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<strong>2018</strong><br />

1<br />

13<br />

ETSON Strategic<br />

Orientations<br />

on Research Activities<br />

Special<br />

Eurosafe 2017<br />

21 ı Environment and Safety<br />

Integrated Approach for Nuclear Safety, Security and Safeguards<br />

ISSN · 1431-5254<br />

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Clearance of Surface-contaminated Objects<br />

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‘Newcomer’ Nuclear Nation Leads Way


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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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<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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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />

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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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />

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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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 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 />

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<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 />

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 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 />

51<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 />

53<br />

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54<br />

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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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 1 ı January<br />

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 />

63<br />

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64<br />

NEWS<br />

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


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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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