
2017, construction work formally<br />
began at Bushehr-2, a pressurised<br />
water reactor unit of the Russian<br />
61<br />
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VVER-1000 design. In September 2017,<br />
Rosatom said site preparation works<br />
had begun for Bushehr-2 and -3, both<br />
of the same Russian design. Rosatom<br />
said at the time that first concrete<br />
for Bushehr-2 was planned for the<br />
third quarter of 2019. Construction of<br />
Bushehr-2 is expected to be completed<br />
in 2024 and of Bushehr-3 in 2026. Iran<br />
and Russia signed an agreement to<br />
build two additional units at Bushehr<br />
in November 2014. Official media in<br />
Iran said the construction of Bushehr-2<br />
and -3 would cost about $10bn<br />
(€ 8.6bn). Bushehr-1 is Iran’s only<br />
commercial nuclear unit. It is a 915-<br />
MW pressurised-water reactor which<br />
was supplied by Russia and began commercial<br />
operation in September 2013.<br />
| | www.rosatom.ru, 8872<br />
Governor approves restart<br />
of Japan’s Ohi-3 and -4<br />
(nucnet) The governor of Fukui<br />
Prefecture in southwest Japan has<br />
approved the restart of the Ohi-3 and<br />
-4 nuclear reactor units, operator<br />
Kansai Electric Power Company said<br />
on 27 November 2017. His decision<br />
clears the final regulatory hurdle for<br />
the restarts of both units early next<br />
year. Ohi-3 is a 1,127-MW pressurised<br />
water reactor that began commercial<br />
operation in 1991. Ohi-4, also a<br />
1,127-MW PWR, began commercial<br />
operation in 1993. All of Japan’s 48<br />
reactors were shut between 2011 and<br />
2012 after the March 2011 Fukushima-<br />
Daiichi accident. Five units have resumed<br />
commercial operation after<br />
meeting revised regulatory standards.<br />
They are: Takahama-3 and -4, Ikata-3<br />
and Sendai-1 and -2. According to the<br />
Japan Atomic Industrial Forum, 12<br />
nuclear units at six sites have now<br />
been approved as meeting new regulatory<br />
standards introduced following<br />
the accident. Ohi-3 and -4 were the<br />
first two reactors to resume operation<br />
in Japan following the Fukushima-<br />
Daiichi accident, but were both taken<br />
offline in September 2013 for scheduled<br />
refuelling and maintenance. But<br />
| | Finland’s Posiva makes progress with<br />
final repository excavation works. Artist’s<br />
view of the encapsulation plant.<br />
(Courtesy: Posiva, 8871)<br />
restarts where delayed when, in May<br />
2014, the Fukui district court ruled<br />
that it would not allow Ohi-3 and -4 to<br />
return to operation. A lawsuit filed by<br />
a group of almost 200 people living<br />
within a 250km radius of the Ohi station<br />
claimed that the plant was sited<br />
near several active seismic faults and<br />
was not adequately protected against<br />
earthquakes. Kansai Electric appealed<br />
the decision and it was overturned by<br />
a higher court in March 2017.<br />
| | www.kepco.co.jp, 8871<br />
Energoatom and Toshiba to<br />
cooperate on modernisation<br />
of Ukraine nuclear plants<br />
(nucnet) Ukraine’s state-owned<br />
nuclear operator Energoatom and<br />
Japan-based Toshiba have signed an<br />
agreement to cooperate on the modernisation<br />
of turbine island equipment<br />
at Ukrainian nuclear power<br />
stations. Energoatom said the modernisation<br />
aims to increase the power<br />
output and efficiency, and improve<br />
the safety of Ukraine’s plants. The<br />
agreement will increase cooperation<br />
in the long-term servicing of existing<br />
plant equipment, a statement said.<br />
Energoatom said a committee will be<br />
formed to ensure the implementation<br />
of the agreement. According to the<br />
International Atomic Energy Agency,<br />
Ukraine has 15 reactors in commercial<br />
operation which produced 52% of the<br />
country’s electricity in 2016.<br />
| | www.energoatom.kiev.ua, 8834<br />
Completion of Vogtle units<br />
is best economic choice<br />
(nucnet) Completing the Vogtle-3 and<br />
-4 AP1000 nuclear reactor units represents<br />
the best economic choice for<br />
customers and preserves the benefits<br />
of carbon-free, baseload generation<br />
for the state of Georgia, Georgia<br />
Power chairman, president and chief<br />
executive officer Paul Bowers told a<br />
Georgia Public Service Commission<br />
(PSC) hearing into the project on<br />
7 November 2017. Mr Bowers said. All<br />
the project owners – Georgia Power,<br />
Oglethorpe Power, MEAG Power and<br />
Dalton Utilities – have agreed to continue<br />
with the project. This decision<br />
was based on the results of a schedule,<br />
cost and cancellation assessment that<br />
was prompted by the bankruptcy of<br />
Westinghouse, supplier of the AP1000<br />
technology being used for the plants.<br />
Mr Bowers said assessments of<br />
the project have included economic<br />
analysis, evaluation of various alternatives<br />
including abandoning one or<br />
both units, and assumptions related to<br />
potential risks. The Georgia PSC will<br />
hear from owners and partners in the<br />
project as well as public witnesses.<br />
The PSC will issue its final recommendation<br />
on 6 February <strong>2018</strong>. Mr Bowers<br />
said construction has continued<br />
uninterrupted at the Vogtle site over<br />
the past six months. Southern Nuclear,<br />
the nuclear operating subsidiary<br />
which operates the existing units<br />
at the Georgia station, is now the<br />
project manager at the site. Bechtel is<br />
managing daily construction efforts.<br />
| | www.georgiapower.com, 8432<br />
Waste Management<br />
Finland’s Posiva makes<br />
progress with final repository<br />
excavation works<br />
(nucnet) Finnish nuclear waste<br />
manage ment company Posiva has<br />
completed the excavations for the<br />
encapsulation plant at the final deep<br />
geologic disposal facility under construction<br />
at Olkiluoto, Posiva’s owner<br />
Teollisuuden Voima Oyj (TVO) said in<br />
a statement. Excavation works began<br />
in October 2016. TVO said Posiva has<br />
also made progress with excavation<br />
work for the vehicle access tunnels<br />
leading to the final disposal facility<br />
itself. TVO said the first phase of excavations<br />
for the final disposal facility is<br />
estimated to take two and a half years.<br />
In December 2016, Posiva was given<br />
regulatory approval to begin construction<br />
of a deep geologic repository at<br />
Olkiluoto on the country’s southwest<br />
coast – the first final repository in the<br />
world to enter the construction phase.<br />
| | www.posiva.fi, 8871<br />
Research<br />
Wendelstein 7-X now ready<br />
for virtual tours!<br />
(ipp-mpg) The new 360-degree panorama<br />
featured on the internet pages of<br />
Max Planck Institute for Plasma<br />
Physics (IPP) leads right into the<br />
plasma vessel of the Wendelstein 7-X<br />
fusion research device at Greifswald.<br />
The address www.ipp.mpg.de/<br />
panoramaw7x takes observers on an<br />
extraordinary tour to the core of the<br />
device, otherwise accessible only to<br />
experts; they can stroll through the<br />
experimentation hall and view the<br />
facilities that heat the plasma to many<br />
millions of degrees.<br />
By way of PC, tablet or smartphone<br />
they can cast an eye at every angle and<br />
zoom in on even tiniest details. Short<br />
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videos in which IPP scientists describe<br />
their workplaces are started and<br />
stopped by mouse click; info panels<br />
can be slotted in to explain important<br />
components. The panorama was<br />
recorded by Munich photographer<br />
Volker Steger, who had already done<br />
the panorama of IPP Garching’s<br />
ASDEX Upgrade fusion device<br />
(www.ipp.mpg.de/panorama).<br />
The objective of IPP’s research is a<br />
fusion power plant to derive energy<br />
from fusion of light atomic nuclei, just<br />
as the sun does. At Garching IPP also<br />
operates the ASDEX Upgrade experiment,<br />
a large-scale device of the tokamak<br />
type. IPP’s branch institute at<br />
Greifswald is conducting research on<br />
the large Wendelstein 7-X stellarator.<br />
As of now both devices are accessible<br />
at any time for a virtual tour.<br />
| | www.ipp.mpg.de, 8892<br />
High performance computing<br />
for energies with EoCoE<br />
(cea) Computer simulation being an<br />
amazing driver of innovation, it is<br />
strategic for Europe to develop supercomputing<br />
resources at the most<br />
advanced level. The UE provides<br />
support to supercomputing infrastructures<br />
(PRACE1), hardware (ETP4H-<br />
PC2) as well as software technologies.<br />
The support to application software<br />
development is spread over nine thematic<br />
centers of excellence – EoCoE<br />
being one of them.<br />
Predicting wind and sunlight intensity,<br />
designing innovative materials to<br />
store electricity, optimizing the management<br />
of water reservoirs, predicting<br />
the performance of geothermal plants<br />
or even stabilizing plasma in a nuclear<br />
fusion reactor are essential tasks to<br />
master in order to accomplish a successful<br />
diversification of the energy mix.<br />
Dedicated to low-carbon energies,<br />
EoCoE, which stands for ‘European<br />
Energy-Oriented Center of Excellence’<br />
(and can be pronounced as “Echo”),<br />
targets the fields of weather forecast,<br />
materials, water management and<br />
nuclear fusion—all of which require<br />
high calculation capacities. The center<br />
brings together twenty-two partners<br />
from eight European countries,<br />
involved in both HPC and energies,<br />
committed to tackling the challenges<br />
in these fields.<br />
“Computer simulation is driven by<br />
the constant upgrades of high-performance<br />
computers,” said Edouard<br />
Audit, the CEA Director of Maison de<br />
la simulation3 and coordinator of<br />
EoCoE. “Yet the challenge is not so<br />
much to gain time than to achieve<br />
things that were previously<br />
inacces sible. In materials science, for<br />
instance, it is now possible to digitally<br />
test a very large number of materials.”<br />
Exascaling the future<br />
The mission of a laboratory such as<br />
Maison de la simulation is to develop<br />
cutting-edge digital tools in close<br />
collaboration with scientists from the<br />
related disciplines, as well as transversal<br />
tools such as linear algebra, input/<br />
output data management, and result<br />
visualization. “We provide support to<br />
researchers as they develop their<br />
own code to help them achieve the<br />
expected result. The help we offer<br />
ranges from applied mathematics to<br />
algorithms and HPC” Mr. Audit<br />
explained. “Meanwhile, we are also<br />
preparing for the future, that is to say<br />
the development of exascale architectures<br />
(1018 operations per second),<br />
that are massively multi-core. They<br />
differ from previous architectures by<br />
the fact that now, not all their processors<br />
are of the same nature. This is<br />
why we must change the way we compute—and<br />
how we manage memory<br />
storage in particular.”<br />
First concrete achievements<br />
Several significant advances have<br />
already been achieved thanks to<br />
EoCoE. During the working sessions,<br />
the scientists learn to “instrument”<br />
their simulation code to monitor the<br />
results step by step, and optimize them.<br />
For nuclear fusion, the Gysela code<br />
developed at CEA (IRFM4) describes<br />
ion transport in plasma inside the<br />
reactor’s toric chamber (tokamak). In<br />
addition to being necessary for the<br />
R&D activities of tokamaks WEST<br />
(CEA) and ITER in Cadarache, this<br />
code also deepens the fundamental understanding<br />
that physicists have of fusion<br />
plasma turbulence. It is now suitable<br />
for hundreds of thousands of computing<br />
cores. The meticulous audit<br />
work accomplished within EoCoE has<br />
saved 10 % in computing time and has<br />
helped prepare for the future upgrade<br />
to the exascale.<br />
| | www.cea.fr, 9983<br />
Company News<br />
MATRIX by Areva TN: a game<br />
changer in used fuel dry storage<br />
(areva) AREVA TN, the nuclear<br />
logistics affiliate of New AREVA, is<br />
launching an advanced used nuclear<br />
fuel storage overpack, NUHOMS®*<br />
MATRIX. With its improved capacity<br />
and performance, NUHOMS® MATRIX<br />
addresses the challenges faced by our<br />
customers when it comes to storing<br />
used fuel safely, efficiently and competitively.<br />
The unique 2-level horizontal and<br />
modular set-up reduces the inde pendent<br />
spent fuel storage installation<br />
(ISFSI) footprint by 45% which in turn<br />
reduces pad construction costs. This<br />
makes NUHOMS® MATRIX the smallest<br />
storage pad on the market for the<br />
same capacity, in a context where space<br />
is at a premium on nuclear sites. Its<br />
design accommodates canisters of<br />
different sizes and it can store high burnup<br />
short cooled fuel, which is of particular<br />
interest for shutdown nuclear<br />
reactors. New features and devices<br />
allow for the complete inspection of the<br />
canister without removing it from the<br />
module, as aging management and<br />
retrieval of the canister for future transport<br />
to a consolidated storage site have<br />
become a challenge for utilities.<br />
NUHOMS® storage systems<br />
securely store the dry fuel storage<br />
containers in a horizontal position<br />
within a sturdy, low-profile, reinforced<br />
concrete structure. This fortress-like<br />
structure serves as a robust barrier.<br />
“As more communities, policymakers<br />
and utilities across the world<br />
discuss securely storing used nuclear<br />
fuel, our NUHOMS® MATRIX system is<br />
a competitive, safe and timely solution<br />
for those needs and concerns,” said<br />
Greg Vesey, president, TN Americas.<br />
With more than 1,250 dry storage<br />
systems loaded worldwide, AREVA TN<br />
offers its customers an unrivaled<br />
experience for the management of<br />
used fuel.<br />
| | www.new.areva.com, 4532<br />
People<br />
Camilla Hoflund new<br />
President and CEO of Studsvik<br />
(studsvik) The current President and<br />
CEO Michael Mononen and the Board<br />
of Directors have together concluded<br />
that a changeover in the chief executive<br />
post is appropriate after the major<br />
changes in the Group that have been<br />
made in recent years. Studsvik’s Board<br />
of Directors has therefore appointed<br />
Camilla Hoflund as new President and<br />
CEO from January 1, <strong>2018</strong>.<br />
Camilla Hoflund is a mining<br />
engineer from the Royal Institute of<br />
Technology (KTH) and has been head<br />
of Studsvik’s Fuel and Materials<br />
Technology business area since 2014.<br />
She has worked at Studsvik since<br />
1994, with a short break in 2000-2003<br />
* NUHOMS: Nuclear<br />
Horizontal Modular<br />
Storage<br />
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when she was a business developer<br />
and consultant for risk management<br />
services at Det Norske Veritas.<br />
| | www.studsvik.com, 983<br />
Westinghouse appoints<br />
Ken Canavan Chief<br />
Technology Officer<br />
(westinghouse) Westinghouse Electric<br />
Company announced that Ken<br />
Canavan has been appointed chief<br />
technology officer (CTO), effective<br />
January 2, <strong>2018</strong>.<br />
Westinghouse’s CTO role has strategic<br />
responsibility to drive nextgeneration<br />
technology and innovation<br />
solutions that align with the com pany’s<br />
global business strategy. Canavan will<br />
lead these efforts, as well as strengthen<br />
Westinghouse with regard to technology<br />
leadership development.<br />
Canavan, 53, previously was director<br />
of engineering for the Electric Power<br />
Research Institute (EPRI). There he<br />
was responsible for turning industry<br />
needs into compelling research and<br />
development plans. These plans improved<br />
safety and performance of the<br />
global nuclear fleet. He has more than<br />
30 years of experience in key engineering<br />
and risk management roles. Prior<br />
to his work at EPRI, Canavan was<br />
responsible for risk applications at<br />
Data Systems and Solutions, ERIN<br />
Engineering and Research and GPU<br />
Nuclear. He also was a safety analysis<br />
engineer with Davis-Besse Nuclear<br />
Power Station in Ohio (USA).<br />
Canavan has a bachelor’s degree in<br />
chemical engineering, with a nuclear<br />
engineering minor, from Manhattan<br />
College, New York. Ken, his wife,<br />
Paula and his two children will relocate<br />
to the Pittsburgh area.<br />
| | www.westinghousenuclear.com,<br />
8831<br />
WANO Nuclear Excellence<br />
Awards 2017<br />
(wano) At the closure of its fourteenth<br />
Biennial General Meeting held in<br />
Gyeongju, the World Association of<br />
Nuclear Operators (WANO) tonight<br />
acknowledged the outstanding contribution<br />
made by nine nuclear professionals<br />
to promote excellence in the<br />
safe operation of commercial nuclear<br />
power.<br />
| | WANO Nuclear Excellence Awards 2017 (873)<br />
The honorary awards were established<br />
in 2003 to recognise individuals<br />
who have made extraordinary contributions<br />
to excellence in the operation<br />
of nuclear power plants, or the infrastructure<br />
that supports the nuclear<br />
power enterprise, or through WANO.<br />
Potential award recipients undergo<br />
a rigorous nomination and selection<br />
process before being approved. The<br />
awards are presented during each<br />
WANO Biennial General Meeting.<br />
This year’s award recipients are:<br />
Brian Cowell, EDF Energy; Bum-nyun<br />
Kim, Korea Hydro & Nuclear Power<br />
Company (KHNP); Pavlo Pavlyshyn,<br />
Rivne Nuclear Power Plant, NNEGC<br />
Energoatom; Pierre Pilon, Bruce<br />
Power; Philippe Sasseigne, Électricité<br />
de France; Debbie Sims, WANO Atlanta<br />
Centre; Jouko Turpeinen, Fortum<br />
Power and Heat Oy; Jean Van Vyve,<br />
ENGIE Electrabel; Makoto Yagi, The<br />
Kansai Electric Power Company, Inc.<br />
| | www.wano.info, 873<br />
Publications<br />
Nuclear Energy Data – 2017<br />
(nea) Nuclear Energy Data is the<br />
Nuclear Energy Agency’s annual compilation<br />
of statistics and country<br />
reports documenting nuclear power<br />
status in NEA member countries and in<br />
the OECD area. Information provided<br />
by governments includes statistics on<br />
total electricity produced by all sources<br />
and by nuclear power, fuel cycle capacities<br />
and requirements, and projections<br />
to 2035, where available. Country<br />
reports summarise energy policies,<br />
updates of the status in nuclear energy<br />
programmes and fuel cycle developments.<br />
In 2016, nuclear power continued<br />
to supply significant amounts<br />
of low-carbon baseload electricity,<br />
despite strong competition from lowcost<br />
fossil fuels and subsidised renewable<br />
energy sources. Three new units<br />
were connected to the grid in 2016, in<br />
Korea, Russia and the United States. In<br />
Japan, an additional three reactors<br />
returned to operation in 2016, bringing<br />
the total to five under the new regulatory<br />
regime. Three reactors were<br />
officially shut down in 2016 – one in<br />
Japan, one in Russia and one in the<br />
United States. Governments committed<br />
to having nuclear power in the energy<br />
mix advanced plans for developing or<br />
increasing nuclear generating capacity,<br />
with the preparation of new build projects<br />
making progress in Finland,<br />
Hungary, Turkey and the United Kingdom.<br />
Further details on these and<br />
other developments are provided in<br />
the publication’s numerous tables,<br />
graphs and country reports. Download<br />
the report at oe.cd/nea-data-2017<br />
| | www.oecd-nea.org, 3342<br />
Market data<br />
(All information is supplied without 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 to December 2016<br />
• Uranium: 18.75–27.80<br />
• Conversion: 5.50–6.50<br />
• Separative work: 47.00–62.00<br />
2017<br />
January 2017<br />
• Uranium: 20.25–25.50<br />
• Conversion: 5.50–6.75<br />
• Separative work: 47.00–50.00<br />
February 2017<br />
• Uranium: 23.50–26.50<br />
• Conversion: 5.50–6.75<br />
• Separative work: 48.00–50.00<br />
March 2017<br />
• Uranium: 24.00–26.00<br />
• Conversion: 5.50–6.75<br />
• Separative work: 47.00–50.00<br />
April 2017<br />
• Uranium: 22.50–23.50<br />
• Conversion: 5.00–5.50<br />
• Separative work: 45.50–48.50<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
May 2017<br />
• Uranium: 19.25–22.75<br />
• Conversion: 5.00–5.50<br />
• Separative work: 42.00–45.00<br />
June 2017<br />
• Uranium: 19.25–20.50<br />
• Conversion: 5.55–5.50<br />
• Separative work: 42.00–43.00<br />
July 2017<br />
• Uranium: 19.75–20.50<br />
• Conversion: 4.75–5.25<br />
• Separative work: 42.00–43.00<br />
August 2017<br />
• Uranium: 19.50–21.00<br />
• Conversion: 4.75–5.25<br />
• Separative work: 41.00–43.00<br />
September 2017<br />
• Uranium: 19.75–20.75<br />
• Conversion: 4.60–5.10<br />
• Separative work: 40.50–42.00<br />
October 2017<br />
• Uranium: 19.90–20.50<br />
• Conversion: 4.50–5.25<br />
• Separative work: 40.00–43.00<br />
November 2017<br />
• Uranium: 19.90–20.50<br />
• Conversion: 4.50–5.25<br />
• Separative work: 40.00–43.00<br />
| | Source: Energy Intelligence<br />
www.energyintel.com<br />
Cross-border Price 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 />
2017:<br />
I. quarter: 95.75; 8,385,071<br />
II. quarter: 86.40; 5,094,233<br />
III. quarter: 88.07; 5,504,908<br />
| | Source: BAFA, some data provisional<br />
www.bafa.de<br />
EEX Trading Results<br />
November 2017<br />
(eex) In November 2017, the European<br />
Energy Exchange (EEX) achieved a<br />
total volume of 276.6 TWh on its<br />
power derivatives markets (November<br />
2016: 423.2 TWh). The November<br />
volume comprised 163.8 TWh traded<br />
at EEX via Trade Registration with<br />
subsequent clearing. Clearing and<br />
settlement of all exchange transactions<br />
was executed by European<br />
Commodity Clearing (ECC).<br />
| | Uranium spot market prices from 1980 to 2017 and from 2007 to 2017. The price range is shown.<br />
In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.<br />
| | Separative work and conversion market price ranges from 2007 to 2017. 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 />
The shift of liquidity from the<br />
German- Austrian power futures into<br />
the German Phelix DE future continued,<br />
resulting in a new record<br />
volume of 72,651,053 MWh traded in<br />
the German contract.<br />
The Settlement Price for base load<br />
contract (Phelix Futures) with delivery<br />
in <strong>2018</strong> amounted to 37.60 €/MWh.<br />
The Settlement Price for peak load contract<br />
(Phelix Futures) with delivery<br />
in <strong>2018</strong> amounted to 45.70 €/MWh.<br />
On the EEX Market for emission<br />
allowances, 144.3 million tonnes of<br />
CO 2 (November 2016: 79.2 million<br />
tonnes of CO 2 ) were traded in<br />
November. The total volume increased<br />
by 82%. Primary market auctions contributed<br />
81.9 million tonnes of CO 2 to<br />
the total volume. On the emission<br />
derivatives market 57.9 million tonnes<br />
of CO 2 were traded which is more than<br />
three times the volume of the same<br />
month of the previous year (November<br />
2016: 18.5 million tonnes of CO 2 ).<br />
The E-Carbix amounted to<br />
7.57 €/EUA, the EUA price with<br />
delivery in December 2017 amounted<br />
to 7.35/7.92 €/ EUA (min./max.).<br />
| | www.eex.com<br />
MWV Crude Oil/Product Prices<br />
October 2017<br />
(mwv) According to information and<br />
calculations by the Association of the<br />
German Petroleum Industry MWV e.V.<br />
in October 2017 the prices for super<br />
fuel, fuel oil and heating oil noted<br />
inconsistent compared with the previous<br />
month September 2017. The<br />
average gas station prices for Euro<br />
super consisted of 134.72 €Cent<br />
(September 2017: 137.12 €Cent,<br />
approx. -1.75 % in brackets: each<br />
information for pre vious month or<br />
rather previous month comparison),<br />
for diesel fuel of 116.19 €Cent (114.36;<br />
+1.60 %) and for heating oil (HEL)<br />
of 57.07 €Cent (55.84, +2.20 %).<br />
The tax share for super with<br />
a consumer price of 134.72 €Cent<br />
(137.12 €Cent) consisted of<br />
65.45 €Cent (48.58 %, 65.45 €Cent)<br />
for the current constant mineral oil<br />
tax share and 21.51 €Cent (current<br />
rate: 19.0 % = const., 21.89 €Cent)<br />
for the value added tax. The product<br />
price (notation Rotterdam) consisted<br />
of 36.20 €Cent (26.87 %, 37.79 €Cent)<br />
and the gross margin consisted of<br />
11.74 €Cent (8.74 %; 11.99 €Cent).<br />
Thus the overall tax share for super<br />
results of 67.58 % (66.73 %).<br />
Worldwide crude oil prices<br />
(monthly average price OPEC/Brent/<br />
WTI, Source: U.S. EIA) were again<br />
approx. +3.27 % (+6.68 %) higher in<br />
September compared to September<br />
2017.<br />
The market showed a stable development<br />
with higher prices; each in<br />
US-$/bbl: OPEC basket: 55.5 (53.44);<br />
UK-Brent: 57.51 (56.15); West Texas<br />
Intermediate (WTI): 51.58 (49.82).<br />
| | www.mwv.de<br />
65<br />
NEWS<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />
66<br />
NUCLEAR TODAY<br />
Links to<br />
reference sources:<br />
Bangladesh<br />
new nuclear project:<br />
http://bit.ly/2BxD8z7<br />
UK nuclear skills<br />
warning:<br />
http://on.ft.com/<br />
2iIIML8<br />
Author<br />
John Shepherd<br />
nuclear 24<br />
41a Beoley Road West<br />
St George’s<br />
Redditch B98 8LR,<br />
United Kingdom<br />
‘Newcomer’ Nuclear Nation<br />
Leads Way into New Nuclear Year<br />
John Shepherd<br />
At the start of a new year, it is appropriate that a ‘newcomer’ nuclear nation has launched work on building its first<br />
nuclear power plant. First nuclear safety-related concrete has been poured for the plant at Rooppur in Bangladesh –<br />
making the South Asia nation the first in 30 years to start building its first commercial reactor unit following the United<br />
Arab Emirates in 2012 and Belarus in 2013.<br />
In Bangladesh, it is Russia’s Atomstroyexport that has been<br />
selected to build two VVER type (AES-2006) pressurised<br />
water reactors, each with a 1,200 MW(e) gross electricity<br />
generating capacity. The units are expected to be commissioned<br />
in 2023 and 2024 respectively.<br />
In addition to supporting the country’s increasing<br />
electricity needs, the reactors will “transform Bangladesh<br />
into a middle income country” and a developed one by<br />
2041, said Prime Minister Sheikh Hasina.<br />
Despite setbacks that nuclear has endured in recent<br />
years, there are nearly 60 reactors under construction<br />
around the world, mostly in Asia, according to the International<br />
Atomic Energy Agency (IAEA). Some 447 commercial<br />
reactor units are in operation in 30 countries.<br />
IAEA director-general Yukiya Amano told the recent<br />
fourth International Ministerial Conference on Nuclear<br />
Power in the 21 st Century in the United Arab Emirates that<br />
the agency’s latest projections showed the global potential<br />
for nuclear energy up to 2050 continues to be high,<br />
although figures show expansion is likely to slow.<br />
Amano warned: “It is difficult to see other low-carbon<br />
energy sources growing sufficiently to take up the slack if<br />
nuclear power use fails to grow.”<br />
But there is cause for optimism, beyond Bangladesh, as<br />
a new nuclear year gets under way. Key developments to<br />
look forward to include a review of the role of nuclear in<br />
France, following a long-overdue acceptance, of sorts, that<br />
the obsession of former president François Hollande to<br />
reduce the national nuclear share to 50 % by 2025 from<br />
the current 75 % was flawed.<br />
France’s grid operator RTE had warned that the country<br />
faced potential supply shortages beyond 2020 – in addition<br />
to increasing CO 2 emissions – if nuclear power were rolled<br />
back. The new administration of President Emmanuel<br />
Macron has chosen to fudge the issue, by saying it remains<br />
committed to reducing nuclear’s role. A new “timetable” to<br />
reduce the nuclear share is being drawn up and environment<br />
minister Nicolas Hulot has indicated that the government<br />
is now considering a period of 2030 to 2035. Therefore,<br />
it will be for a future leader of France to potentially<br />
revisit the issue.<br />
Another highlight of this new nuclear year will be in<br />
Pakistan, which is set to see construction start on a Chinese<br />
Generation III HPR1000 Hualong One reactor at the<br />
country’s Chashma nuclear power plant. This follows a<br />
cooperation agreement signed recently by the China<br />
National Nuclear Corporation and the Pakistan Atomic<br />
Energy Commission.<br />
China is also making strides in the UK, where regulators<br />
have begun the second stage of a generic design assessment<br />
that could see a version of the HPR1000 being built at<br />
the Bradwell B site in Essex, in eastern England.<br />
Meanwhile, the UK government has unveiled its<br />
Industrial Strategy white paper, with proposals to be<br />
fleshed out during <strong>2018</strong> aimed at seeking cost reductions<br />
across new build and decommissioning programmes.<br />
However, the chairman of the UK’s Nuclear Industry<br />
Association, Lord Hutton, said: “As we build new capacity to<br />
replace retiring power stations, and decommission old<br />
ones, the UK is well placed to develop supply chains, skills<br />
and international opportunities for the long term.”<br />
The UK has an ambitious domestic programme of<br />
nuclear new build, but industry and labour leaders warned<br />
last summer that the country would not have enough<br />
skilled workers to build the plants planned unless ministers<br />
removed uncertainty hanging over national energy policy.<br />
Recruitment in the UK is ramping up to complete the<br />
Hinkley Point C EPR nuclear plant in Somerset, along with<br />
several other reactors planned around the UK over the next<br />
20 years. Thousands more will be needed with the expertise<br />
to decommission the UK’s existing fleet of reactors.<br />
The nuclear development director at engineering giant<br />
Costain, Alistair Smith, told the Financial Times: “It’s<br />
20 years since we built a nuclear power station. These<br />
people are not just sitting around waiting to start again.<br />
We’ve just got Hinkley C started and resources on that<br />
project are already starting to look scarce.”<br />
So as the new year gets under way, questions will rise<br />
again as to whether the world has the skilled workforce<br />
needed to operate the nuclear stations of the future. As the<br />
IAEA has rightly pointed out, the availability of skilled staff<br />
is a cornerstone of the sustainability of the civil nuclear<br />
sector – and this will be the focus of a conference to be held<br />
in May in South Korea.<br />
The Third International Conference on Human Resource<br />
Development for Nuclear Power Programmes: Meeting Challenges<br />
to Ensure the Future Nuclear Workforce Capability,<br />
will review progress since the last IAEA conference held<br />
on the issue in 2014.<br />
Sustainable nuclear power relies on a sustainable<br />
workforce, which means investing in the recruitment and<br />
training of tomorrow’s nuclear generation.<br />
The IAEA is also developing a ‘SAT (systematic approach<br />
to training) Nuclear Training Effectiveness Evaluation’<br />
model that is designed to support member states. The<br />
agency said the model is “designed around a self- assessment<br />
process, together with the option to establish some form of<br />
independent validation capability”.<br />
Towards the end of 2017, a new nuclear training centre<br />
was launched in France by Trihom, a training organisation<br />
jointly owned by New Areva and Engie's industrial maintenance<br />
subsidiary Endel. The new centre in Normandy is<br />
said to be the largest nuclear training centre in France.<br />
The world’s nuclear industry understands the urgent<br />
need to nurture a new generation of nuclear professionals<br />
and equip them with the expertise they will need. Opponents<br />
of nuclear power will be quick to stoke up fears about a lack<br />
of skills in an attempt to halt progress on the development of<br />
new reactors. They should be denied that opportunity, so<br />
there is no time to lose. The start of a new year represents an<br />
ideal opportunity for nuclear industry leaders to renew<br />
their commitment in this area.<br />
Nuclear Today<br />
‘Newcomer’ Nuclear Nation Leads Way into New Nuclear Year ı John Shepherd
nucmag.com<br />
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The International Expert Conference on Nuclear Technology<br />
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The 49 th AMNT offers many various formats such as Focus Sessions, Technical Sessions<br />
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3 Carsten Haferkamp ı Managing Director, New NP GmbH, Germany<br />
3 Ursula Heinen-Esser ı Managing Director, Bundesgesellschaft für Endlagerung (BGE),<br />
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3 Dr. Willibald Kohlpaintner ı Head for Nuclear Energy Division, Axpo Holding AG, Switzerland<br />
3 David Peattie ı CEO, Nuclear Decommissioning Authority (NDA), United Kingdom<br />
3 Dr. Maria J. Rodriguez ı Research Group Leader, Gravitation and Black Hole Theory,<br />
Max Planck Institute for Gravitational Physics, Germany<br />
3 Dr. Dirk Stenkamp ı CEO, TÜV Nord Group, Germany<br />
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