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ETSON Strategic

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21 ı Environment and Safety

Integrated Approach for Nuclear Safety, Security and Safeguards

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29 ı Operation and New Build

Clearance of Surface-contaminated Objects

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3Energy Economy: Under the Banner of Jobs

atw Vol. 63 (2018) | Issue 1 ı January

Dear reader, Truly, these are no good news the two large integrated tech companies General Electric (GE) and Siemens

released to their employees right before Christmas: GE announced to cut 12,000 jobs worldwide in its power plant division,

6,800 in Europe alone. Siemens plans to cut 6,900 jobs, 6,100 of them in the power plant division. Thus, both groups

­continue job cutback in this field. GE has already massively cut jobs in Europe when having acquired the power plant

division of the French company Alstom. And at Siemens job cutbacks in the energy sector is almost an ongoing theme.

Both companies quote that this is ultimately a reaction to

the changed worldwide investment landscape. The

number of projects for these division-driving conventional

gas and coal-fired power plants would decline distinctly so

that with a constant offer a further decline in prices is

inevitable not leaving enough margins anymore.

With few exceptions blanket accusations, as common in

the past, with the hint “one has not early enough and not

sufficiently focused on other products such as renewable

energies” stayed out. Possibly, political protagonists were

aware of recent facts, which acknowledge renewable

energies a rather moderate perspective as alleged job

creator. Insolvency of the German Solarworld AG, decline of

the Chinese “Solar Valley” as well as only few weeks earlier

announced downsizing of around 6,000 out of 27,000 jobs

at Siemens wind energy subsidiary Siemensgamesa are

individual examples.

Therefore, rather the overall situation should be

considered which; according to the “World Energy Investment

2017” report of the International Energy Agency has

changed in the past years. For the first time since the year

2000 the major share of all investments worldwide in the

energy sector of 1,700 bn. $ were made in electricity

production, superseding considerably higher investments

in oil and gas in the prior years. This represents 2.2 % of

the Gross World Product (GWP), however, a decrease by

12 % from 1,900 bn. $ in the past year.

The total of roughly 724 bn. US $ consists of 277 bn. US

$ for grids (39 %), 297 bn. US $ for renewables (41 %) and

143 bn. US $ for conventional production and as a total are

situated clearly in the upper range of the past 20 years. The

total investments inflation-adjusted were 750 bn. US $ in

the year 2000. It has to be taken into account that these

numbers do not display the actual contribution to the

electricity supply and the supply security. Measured in

­production capacity these are 165,000 MW of newly

­installed plants in the renewables and around 95,000 MW

of conventional plants.

Eventually, the actual possible contribution to power

supply reflects an entirely different relation. Natural availability

of the renewables and with the technical avail ability

of the conventional these 165,000 MW would approximately

adequate 25,000 MW of conventional power.

Regarding the distribution of investments and

mentioned further high investments the question remains

for the reasons of job downsizing.

Here, my unloved, because rather nondescript term of

globalisation plays a role.

The market for all facilities and establishments in

electricity production has transformed massively. For one,

the investments have significantly shifted regionally.

­Nowadays it is invested in countries other than Europe or

North America, in Asia, in Africa and in South America.

And, even more essential, also the landscape of manufacturing

has shifted towards Asia.

Additionally, there is a noticeable deterioration of the

political investment setting for conventional electricity

production in western countries, even though it is about

exports and therefore own local jobs. Lacking loan

guarantees and for instance prohibition initiated by French

politicians for any governmental subsidies for the export of

conventional technology aggravates the situation. By the

generalising term “Green Investments” it is hoped for

popularity. It should be questioned if the stakeholders

know that “new” players from Asia promptly close such a

gap by not only bringing along the technology but also

­required financial management for foreign investment.

Neither for the environment nor for jobs in this country

such a general actionism is of any help. Still, politics needs

to turn to indeed difficult conflicting priorities of politics,

economy and citizens (voters).

And this by not only the dimension of “environment”

but also equally several other dimensions such as business

and economic aspects, social interests, jobs and responsibility

for future generations.

The politics has to be granted that of course it is difficult

to nearly impossible to foresightfully valuate single

measures especially for the job market. A reliable economic

model for governmental intervention does not exist.

Diverse models between centrally planned economy and

market liberalism is subject of discussion of the savants

and belongs to the catalogue of the politics.

There is one thing experience and common sense show:

Permanent measures guided by governmental intervention,

be it directly for jobs or particular industries are not

beneficial. They only lead to distortion of the national

performance and will place growing strains on the national

economy. This gradual loss of in the end social security is

dramatic.

In the past decades the peaceful use of nuclear energy

has contributed a considerable share to jobs, social security

and the environment. In the 1990s and the 2000s around

40,000 people were working directly in nuclear energy. In

France, as country with the highest ratio of nuclear energy

it is said that 400,000 jobs are directly or induced related

to nuclear energy. The greatest, but not directly perceptible

benefit of nuclear power and the performance of its

employees lies in their macro-economic contribution. The

non-subsidized jobs contribute significantly to an economically

stable and attractive investment and market environment

through favourable electricity generation costs and

are thus a basis for a secure and viable infrastructure.

However, it is the politics themselves that is called upon

for work and social welfare in an increasingly distorted

energy policy: only a sustainable and fair framework without

permanent money transfers for all technologies creates jobs

and promotes social development before the reality of

globalisation with all its negative consequences will catch up.

Christopher Weßelmann

– Editor in Chief –

EDITORIAL

Editorial

Energy Economy: Under the Banner of Jobs


atw Vol. 63 (2018) | Issue 1 ı January

4

EDITORIAL

Christopher

Weßelmann

– Chefredakteur –

Energiewirtschaft:

Im Zeichen von Arbeitsplätzen

Liebe Leserin, lieber Leser, es sind wahrlich keine guten Nachrichten, mit denen die zwei großen integrierten

Technologieunternehmen General Electric (GE) und Siemens kurz vor Weihnachten und Jahreswechsel an ihre Belegschaft

gingen: GE kündigte an weltweit rund 12.000 Arbeitsplätze in seiner Kraftwerkssparte zu streichen, 6.800 davon in Europa;

Siemens plant, 6.900 Stellen zu streichen, 6.100 davon im Kraftwerksbereich. Damit setzen beide Konzerne ihren Arbeitsplatzabbau

in diesem Bereich fort. GE hatte schon mit der Übernahme des Kraftwerksbereichs der französischen Alstom massiv in

Europa Stellen gestrichen und bei Siemens ist Arbeitsplatzabbau im Energiesektor fast schon ein Dauerthema.

Von beiden Unternehmen wird angeführt, dass dies letzt endlich

eine Reaktion auf die veränderte weltweite Investitionslandschaft

sei. Die Anzahl von Projekten für die vor allem diese

Sparten tragenden konventionellen Gas- und Kohlekraft werke

würden deutlich zurück gehen, sodass bei gleich bleibendem

Angebot der ein weiterer Preisverfall unausweichlich sei und

damit nicht mehr ausreichend Margen gegeben seien.

Mit wenigen Ausnahmen waren pauschale Schuldzuweisungen,

wie in der Vergangenheit üblich, mit dem Hinweis

„man habe sich nicht frühzeitig genug auf andere Produkte,

sprich erneuerbare Energien, gestützt,“ ausgeblieben. Vielleicht

waren hier politischen Protagonisten doch einige Fakten aus

dem Jahresverlauf präsent, die auch für die Erneuerbaren als

vermeintlicher Arbeitsplatzmotor eher dämpfende Aussichten

bescheinigen. Die Insolvenz der deutschen Solarworld AG, der

Niedergang des chinesischen „Solar Valley“ und auch der nur

wenige Wochen vorher seitens der Siemens Wind- Tochter

„ Siemensgamesa“ angekündigte weltweite Stellen abbau im

Umfang von voraussichtlich 6.000 Arbeitsplätzen – bei einer

Gesamtbelegschaft von rund. 27.000 noch wesentlich eingreifender

als beim konventionellen Geschäft – sind Einzelbeispiele.

Zu betrachten ist also eher die Gesamtsituation, die sich

gemäß dem aktuellen „World Energy Investment 2017“ ­Report

der Internationalen Energie Agentur (International Energy

Agency) deutlich in den vergangenen Jahren gewandelt hat.

Von den erfassten weltweiten Gesamtinvestitionen im

Energie sektor in Höhe von 1.700 Mrd. US-$ im Jahr 2016

(dies entspricht rund 2,2 % des globalen Bruttosozialproduktes,

bedeutet im Vorjahresvergleich mit 1.900 Mrd. US-$

aber auch einen Rückgang um 12 %) entfällt erstmals seit dem

Jahr 2000 der größte Anteil auf die Stromerzeugung, die

damit die in den Vorjahren deutlich höheren Investitionen in

den Öl & Gas Sektor ablöst. Die rund 724 Mrd. US-$ teilen sich

auf in 277 Mrd. US-$ für Netze (39 %), 297 US-$ für Erneuerbare

(41 %) und 143 Mrd. US-$ für konventionelle Erzeugung

und sie liegen noch deutlich im oberen Bereich der vergangenen

20 Jahre – Inflationsbereinigt lagen z.B. die Gesamtinvestitionen

im Jahr 2000 bei rund 750 Mrd. US-$. Zu beachten

ist, dass diese Zahlen nicht den tatsächlichen Beitrag für

Stromversorgung und Stromversorgungssicherheit abbilden.

In Erzeugungsleistung gemessen ergeben sich für das Jahr

2016 rund 165.000 MW an neu installierten Anlagen im

­Bereich der Erneuerbaren und rund 95.000 MW an konventionellen

Anlagen. Der schlussendlich tatsächliche, mögliche

Beitrag für die Energieversorgung spiegelt dann noch ein ganz

anderes Verhältnis wider, denn aufgrund der natürlichen

Verfügbarkeit bei den Erneuerbaren und mit den technischen

Verfügbarkeiten der Konventionellen würden die 165.000 MW

in etwa 25.000 MW an konventioneller Leistung entsprechen.

In Summe einer Betrachtung des Investitionskuchens

sowie der erwähnten weiter hohen Investitionen verbleibt die

Frage nach den Gründen für den Stellenabbau.

Hier spielt dann doch einmal mein ungeliebter, weil meist

ohne Inhalte gefüllter Begriff der Globalisierung die Rolle.

Der Markt für alle Anlagen und Einrichtungen in der Stromerzeugung

hat sich stark gewandelt. Zum einen, weil sich die

Investitionen regional erheblich verschoben haben. Investiert

wird heute außerhalb von Europa und Nord­amerika, in Asien, in

Afrika, in Südamerika. Und, was noch wesentlicher ist, auch die

Herstellerlandschaft hat sich in Richtung Asien verschoben.

Hinzu kommt eine erkennbare Verschlechterung des poli tischen

Investitionsumfeld für die konventionelle Stromer zeugung in

westlichen Ländern, auch wenn es um Exporte und damit

eigene, heimische Arbeits plätze geht. Fehlende Kreditbürgschaften

und eine z.B. von Politiken in Frankreich gebotenes

Verbot für jegliche staat liche Unterstützung beim Export

konventioneller Technologie verschärfen die Situation. Unter

dem pauschalisierenden Begriff „Grüner Investitionen“ erhofft

man sich Popularität. Ob die Akteure wissen, dass „neue“ Akteure

aus Asien prompt eine solche sich auftuende ­Lücke schließen

und nicht nur die Technologie mitbringen sondern auch das für

Auslands investitionen erforderliche Finanzmanagement, sollte

gefragt werden. Für die Umwelt bringt solcher pauschaler

Aktio nismus jedenfalls nichts, und für Arbeitsplätze hierzulande

auch nicht. Dennoch muss sich Politik im zugegeben

schwie rigen Spannungsfeld von Politik, Wirtschaft und Bürger

(Wähler) nicht nur der Dimension „Umwelt“ zuwenden,

sondern weitere wie Volk- und Betriebswirtschaftliche Aspekte,

soziale Interessen, Arbeitsplätze und Verantwortung für

­zukünftige Generationen gleichermaßen berücksichtigen.

Dabei ist der Politik zugute zu halten, dass es natürlich

schwierig bis unmöglich ist, Einzelmaßnahmen gerade für

den Arbeitsmarkt vorausschauend zu bewerten. Ein ver lässliches

Volkswirtschaftliches Modell für staatliche Interventionen

gibt es nicht. Die verschiedensten Modelle zwischen

Planwirtschaft und vollständigem Marktliberalismus sind seit

jeher Diskussionsgegenstand der Gelehrten und gehören zum

Katalog der Politik. Eines zeigen die einfache Erfahrung und

der gesunde Menschenverstand: Dauerhaft durch staatliche

Interventionen gelenkte Maßnahmen, sei es direkt für Arbeitsplätze

oder einzelne Wirtschaftszweige, zeichnen sich

nicht aus. Diese führen nur zu Verzerrungen der nationalen

­Leistung und setzen die Nationalökonomie im heutigen

globalen Wettbewerb später unter Druck. Dieser schleichende

Verlust von am Ende sozialer Sicherheit ist dramatisch.

Die friedliche Nutzung der Kernenergie hat in den vergangenen

Jahrzehnten ihren volkswirtschaftlich bedeutenden

Beitrag für Arbeitsplätze, soziale Sicherung und Umwelt

geleistet. In den 1990er und 2000er Jahren waren in Deutschland

rund 40.000 Menschen direkt für die Kernenergie tätig. In

Frankreich, dem Land mit dem weltweit höchsten Kernenergieanteil,

wird von 400.000 Arbeitsplätzen gesprochen, die direkt

oder induziert in Zusammenhang mit der Kernenergie stehen.

Der weitaus größte aber nicht direkt fühlbar Nutzen der Kernenergie

und der Leistung ihrer Beschäftigten liegt im volkswirtschaftlichen

Beitrag. Die nicht subventionierten Arbeitsplätze

tragen über günstige Stromerzeugungskosten wesentlich für

ein ökonomisch stabiles und attraktives Investitions- und

Marktumfeld bei und sind damit eine Grundlage für eine

sichere und Erfolg versprechende Infrastruktur.

Doch gefordert ist für Arbeit und Soziales in einer immer

mehr und mehr verzerrten Energiepolitik dann doch die Politik

selbst: Nur zukunftsfähige, auf Dauerhaftigkeit zielende und

faire Rahmenbedingungen ohne dauerhafte Geldtransfers für

alle Technologien schaffen Arbeitsplätze und fördern soziale

Entwicklung, bevor einen die Realität der Globalisierung mit

allen negativen Konsequenzen einholen wird.

Editorial

Energy Economy: Under the Banner of Jobs


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atw Vol. 63 (2018) | Issue 1 ı January

6

Issue 1

January

CONTENTS

13

ETSON Strategic

Orientations

on Research Activities

| | View of two of four reactors at the Ringhals nuclear power plant site in the Varberg Municipality approximately 65 km south

of Gothenburg, Sweden. (Courtesy: Vattenfall AB)

Editorial

Energy Economy: Under the Banner of Jobs 3

Energiewirtschaft:

Im Zeichen von Arbeitsplätzens 4

Abstracts | English 8

Abstracts | German 9

Energy Policy, Economy and Law

ETSON Strategic Orientations

on Research Activities.

ETSON Research Group Activity 13

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

Spotlight on Nuclear Law

Council Regulation of the European Dual Use

Regulation – A Never Ending Story? 19

Die Novellierung der europäischen Dual-Use

Verordnung – eine unendliche Geschichte? 19

Ulrike Feldmann

10

DAtF Notes 20

| | AP1000 new build in Haiyang, China.

Inside Nuclear with NucNet

UK Is Leading the Way

With Clear Strategy for Nuclear 10

NucNet

Calendar 12

21

| | Nuclear Triple “S”.

Contents


atw Vol. 63 (2018) | Issue 1 ı January

7

Environment and Safety

Nuclear Safety, Security and Safeguards:

An Application of an Integrated Approach 21

Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy,

Robert Rodger and Jonathan Scott

Fuel

Review of Fuel Safety Criteria in France 38

Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne

and Olivier Dubois

CONTENTS

Operation and New Build

Clearance of Surface-contaminated Objects

from the Controlled Area of a Nuclear Facility:

Application of the SUDOQU Methodology 29

F. Russo, C. Mommaert and T. van Dillen

AMNT 2017

Key Topic |

Outstanding Know-How & Sustainable Innovations

Technical Session:

Reactor Physics, Thermo and Fluid Dynamics

Neutron Flux Oscillations Phenomena 44

Joachim Herb

Key Topic |

Enhanced Safety & Operation Excellence

Focus Session:

Radiation Protection 46

29

Erik Baumann and Angelika Bohnstedt

| Variation of the total dose values in the analysed scenarios.

AMNT 2018

Preliminary Programme 47

|38

Decommissioning and Waste Management

Carbon-14 Speciation During Anoxic Corrosion

of Activated Steel in a Repository Environment 34

KTG Inside 54

E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat

34

54

| | KTG Inside. Horst Kemmeter, speaking at a WiN meeting in Biblis.

| | Sketch of the reactor.

News 57

Nuclear Today

‘Newcomer’ Nuclear Nation

Leads Way into New Nuclear Year 66

John Shepherd

Imprint 11

| | Topics reviewed in the frame of French rulemaking

on fuel safety criteria.

AMNT 2018: Registration Form . . . . . . . . . . . Insert

Contents


atw Vol. 63 (2018) | Issue 1 ı January

8

ABSTRACTS | ENGLISH

UK Is Leading the Way With Clear Strategy

for Nuclear

NucNet | Page 10

The UK is Europe’s most prominent leader in nuclear

development because of the government’s clear

strategy of supporting nuclear energy as part of its

future energy mix, a senior official from ­US-based

nuclear equipment manufacturer Westinghouse

Electric Company said. Mr Kirst told that the UK

­government’s decision to support the financing of

new energy projects, including nuclear, by way of a

contract for difference scheme was a breakthrough.

Additionally potential for nuclear development in

other EU member states is possible in Poland and the

Czech Republic where also new nuclear capacities

are possible. Potential exists also in non-EU countries

like Turkey and the Ukraine.

ETSON Strategic Orientations on Research

Activities. ETSON Research Group Activity

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 | Page 13

In 2011, ETSON published the “Position Paper of

the Technical Safety Organizations: Research Needs

in Nuclear Safety for Gen 2 and Gen 3 NPPs”. This

paper, published only a few months after the

Fukushima- Daiichi severe accidents, presented the

priorities for R&D on the main pending safety

­issues. It was produced by the ETSON Research

Group (ERG) that has the mandate of identifying

and prioritizing safety research needs, sharing

­information on research projects in which ETSON

members are involved, defining and launching new

research projects and disseminating knowledge

among ETSON members. Six years after this

publication, many R&D international projects

­finished in diverse frames, and other ones have

started. In particular a lot of work was done (and is

going on…) on the analysis of the Fukushima-

Daiichi severe accidents. Meanwhile a roadmap on

research on Gen. 2 and 3 nuclear power plants

(NPP), including safety aspects, was produced by

the NUGENIA association, followed by a more

­detailed document as “NUGENIA global vision”. It

was also demonstrated that the ETSON R&D

priorities were consistent with the implementation

of the 2014 Euratom Directive on safety of nuclear

installations.

Council Regulation of the European Dual

Use Regulation – A Never Ending Story?

Ulrike Feldmann | Page 19

For the first time, the EC Council Regulation of

19 December 1994 established a Community ­regime

for the control of exports of dual-use items. In 2000,

the first major revision of the dual-use regime came

into force, subjecting not only sensitive material, i.

e. plutonium and highly enriched uranium, but also

the entire category 0 (nuclear material, installations,

equipment) to a licensing requirement for intra-

Community shipments. This revision was revised a

few months later due to inappropriate content by

removing a small proportion of nuclear goods. A

further comprehensive new revision was published

in 2009. However, the EU Commission’s current

proposal to revise Annex IV of the regulation does

not do justice to the objective of free trade of goods

and the maintenance of the competitiveness of

European industry from the point of view of the

European nuclear industry, as well as from the point

of view of the non-nuclear industry in the EU.

Nuclear Safety, Security and Safeguards:

An Application of an Integrated Approach

Howard Chapman, Jeremy Edwards,

Joshua Fitzpatrick, Colette Grundy,

Robert Rodger and Jonathan Scott | Page 21

National Nuclear Laboratory has recently produced

a paper regarding the integrated approach of

nuclear safety, security and safeguards. The paper

considered the international acknowledgement of

the inter-relationships and potential benefits to be

gained through improved integration of the nuclear

‘3S’; Safety, Security and Safeguards. It considered

that combining capabilities into one synergistic

team can provide improved performance and value.

This approach to integration has been adopted, and

benefits realised by the National Nuclear ­Laboratory

through creation of a Safety, Security and

Safeguards team. In some instances the interface is

clear and established, as is the case between safety

and security in the areas of Vital Area Identification.

In others the interface is developing such as the

utilisation of safeguards related techniques such as

nuclear material accountancy and control to

enhance the security of materials. This paper looks

at a practical example of the progress to date in

implementing Triple S by a duty holder.

Clearance of Surface-contaminated Objects

from the Controlled Area of a Nuclear

Facility: Application of the SUDOQU

Methodology

F. Russo, C. Mommaert and T. van Dillen | Page 29

The lack of clearly defined surface-clearance levels in

the Belgian regulation led Bel V to start a collaboration

with the Dutch National Institute for Public

Health and the Environment (RIVM) to evaluate the

applicability of the SUDOQU methodology for the

derivation of nuclide-specific surface-clearance

criteria for objects released from nuclear facilities.

SUDOQU is a methodology for the dose assessment

of exposure to a surface-contaminated object, with

the innovative assumption of a time-dependent

­surface activity whose evolution is influenced by

removal and deposition mechanisms. In this work,

calculations were performed to evaluate the annual

effective dose resulting from the use of a typical

­office item, e.g. a bookcase. Preliminary results ­allow

understanding the interdependencies between the

model’s underlying mechanisms, and show a strong

sensitivity to the main input parameters. The results

were benchmarked against those from a model described

in Radiation Protection 101, to investigate

the impact of the model’s main assumptions. Results

of the two models were in good agreement.

The SUDOQU methodology appears to be a flexible

and powerful tool, suitable for the proposed application.

Therefore, the project will be extended to

more generic study cases, to eventually develop surface-clearance

levels applicable to objects leaving

nuclear facilities.

Carbon-14 Speciation During Anoxic

Corrosion of Activated Steel in a Repository

Environment

E. Wieland, B.Z. Cvetkovic, D. Kunz,

G. Salazar and S. Szidat | Page 34

Radioactive waste contains significant amounts

of 14 C which has been identified a key radionuclide

in safety assessments. In Switzerland, the 14 C inventory

of a cement-based repository for low- and

intermediate-level radioactive waste (L/ILW) is

mainly associated with activated steel (~85 %). 14 C

is produced by 14 N activation in steel parts exposed

to thermal neutron flux in light water reactors.

Release of 14 C occurs in the near field of a deep

geological repository due to anoxic corrosion of

activated steel. Although the 14 C inventory of the

L/ILW repository and the sources of 14 C are well

known, the formation of 14 C species during steel

corrosion is only poorly understood. The aim of the

present study was to identify and quantify the

14 C-bearing carbon species formed during the

anoxic corrosion of iron and steel and further to

determine the 14C speciation in a corrosion experiment

with activated steel. All experiments were

conducted in conditions similar to those anticipated

in the near field of a cement-based repository.

Review of Fuel Safety Criteria in France

Sandrine Boutin, Stephanie Graff,

Aude Foucher-Taisne and Olivier Dubois | Page 38

Fuel safety criteria for the first barrier, based on

state-of-the-art at the time, were first defined in the

1970s and came from the United States, when the

French nuclear program was initiated. Since then,

there has been continuous progress in knowledge

and in collecting experimental results thanks to the

experiments carried out by utilities and research

institutes, to the operating experience, as well as to

the generic R&D programs, which aim notably at

improving computation methodologies, especially

in Reactivity-Initiated accident and Loss-of-Coolant

Accident conditions. In this context, the French

utility EDF proposed new fuel safety criteria, or

reviewed and completed existing safety demonstration

covering the normal operating, incidental

and accidental conditions of Pressurised Water

­Reactors. IRSN assessed EDF’s proposals and presented

its conclusions to the Advisory Committee

for Reactors Safety of the Nuclear Safety Authority

in June 2017. This review focused on the relevance

of historical limit values or parameters of fuel safety

criteria and their adequacy with the state-of-the-art

concerning fuel physical phenomena (e.g. Pellet-

Cladding Mechanical Interaction in incidental conditions,

clad embrittlement due to high temperature

oxidation in accidental conditions, clad ballooning

and burst during boiling crisis and fuel melting).

AMNT 2017: Outstanding Know-How &

Sustainable Innovations – Technical Session:

Reactor Physics, Thermo and Fluid Dynamics

Enhanced Safety & Operation Excellence –

Focus Session: Radiation Protection

Joachim Herb, Erik Baumann and

Angelika Bohnstedt | Page 44

Summary report on the Key Topics “Outstanding

Know-How & Sustainable Innovations – Technical

Session: Reactor Physics, Thermo and Fluid

Dynamics” and “Enhanced Safety & Operation Excellence

– Focus Session: Radiation Protection” of

the 48 th Annual Meeting on ­Nuclear Technology

(AMNT 2017) held in Berlin, 16 to 17 May 2017.

‘Newcomer’ Nuclear Nation Leads Way Into

New Nuclear Year

John Shepherd | Page 66

At the start of a new year, it is appropriate that a

‘newcomer’ nuclear nation has launched work on

building its first nuclear power plant. First nuclear

safety-related concrete has been poured for the

plant at Rooppur in Bangladesh – making the South

Asia nation the first in 30 years to start building its

first commercial reactor unit following the United

Arab Emirates in 2012 and Belarus in 2013.

Despite setbacks that nuclear has endured in recent

years, there are nearly 60 reactors under construction

around the world, mostly in Asia. Some

447 commercial reactor units are in operation in

30 countries.

Abstracts | English


atw Vol. 63 (2018) | Issue 1 ı January

Großbritannien ist führend mit

seiner klares Strategie für die Kernenergie

NucNet | Seite 10

Großbritannien ist in Europa führend bei der

zukünftigen Kernenergieentwicklung aufgrund der

klaren Strategie der Regierung, die Kernenergie als

Teil ihres zukünftigen Energiemixes zu unterstützen.

Dies hob Michael Kirst voms US-Kern technik

unternehmen West inghouse Electric Company

hervor. Die Entscheidung der britischen Regierung,

die Finanzierung neuer Energieprojekte, einschließlich

der Kernenergie, im Wege eines

Differenz vertrags zu unterstützen, sei ein Durchbruch

gewesen. Darüber hinaus sind in anderen

EU-Mitgliedsstaaten, wie Polen und Tschechien,

Potenziale auch für neue Kernkraftwerke vorhanden.

Potenziale bestehen auch in Nicht-EU-­

Ländern, so in der Türkei und der Ukraine.

ETSON Strategische Ausrichtung

für Forschungsaktivitäten.

Aktivitäten der ETSON-Forschungsgruppe

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 und P. Liska | Seite 13

Im Jahr 2011 veröffentlichte ETSON das „Positionspapier

der Technischen Sicherheitsorganisationen:

Forschungsbedarf für die nukleare Sicherheit für

die Kernkraftwerke der Generation 2 und 3“.

Nur wenige Monate nach den schweren Unfällen

von Fukushima-Daiichi wurden Prioritäten für

Forschung und Entwicklung in Bezug auf wichtige

noch offene Fragen zur Sicherheit vorgestellt. Das

Positionspapier wurde von der ETSON Research

Group (ERG) erstellt, die den Auftrag hat, den

Bedarf an Sicherheitsforschung zu ermitteln und

zu priorisieren, Informationen über Forschungs projekte

von ETSON-Mitgliedern auszutauschen, neue

Forschungsprojekte zu definieren und zu lancieren

und den ETSON-Mitgliedern Informationen bereit

zu stellen. Sechs Jahre nach dieser Veröffentlichung

sind viele internationale F&E-Projekte abge schlossen,

andere haben begonnen. Insbesondere an der

Analyse der schweren Unfälle von Fukushima-

Daiichi wurde gearbeitet. Zwischenzeitlich hat

­NUGENIA einen Fahrplan für die Sicherheitsforschung

erstellt und das detaillierte Dokument

„­NUGENIA Global Vision“ veröffentlicht. Die F&E-­

Prioritäten von ETSON stehen zudem in ­Einklang

mit der Umsetzung der Euratom-Richt linie 2014.

Die Novellierung der europäischen

Dual Use-Verordnung – eine unendliche

Geschichte?

Ulrike Feldmann | Seite 19

Erstmalig wurde mit der Verordnung des Rates vom

19.12.1994 eine Gemeinschaftsregelung für die

Ausfuhrkontrolle von Gütern mit doppeltem Verwendungszweck

geschaffen. Im Jahr 2000 fand die

erste größere Revision der Dual-Use Regelungen

statt, mit der für den Nuklearbereich nicht nur

sensitives Material, d.h. Plutonium und hochangereichertes

Uran sondern die gesamte Kategorie

0 (Nuklearmaterial, Anlagen, Ausrüstung) auch

einer Genehmigungspflicht für die innergemeinschaftliche

Verbringung unterworfen wurde, die

aufgrund nicht angebrachter Inhalte wenige

Monate später revidiert wurde durch Herausnahme

eines kleinen Teils von Nukleargütern. 2009

erschien eine weitere umfassende neue Revision.

Der aktuelle Revisionsvorschlag der EU-Kommission

zum Annex IV der Verordnung wird dem

Ziel des freien Warenverkehrs und dem Erhalt der

Wettbewerbsfähigkeit der europäischen Industrie

jedoch aus Sicht der europäischen Nuklearindustrie

wie auch aus Sicht der nicht-nuklearen Industrie in

der EU nicht gerecht.

Nukleare Sicherheit, Gefahrenabwehr und

Safeguards: Anwendung eines integrierten

Ansatzes

Howard Chapman, Jeremy Edwards,

Joshua Fitzpatrick, Colette Grundy,

Robert Rodger und Jonathan Scott | Seite 21

Das National Nuclear Laboratory hat eine Studie

über einen integrierten Ansatz zur nuklearen

Sicherheit, sowie Gefahrenabwehr und Safeguards

erstellt. Vorgestellt werden die Wechselbeziehungen

und Vorteile, die durch eine bessere

Integration der nuklearen“3S“ (Safety, Security and

Safeguards) erzielt werden können. Ein integrierter

Anssatz kann dabei potenzielle Synergien schöpfen

und Vorteile erschließen. Dieser integrierte Ansatz

wurde bei der Bildung eines Teams für Sicherheit,

Gefahrenabwehr und Safeguards des NNL übernommen.

In einigen Anwendungsfällen sind die

Schnittstellen eindeutig in anderen müssen sie

weiter entwickelt werden. Vorgestellt wird ein

praktisches Beispiel für die bisherigen Fortschritte

bei der Umsetzung von Triple S anhand eines

Sicherheitsbeauftragen.

Freigabe oberflächenkontaminierter

Objekte aus dem Kontrollbereich

eines Kernkraftwerkes

Anwendung der SUDOQU-Methode

F. Russo, C. Mommaert und T. van Dillen | Seite 29

Das Fehlen definierter Grenzwerte für die Oberflächen­kontamination

in der betreffenden bel­gischen

Verordnung veranlasste Bel V in Zusammenarbeit mit

dem National Institute for Public Health and the

­Environment (Niederlande) die Anwendung der

SUDOQU-Methode für die Ableitung nuklidspezifischer

Oberflächendosiskriterien für Objekte zu

evaluieren, die aus kerntechnischen Anlagen freigemessen

werden sollen. SUDOQU ist eine Methode zur

Dosisbewertung der Exposition eines oberflächenkontaminierten

Objekts unter der Annahme einer

zeitabhängigen Oberflächenaktivität, deren Entwicklung

von Entfernungs- und Ablagerungsmechanismen

beeinflusst wird. Berechnungen zur Ermittlung

der effektiven Jahresdosis werden vorgestellt,

die sich aus der Verwendung eines typischen Büroartikels

ergibt. Vorläufige Ergebnisse erlauben es,

die Wechselwirkungen zwischen den zugrunde

liegenden Mechanismen des Modells zu verstehen

und zeigen eine starke Sensitivität gegenüber den

wichtigsten Eingangsparametern. Die Ergebnisse

wurden mit denen eines weiteren beschriebenen

Modells verglichen. Die Ergebnisse der beiden

Modelle stimmten gut überein.

Die SUDOQU-Methode scheint ein flexibles und

leistungsfähiges Werkzeug zu sein, das für die

vorgeschlagene Anwendung geeignet ist. Das

Projekt wird auf allgemeinere Fälle ausgeweitet, um

Oberflächenfreigabekriterien zu entwickeln, die für

Objekte aus kerntechnischen Anlagen anwendbar

sind.

Kohlenstoff-14-Verhalten bei der

anaerober Korrosion von aktiviertem Stahl

in einer Endlagerumgebung

E. Wieland, B.Z. Cvetkovic, D. Kunz,

G. Salazar und S. Szidat | Seite 34

Radioaktive Abfälle enthalten signifikante Mengen

von 14 C, die in Sicherheitsbewertungen als ein

­Leitradionuklid identifiziert wurden. In der Schweiz

wird das 14 C-Inventar eines Endlagers für mit Zement

konditionierte schwach- und mittelradioaktive

Abfälle hauptsächlich von aktiviertem Stahl (~85 %)

dominiert. 14 C wird durch 14 N-Aktivierung in Stahlkomponenten

gebildet, die dem ther­mischen Neutronenfluss

in Leichtwasserreaktoren ausgesetzt

sind. Die Freisetzung von 14 C erfolgt im Nahfeld eines

geologischen Tiefenlagers durch anaerobe Korrosion

des aktivierten Stahls. Obwohl das 14 C-Inventar des

Endlagers und die Quellen von 14 C bekannt sind, ist

zur Bildung von 14 C-Ver bindungen bei der Korrosion

von Stahl nur wenig bekannt. Das Ziel der vorliegenden

Studie war es, die 14 C-haltigen Kohlenstoffver

bindungen, die während der anaeroben

Korrosion von Eisen und Stahl gebildet werden, zu

identifizieren und quan­tifizieren und die 14 C-Verbindungen

in einem Korrosionsexperiment mit

aktiviertem Stahl zu bestimmen. Alle Experimente

wurden unter ähn lichen Bedingungen wie im

­Nahfeld eines End­lagers durchgeführt.

Überprüfung der Kriterien für die Sicherheit

von Kernbrennstoff in Frankreich

Sandrine Boutin, Stephanie Graff,

Aude Foucher-Taisne und Olivier Dubois | Seite 38

Die Kriterien für die Sicherheit der ersten Barriere

des Kernbrennstoff gegenüber Spalt produkt freisetzung

wurden in den 1970er Jahren definiert als

das französische Atomprogramm initiiert wurde.

Seitdem haben sich Wissen und Erfahrungen

dank der von den Kernkraftwerksbetreibern und

Forschungsinstituten durchgeführten Experimente,

Betriebserfahrungen sowie generischer F&E-

Programme, die insbesondere auf die Verbesserung

der Berechnungsmethoden abzielen, kontinuierlich

weiterentwickelt. Der französische Energieversorger

EDF schläg neue Kriterien für die Brennstoffsicherheit

vor und überprüft und ergänzt

be stehende Sicherheitskriterien, die sich auf

die normalen Betriebs-, Ereignis- und Unfallbedingungen

von Druckwasserreaktoren beziehen.

IRSN hat die Vorschläge des EDF bewertet und seine

Schlussfolgerungen im Juni 2017 dem Beratenden

Ausschuss für Reaktorsicherheit der Französischen

Behörde für nukleare Sicherheit vorgelegt.

AMNT 2017: Outstanding Know-How &

Sustainable Innovations – Technical Session:

Reactor Physics, Thermo and Fluid Dynamics

Enhanced Safety & Operation Excellence –

Focus Session: Radiation Protection

Joachim Herb, Erik Baumann und

Angelika Bohnstedt | Seite 44

Zusammenfassender Bericht zu den Sessions der

Key Topics „Outstanding Know-How & Sustainable

Innovations – Technical Session: Reactor Physics,

Thermo and Fluid Dynamics“ und „Enhanced Safety

& Operation Excellence – Focus Session: Radiation

Protection“ des 48 th Annual Meeting on Nuclear

Technology (AMNT 2017), Berlin, 16 bis 17 Mai

2017.

Ein Newcomer führt die Kernenergie

in das Neue Jahr

John Shepherd | Seite 66

Zu Beginn des neuen Jahres weist ein “Newcomer“

mit dem Bau des ersten Kernkraftwerks den Weg.

Für das Fundament des Kernkraftwerks in Rooppur

in Bangladesch wurde der erste Beton gegossen.

­Damit ist die südasiatische Nation eine weitere, die

nach den Vereinigten Arabischen Emiraten 2012

und Weißrussland 2013, mit dem Bau eines ersten

kommerziellen Reaktors begonnen hat.

Trotz der Rückschläge für die Kernenergie in den

letzten Jahren, sind weltweit fast 60 Reaktoren in

Bau, vor allem in Asien. 447 kommerzielle Reaktorblöcke

sind in 30 Ländern in Betrieb.

9

ABSTRACTS | GERMAN

Abstracts | German


atw Vol. 63 (2018) | Issue 1 ı January

10

INSIDE NUCLEAR WITH NUCNET

UK Is Leading the Way

With Clear Strategy for Nuclear

NucNet

The UK is Europe’s most prominent leader in nuclear development because of the government’s clear

­strategy of supporting nuclear energy as part of its future energy mix, a senior official from US-based nuclear

equipment manufacturer Westinghouse Electric Company said.

Michael Kirst, Westinghouse’s vice-president of

strategy for Europe, Middle East and Africa

(EMEA), warned, however, that choices about nuclear

development must be based on technology, and not on the

type of financing package. “We now have a banking ­contest

and not a technology contest and this is not healthy for the

industry or the energy system,” he said.

Mr Kirst told reporters in Brussels that the UK government’s

decision to support the financing of new energy

projects, including nuclear, by way of a contract for

difference (CfD) scheme was a breakthrough.

“The UK government made it clear they need these new

nuclear capacities”, he said. The UK model provides a “fair

foundation” where all low-carbon technologies were given

exactly the same access to state support.

Mr Kirst said Westinghouse, a privately owned company,

does not have access to state support on demand, unlike its

major competitors in the nuclear industry, which are

“somehow state-owned or state-controlled”. A clear market

signal for private investors in nuclear development is therefore

essential because it allows choices based on technology,

rather than on a financing package, Mr Kirst said.

Speaking about NuGen’s planned three-unit Moorside

nuclear project in Cumbria, northwest England, the

company’s president for EMEA, Luc Van Hulle, said there

are “a couple of options on the table” and Westinghouse’s

AP1000 Generation III+ pressurised water reactor

technology is still potentially one of these options.

The future of the Moorside project to build three

AP1000s has been overshadowed by Westinghouse’s filing

for Chapter 11 bankruptcy protection in the US in March

2017, along with Westinghouse owner Toshiba’s financial

woes and its decision to no longer serve as a contractor of

engineering, procurement and construction for overseas

nuclear projects.

Mr Van Hulle said the Moorside project became “more

complicated” after Engie sold its 40 % stake in NuGen to

Toshiba in April 2017, making the Japanese company the

sole owner of the project. But he said Westinghouse is

­confident that the project will proceed “one way or

another”. He said the fate of the project is in the hands of

the UK government and NuGen’s owner Toshiba.

Last month state media reported that China General

Nuclear Power Corporation (CGN) is considering investing

in Moorside, while in March 2017, South Korea’s Korea

Electric Power Corporation (Kepco) expressed an interest in

taking a stake in NuGen.

Mr Van Hulle said that holding on to the AP1000 design

will be the securest and fastest way to realise the Moorside

project because the plant completed the UK’s generic

design assessment (GDA) review by regulators in the UK in

March 2017.

If NuGen chooses another technology, the process of

going through another GDA process could delay the project

by four or five years, he said.

“Clearly there will be a shift in the start date from 2025

to later in the 2020s, but the plant could still be up and

running before 2030,” NuGen’s chief executive officer Tom

Samson told Reuters last week.

Mr Samson said the timing will largely depend on the

technology choice, because the new bidders may want to

bring in their own designs. However, Mr Samson said:

“We are not ruling out any technology at this stage.”

In the US, the expected delay to the Vogtle nuclear

project and the cancellation of the Summer project in

South Carolina was not related to the AP1000 technology,

Mr Van Hulle said.

He said the AP1000 design is “safe and sound” and the

AP1000 reactor units being built in China will prove this

once they enter commercial operation.

There are four AP1000 nuclear units under construction

in China – two at Sanmen and two at Haiyang – all expected

to become commercially operational in 2018.

| | AP1000 new build in Haiyang, China.

South Carolina Electric and Santee Cooper, the two US

utilities that co-own the Summer AP1000 project, decided

to suspend its construction in July 2017 quoting cost

overruns and schedule delays.

Mr Van Hulle said the utilities’ decision to stop construction

was “saddening” because of the advanced stage

of development, with all nuclear steam supply systems

having been installed. He said the Summer units will not be

completed in the “foreseeable future”, but there is a

possibility that a new owner could take over the project.

In September 2017, the owners of the two-unit Vogtle

AP1000 project in Georgia recommended completing

construction, despite Westinghouse’s financial woes and

increased costs.

The two new reactors at Vogtle, units 3 and 4, under

construction since 2013, represent the first US deployment

of the AP1000 technology.

According to Mr Van Hulle, despite its current difficulties

in the US, Westinghouse has a “very sound base

business” which will serve as the backbone of the

company’s future.

In August 2017, Westinghouse submitted a five-year

business plan to the company’s debtor-in-possession (DIP)

financing lenders and the unsecured creditors committee.

Inside Nuclear with NucNet

UK Is Leading the Way With Clear Strategy for Nuclear ı NucNet


atw Vol. 63 (2018) | Issue 1 ı January

The company said at the time that this marked a critical

milestone in the Chapter 11 bankruptcy process.

The plan integrates Westinghouse’s initiatives into a

five-year financial forecast and would result in projected

savings of $20 5m (€ 174 m) expected to improve earnings

before interest, taxes, depreciation and amortisation

(EBITDA) over the five-year term.

Westinghouse said the plan supports the operation of its

core businesses and its new projects business. One component

of the savings will be global staff reductions, starting

with 7 % of staff being made redundant in fiscal year 2017.

Since filing for Chapter 11 in March 2017, Westinghouse

has obtained approval of an $ 800 m DIP financing ­package

and has negotiated a long-term services agreement with

Southern Nuclear Company for the two AP1000 plants

under construction at Vogtle.

“We are well on track with exiting the Chapter 11

process”, Mr Van Hulle said.

Asked to comment on the potential for nuclear

development in other EU member states, Mr Van Hulle said

Bulgaria, Hungary, Poland, and the Czech Republic could

be expected to develop existing or new nuclear capacities.

Potential exists also in non-EU countries like Switzerland,

Turkey and particularly Ukraine, he said.

According to Mr Kirst, Ukraine’s reactor fleet operates

at an average load factor of about 70 % compared to 85 to

90 % in the US and EU. “There is a lot of untapped energy

that can come online at a very low cost and this is what

we have been suggesting to the Ukrainian government”,

Mr Kirst said.

Mr Van Hulle said there is also an opportunity for

Westinghouse to expand its business relationships in

Ukraine in terms of fuel supplies and plant operation,

availability and energy distribution.

“With the amount of reactors they have they can be

­really influential in non-Russia based VVER technology”,

he noted.

Westinghouse has contracts to supply nuclear fuel for six

VVER reactor units in Ukraine, as well as core monitoring

systems for Zaporozhye-5, and a potential uprate project at

South Ukraine-3.

Ukraine operates a fleet of 15 commercial units, all of

the VVER pressurised water reactor design and built

during the Soviet Era.

Mr Kirst said Ukraine is the only country which has

­significantly diversified its nuclear fuel supply away from

Russia, while EU counties which use VVER reactors remain

completely dependent on Russian supply.

“There have not been significant efforts in Brussels to

address that issue, which is interesting considering that

they are talking about an energy union and the need for

secure and diverse energy supplies”, he said.

Author

NucNet

The Independent Global Nuclear News Agency

Editor responsible for this story: Kamen Kraev

Avenue des Arts 56

1000 Brussels, Belgium

www.nucnet.org

INSIDE NUCLEAR WITH NUCNET 11

| | Editorial Advisory Board

Frank Apel

Erik Baumann

Dr. Maarten Becker

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Dr. Ralf Güldner

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Prof. Dr. Marco K. Koch

Dr. Willibald Kohlpaintner

Ulf Kutscher

Andreas Loeb

Jörg Michels

Roger Miesen

Dr. Thomas Mull

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Dr. Joachim Ohnemus

Prof. Dr. Winfried Petry

Dr. Tatiana Salnikova

Dr. Andreas Schaffrath

Dr. Jens Schröder

Dr. Wolfgang Steinwarz

Prof. Dr. Bruno Thomauske

Dr. Walter Tromm

Dr. Hans-Georg Willschütz

Dr. Hannes Wimmer

Ernst Michael Züfle

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Inside Nuclear with NucNet

UK Is Leading the Way With Clear Strategy for Nuclear ı NucNet


atw Vol. 63 (2018) | Issue 1 ı January

12

CALENDAR

Calendar

2018

30.01.-31.01.2018

NNBS Egypt 2018 — Nuclear New Build Summit

Egypt 2018. Cairo, Egypt, InforValue Consulting

Company, nuclearegypt.com

05.02.-07.02.2018

Components and Structures under Severe

Accident Loading Cossal (COSSAL). Cologne,

Germany. OECD/NEA, GRS,

www.grs.de, www.oecd-nea-org

07.02.-08.02.2018

8. Symposium Stilllegung und Abbau

kerntechnischer Anlagen. Hanover, Germany.

TÜV Nord, www.tuev.nord.de

26.02.-01.03.2018

Nuclear and Emerging Technologies for Space

2018. Las Vegas, NV, USA. American Nuclear Society

(ANS), www.ans.org

01.03.2018

7. Fachgespräch Endlagerbergbau. Essen,

Germany, DMT, GNS, www.dmt-goup.com

04.03.-09.03.2018

82. Jahrestagung der DPG. Erlangen, Germany,

Deutsche Physikalische Gesellschaft (DPG),

www.dpg-physik.de

11.03.-17.03.2018

International Youth Nuclear Congress (IYNC).

Bariloche, Argentina, IYNC and WiN Global,

www.iync.org/category/iync2018/

26.03.-27.03.2018

Fusion energy using tokamaks: can development

be accelerated? London, United Kingdom,

The Royal Society, royalsociety.org

08.04.-11.04.2018

International Congress on Advances in Nuclear

Power Plants – ICAPP 18. Charlotte, NC, USA,

American Nuclear Society (ANS), www.ans.org

08.04.-13.04.2018

11 th International Conference on Methods and

Applications of Radioanalytical Chemistry –

MARC XI. Kailua-Kona, HI, USA, American Nuclear

Society (ANS), www.ans.org

17.04.-19.04.2018

World Nuclear Fuel Cycle 2018. Madrid, Spain,

World Nuclear Association (WNA),

www.world-nuclear.org

22.04.-26.04.2018

Reactor Physics Paving the Way Towords More

Efficient Systems – PHYSOR 2018. Cancun, Mexico,

www.physor2018.mx

08.05.-10.05.2018

29 th Conference of the Nuclear Societies in Israel.

Herzliya, Israel. Israel Nuclear Society and Israel

Society for Radiation Protection, ins-conference.com

13.05.-19.05.2018

BEPU-2018 — ANS International Conference on

Best-Estimate Plus Uncertainties Methods. Lucca,

Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,

ANS, IAEA, NEA, www.nineeng.com/bepu/

13.05.-18.05.2018

RadChem 2018 — 18 th Radiochemical

Conference. Marianske Lazne, Czech Republic,

www.radchem.cz

14.05.-16.05.2018

ATOMEXPO 2018. Sochi, Russia, atomexpo.ru

15.05.-17.05.2018

11 th International Conference on the Transport,

Storage, and Disposal of Radioactive Materials.

London, United Kingdom, Nuclear Institute,

www.nuclearinst.com

20.05.-23.05.2018

5 th Asian and Oceanic IRPA Regional Congress on

Radiation Protection – AOCRP5. Melbourne,

Australia, Australian Radiation Protection Society

(ARPS) and International Radiation Protection

Association (IRPA), www.aocrp-5.org

29.05.-30.05.2018

49 th Annual Meeting on Nuclear Technology

AMNT 2018 | 49. Jahrestagung Kerntechnik.

Berlin, Germany, DAtF and KTG,

www.nucleartech-meeting.com

03.06.-07.06.2018

38 th CNS Annual Conference and 42nd CNS-CNA

Student Conference. Saskotoon, SK, Canada,

Candian Nuclear Society CNS, www.cns-snc.ca

03.06.-06.06.2018

HND2018 12 th International Conference of the

Croatian Nuclear Society. Zadar, Croatia, Croatian

Nuclear Society, www.nuklearno-drustvo.hr

04.06.-07.06.2018

10 th Symposium on CBRNE Threats. Rovaniemi,

Finland, Finnish Nuclear Society, ats-fns.fi

04.06.-08.06.2018

5 th European IRPA Congress – Encouraging

Sustainability in Radiation Protection. The Hague,

The Netherlands, Dutch Society for Radiation

Protection (NVS), local organiser, irpa2018europe.com

06.06.-08.06.2018

2 nd Workshop on Safety of Extended Dry Storage

of Spent Nuclear Fuel. Garching near Munich,

German, GRS, www.grs.de

17.06.-21.06.2018

ANS Annual Meeting “Future of Nuclear in the

Shifting Energy Landscape: Safety, Sustainability,

and Flexibility”. Philadelphia, PA, USA, American

Nuclear Society (ANS), www.ans.org

25.06.-26.06.2018

index2018 – International Nuclear Digital

Experience. Paris, France, Société Française

d’Energie Nucléaire, www.sfen.org,

www.sfen-index2018.org

27.06.-29.06.2018

EEM — 2018 15 th International Conference

on the European Energy Market. Lodz, Poland,

Lodz University of Technology, Institute of Electrical

Power Engineering, Association of Polish Electrical

Engineers (SEP), www.eem18.eu

29.07.-02.08.2018

International Nuclear Physics Conference 2019.

Glasgow, United Kingdom, www.iop.org

05.08.-08.08.2018

Utility Working Conference and Vendor

Technology Expo. Amelia Island, FL, USA,

American Nuclear Society (ANS), www.ans.org

22.08.-31.08.2018

Frédéric Joliot/Otto Hahn (FJOH) Summer School

FJOH-2018 – Maximizing the Benefits of Experiments

for the Simulation, Design and Analysis of

Reactors. Aix-en-Provence, France, Nuclear Energy

Division of Commissariat à l’énergie atomique et aux

énergies alternatives (CEA) and Karlsruher Institut

für Technologie (KIT), www.fjohss.eu

28.08.-31.08.2018

TINCE 2018 – Technological Innovations in

Nuclear Civil Engineering. Paris Saclay, France,

Société Française d’Energie Nucléaire, www.sfen.org,

www.sfen-tince2018.org

05.09.-07.09.2018

World Nuclear Association Symposium 2018.

London, United Kingdom, World Nuclear Association

(WNA), www.world-nuclear.org

09.09.-14.09.2018

21 st International Conference on Water Chemistry

in Nuclear Reactor Systems. EPRI – Electric Power

Research Institute, San Francisco, CA, USA,

www.epri.com

09.09.-14.09.2018

Plutonium Futures – The Science 2018. San Diego,

United States, American Nuclear Society (ANS),

www.ans.org

10.09.-13.09.2018

Nuclear Energy in New Europe – NENE 2018.

Portoroz, Slovenia, Nuclear Society of Slovenia,

www.nss.si/nene2018/

17.09.-21.09.2018

62 nd IAEA General Conference. Vienna, Austria.

International Atomic Energy Agency (IAEA),

www.iaea.org

17.09.-20.09.2018

FONTEVRAUD 9. Avignon, France, Société Française

d’Energie Nucléaire (SFEN), www.sfen-fontevraud9.org

17.09.-19.09.2018

4 th International Conference on Physics and

Technology of Reactors and Applications –

PHYTRA4. Marrakech, Morocco, Moroccan

Association for Nuclear Engineering and Reactor

Technology (GMTR), National Center for Energy,

Sciences and Nuclear Techniques (CNESTEN) and

Moroccan Agency for Nuclear and Radiological

Safety and Security (AMSSNuR), phytra4.gmtr.ma

30.09.-05.10.2018

Pacific Nuclear Basin Conferences – PBNC 2018.

San Francisco, CA, USA, American Nuclear Society

(ANS), www.ans.org

02.10.-04.10.2018

7 th EU Nuclear Power Plant Simulation ENPPS

Forum. Birmingham, United Kingdom, Nuclear

Training & Simulation Group, www.enpps.tech

14.10.-18.10.2018

12 th International Topical Meeting on Nuclear

Reactor Thermal-Hydraulics, Operation and

Safety – NUTHOS-12. Qingdao, China, Elsevier,

www.nuthos-12.org

14.10.-18.10.2018

NuMat 2018. Seattle, United States, www.elsevier.com

16.10.-17.10.2018

4 th GIF Symposium 16-17 Oct. 2018 at the

8 th edition of Atoms for the Future. Paris, France,

www.gen-4.org

22.10.-24.10.2018

DEM 2018 Dismantling Challenges: Industrial

Reality, Prospects and Feedback Experience. Paris

Saclay, France, Société Française d’Energie Nucléaire,

www.sfen.org, www.sfen-dem2018.org

22.10.-26.10.2018

NUWCEM 2018 Cement-based Materials

for Nuclear Wates. Avignon, France, French

Commission for Atomic and Alternative Energies

and Société Française d’Energie Nucléaire,

www.sfen-nuwcem2018.org

24.10.-25.10.2018

Chemistry in Power Plant. Magdeburg, Germany,

VGB PowerTech e.V., www.vgb.org

11.11.-15.11.2018

ANS Winter Meeting. Orlando, FL, USA, American

Nuclear Society (ANS), www.ans.org

Calendar


atw Vol. 63 (2018) | Issue 1 ı January

ETSON Strategic Orientations on Research

Activities. ETSON Research Group Activity

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

1 Introduction In October 2011, ETSON published the “Position Paper of the Technical Safety Organizations:

Research Needs in Nuclear Safety for Gen 2 and Gen 3 NPPs” [1]. This paper, published only a few months after the

Fukushima-Daiichi severe accidents in Japan, presented the R&D priorities on the main pending safety issues. It was

produced by the ETSON Research Group (ERG) that has the mandate of identifying and prioritizing safety research

needs, sharing information on research projects in which ETSON members are involved, defining and launching new

research projects and disseminating knowledge among ETSON members.

Six years after the above publication, many R&D international

projects in frames such as OECD/NEA/CSNI and

Euratom have finished and others have started. In

particular a lot of work was done (and is going on…) on

the analysis of the Fukushima-Daiichi severe accidents.

Meanwhile a roadmap on research on Gen.II and III

­nuclear power plants (NPP), including safety aspects,

was elaborated by the NUGENIA association and published

in 2013 [2], followed in April 2015 by a more detailed

­document as “NUGENIA global vision” [3].

Thus in 2016-2017, the ERG judged it necessary to

perform an update of the ETSON ranking of R&D priorities,

accounting for recent outcomes of research projects (and,

for severe accidents, knowledge gained on the Fukushima-

Daiichi accidents) and for the NUGENIA R&D roadmaps.

The main objective was to underline a possible convergence

of topics for further R&D, but accounting for current

international R&D projects to avoid duplication of efforts.

2 Process of ranking of priorities

Thirteen ETSON members participated to the exercise

focusing on the safety aspects with the challenge to agree

on a short list of high priority topics and avoid the topics

where significant R&D is ongoing. A good example of

the latter case is In-Vessel-Melt-Retention during a severe

accident where many organizations from Europe (and

beyond) participate in the IVMR H2020 project [4]. For

the sake of simplification, the process was based on the

list of R&D challenges and issues from the NUGENIA

roadmap (each challenge includes several specific issues).

The partners were asked to:

• Select up to 10 highest-priority challenges: give

the mark 1 for the most important,…, 10 for the less

important,

• Then, for each of them, select up to 3 issues: give

the mark 1 for the most important..., 3 for the less

important.

The ranking process was based on the list of R&D highpriority

issues (around 150) from the latest NUGENIA

R&D roadmap. This list covers the 6 following topical

areas: plant safety and risk assessment, severe accidents,

improved reactor operation, integrity assessment of

systems, structures and components, fuel development,

waste and spent fuel management and decommissioning,

innovative LWR design and technology.

The results indicated a rather large scattering of votes

on issues but also the possibility of identifying issues with

a majority of votes. The average ranking was the sum of

marks divided by number of votes. The combined ranking

of challenges and issues was then obtained as “challenge

average ranking” multiplied by the “issue average ranking”.

The smallest figures have the highest priority.

Eight issues, described in the Section 3, were selected

as the highest priority (the order of presentation does not

represent a decreasing order of priority, the issues are in

the order of the NUGENIA roadmap). This Section

summarizes the importance of the issue for safety, the

state of knowledge and the remaining gaps, and the international

context such as ongoing or starting R&D projects.

3 High priority issues

3.1 Improved thermal-hydraulics evaluation

for the existing plants

Most of the thermal-hydraulic phenomena during

­accidents in NPPs occur at the scale of NPP cooling

systems (thermal-hydraulics in Spent Fuel Pools or SFP is

­addressed in § 3.5). The NPP response often represents a

complex interplay of the processes and phenomena in the

subsystems, which can be reproduced or analyzed only

with an experimental facility with a similar complexity or

with a simulation system code that contains models of all

relevant subsystems. Large integral facilities and system

codes thus represent a basis for NPP safety analyses. More

or less an integral facility was built in the past (or is being

built) to correspond to every major NPP type, and thus was

(or is) used to examine the plant performance during

safety relevant scenarios. Such review of integral facilities

and experiments was prepared by OECD/NEA/CSNI [5].

Some of these facilities have already been dismantled,

some of them are maintained (PKL in Germany, as well as

INKA for Gen.3+ BWR safety systems, and LSTF in Japan),

while the countries with long term nuclear goals upgrade

(MOTEL in Finland) or build entirely new (ACME in China)

facilities. These experiments were and are still used for

­validation and verification of system codes (CATHARE,

ATHLET, TRACE, RELAP ...) that represent indispensable

tools for safety analyses.

A complementary approach to the integral thermalhydraulics

testing is the “bottom-up” approach, which

actually means experimental and numerical studies of

­separate effects at larger scales under well-defined initial

and boundary conditions. These test facilities are more

accessible for academic institutions and can be roughly

divided into problems of single-phase and two(multi)-

phase flow phenomena. Single-phase experiments and

computational fluid dynamics (CFD) can be considered a

mature research field, where even blind predictions of

rather complex flows with heat transfer (pressurized

thermal shock, natural convection) and mixing of species

atw-Special „Eurosafe

2017“. In cooperation

with the EUROSAFE

2017 partners,

Bel V (Belgium),

CSN (Spain), CV REZ

(Czech Republic),

MTA EK (Hungary),

GRS ( Germany), ANVS

(The Netherlands),

INRNE BAS (Bulgaria),

IRSN (France), NRA

(Japan), JSI (Slovenia),

LEI (Lithuania),

PSI (Switzerland),

SSM (Sweden),

SEC NRS (Russia),

SSTC NRS (Ukraine),

VTT (Finland),

VUJE (Slovakia),

Wood (United

Kingdom).

Revised version

of a paper presented

at the Eurosafe,

Paris, France, 6 and

7 November 2017.

13

ENERGY POLICY, ECONOMY AND LAW

Energy Policy, Economy and Law

ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity

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


atw Vol. 63 (2018) | Issue 1 ı January

ENERGY POLICY, ECONOMY AND LAW 14

(boron dilution) will closely approach the measurements.

Tackling the two(multi)-phase phenomena is much more

difficult. Just like in 1D system codes, detailed 3D twophase

flow models still rely on a number of ­(semi)­empirical

closure relations, which must be carefully considered

for each particular geometry and phenomenon. Blind predictions

are successful only in some simple configurations,

while predictions of complex phenomena like critical heat

flux with 3D CFD models are not much more accurate than

with 1D sub-channel codes (NURESAFE project, Section

2.4 of [1]).

From the TSO point of view, one of the most important

research directions is upgrading of system codes with

(quasi)3D modules for 3D components, especially the

reactor vessel [6]. These coarse grid models can be tuned

with CFD results and high-resolution experiments. These

activities aim at coupling with 3D neutronic models and

more detailed description of the heat transfer and mixing

in the core region. Rough 3D approximations are used

and are applicable also in the simulations of SFP and

containment thermal-hydraulics, and represent a basis for

severe accident simulations.

For TSOs, more attention should probably be focused

on integral studies, which are typically much more

expensive, and can be as such seen as a critical infrastructure

[1]. Research in smaller test facilities on

phenomena such as single bubble, smallest turbulent

eddy,... will not disappear, as they are relevant for many

non-nuclear problems, while the equipment, knowledge

and experts in the field and the integral facilities are much

more difficult to maintain.

3.2 Impact of single or multiple external events

Many ETSON members contributed to the EU FP7

ASAMPSA_E project [7], which began in 2013 and

­concluded at the end of 2016; the project was led by IRSN

with 28 partners in 18 European countries. The aim was to

support the systematic extension of PSA to all potential

natural or man-made external and internal hazards.

Documents were developed to guide European stakeholders

in conducting extended PSAs and ensuring that all

dominant risks are identified and managed. The project

identified areas for future development relating to ­external

hazards; the majority of these also apply to deterministic

methods, which with PSA form the key aspects of hazard

analysis.

For the external flooding hazard, work identified to

address the following shortfalls of current methodologies

included:

• Limitations in modelling and forecasting the physical

phenomena and conditions leading to external flooding

hazard,

• Uncertainties in estimation of the impact of climate

change on external flooding events,

• Lack of site-specific data and limitations of spatial

modelling and downscaling methods,

• Difficulties in quantification of uncertainties for

common-cause failures,

• Difficulties in integrated modelling of hazard internal

and external impact assessment,

• Modelling of water propagation on the site and inside

the buildings.

For meteorological hazards, the recommendations

included:

• The provision of a better understanding and means

for quantifying the correlation mechanisms between

extreme weather events,

• An analysis of the time of the occurrence of extreme

hazard events and simultaneous evaluation of the

atmospheric states at the time of the hazard,

• More accurate estimation of the impact of climate

change on extreme meteorological events,

• Development and validation of downscaling methods

and tools for analysing and characterizing spatially

distributed extreme data.

For the seismic hazard, the recommendations included:

• The reduction of aleatory and epistemic uncertainties

in both the derivation of the seismic hazard and the

methods used to derive fragility curves,

• Improved methods for deriving conditional probabilities

of seismically induced consequential events such as

fire and flood.

For many hazard types, the need for work on treatment of

hazard combinations was also identified. There is a need

for a formalised approach for assessing and screening

hazards in which a primary external hazard would cause

one or more secondary hazards, or in which multiple

hazards occur together as a result of a common event or

underlying cause. Combinations of external and internal

hazards also need to be considered more rigorously and

systematically.

The need for integration of natural external hazards

in the plant safety case and PSA was identified previously

[1] which recommended that the identification of a

­comprehensive list of hazards, the site specific screening of

hazards, and the definition of the design basis hazards and

hazard combinations are required. It recommended that a

methodology or procedure was needed to integrate these

into the overall safety case and PSA. The ASAMPSA_E

project went some way towards achieving this objective.

The needs identified in ASAMPSA_E are partly covered

by the NARSIS (New Approach to Reactor Safety Improvement)

H2020 project [8], coordinated by CEA (France),

which recently started with the contribution of ENEA, JSI,

IRSN and VTT. In summary, it addresses improvements on

characterization of natural external hazards (concomitant

external events…), on the fragility of NPP Structures,

Systems and Components (SSC), on a combination of

risk integration with uncertainty quantification, and on

integration of expert-based information within PSA

methodology.

3.3 Methodologies for beyond design basis

assessments

Until recently, the safety assessment in the design of NPPs

was mainly focused on evaluation of postulated transients

and design basis accidents (DBA) and demonstration that

the systems of the plant can ensure that the prescribed

limits for fuel damage and radiation consequences are

not exceeded. Analysis of beyond design conditions was

generally treated as a complementary one and was mainly

used to evaluate the progression of the accident sequences

accompanied by multiple failures of systems, equipment

and components, for a more precise definition of accident

end states in the framework of probabilistic risk assessment

and for identification of operator actions for bringing

the plant into the controlled state and/or mitigating the

consequences.

With update of the IAEA requirements ([9] in particular),

the analysis of design extension conditions that

include multiple failure events without nuclear fuel

melting, as well as severe accidents, becomes an intrinsic

part of the plant safety assessment, and appropriate safety

features for preventing such conditions from arising, or,

Energy Policy, Economy and Law

ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity

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


atw Vol. 63 (2018) | Issue 1 ı January

if they do arise, for controlling them and mitigating their

consequences, are required to be included in the NPP

design.

Individual aspects of the methodology for beyond design

basis accidents (BDBA), with different levels of detail, are

reflected in national regulations of the ETSON member

states. Basic considerations on assessing the design extension

conditions can also be found in IAEA documents [10,

11]. However some questions still need to be addressed, e.g.

how to ensure that all relevant scenarios are considered,

what is the extent of failures to be considered, how the

­uncertainties shall be identified and accounted for? Therefore

there is a need to provide TSOs with unified detailed

guidelines that cover all BDBA stages, starting from the

deterministic and probabilistic criteria for selection of

corresponding scenarios, assumptions on systems/equipment

operability (including non-safety graded systems),

and ending with the evaluation of assessment results, and

establishes interfaces with practical applications of assessment

results, including identification and justification of the

provisions which are incorporated in the plant design or to

be implemented as safety upgrade measures for mitigating

the consequences of such events, identification of operator

preventive and mitigating actions, etc.

The important aspects to be addressed in the guidelines

are incorporation of up-to-date results of R&D in the area

of phenomenology, validation of computer codes and

models and procedure for treatment of inherent uncertainties

associated with current knowledge.

The development of such guidelines is a complex and

rather immense task. Therefore, a possibility for a first phase

could be to collect information on respective experience

of the participants, systemize and critically analyse this

information to identify the existing gaps and then to elaborate

solutions for enhancing the BDBA methodology.

3.4 Development and validation of severe

accident integral codes

Considering the complexity and different mutual interacting

phenomena in severe accident (SA) progression and

the possible source term release to the environment,

research is fundamental in order to characterize the main

phenomena determining the NPP transient evolution and

to support severe accident management (SAM). With this

in mind, a key role is given to the state-of-art SA integral

codes (as ASTEC [12] and MELCOR [13] that are mostly

used within TSOs and safety authorities, but also MAAP

used mainly by the industry) that store all the knowledge

developed in the last decades from the experimental

activities.

With the target of assessing SAM, some modelling

uncertainties still present, sometimes closely linked to

remaining uncertainties on the knowledge of phenomena

itself, should be addressed. The latest status of SA research

highest priorities issued from the SARNET European

network is presented in [14, 15]. Among them, the

modelling improvements must address in priority:

• The coolability of the degraded core and the phenomena

necessary to assess the In-Vessel Melt Retention

strategy,

• The coolability of corium during Molten Core Concrete

Interaction in the NPP cavity after a possible vessel

failure,

• The mitigation of potential source term (mainly

­ruthenium and iodine), in particular the use of filtered

containment venting systems (FCVS) and the related

efficiency, including the accident long term situation.

An essential field of applications of such codes in the

next years concerns the need to improve SAM guidelines.

In addition, for plant applications, uncertainty analysis

should be systematically performed (e.g. by using tools

such as DAKOTA, RAVEN, SUNSET, SUSA, etc). More and

more code-to-code exercises called “crosswalk” activities

(e.g. involving the teams of code developers and thus

going much more deeply than classical benchmark

exercises) should be continued (see examples in [16, 17])

in order to identify the modelling differences affecting

code prediction results.

In order to reduce the code user-effect [18], considering

the SA complexity, a high level understanding of the

phenomena/processes and of the use of such codes is

required from code users. It is important to continue three

types of ongoing actions:

• Users’ training programs, led by international

recognized experts,

• Well-defined international cooperation platform of

research activities where exchange of opinions,

methods, experimental/calculated data, ideas and

possible interactions between code users and developers

take place (e.g. SARNET network set up under the

European Commission FP, ASTEC-User Club sponsored

by IRSN, CSARP/MCAP organized by USNRC, OECD/

NEA/CSNI ISP, IAEA ICSP and research and innovation

through EU-FP). In this framework the code-to-code

benchmark exercises (such as the exercise in [19]), as

well as independent user crosswalk activities, will allow

to characterize also the influence of user effect on the

different code predictions.

• Availability of user manual and guidelines to be

provided to the user, in addition to the complete code

documentation (models, numerics, assessment), as

well as development of graphical-user-interfaces [20]

to support the user in the input-deck preparation and to

make the post-processing of the data easier (a good

­example of such tool is SNAP developed for USNRC for

use with MELCOR and other codes).

In relation to the extension of SA prediction capability,

another useful action is coupling of SA integral codes with

specialized codes designed to predict the impact of the

source term in the surrounding environment (source term

release, transport, dispersion). This permits a best estimate

evaluation of the source term and a consequent detailed

consequence analyses to support emergency preparedness

and response. An example is the coupling between

­MELCOR and MACCS SNL tools developed for USNRC.

A long term consideration could be related to the development

of advanced software platforms where SA integral

codes can be coupled with specific detailed codes (e.g.

CFD) to get a more detailed characterization of SA

­progression, in terms of single specific phenomenon or/

and 3D nature predictability.

Finally, as an essential activity, validation of codes vs.

experiments should obviously be performed in the future

as a continous process, on current experiments often

dedicated to mitigation aspects but also on the huge

amount of SA experiments that were performed during

more than 30 years. The codes application to the

Fukushima-Daiichi accidents is also an important task

planned in the next years.

3.5 Spent fuel pool accident scenarios

SFPs are large accident-hardened structures that are used

to temporarily store irradiated nuclear fuel [20, 22]. Safety

and security are continuously reassessed [23], e.g. after

ENERGY POLICY, ECONOMY AND LAW 15

Energy Policy, Economy and Law

ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity

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


atw Vol. 63 (2018) | Issue 1 ı January

ENERGY POLICY, ECONOMY AND LAW 16

the terrorist attacks in the USA on September 11, 2001 and

the Fukushima Daiichi accident in March 2011 [24],

although the SFPs and the fuel stored in the pools remained

safe during the accident. Considering all possible initiating

events from safety as well as security perspectives, and the

assumption that the accident cannot be prevented or

mitigated, some SFP scenarios could possibly lead to large

radiological consequences on-site and off-site.

The main knowledge gaps are identified thanks to a

recently completed OECD/NEA/CSNI activity, led by IRSN

with the participation (among the international panel

of experts) of ETSON members Bel V, GRS, PSI and NRA,

on applying a Phenomena Identification and Ranking

Technique (PIRT) on SFPs under loss-of-cooling and

loss-of-coolant accidents conditions [20]. The resulting

phenomena of primary interest for further research can be

summarized as follows:

• Cladding chemical reactions with mixed steam-air

environments for all type of fuel claddings present in

SFPs and also the low temperature range,

• Thermal-hydraulic and heat transfer phenomena for

the coolability of partly or completely uncovered fuel

assemblies,

• Thermal-hydraulic behaviour and large-scale natural

circulation flow pattern that evolves in the SFP with

fuel assemblies covered with water,

• Spray cooling of uncovered spent fuel assemblies in

typical storage rack designs.

Quite a few experiments, specifically targeted to SFP

accidents, are underway or planned. Improvements of

models and simulation codes are still necessary, and their

validation will continue against the produced data.

Regarding applicability of codes, sensitivity and uncertainty

analyses should be considered an integral part of

their applications for SFPs accidents conditions.

National projects focusing on SFP issues are addressed

by several ETSON members, e.g. in cooperation with

universities and research institutes in case of Bel V [25], by

launching experimental programs by IRSN [26], related

to analysis of processes in SFP for LEI, sensitivity analysis

of various modelling options on SFP accidents in SSTC

NRS etc.

3.6 Corium thermophysical and thermodynamic

properties

During a severe accident sequence in LWRs, thermodynamic

models are required to predict the behaviour of

the melts (so-called corium) formed from the degradation

of the core materials, the fission product (FP) releases and

the residual power within the corium different phases.

Data such as the composition of the phases present in the

corium and its physical-chemical properties (solidus and

liquidus temperatures, heat capacities, enthalpies …) are

key parameters for modelling, among other things, the

­corium flow properties, the FP distribution between the

gas and the condensed phases and then for modelling of

the progression of the accident.

Since 1990’s, in the framework of projects in the frame

of the EC (COLOSS, SARNET…), the International Science

and Technology Center (CORPHAD and PRECOS) and the

OECD (MASCA [27]), SA experts have been interested in

the assessment of thermodynamic data for a number of

compounds of reactor materials and fission products and

more complex phases. The most common thermodynamic

data assessment approach for the chemical species of

interest is the CALPHAD method [28]. All properties are

derived from the Gibbs energy expression for each phase.

Based on physical models of the different phases, such

expression depends on various parameters, the values

of which are optimised in order to best fit available

experimental data.

Databases thus obtained are more than mere compilations

of thermodynamic data from various sources.

Their constitution and maintenance needs considerable

work for self-consistency analysis, to ensure that all

the available experimental information is satisfactorily

reproduced. Updating and improving the database

becomes then a regular task, tightly linked to the needs of

end-users.

IRSN is developing, with the SIMAP French Laboratory

scientific support, two consistent thermodynamic

data bases for use for the interpretation of SA experiments

and modelling. NUCLEA [29] is mainly used in research

related to the core degradation (in- and ex-vessel) while

MEPHISTA addresses the fuel and FP behaviour in normal

and off-normal conditions. Both databases are currently

used by a large number of institutes, industrial partners,

and universities, including a few ETSON partners (VTT,

soon PSI), EDF, CEA, Areva, KAERI (South Korea), JAEA

(Japan) and others. The OECD-NEA Thermodynamics of

Advanced Fuels – International Database (TAF-ID) project

[30] (2013-2016) made available a comprehensive,

internationally recognized and quality-assured database

of phase diagrams and thermodynamic properties of

advanced nuclear fuels. Its main goal consists in providing

a computational tool to perform thermodynamic calculations

on both fuel and structural materials for SA in

LWRs and for the design of advanced fuel materials (MOX,

metallic, carbide, nitride fuels) for Generation IV reactors.

The recently launched OECD/NEA Thermodynamic

Characterisation of Fuel Debris and Fission Products

(TCOFF) project (2017-2019), involving 16 partners, aims

at improving the existing thermodynamic databases

(e.g. NUCLEA and TAF-ID) for scenario analyses of SA

progression, looking particularly at the Fukushima-Daiichi

accident.

To date, the main gaps of knowledge in databases are

the following ones:

• The interactions between molten U-Zr-O and iron (and

steel) within the vessel since they impact the heat flux

to the vessel in order to determine the conditions (in

particular time and location) of an eventual rupture, in

particular for a molten metal layer located on top of the

oxide one. Some work has been done in the framework

of the MASCA and MASCA2 projects but it would be

necessary to extend it to MOX fuel.

• The impact of the stainless steel oxide components on

the thermochemistry of the corium-concrete mixtures

which should be experimentally investigated.

• The activity coefficients of the Ag-In-Cd control rod

elements in the melts are a very important item to

derive reliable expressions for vapor pressures of

absorber elements. Vaporization of these elements

during a SA is of prime interest for reactors with

Ag-In-Cd control rods. They actually constitute the

main contributors in terms of mass of the aerosol

release into the reactor coolant system and overall,

they greatly impact the aerosol deposition and

the source term behaviours. In fact, silver and cadmium

are very reactive with iodine which is known to

be a major contributor to the gaseous source term

to environment.

Energy Policy, Economy and Law

ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity

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


atw Vol. 63 (2018) | Issue 1 ı January

3.7 Ageing/degradation mechanisms,

modelling and materials properties

for metallic components

Many operating NPPs are nearing or have exceeded

their original design lifetime (often 40 years). To safely

continue operation beyond that, i.e. to enter the long term

operation (LTO) period, necessitates considerable technical

preparations and permission from the domestic

regulator. This needs to take thoroughly into account the

ageing/degradation mechanisms, through e.g. knowledge

of materials properties and computational modelling.

There are several potential ageing/degradation

­mechanisms affecting metallic NPP components, mainly:

irradiation and thermal embrittlement, fatigue, stress

corrosion cracking (SCC), general and local corrosion,

flow accelerated corrosion, creep and mechanical wear.

The commonly used steel types include: ferritic steels,

austenitic stainless steels and nickel-base alloys. The

susceptibility to degradation mechanisms depends mainly

on physical loads, material properties and process

environment. Important material properties include yield

& tensile strength, fracture toughness and carbon content.

There are still knowledge gaps concerning understanding

irradiation embrittlement, fatigue, SCC and mechanical

wear as well as joint action of degradation mechanisms.

It is necessary to computationally model the propagation

of degradation in metallic NPP components. When considering

LTO, this is called “time limited ageing analysis”

(TLAA). Most TLAAs necessitate knowing the temperature

and stress distributions across the components. These are

computed with heat transfer and structural mechanics

analyses, typically applying ­numerical finite element (FE)

codes. These results are used as input data in the ensuing

degradation propagation analyses. For local flaws, e.g.

cracks, these analyses are carried out applying fracture

mechanics and empirically derived crack growth correlations.

There are still gaps concerning modelling of irradiation

embrittlement, thermal fatigue, SCC and mechanical

wear as well as joint action of degradation mechanisms.

Current research in Europe is performed in the frame of

Euratom projects: irradiation embrittlement in SOTERIA,

joint action of corrosion and fatigue in INCEFA+, thermal

fatigue and fracture mechanics based modelling of

degradation mechanisms in ATLAS+. Despite this

intensive activity in Europe, this issue selection underlines

the very high importance given by TSOs on such issue.

3.8 Small modular reactors

Currently Small Modular Reactor (SMR) concepts are

discussed as one main option for new builds worldwide.

This revival in SMRs is driven by the potential for enhanced

safety and security while reducing capital costs and thus

investment risks, through design simplification. SMRs

­introduce flexibility on locations unable to accommodate

larger NPPs and can be operated under onshore, offshore

and subsea-based conditions. Improved technologies and

methods will be implemented, thus contributing to the

­demand of higher safety and reliability without sacrificing

the long lasting operation experience of LWR technology.

The European nuclear industry has developed no

near-term feasibly deployable SMR [31] and countries

have just begun to build-up the necessary regulatory

structures and capacities. SMR based on LWR technology

offer advantages due to the experience of the nuclear

stakeholders (especially of the regulators) with LWR

technology collected in the last decades. Therefore for

­ETSON, the priority concerns are LWR-type SMRs, and the

basis for further success is the edge in knowledge, which

also includes validated simulation tools.

Several international activities were initiated concerning

the identification and closure of open SMR issues.

Several workshops and studies took place in the IAEA and

OECD/NEA frame [32, 33, 34, 35]. In the UK a feasibility

Study on SMR was published in 2014 to identify inter alia

the best value for the UK. Several European TSOs deal with

this issue, whereby the GRS study [36] is recognized as

one of the most extensive works on this topic. The aims of

the latter were to set-up a sound overview on current SMR,

to identify essential issues of reactor safety research and

future R&D projects, and to identify needs for adaption of

system codes used in reactor safety research as well as

approval and supervisory procedures. This overview

consists of the description of 69 SMR diverse concepts (32

LWR, 22 liquid metal cooled reactors, 2 heavy water cooled

reactors, 9 gas cooled reactors and 4 molten salt reactors).

It provides information e.g. about the core, the cooling

circuits and the safety systems. The safety relevant issues

of the selected SMR concepts were identified on the basis

of the defense-in-depth concept, which is one core issue of

the new Euratom Safety Directive 2014 (see the ETSON

paper [37]). Further on, it was evaluated whether these

safety systems and measures can already be simulated

with the existing nuclear simulation chains and where

further code development and validation are necessary.

In general the existing codes are a good basis for the

simulation of SMR. However the safety-related im provements

of these advanced reactors, in general, still require a

considerable effort for further development and validation.

Both require new experiments with advanced (two-phase

flow) measuring techniques. In addition to component tests,

in which the start-up and operating behaviour has to be investigated

under defined and ­idealized initial and boundary

conditions, integral tests are required for the investigation of

the mutual interaction of different passive safety systems or

different trains of one passive safety system required for

( severe) accident control. For such investigations, already

existing large European experimental facilities for the investigation

of passive safety systems (such as INKA in AREVA

GmbH, PANDA in PSI or SPES in SIET-ENEA) can be applied.

Main topics for improvements are e.g. advanced fuel

patterns, innovative fuel and cladding design, increase of

enrichment and burn-up, longer fuel cycles, boron-free

cores, (new) working fluids with extended scopes, passive

safety systems and their mutual interactions, natural

­circulation and flow instabilities, innovative heat

exchanger designs (such as plate and helically coiled heat

exchangers, heat pipes), 2D/3D models for simulation of

temperature and velocity fields in large water pools.

4 Conclusion

The R&D highest priority needs that are described in this

paper correspond mostly to the objectives of the new

2014/87 Euratom Directive on the safety of nuclear

­installations, as shown in the ETSON EUROSAFE-2015

paper [37]. In particular they aim at preventing accidents

through defence in depth and at avoiding radioactive

releases outside a nuclear installation. They were also

­already identified in the ETSON 2011 position paper [1].

This ranking will first serve as basis for new potential

research projects, either to be performed by ETSON

partners only or as a kernel to be proposed in a larger

frame such as NUGENIA or H2020. The ranking may

also serve as the ETSON input to future roadmaps or to

inter national R&D projects.

ENERGY POLICY, ECONOMY AND LAW 17

Energy Policy, Economy and Law

ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity

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


atw Vol. 63 (2018) | Issue 1 ı January

ENERGY POLICY, ECONOMY AND LAW 18

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of-Coolant Accident Conditions, NEA/CSNI/R(2015)2, May 2015.

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(PIRT) on Spent Fuel Pools under Loss-of-Cooling and Loss-of-

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investigation committee. 2015, Tokyo, Japan: Springer.

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issues related to some specific phenomena under natural

circulation flow conditions. In: EUROSAFE 2012, November 5-6,

2012 Brussels, Belgium.

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database NUCLEA, Progress in Nuclear Energy, volume 52,

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System Codes for Design and Safety Analysis of Integral Type

Reactors, Vienna, 2014.

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(2015).

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A Schaffrath, A. Nieminen, ETSON views on R&D priorities for

implementation of the 2014 Euratom Directive on safety of

nuclear installations, Kerntechnik: Vol. 81, No. 5, pp. 527-534.

Authors

J.P. Van Dorsselaere (Contact author)

M. Barrachin

IRSN, Centre de Cadarache, BP3,

13115 Saint Paul les Durance Cedex, France

D. Millington

Wood RSD, 305 Bridgewater Place, Birchwood Park,

Warrington WA3 6XF, UK

M. Adorni

BelV, 148 Walcourtstraat, B-1070 Brussels, Belgium

M. Hrehor

CV REZ, Centrum Vyzkumu Rez, Husinec – Rez 130,

250 68 Rez, Czech Republic

F. Mascari

ENEA, Via Martiri di Monte Sole, 4, 40129 Bologna, Italy

A. Schaffrath

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS)

gGmbH, Forschungszentrum, Boltzmannstraße 14,

85748 Garching bei München, Germany

I. Tiselj

JSI, Jozef Stefan Institute, Jamova cesta 39,

SI-1000 Ljubljana, Slovenia

E. Uspuras

LEI, Lithuanian Energy Institute, Breslaujos 3,

LT-44403 Kaunas, Lituania

Y. Yamamoto

NRA, Nuclear Regulation Authority, Toranomon Towers

Office, 4-1-28 Toranomon Minato-ku, Tokyo 105-0001, Japan

D. Gumenyuk

SSTC-NRS, State Scientific and Technical Center,

35-37 Radhospna Str., 03142 Kiev, Ukraine

N. Fedotova

SEC-NRS, Scientific and Engineering Center for Nuclear and

Radiation Safety, Malaya Krasnoselskaya st. 2/8, building 5,

Moscow, 107140, Russia

O. Cronvall

VTT Technical Research Centre of Finland Ltd, Vuorimiehentie

5, P.O.Box 1000, FI-02044, Finland

P. Liska

VUJE, Okruzna 5, 91864 Trnava, Slovakia

Energy Policy, Economy and Law

ETSON Strategic Orientations on Research Activities. ETSON Research Group Activity

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


atw Vol. 63 (2018) | Issue 1 ı January

Die Novellierung der europäischen Dual-Use Verordnung –

eine unendliche Geschichte?

19

Ulrike Feldmann

Entwicklung der europäischen Dual-Use Verordnung Erstmalig wurde mit der Verordnung (EG) Nr. 3381/94

des Rates vom 19. 12.1994 (ABl. Nr. L 367 vom 31.12.1994, S. 1) eine Gemeinschaftsregelung für die ­Ausfuhrkontrolle

von Gütern mit doppeltem Verwendungszweck geschaffen. Mit der Verordnung (EG) Nr. 1334/2000 vom 22.06.2000

(ABl. Nr. L 159 vom 30.06.2000, S. 1) fand die erste größere Revision der Dual-Use Regelungen statt, mit der für den

Nuklearbereich nicht – wie bis dato – nur sensitives Material, d.h. Plutonium und hochangereichertes Uran, sondern die

gesamte Kategorie 0 (Nuklearmaterial, Anlagen, Ausrüstung) auch einer Genehmigungspflicht für die innergemeinschaftliche

Verbringung unterworfen wurde. Außerdem wurde mit der Verordnung 1334/2000 in Art. 21 Abs. 1

bestimmt, dass die Nukleargüter der Kategorie 0 nicht Gegenstand einer Allgemeingenehmigung sein können. Die

EU-Kommission erkannte dann schnell, dass das „Kind mit dem Bade ausgeschüttet“ und mit der rigorosen Erfassung

aller Nukleargüter der Kategorie 0 der innergemeinschaftliche Handel unnötig behindert wurde und nahm wenige

Monate später mit der Verordnung (EG) Nr. 2889/2000 vom 22.12.2000 einen kleinen Teil von Nukleargütern aus der

innergemeinschaftlichen Verbringungsgenehmigungspflicht wieder aus.

Ab 2006 arbeitete die Kommission an einer weiteren

­umfassenden neuen Revision, um u.a. die UN Resolution

1540 vom 28.04.2004 zur Nichtverbreitung von chemischen,

biologischen, nuklearen Waffen und ihrer Trägersysteme

durch Verschärfung der Exportkontrolle umzusetzen

(z.B. durch Ausweitung des Geltungsbereichs auch

auf Vermittlungsdienstleistungen und Einbeziehung des

Technologietransfers, d.h. Bereitstellen von Software und

Technologie), aber auch um das Genehmigungsverfahren

zu beschleunigen und zu ver einfachen (z.B. durch die Einführung

neuer Allgemeingenehmi gungen der Gemeinschaft

für nicht-nukleare Dual-Use Güter). Nachdem die EU-­

Kommission aufgrund massiver Kritik aus den Mitgliedstaaten

wie auch von Seiten der Industrie einen Teil

ihrer –praxisuntauglichen – Novellierungsvorschläge zurück

gezogen hatte, konnte die Revision verabschiedet

werden und erschien im Amtsblatt der EU als Verordnung

(EG) 428/2009 (ABL. Nr. L 134 vom 29.05.2009).

Novellierung der Dual-Use-Verordnung

428/2009/EG

Bereits vor der Verabschiedung der Verordnung 428/2009

hatte die EU-Kommission angekündigt, in einem weiteren

Schritt den Annex IV der Verordnung zu novellieren.

Sicherlich auch bedingt durch den Wechsel in der EU-

Kommission legte die derzeit amtierende EU-Kommission

erst im Herbst 2016 einen Revisionsentwurf vor, der jedoch

über eine bloße Überarbeitung des Annex IV weit hinaus

geht. Angedacht war von der Vorgänger-Kommission,

mit der Novellierung die gestiegenen Sicherheitsanforderungen

mit dem Grundsatz des freien Warenverkehrs

und dem Erhalt der Wettbewerbsfähigkeit der europäischen

Industrie zu einem besseren Ausgleich zu bringen

als bisher. Der Revisionsvorschlag der jetzigen EU-

Kommission wird diesem Ziel jedoch aus Sicht der

­europäischen Nuklearindustrie wie auch aus Sicht der

nicht-nuklearen Industrie in der EU nicht gerecht.

Schutz von Menschenrechten und Cyber-Überwachungstechnologien

Im Vordergrund der Kritik steht sowohl der Vorschlag, in

die Dual-Use Verordnung den Schutz von Menschenrechten

aufzunehmen als auch der Vorschlag, Cyber-Überwachungstechnologien

als neuen Typus eines Dual-Use

Gutes in die Verordnung zu integrieren. Der Export von

Technologien soll stärker kontrolliert werden, wenn das

Risiko besteht, dass diese Technologien zur Überwachung

von Menschen genutzt werden können. Zweifellos ist der

Schutz von Menschenrechten ein hohes Gut. Angesichts

der weitreichenden und rasanten geopolitischen Veränderungen

wie auch angesichts ständig sich erweiternder

Möglichkeiten zur digitalen Überwachung muss die

Exportpolitik dieser Entwicklung zweifellos Rechnung

tragen. Dies sollte allerdings auf gesicherter gesetzlicher

Grundlage erfolgen. Zudem sollten verschärfte Kontrollregelungen

praktikabel und sinnvoll sein und mit ­Augenmaß

festgelegt werden. Zu bedenken ist dabei, dass heutzutage

Überwachungstechnologie in vielen Produkten enthalten

ist und zahlreiche Unternehmen ihre Waren weltweit vermarkten.

Des weiteren sollten verschärfte Kontrollregelungen

nicht dazu führen, dass Verbringung und Export von

Nukleargütern grundlos strengeren ­Kontrollen unterworfen

werden als andere Dual-Use- Güter.

Bedenken gegen den Kommissionsvorschlag

Jedoch bestehen zum einen Zweifel an der Mandatierung

der EU-Kommission. Zum anderen fehlt es an einer klaren

Definition der Menschenrechte im Kommissionsentwurf

selber. Außerdem divergieren die Definitionen im Katalog

der Menschenrechte in der Europäischen Menschenrechtskonvention

und in der UN-Menschenrechtscharta. Hinzu

kommt, dass der Kommissionsentwurf dem Exporteur, dem

Broker und/oder demjenigen, der technische Überwachung

zur Verfügung stellt, eine Prüfungs- und ­Informationspflicht

auferlegt, deren Erfüllung jedenfalls ohne nähere Erläuterung

(z.B. durch einen ent sprechenden Leitfaden) in vielen

Fällen nicht leistbar ist. Insbesondere kleinere Unternehmen

werden fachlich, zeitlich und personell nicht in der

Lage sein zu beurteilen, ob das zu exportierende Gut in

dem Empfängerland z.B. im Zusammen hang mit einem

­bewaffneten Konflikt oder einem terroristischen Akt oder

von einem Dritten dazu benutzt werden soll, schwerwiegende

Menschenrechts verletzungen zu begehen. Mit

einem noch so guten „ Internal Compliance Programme“

(ICP) werden sich diese Fragen oftmals nur unzureichend

lösen lassen. Der Schutz von Menschenrechten ist im Inund

Ausland im Übrigen zuvörderst eine Staatsaufgabe.

Die Rechts unsicherheit auf Seiten der Unternehmen dürfte

– auch nach Einschätzung der deutschen Behörden – dazu

führen, dass sich die Unternehmer vermehrt ratsuchend an

die zuständige Genehmigungsbehörde wenden werden, so

dass deren Fallzahlen und damit die Wahrscheinlichkeit für

längere Genehmigungsverfahren steigen werden. Ähnliche

Bedenken bestehen gegen die Einführung einer „Catch-All“

Regelung, nach der alle Internet–Überwachungstechnologien

prinzipiell einer Exportgenehmigung bedürfen.

SPOTLIGHT ON NUCLEAR LAW

Spotlight on Nuclear Law

Council Regulation of the European Dual Use Regulation – A Never Ending Story? ı Ulrike Feldmann


atw Vol. 63 (2018) | Issue 1 ı January

20

DATF NOTES

Die EU-Kommission hat im Laufe 2017 zwar einige

Änderungen an ihrem Entwurf konzediert, darunter auch

den Vorschlag für eine Verlängerung der – zunächst im

Kommissionsentwurf auf ein Jahr festgelegten – Genehmigungsdauer

sowie die Einführung einer Allgemeingenehmigung

für Großprojekte aufgegriffen, ist aber z.B. auf

Vorschläge für einen mehr risikobasierten Ansatz bei

­Nukleargütern oder für die Einführung von EU-Allgemeingenehmigungen

soweit ersichtlich bisher nicht eingegangen.

Allerdings beabsichtigt die Kommission, in der

zweiten Dezemberhälfte wieder ein Exportkontrolle-

Forum unter Beteiligung der Industrie zu veranstalten. In

Fachkreisen wird es jedoch für wenig wahrscheinlich

gehalten, dass die Kommission ihre Position aufgrund des

Exportkontrollforums noch wesentlich ändern wird.

Haltung des Europäischen Rates und

des Parlaments

Während der Europäische Rat sich bisher eher abwartend

verhalten hat, hat sich das Europäische Parlament (EP)

intensiv mit dem Vorschlag der EU-Kommission befasst.

Zum Berichterstatter für die Revision der Dual-Use-

Verordnung hatte das EP in 2017 MdEP Prof. Dr. Klaus Buchner

bestimmt. Buchner ist u.a. Mitglied im EP-Ausschuss für

auswärtige Angelegenheiten sowie in den EP-Unterausschüssen

für Menschenrechte, Sicherheit und Verteidigung. Außerdem

ist er stellvertretendes Mitglied im EP-Ausschuss für

internationalen Handel (INTA), der federführend für die

Revision der Dual-Use-Verordnung ist. Der Ausschuss INTA

hat in seinem Berichtsentwurf zu dem Kommissionsentwurf

424 Änderungsvorschläge gemacht (z.B. Ausdehnung

des Schutzes der Menschenrechte, Veröffentlichung der

Abwägungskriterien für die Exportkontrolle von Dual-Use-

Gütern, Klarstellung des Begriffs des Exporteurs sowie Ablehnung

einer Allgemeingenehmigung für Großprojekte als

zu nuklearbezogen). Am 23. November 2017 hat der Ausschuss

INTA in erster und einziger Lesung mit der ganz

überwiegenden Mehrheit der Stimmen dafür gestimmt, die

Exportkontrollen von Überwachungstechnologien deutlich

auszuweiten und die Menschenrechte zum zentralen Bestandteil

der Exportkontrolle zu machen. Berichterstatter

Buchner befürchtet jedoch, wie sich seiner Presseerklärung

vom 23.11.2017 zu der Beschlussfassung im INTA-Ausschuss

vom selben Tag entnehmen lässt, „dass die Industrie,

die um ihre Geschäfte bangt, massiven Druck ausübt, und

mithilfe ihrer Lobbyisten die Verabschiedung der

Ver ordnung bremst und versucht abzuschwächen.“ Es

besteht die Gefahr so Buchner, „dass die gute, heute vom

­INTA-­Ausschuss beschlossene Reform von der Industrie

mit der willfährigen Unterstützung konservativer Abgeordneter

im Plenum verwässert wird“.

Wer solcherart versucht, einen stärkeren Schutz von

Menschenrechten einzufordern, dürfte damit allerdings

sich und seiner Sache einen Bärendienst erweisen.

Ausblick

Sollte das Plenum, wie terminiert am 16. Januar 2018 einen

Beschluss zum Novellierungsentwurf fassen (was nicht

sicher ist), dürfte sich der Rat vermutlich ab Februar/März

2018 intensiver mit der Thematik befassen. Die Geschichte

um die Novellierung der Dual-Use Verordnung geht also

zumindest noch ein Weilchen weiter.

Autorin

Ulrike Feldmann

Berlin, Germany

Notes

For further details

please contact:

Nicolas Wendler

DAtF

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

kernenergie.de

www.kernenergie.de

New Explanatory Video:

Dismantling – 60 Seconds

The essentials of decommissioning and dismantling a nuclear

power plant in 60 seconds:

• Who is responsible?

• Who supervises it?

• What happens with the material?

You get brief answers on these and more questions

in this explanatory video from DAtF (in German).

3 The complete video can be watched at www.kernenergie.de

or at the DAtF YouTube channel.

3 A more comprehensive explanatory video, a brochure of DAtF

on Decommissioning of NPPs and additional Information

(all in German) are available on www.kernenergie.de.

New Edition of the Brochure

on the Final Disposal

of High Radioactive Waste

The brochure “Endlagerung hochradiaoktiver Abfälle” (in German)

gives you a comprehensive overview on:

• The history of final disposal in Germany

and current waste management

• How the new site selection process will run

and what are the safety criteria

• Who will run the process, who will be involved

and how it is paid for

3 These and other issues surrounding the management

of highly active waste in Germany are addressed

in the brochure available online and in print.

DAtF Notes


atw Vol. 63 (2018) | Issue 1 ı January

Nuclear Safety, Security and Safeguards:

An Application of an Integrated Approach

Howard Chapman, Jeremy Edwards, Joshua Fitzpatrick, Colette Grundy, Robert Rodger

and Jonathan Scott

1 Introduction At the 34 th G8 1 Summit in Japan in 2008 the assembled leaders acknowledged the role of

nuclear power in reducing CO 2 emissions. Part of the final communique stated their commitment to the highest ­possible

standards on “nuclear non-proliferation, safeguards, safety and security” [2]. They recognised that synergies exist

between the 3Ss, (nuclear safety, nuclear security, and nuclear safeguards) and considered it was important that the

separate disciplines are integrated, and that the 3S infrastructure is strengthened through international cooperation

and assistance.

In order to identify the synergies

between the individual specialisms,

international legislation and regulatory

regimes are reviewed before

considering the methods and assessment

techniques used. We then

consider which approaches can contribute

most to improving the integration

of the nuclear 3S, and recount

practical experience of implementing

the Triple S approach.

The aims for the individual

specialisms are:

• Safety is aimed at protecting

workers and the public from the

harmful effects of radiation (or

chemicals or other hazards);

• Security is aimed at preventing

malicious acts that might harm a

nuclear facility (sabotage) or result

in the loss (theft) of nuclear

materials; and

• Safeguards are aimed at preventing

the diversion of nuclear materials

from a civil nuclear programme

to nuclear weapons purposes.

The 3Ss share the same overall objectives

of protecting the public and the

environment from potential hazards.

They use similar principles to achieve

protection; multiple barriers, defence

in depth, decision analysis and consequence

assessment. The regulatory

regimes for all 3Ss use, in the main,

the same processes; assessment, permissioning,

inspection, enforcement

and influence [3].

3.1 Safety

The International Atomic Energy

Agency (IAEA) Fundamental Safety

Principles document [4] states “The

fundamental nuclear safety objective

is to protect people and the environment

from the harmful effects of

ionising radiation”

“To ensure that facilities are operated

and activities conducted so as to

achieve the highest standards of safety

that can reasonably be achieved,

measures have to be taken:

a) To control the radiation exposure

of people and to prevent the release

of radioactive material to the

environment;

b) To restrict the likelihood of events

that might lead to a loss of control

over a nuclear reactor core, nuclear

chain reaction, radioactive source

or any other source of radiation;

and

c) To mitigate the consequences of

such events if they were to occur”.

3.2 Security

Nuclear security focuses on the prevention,

detection and response to

malicious acts involving or directed at

nuclear material, other radioactive

material, associated facilities, or

associated activities [5]. The objectives

of a State’s Physical Protection

Regime [6] should be:

a) To protect against unauthorised

removal;

b) To locate and recover missing

nuclear material;

c) To protect against sabotage; and

d) To mitigate or minimize effects of

sabotage.

3.3 Safeguards

The objective of Safeguards is to prevent

the diversion of nuclear material

from peaceful use to nuclear weapons

or other nuclear explosive devices

(Article III.1 of the Non-Proliferation

Treaty (NPT)).

4 Approaches

4.1 Safety

The concept of defence in depth is

fundamental to nuclear safety to

prevent accidents and if prevention

fails, to limit potential consequences.

Nuclear Safety Assessment has a

number of complementary analysis

Safety Security Safeguards

Convention on Nuclear Safety

Convention on Assistance

in the Case of a Nuclear Accident

Convention on the Physical Protection

of Nuclear Materials (CPPNM)

United Nations (UN) International

Convention for the Suppression

of Acts of Nuclear Terrorism

IAEA Statute

atw-Special „Eurosafe

2017“. In cooperation

with the EUROSAFE

2017 partners,

Bel V (Belgium),

CSN (Spain), CV REZ

(Czech Republic),

MTA EK (Hungary),

GRS (Germany), ANVS

(The Netherlands),

INRNE BAS (Bulgaria),

IRSN (France),

NRA (Japan),

JSI (Slovenia),

LEI (Lithuania),

PSI (Switzerland),

SSM (Sweden),

SEC NRS (Russia),

SSTC NRS (Ukraine),

VTT (Finland),

VUJE (Slovakia),

Wood (United

Kingdom).

Revised version

of a paper presented

at the Eurosafe,

Paris, France, 6 and

7 November 2017.

1) Canada, France,

Germany, Italy,

Japan, Russia,

United Kingdom,

United States

and European

Commission

Non Proliferation Treaty

(NPT)

21

ENVIRONMENT AND SAFETY

2 International statues and

agreements

Some of the main international

statutes (written law passed by a

legislative body) and agreements for

the 3Ss is presented in Table 1.

3 Nuclear 3S objectives

By considering the objectives of each

of the 3Ss it becomes clear that they

share the same broad aim and desired

outcomes.

Convention on the Early

Notification of a Nuclear Accident

or Radiological Emergency

Threats to International Peace and

Security caused by Terrorist Acts –

UN Resolution 1373

Code of Conduct on the Safety and Security of Radioactive Sources

Joint Convention on the Safety

of Spent Fuel Management and

on the Safety of Radioactive Waste

Management

Code of Conduct on the Safety

of Research Reactors

| | Tab. 1.

International Legislation and Agreements.

Safeguards Agreements

Additional Protocols

Non-proliferation of Weapons

of Mass Destruction –

United Nations Security Council

(UNSC) Resolution 1540

Comprehensive Test Ban Treaty

(CTBT)

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atw Vol. 63 (2018) | Issue 1 ı January

ENVIRONMENT AND SAFETY 22

| | Fig. 1.

Schematic showing the general ranges of applicability of the 3 methods of Fault Analysis 2,3 .

attitudes in organizations and individuals

which establishes that, as an

overriding priority, protection and

safety issues receive the attention

­warranted by their significance”

[9]. The development of a good

safety culture requires a transparent

approach to information sharing and

dissemination. This helps ensure that

incident reoccurrences can be prevented,

and others who may be using

the same or similar equipment,

techniques or procedures can review

their arrangements to prevent a

similar incident.

“The existence of a good safety

culture is a prerequisite for the

implementation of a good safety case.

The converse is also true” [10]. This

enables a good safety case to be

­translated into beneficial changes

in behaviour associated with the

existing safety culture and arrangements

for the management of safety.

Practicing a graded approach

to safety ensures that the effort

expanded is proportionate to the

possible consequences. Figure 1 is

from the Office for Nuclear Regulation

Safety Assessment Principles [11]

and shows the applicability for

the methods of fault analysis; PSA,

DBA and SAA. Thus more assessment

effort is expended on those higher

consequence and higher frequency

events.

2) Office for Nuclear

Regulation [11].

3) Target 4 (BSL):

‘ Target 4 is

intended to provide

a broad indication

of where DBA might

be expected to be

applied’ [11]. BSL –

Basic Safety Level

4) Based upon a

Sandia National

Laboratories

diagram

| | Fig. 2.

Design and Evaluation Process Outline 4 .

techniques to demonstrate the

effectiveness of defence in depth,

such as:

• Design Basis Analysis (DBA): to

ensure that the design is robust,

fault tolerant and has effective

safety measures;

• Probabilistic Safety Analysis (PSA):

to ensure risks are acceptable,

understand inter-dependencies

and to evaluate failures; and

• Severe Accident Analysis (SAA): to

determine further practicable

measures to improve defence in

depth.

The hierarchical view deviations,

incidents and accidents for nuclear

­facilities is compared against five

levels of defence in depth [7] for

safety:

• Preventing deviations from normal

operations;

• Controlling deviations from operational

states;

• Controlling accidents within the

design basis;

• Mitigating accidents and ensuring

confinement of radioactive materials;

and

• Mitigating the radiological consequences

of radioactive releases.

This hierarchical view allows

designers, operators and others to

identify where they can most effectively

contribute to maintaining safety.

The Safety Case is a well-documented

approach normally used by

regulators for proportionally assessing

the safety submissions against

the radiological hazards presented.

Safety cases are typically defined as a

“ structured argument, supported by a

body of evidence that provides a

compelling, comprehensible and valid

case that a system is safe for a given

application in a given operating

environment” [8].

For the safe operation of a nuclear

site, facility or activity an effective

safety culture needs to be in-place and

­fostered. Safety culture is defined as

“The assembly of characteristics and

4.2 Security

A number of methodologies are used

in security to increase the likelihood

of creating and maintaining secure

operations. An example holistic

approach is the Design and Evaluation

Process Outline (DEPO) (Figure 2)

[12]. The physical protection system

(PPS) is developed from determining

the targets to be protected from the

postulated malicious capabilities, and

then designing for delay, detection,

assessment and response. Vulnerability

assessment is undertaken to

ensure that the PPS is likely to be

effective and depending on the outcome

the design will be refined or

implemented.

However, a number of assessment

techniques need to be deployed

and the associated performance

measures calculated and considered

for operational acceptance. For

example, a sensitive detector with a

high probability of detection may

detect all intrusions but have a high

false alarm rate such that responders

ignore the alarms being received.

Defence in depth for security

[7] comprise layers of physical and

Environment and Safety

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atw Vol. 63 (2018) | Issue 1 ı January

| | Fig. 3.

Schematic showing the general ranges of applicability for 3 security assessment methods.

technical measures, along with

operational and procedural protection

that have to be overcome or circumvented

by an adversary. The defence

in depth approach should be applied

in the following order:

• Detecting a potential malicious act;

• Delaying the adversary to allow an

appropriate response; and

• Responding to or neutralising the

attack.

Risk management techniques also

need to be used to balance investment

required to prevent high consequence

low probability events and low consequence

high probability events without

putting an unnecessary burden on

operational processes.

A State’s Design Basis Threat

(DBT) [13] details the capability and

capacity of the malicious actors that

any PPS should counter to maintain

the security of nuclear material

and other radioactive materials.

Using that as a basis for the threats,

scenarios should be developed that

will be used in vulnerability assessments

to determine how effective

the security arrangements are likely to

be in practice. Any weaknesses should

be identified so that compensatory or

enhancements in the arrangements

can be implemented. The type and

quality of assessment techniques that

should be undertaken is suggested in

Figure 3.

The Nuclear Security Case 5 , part of

the Nuclear Site Security Plan, is a

more recent development than the

Safety Case. Like the safety case the

security case “should justify the

claims, arguments and rationale for

the ‘duty holders’ security regime by

substantiating the security arrangements

for a site, plant, activity, operation

or modification. It should provide

written evidence that the relevant

security standards have been or are

going to be met. It should also demonstrate

that the risk posed by malicious

activity has been reduced as far as

could be reasonably expected” [14].

As with the safety case effort should

be expended to reviewing security

as a system rather than as individual

components.

Risk management techniques

are used to manage any variations

between the optimal arrangements

and what is in currently in place,

particularly when a possible vulnerability

is identified.

Vulnerability assessment techniques

to determine the performance

of security arrangements involve

many aspects of the system performance

including the probability of

detection, probability of interruption,

probability of neutralisation and

probability of effectiveness. The

IAEA Nuclear Security Assessment

Methodology (NUSAM) Co-ordinated

Research Programme (CRP) has been

establishing a risk-informed, performance-based

methodological framework

for nuclear security assessment

at sites, facilities and activities so that

practitioners will be better informed

of the approaches and techniques that

can be used, and those that provide

the most effective assessment and value

for the different facility type. The

CRP also allows the different methods

to be compared and helps to identify

the comparative strengths, weaknesses

and limitations of the alternative

approaches. This should ensure

a consistency of approach in security

assessment, and therefore by implication

a baseline standard for international

approaches.

As in the case of safety for the

secure operation of a nuclear site,

facility or activity, an effective security

culture needs to be in-place and

­fostered. Security culture is defined

as “The assembly of characteristics,

attitudes and behaviour of individuals,

organizations and institutions

which serves as a means to support

and enhance nuclear security” [15].

Security and safety culture are

both based upon the principles of

adopting a questioning attitude, rigorous

and prudent approaches, and

effective communication.

4.3 Safeguards

Underpinning and implementing

the principles within the Non-Proliferation

Treaty (NPT) the main

approaches used by the safeguards

community for the protection of civil

nuclear material preventing it from

being redirected into weapons activities

is ‘Safeguards by Design’ [16]

and nuclear materials accountancy

and control (NMAC). The physical

arrangements including Tamper

Indicator Devices, multiple barriers,

NMAC and facility arrangements such

as Material Balance Areas provide

additional measures for defence

in depth aiding the inspection of

material and the ability to detect

potential diversion.

Inspection and material characterisation

activities are used in decision

analysis to determine whether

the plant or facility is operating to

specification and agreement.

Safety, security and safeguards

broadly follow the same principles to

achieve protection.

5 3S synergies

The synergies and major considerations

in the nuclear 3S are shown in

Figure 4 [17]. This identifies the main

issues and considerations within

the 3S and where they intersect and

overlap, irrespective of the type of

regulatory regime.

5.1 Triple S

Moving into the practice of 3S,

through the applications of methods

and techniques we use the term Triple

S. Thus, when Figure 4 is revised

with a selection of typical, but not

exhaustive, activities and assessments

5) Part of NNL’s

approach to

demonstrating

compliance with

ONR’s Nuclear

Security Assessment

Principles (SyAPs).

ENVIRONMENT AND SAFETY 23

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atw Vol. 63 (2018) | Issue 1 ı January

ENVIRONMENT AND SAFETY 24

| | Fig. 4.

Nuclear 3S 6 .

| | Fig. 5.

Nuclear Triple S.

6) Based upon ‘An

integrated approach

to nuclear safety

and security: in the

context of 3S’,

Jor-Shan Choi,

Tokyo, Japan,

9 December 2011

7) An area inside a

protected area containing

equipment,

systems or devices,

or nuclear material,

the sabotage of

which could directly

or indirectly lead to

high radiological

consequences [18].

that are undertaken the resulting

synergies, overlaps and interaction

are presented in Figure 5. For

­example, Vital Area Identification

(VAI) is a process to identify potential

high consequence targets so that

protection can be provided to prevent

or reduce the likelihood of sabotage

[18]. Although VAI is primarily driven

by a security need the contribution

from safety specialists’ considering

the potential consequences and

operational limitations is important,

and is therefore shown in the intersection

of safety and security. There

are some activities that all specialisms

contribute to, such as new nuclear

build and stakeholder engagement,

and is shown across all three sections

of the diagram requiring input from

all three. It is where activities fall into

more than one area that deliberate

and positive interactions between all

the specialisms will provide added

value and potential conflicts are

averted or minimised. Each specialist

develops a clearer understanding of

the needs, intentions and priorities of

the other specialists, resulting in an

integrated approach to Triple S. Thus

time, effort and cost are minimised as

plant workarounds, reworks or design

changes are prevented, and operational

arrangements can be considered

earlier in the project.

Exploring this in further detail,

safety and security, followed by

security and safeguards, is where the

largest interaction, potential synergies

and similar approaches are to be

found.

5.2 Safety and security

Security requires extensive safety

­input for the identification of Vital

Areas 7 . The safety assessments including

radiological consequence

modelling, radiological hazard analysis,

PSA, SAA, internal and external

hazards, and layout design all contribute

to identifying potential Vital

Areas.

The design basis accidents and

design basis threats (DBT) approaches

in both specialisms guide designers,

practitioners and assessors to adequately

consider those threats that

may need to be countered.

Safety and security both use a

graded approach. The relative importance

of accident prevention and

mitigation measures is expressed in

terms of the adverse consequences for

public and worker health. Likewise

the relative importance of security

measures is directed towards preventing

and limiting what are considered

high and low consequence

events.

Prevention, Response, Control

and Management effort to counter

malicious attack for security, or accidents

in safety require considerations

on the speed of progress of an incident,

the potential consequences of

those responses and management

actions and how to minimise the

impact on the plant, people, public

and environment.

Approaches and methods used in

the minimisation of impact for radiological

consequence through ‘As Low

As Reasonably Practicable’ (ALARP)

practices in safety are commensurate

with those used by security not to

create ‘As Secure As Reasonable

Practicable’ (ASARP) but rather the

introduction of risk management

practices to manage potential vulnerabilities

identified through PPS

evaluation activities.

Safety and security both encourage

and embrace Advisory Missions and

inspections; from the World Association

of Nuclear Operators (WANO),

Integrated Regulatory Review Service

(IRRS) and Operational Safety Review

Team (OSART) for safety; from the

International Physical Protection

Advisory Service (IPPAS) for security,

and which is understood to potentially

be expanded to include a module on

NMAC.

Safety and security both attempt to

foster positive cultures that identify

and report problems and issues. However

the transparent and open communications

of safety may conflict with

the ‘need to know’ principles employed

in security. Appropriate implementation

of ‘need to know’ principle where

consideration is given to what is

‘ needed to be known’ can d irect appropriate

filtering and ­redaction so that

appropriate inter actions can occur

without compro mising security of

materials or information. For example,

consequence assessors do not need to

know the locations or means that

material can be acquired by a perpetrator

to undertake the assessment.

Environment and Safety

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atw Vol. 63 (2018) | Issue 1 ı January

Thus, by fostering an approach

that integrates both safety and

security in a mutually supporting

manner through peer-to-peer and

other challenges of behaviours creates

the opportunity for reinforcement of

positive behaviours.

5.3 Safety and safeguards

Safety and safeguards particularly

interact during the design activities

around ‘Safeguards by Design’ [19],

and then through the construction

phase. One example of this interaction

is preventing diversion through layout

design, aiding inventory control

(­Nuclear Material Accounting and

Control – NMAC) through criticality

control accountancy measures.

Inspections are carried out as part

of an integrated safeguards regime

and undertaken by inspectors from

the IAEA, Euratom or the Brazilian-

Argentine Agency for Accounting and

Control of Nuclear Materials (ABACC).

5.4 Security and safeguards

Security and safeguards should interact

during the design activities around

‘Safeguards by Design’ [19], and then

through the operational phase of the

facility or plant. Both safeguards and

security [20] are aimed at deterring

and detecting unauthorised removal

of nuclear material, providing assurance

that all nuclear material is secure

and timely detection of any material

loss.

There are areas where security and

safeguards can interact to improve

­effectiveness and efficiency in achieving

their objectives such as research

and development of Non-destructive

Assay equipment and surveillance

system, analysis capability (i.e.

­Nuclear Forensics, Destructive Analysis)

and, Security- and Safeguardsby-

design. Further, enhancing nuclear

security may be achieved through the

use of nuclear material accounting

and control systems [21]. This is an

approach being advocated by the IAEA

and clearly demonstrates how existing

accountancy measures can be utilised

to provide a potential additional

means through which the theft of

material by an insider can be detected.

6 Duty holder application

of triple S integration

The integration of Nuclear Safety,

­Nuclear Security and Nuclear Safeguards

can be beneficial for nuclear

site duty holders, operators and

tenant organisations. More broadly,

the early integration and interaction

of safety and security in critical

national infrastructure (CNI) and other

projects that require a security input

is of value. A duty holder can begin

the integration of Safety, Security

and Safeguards (SSS) by the formation

of a SSS team, bringing together

Safety, Security, Safeguards and the

broader safety disciplines. The following

application of Triple S integration

shows how a security technique can be

applied to a recent change in nuclear

registration within the UK and how

this technique can be bolstered by the

Safety and Safeguards disciplines.

6.1 NNL application and

experience

Returning to the UK nuclear industry,

the Office for Nuclear Regulation

(ONR), recently replaced its security

guidance to support the regulations

by introducing the Security Assessment

Principles (SyAPs) as a replacement

for the National Objectives,

Requirements and Model Standards

(NORMS). NORMS was considered by

some to be a prescriptive approach to

nuclear security regulation. It set out

security objectives that dutyholders

were expected to meet. However,

some in the industry viewed the

suggested Model Standards, that were

presented as what may allow a facility

or site to meet regulatory compliance

was provided as guidance. The introduction

of SyAPs is a move to an outcomes-based

regulatory regime and a

non-prescriptive approach to nuclear

security, giving duty holders more

freedom and therefore more space for

Triple S integration. Importantly in

the context of 3S principles, SyAPs are

more in line with the Safety Assessment

Principles (SAPs), reinforcing

the benefits of adopting an integrated

approach to safety and security, and

working together, learning for each

other, and adapting methodologies to

meet similar regulatory expectations.

The integration of Triple S has been

recognised by the ONR as an efficient

way of thinking, this is reflected in the

formation of the Security Informed

Nuclear Safety (SINS) team within

ONR.

However, with the introduction of

SyAPs, duty holders across the UK

must review their current security

arrangements so that the requirements

of SyAPs can be met. Reviewing

nuclear site security measures in line

with SyAPs using a team that includes

specialists from the three disciplines;

Safety, Security and where appropriate

Safeguards; will allow duty

holders to better address the principles

and gain organisational value.

6.2 Operational requirements

The Centre for the Protection of

National Infrastructure (CPNI) is the

government authority for protective

security advice to the UK national

infrastructure. [22]. CPNI provides

tools to help CNI companies and

organisations undertake an improved

security assessment of their sites, and

their methods are often considered

‘best relevant practice’ and serve as a

logical approach to an outcomes-based

regulatory regime such as SyAPs.

One such method promulgated by

CPNI is the Operational Requirements

(ORs) process. [23] The OR process

identifies, develops and aids justification

of actions to be taken and

investments to be made to protect

assets. [24] The OR process consists of

two levels; Level 1 OR seeks to:

• Identify assets and critical infrastructure

• Identify threats and vulnerabilities

• Assess possible risks

• Identify risk mitigation options

and develop a Strategic Security

Plan (SSP) Review organisational

readiness to deliver the developed

SSP.

Level 2 OR is a continuation of the

Level 1 OR. It is concerned with

in-depth analysis of requirements

suggested as a result of the security

posture formed from the Level 1 OR

process. An example application of

the OR process with regards to SyAPs

can be seen below (Figure 6).

The OR process provides a useful

vehicle for the integration of Safety,

Security and Safeguards, from the

perspective of ‘Safety and Safeguards

informed Security’ (SSIS 8 ). SyAPs

requires the categorisation of nuclear

sites and facilities, and nuclear

­material (NM) and other radioactive

materials (ORM) for both theft and

sabotage; it follows logically to

integrate safety specialisms when

considering potential consequences

(categorisation) and the malicious

actions that may be undertaken to

achieve such consequences, as well as

considering the implications of operational

‘flow’ of material around a

­proposed facility from an NMAC

perspective. Such perspectives may

further inform the design process at a

high level (remembering the purpose

of OR1).

Using the principles behind the OR

process and SSIS in conjunction,

assets, vulnerabilities, risks and

mitigations are found, resulting in

a security posture for the site that

takes due account of safety and safeguards.

Triple S integration allows

8) Coined herein to

describe the

intermediate stage

between individual,

isolated Safety,

Security and

Safeguards

functions and the

notion of fully

integrated ‘SSS’.

ENVIRONMENT AND SAFETY 25

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ENVIRONMENT AND SAFETY 26

| | Fig. 6.

Example of the integration of Triple S into phases of an OR process.

| | Fig. 7.

Level 1 OR, Categorisation.

reinforcement of decisions, shown

above as the SAPs supports the regulatory

guidance for security categorisation.

Additionally, the use of nuclear

and radiological safety consequence

analysis supports the security categorisation

for sabotage. The Level 1

OR process can be used for SyAPs

reviews as shown in Figure 7.

Once a duty holder has categorised

their nuclear site, the Level 2 OR

process can assess the individual

security requirements of site and

facility areas (Figure 8).

The Level 2 OR process assesses

the site in terms of its security

capabilities. The outcomes of the

Level 2 OR process are a set of

performance requirements against

the defined functions that the Physical

Protection System (PPS) must meet in

order to be compliant with the

standards held by SyAPs; those being

Delay, Detect, Assess, Control of

Access and Insider Mitigation.

SAPs feeds into the regulatory

guidance that underlies the security

assessment. Initial attempts at applying

the safety methodology of HAZOPs

(using keywords to explore potential

issues in the design and test for ‘compliance’)

resulted in a level of success.

However, this experiment highlighted

the fundamental differences between

safety and security, in that safety

can be probabilistically assessed and

security cannot. Said differently, the

laws of physics and attributes of

systems/components determine what

is and is not possible in the world of

safety. In the world of security, outcomes

are more strongly determined

by malicious capabilities (knowledge

and resources) and their imagination;

as such security scenarios cannot be

conceptualised deterministically and

calculated probabilistically.

The outcomes of the Level 2 OR

process are defined and communicated

in a Performance Specification.

The Performance Specification relays

the PPS specifications that the duty

holder requires to the design process.

The PPS design process is carried out

using aspects of Safety, Security and

Safeguarding to update the nuclear

facility and maintain high standards

in all 3 fields (Figure 9).

SSS (or SSIS) can influence ­specific

design aspects, such as turnstile

requirements (linking access control

and emergency egress), material store

access and surveillance features

( security and safeguards) and material

handling limits in specified areas

(radio logical protection and counterdiversion/insider

threat mitigation).

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Duty Holders can validate the

results of the OR process through

further vulnerability assessments of

the derived PPS. This will require

further Triple S integration as each

field will require an analysis of

updated facilities to ensure that high

standards continue (Figure 10).

Integrating Triple S in the early

stages of a design project (similarly to

the application above) can prevent

future costs and save time. Our

experience has shown a greater

engagement and interaction of previously

disparate disciplines, whose

assessment needed to be rationalised

and integrated, often leading to

re-assessment/re-work to ensure

consistency of assessment boundaries,

assumptions, etc.

If integration is not considered,

then designs implemented by one

discipline can interfere with designs

implemented by another and the

­earlier examples of beneficial integration

may be pre-empted by the

need for avoidance of conflicts or

issues. A basic example of issues raised

by the lack of integration can be

shown by the installation of security

fences interfering with on-site fire

safety (evacuation routes), forcing an

expensive retrofit on the security

fence.

7 Conclusion

This paper has covered NNL’s progress

to date in Triple S integration ( referred

to as SSIS rather than SSS) and its

implementation in a new concept

design project.

Whilst the potential benefits of an

integrated Triple S approach are

abundantly clear, it is somewhat more

difficult to realise these conceptual

benefits practically. The National

Nuclear Laboratory (NNL) has made

significant progress in its own

approach to aligning the three

disciplines, though the approach could

still be described more as ‘Safety and

Safeguards Informed Security’ (SSIS).

Experience thus far has ­identified that

specialists in Triple S disciplines need

to become more aware of the priorities,

approaches, methods and drivers

of other specialists delivering their

respective objectives to develop and

promote an integrated approach.

NNL has observed more effective

cross-specialism communication and

interactions and much heightened

awareness and interaction between

the broader organisation and Triple S

functions. Triple S can lead to increasing

professionalism as methods

and techniques used by one group of

| | Fig. 8.

Level 2 OR, specific requirements of PPS.

| | Fig. 9.

Performance Specification and Design Stages of OR.

| | Fig. 10.

Vulnerability Assessment of new Facility Design.

specialists are adapted and used by

others through sharing of knowledge

and learning from experience. Interaction

with the other specialists can

lead individuals to reconsider how to

undertake work and what information

is important such that safety, security

and safeguards are integrated in a

holistic manner.

Further, integration of 3S is more

likely to be achieved and be effective

in the early design and construction

phases of a project, with the positive

effects being realised as cost and

­efficiency benefits throughout operation.

Early interaction reduces the

­potential for conflict by identifying

where negative interactions might

ENVIRONMENT AND SAFETY 27

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atw Vol. 63 (2018) | Issue 1 ı January

ENVIRONMENT AND SAFETY 28

occur, thus, potentially expensive

rework or compromises are removed.

The application of the OR process

to a SyAPs review shows that safety,

security and safeguards can bolster

the effectiveness of new design projects.

It shows the importance of

integration and its cost and time

saving potential, and that the

legitimacy of the Triple S approach

spans beyond the conceptual stage.

Abbreviations

3S Safety, Security and Safeguards

ABACC Brazilian-Argentine Agency for

Accounting and Control of Nuclear

Materials

ALARP As Low As Reasonably Practicable

ASARP As Secure As Reasonably Practicable

CPPNM Convention on the Physical

Protection of Nuclear Materials

CRP

DBA

DBT

Co-ordinated Research Programme

Design Basis Analysis

Design Basis Threat

DEPO Design and Evaluation Process

Outline

IAEA International Atomic Energy Agency

IPPAS International Physical Protection

Advisory Service

IRRS Integrated Regulatory Review Service

NMAC Nuclear Material Accountancy and

Control

NPT Non Proliferation Treaty

NUSAM Nuclear Security Assessment

Methodologies

OSART Operational Safety Review Team

PPS Physical Protection System

PSA Probabilistic Safety Analysis

SAA Severe Accident Analysis

Triple S Safety, Security and Safeguards

UN United Nations

UNSC United Nations Security Council

VAI Vital Area Identification

WANO World Association of Nuclear

Operators

References

[5] International Atomic Energy Agency,

Objectives and Essential Elements of a

State's Nuclear Security Regime, Vienna:

International Atomic Energy Agency,

2013.

[6] International Atomic Energy Agency,

Nuclear Security Recommendations on

Physical Protection of Nuclear Material

and Nuclear Facilities, Vienna: International

Atomic Energy Agency,

January 2011.

[7] International Atomic Energy Agency,

Management of the Interface between

Nuclear Safety and Security for Research

Reactors, International Atomic Energy

Agency, Vienna, 2016.

[8] J. Inge, The Safety Case, Its Development

and Use in the United Kingdom 2 (n.d.),

Ministry of Defence.

[9] International Atomic Energy Agency,

IAEA Safety Glossary, International

Atomic Energy Agency, Vienna, 2007.

[10] R. J. Cullen, Safety Culture: Cornerstone

of the Nuclear Safety Case, in Hazards

XXI: Process Safety and Environmental

Protection in a Changing World,

Manchester, 2009.

[11] Office for Nuclear Regulation, Safety

Assessment Principles for Nuclear

Facilities, Office for Nuclear Regulation,

Bootle, 2014.

[12] Sandia National Laboratory and Japan

Atomic Energy Agency, Security by

Design Handbook, 2013.

[13] International Atomic Energy Agency,

Development, Use and Maintenance of

the Design Basis Threat: Implementing

Guide, Vienna: International Atomic

Energy Agency, 2009.

[14] Office for Nuclear Regulation, Guidance

on the Purpose, Scope and Quality of a

Nuclear Site Security Plan, Office for

Nuclear Regulation, Bootle, 2016.

[15] International Atomic Energy Agency,

Nuclear Security Culture, International

Atomic Energy Agency, Vienna, 2008.

[16] R. S. Bean, T. A. Bjornard und

D. J. Hebditch, Safeguards-by-Design:

An Element of 3S Integration, in IAEA

Symposium on Nuclear Safety, April

2009.

[17] J.-S. Choi, An integrated approach to

nuclear safety and security: in the

context of 3S, in JAEA International

Forum on Peaceful Use of Nuclear

Energy and Nuclear Security, Tokyo,

2011.

[22] CPNI, About CPNI, 2017. [Online].

Available:

https://www.cpni.gov.uk/about-cpni.

[Zugriff am 10 October 2017].

[23] CPNI, Operational Requirements, 2017.

[Online]. Available:

https://www.cpni.gov.uk/operationalrequirements.

[Zugriff am 10 October 2017].

[24] CPNI, Guide to Producing Operational

Requirements for Security Measures,

2016.

Authors

Howard Chapman

Jeremy Edwards

Joshua Fitzpatrick

Colette Grundy

Robert Rodger

Jonathan Scott

National Nuclear Laboratory

Fifth Floor, Chadwick House

Warrington Road, Birchwood Park,

Warrington, WA3 6AE,

United Kingdom

[1] NNL, Nuclear Safety, Security and

Safeguards: An Integrated Approach,

2017.

[2] Ministry of Foreign Affairs of Japan,

International Initiative on 3S-Based

Nuclear Energy Infrastracture, G8

Hokkaido Toyako, Institute of Oriental

Culture, University of Tokyo, Hokkaido

Toyako, 2008.

[3] M. Weightman, Leadership and

Organisational Aspects, Bootle: Office

for Nuclear Regulation, 2011.

[4] International Atomic Energy Agency,

Fundamental Safety Principles, Vienna:

International Atomic Energy Agency,

November 2006.

[18] International Atomic Energy Agency,

Identification of Vital Areas at Nuclear

Facilities, International Atomic Energy

Agency, Vienna, 2012.

[19] R. S. Bean, J. W. Hockert und

D. J. Hebditch, Integrating Safeguards

and Security with Safety into Design,

in 19 th Annual EFCOG Safety Analysis

Workshop, 2009.

[20] K. Murakami, Nuclear Safeguards

Concepts, Requirements, and Principles

applicable to Nuclear Security, July 2012.

[21] International Atomic Energy Agency,

Use of Nuclear Material Accounting and

Control for Nuclear Security Purposes at

Facilities, International Atomic Energy

Agency, Vienna, 2015.

Environment and Safety

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


atw Vol. 63 (2018) | Issue 1 ı January

Clearance of Surface-contaminated

Objects from the Controlled Area

of a Nuclear Facility: Application of the

SUDOQU Methodology

F. Russo, C. Mommaert and T. van Dillen

1 Introduction During and after the Fukushima nuclear accident, the possibility existed that surface-contaminated

consumer goods, freight containers and conveyances would be imported from Japan, which revealed the need

for proper criteria and screening levels for surface contamination of these items, to insure protection of the public.

In this framework, it was concluded

that the then existing dose-calculation

models mostly addressed exposure

scenarios for occupationally exposed

workers, which were generally not

aimed at properly evaluating the

effective dose incurred by members of

the public exposed to surface-contaminated

objects. The main difference

between occupational and public

exposure scenarios is that, while

workers may frequently be exposed

to freshly contaminated objects,

members of the public are likely to

come in contact with only one (same)

object during a prolonged period of

time. Therefore, while the hypothesis

of a constant contamination level may

suffice for occupationally exposed

workers, it is less realistic for objects

handled by members of the public,

where the initial contamination

present on the object will be affected

by several removal mechanisms,

which need to be considered when

evaluating the annual effective dose.

Based on these findings, the Dutch

National Institute for Public Health

and the Environment (RIVM) developed

the SUDOQU (SUrface DOse

QUantification) methodology [1] for

the evaluation of the annual effective

dose for members of the public

resulting from exposure to surfacecontaminated

objects. It assumes

time-dependent surface- and air- contamination

levels, whose evolution is

governed by a system of coupled

differential equations, describing the

mass balance imposed by the involved

mechanisms. The surface-activity concentration

(Bq/cm 2 ) is considered to

decrease by radioactive decay,

resuspension and wipe-off (transfer

of activity to the hands). The resuspended

activity contributes to the

(increase in) air-activity concentration

(Bq/m 3 ) and can, in turn, partly re-deposit

onto the object surface. The air

activity concentration is further

affected by radioactive decay and

ventilation. Different exposure pathways

are considered: external-gammaradiation

exposure, inhalation, indirect

ingestion and skin contamination

through wipe-off. The effective dose

can then be calculated as the sum of

the contributions of the exposure

pathways. Based on these intrinsic

properties, the SUDOQU methodology

is particularly attractive for clearance

and exemption calculations, especially

when considering public reuse

scenarios, because they often involve

the prolonged use of the same object.

Therefore, in 2016, a collaboration

was started between Bel V and RIVM,

to extend the scope of the SUDOQU

model, and to test its suitability for the

derivation of surface-clearance levels

for objects released from the controlled

area of a nuclear facility.

2 Objectives and

methodology

The results presented in this paper

were obtained in the framework of a

pilot project, having as main objective

to investigate the applicability of the

SUDOQU methodology for clearance

calculations, and to gain a better

understanding of the interplay among

the involved mechanisms and how

this affects the resulting total effective

dose. This was achieved by performing

deterministic calculations

of the annual effective dose resulting

from exposure to a typical office

item, i.e. a bookcase, considering

different scenarios of use and different

nuclides.

2.1 Reference scenario

In the reference scenario (scenario 1),

a bookcase is considered that leaves

the controlled area of a nuclear facility

with a homogeneous surface contamination

of 1 Bq/cm 2 (different

radionuclides are considered, as

explained further in this Section).

Next, the bookcase is placed in an­

­office with a 50-m 2 area and a 2.5-m

height and is used by an “average”

­office worker, who will be exposed to

the contaminated surface. During

working hours (i.e. 8 h/d, 5 d/w, and

52 w/y, resulting in 2080 h/y, thus

in a duty factor f exp =0.24 [1]) the

worker is in the office at a distance of

3 m from the contaminated bookcase,

by which he incurs a certain exposure

by external (gamma) radiation. The

bookcase is assumed to be contaminated

only on its front panel,

characterised by a 6-m 2 surface.

For the calculation of the externalradiation

dose contribution, the

conversion factor from ambient

dose equivalent to effective dose

(E/H*(10)) is set equal to one, which

is conservative for any irradiation

geometry in the photon energy range

of the considered nuclides. During

­office hours, the worker is assumed

to occasionally touch the bookcase,

thereby wiping off some activity from

its surface, with a frequency of

approximately once every three

hours (ϕ = 0.31 h -1 during use) and

an ­efficiency of 20% (f oth = 0.2, corresponding

to the ratio of the contamination

level of the hands after a

wipe-off event and that of the bookcase).

Activity is also transferred

indirectly to the face after contact

with the hands. This transfer is

­modelled by an efficiency of f htf =0.2

(ratio of contamination levels of face

and hands). The individual will thus

incur a skin equivalent dose following

contamination of the skin area of

the hands (A hands =400 cm 2 ) and

of the face (A face =100 cm 2 ), which

eventually also contributes to the

effective dose. Furthermore, part of

the activity on the hands will be

transferred to the mouth (indirect

ingestion): this is assumed to occur

with a frequency equal to that of

wipe-off (0.31 h -1 ). The activity transferred

from the hands to the mouth

per ingestion event is set equal to

100 % (f htm =1) of the activity present

atw-Special „Eurosafe

2017“. In cooperation

with the EUROSAFE

2017 partners,

Bel V (Belgium),

CSN (Spain), CV REZ

(Czech Republic),

MTA EK (Hungary),

GRS (Germany), ANVS

(The Netherlands),

INRNE BAS (Bulgaria),

IRSN (France),

NRA (Japan),

JSI (Slovenia),

LEI (Lithuania),

PSI (Switzerland),

SSM (Sweden),

SEC NRS (Russia),

SSTC NRS (Ukraine),

VTT (Finland),

VUJE (Slovakia),

Wood (United

Kingdom).

Revised version

of a paper presented

at the Eurosafe,

Paris, France, 6 and

7 November 2017.

29

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OPERATION AND NEW BUILD 30

on the ingested area, but it is assumed

that, per ingestion event, indirect

ingestion occurs only from a limited

fraction of the surface of the hands,

i.e. f ing A hands , with f ing =0.01. Activity

on the hands is assumed not to be

affected by removal through indirect

ingestion or transfer to the face

( conservative approach). Moreover, it

is assumed that a certain fraction of

activity is re-suspended from the

object surface and becomes airborne,

therefore producing an effective dose

contribution through inhalation. The

dose conversion factors for the inhalation

and ingestion pathways are

those indicated in ICRP Publication 72

[3] for an adult member of the public.

More specifically regarding the inhalation

dose, when dose conversion

­coefficients for different lung absorption

types are available, the most

conservative value is chosen.

To study the role of the characteristics

(type and energy) of the

emitted radiation, the dose calculation

for this scenario was performed

for several radionuclides (βγ- or pure

β-emitters) and for a radioisotopic

composition typical of a nuclear

­power plant, as indicated in the first

row of Table 1. The latter is labelled

“NPP” in Table 1 and Figure 1, and

corresponds to the nuclide vector of

the whole site of the nuclear power

plant in Doel, Belgium, in the year

2015-2016. Radioactive progeny is

here considered to contribute to the

dose only if equilibrium can be

reached within the time-integration

period of one year and for sufficiently

large branching ratios. For the list of

considered nuclides, this is the case

for Cs-137 (including Ba-137m) and

Sr-90 (including Y-90).

2.2 Alternative scenarios

Starting from the reference scenario

described in Sect. 2.1, several alternative

scenarios were developed, by

varying one parameter at a time. This

was done to analyse the effect of

separate parameter variations on

the effective dose and therefore to

identify the most relevant parameters,

to which the results are most sensitive.

This study then serves as basis for a

more detailed sensitivity analysis.

In this study, five alternative

scenarios were developed from the

reference scenario. In scenario 2, the

distance of the worker to the contaminated

bookcase is increased from

3 m to 4.5 m. In scenario 3, wipe-off

events are assumed to occur with an

increased frequency of once per hour

(during use), instead of once every

three hours (the ingestion frequency,

instead, remains unvaried with

respect to the reference scenario). In

scenarios 4 and 5, the transfer

­efficiency f oth is decreased from 0.2 to

0.1 and 0.05, respectively. In scenario

6, the worker benefits from six weeks

of holiday, thus is only exposed during

46 weeks per year. As a result, the

duty factor decreases from 0.24 to

0.22.

3 Preliminary results

The obtained results are summarised

in Table 1, reporting the total annual

effective dose in the six scenarios for

all considered nuclides and for the

NPP nuclide vector.

It can be noticed from Table 1 that

the dose values for the considered

nuclides range from about 10 -1 µSv/y

for isotopes as Ni-63 and Co-57, to

values as high as 10² µSv/y for Pu-241.

The dose values resulting from exposure

to the NPP isotopic vector are

similar to those of its most abundant

radionuclide, i.e. Co-60.

The (large) differences among

the considered nuclides are related

to the characteristics of the emitted

radiation (type and energy of emitted

particles), the half-life of the nuclides

and the metabolic behaviour of these

elements when ingested or inhaled.

Note that, in general, results of a dose

evaluation will also strongly depend

on the type of object (geometry, surface

area, distance) and how exactly it

is used or handled. Effective doses

presented in Table 1 for the bookcase

may thus differ significantly from

those for other objects released from a

nuclear facility, because the relevant

exposure pathways may contribute

differently to the effective dose, in

absolute and relative sense. Variations

between nuclides may then also be

different from those observed in

Table 1, depending on their dominant

exposure pathways. The comparison

of results for several objects is currently

under investigation.

Furthermore, Figure 1 illustrates,

for each nuclide, the relative dose,

­defined as the ratio of the dose in a

specific scenario and the dose in the

reference scenario. Elimination of the

absolute differences by such normalisation

enables a way to compare the

relative impact of parameter changes

for the considered nuclides, thus a

comparison of parameter sensitivity

between nuclides. It can be observed

that, in most of the alternative

scenarios, the variation of the dose

with respect to the reference scenario

is rather heterogeneous for the considered

radionuclides. For example,

considering scenario 2, in which the

distance to the object is increased,

the total dose for βγ-emitters decreases

as a consequence of the reduction

of the external-gamma-radiation

dose, which is here the only contribution

affected by (a change in)

distance. The relative decrease, however,

is not the same for all nuclides,

as it depends on the relative contribution

of the external-gamma-radiation

pathway to the total dose, which

differs per nuclide. Accordingly,

reducing the distance with respect to

the object would lead to an increase

of the total dose, which is more pronounced

when the external-gammaradiation

exposure term is more

dominant: it can be shown that,

for the βγ-emitters considered here,

the total dose increases by a factor

between two and four when the

distance is reduced to 1 m. For purebeta

emitters, in which the externalgamma-radiation

component is absent,

the dose is not affected by a

variation of the distance. A certain

dose contribution could result from

external-beta radiation, but is not

considered here. In scenario 3, in

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

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

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

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

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

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

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

| | Tab. 1.

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.

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atw Vol. 63 (2018) | Issue 1 ı January

| | Fig. 1.

Variation of the total dose values in the six analysed scenarios for the bookcase with respect to the reference scenario (i.e. scenario 1).

which the wipe-off frequency is increased,

a decrease of the total dose is

observed for all nuclides. This can be

attributed to a more rapid removal

from the surface, which leads to a

reduction of the time-integrated

surface- and air-contamination levels,

and thus to a decrease of all dose

contributions. This decrease is of

course larger when wipe-off is a more

dominant mechanism for removal of

surface activity. As a result, in the case

of long-lived radionuclides, for which

radioactive decay does not constitute

a competing removal mechanism, the

wipe-off process will have a larger

­relative contribution, and the final

result will be more sensitive to a variation

in this mechanism: such radionuclides

will, therefore, show a larger

decrease than shorter-lived nuclides

as Co-56 and Co-58. In scenarios 4

and 5, a decrease in the transfer

­efficiency has two opposite effects. On

the one hand, activity residing on the

object surface will be removed at a

slower rate, leading to an increase of

the time-integrated surface-contamination

level (TISC). As a result, more

activity is available for resuspension,

thus the time-integrated air-contamination

level (TIAC) also increases.

Since the external-gamma-radiation

dose is proportional to TISC and the

committed effective dose from inhalation

is proportional to TIAC, both dose

contributions increase with respect to

the reference scenario. On the other

hand, the effective-dose contributions

from indirect ingestion and skin contamination

are both proportional to

the product f oth TISC (f oth decreases,

TISC increases). For the assumptions

made here, the product f oth TISC

decreases, thus the latter dose contributions

decrease. Altogether, the

total annual effective dose is the result

of the balance between the opposite

trends of these considered dose

contributions. For some nuclides (e.g.

­Co-60, Mn-54, Pu-241, and Eu-152)

the total dose increases as a result of

the increase of the external-radiation

exposure or inhalation contribution

(or a combination of both). For other

nuclides (e.g. Cs-137, Cs-134, Zn-65,

Sr-90) the total dose follows the

decreasing trend of its leading contribution,

i.e. ingestion. In other cases

(Co-56 and Co-57), the total dose

marginally changes, due to the fact

that the opposite effects approximately

cancel each other out. Finally,

in scenario 6, a decrease of the exposure

duration leads to an (approximately)

identical decrease in the total

dose for all nuclides (the relative

values in this scenario range between

0.90 and 0.95).

3.1 Benchmarking study

The results obtained with SUDOQU

were compared to the results obtained

with the model described in RP101

[2]. A graphical illustration of this

comparison is provided in Figure 3.2.

The RP101-model was chosen for the

benchmarking study because one of

the scenarios studied in RP101 considers

a surface-contaminated tool

cabinet, which is comparable to the

bookcase considered in this paper.

Moreover, like SUDOQU, the RP101-

model assumes a non-constant surface

activity. However, a fundamental

difference between the two models

is that the RP101-model only considers

radioactive decay as a removal

mechanism, whereas the SUDOQU

model considers other processes

affecting the evolution of the contamination

level (Sect. 1). Another

important difference concerns the

removability of surface contamination:

in SUDOQU, 100 % of the surface

activity is assumed to be remov able,

with a transfer efficiency of 20 %; in

RP101, only 10 % of the total surface

activity is removable, and the transfer

efficiency is equal to 10 %. These

differences lead to dissimilar (relative)

contributions of the exposure

pathways in the two models.

In this study, parameter values

­defining the exposure geometry and

duration in SUDOQU were harmonised

with those in RP101. In this way,

differences in dose results between

the two models are only related to

differences in model construction

and the (remaining) underlying

assumptions.

As a first step of the benchmarking

study, values of the remaining parameters

were left unvaried in SUDOQU

(i.e. values from Sect. 2.1), with the

aim of comparing the two models

based on their main, default assumptions

and to investigate their impact

on the results. The assumption in

RP101 that only 10 % of the total

surface activity is removable enhances

the dose contribution from externalgamma-radiation

exposure, as the

remaining 90 % of the surface activity

contributes exclusively to this pathway,

while only being modified by

radioactive decay. On the other hand,

the contribution of the other exposure

pathways, related to activity removal

from the surface (resuspension and

wipe-off), will be reduced in RP101

with respect to those in SUDOQU, for

which 100 % of the surface activity is

removable and may thus contribute to

these pathways (inhalation, ingestion

and skin contamination). Again, the

net outcome depends on the balance

OPERATION AND NEW BUILD 31

Operation and New Build

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


atw Vol. 63 (2018) | Issue 1 ı January

OPERATION AND NEW BUILD 32

| | Fig. 2.

Comparison of the total annual effective dose obtained for several radionuclides with the model in RP101 [2] and with SUDOQU. Values labelled as SUDOQU*

are obtained by applying the same removable fraction and wipe-off efficiency as those used in RP101.

between these opposite effects. For

Co-60 and Na-22, the smaller value

of the external-radiation dose in

SUDOQU (with respect to that in

RP101) is not fully compensated by

the larger values of the other dose

contributions, leading to a slightly

smaller total dose in SUDOQU. For

the other considered nuclides ( Cs-137,

Sr-90 and Pu-241), the opposite

occurs, leading to more conservative

results in SUDOQU.

An additional comparison was

made by implementing in SUDOQU

the same assumptions as in RP101

concerning the removable fraction

and the transfer efficiency. These

results are shown in Figure 2 as well,

indicated by the label SUDOQU*.

Due to these assumptions, the

external-gamma-radiation exposure

in SUDOQU* now increases to values

larger than those in RP101, while the

other dose contributions decrease,

although still being larger than the

values obtained in RP101. As a result,

the annual dose values obtained with

SUDOQU* are more conservative for

all considered nuclides, but in good

agreement with the RP101-results.

4 Conclusions

The SUDOQU model [1] enables dose

evaluations for exposure to a surfacecontaminated

object. It is characterised

by the innovative and distinctive

assumption of time-dependent

surface- and indoor air-contamination

levels governed by mass-balance

equations based on the following

mechanisms: radioactive decay,

resuspension, wipe-off, deposition

and ventilation. These features make

the SUDOQU methodology a suitable

candidate for performing clearance

calculations based on reuse scenarios,

where the individual is likely to be

exposed to the same object throughout

the year, and for which the

assumption of constant contamination

levels would be unrealistically

conservative. In this work, a surfacecontaminated

bookcase released from

the controlled area of a nuclear facility

is studied, with the aim of assessing

the applicability of SUDOQU for

the development of surface-clearance

criteria for nuclear facilities. Deterministic

calculations of the annual

effective dose were thus conducted for

several nuclides in different scenarios

of use. First, the results in this paper

reveal a strong nuclide dependency:

even within the same category of

emitters there can be pronounced

differences in absolute dose values,

depending on the radiological characteristics

of the nuclides and their metabolic

behaviour and radiobiological

impact on the human body. Moreover,

the consideration of a mass balance

describing the time evolution of the

contamination levels causes the total

annual dose to be the result of

a delicate interplay of the involved

elements. In this way, a variation of a

certain input parameter may lead to

opposite effects on the various dose

contributions, and thus to a total dose

that either decreases, increases or

remains constant. The net outcome

again depends on the characteristics

of the nuclide and on the specifics of

the exposure scenario. The results

obtained with SUDOQU were benchmarked

against the results reported in

RP101 [2] for the reuse scenario of a

tool cabinet, and the two models

proved to be in good agreement.

The results presented in this paper

not only demonstrate the suitability of

SUDOQU for dose assessments related

to clearance of objects from nuclear

facilities, but they are also a good

starting point to better understand

the intricate interplay among the

involved mechanisms. Their interaction

also disclosed the importance and

difficulty of a detailed sensitivity

analysis. Future work will focus on the

development of surface clearance

levels based on probabilistic and

realistically conservative dose assessments.

References

[1] T. van Dillen, SUDOQU: a new dose

model to derive criteria for surface

contamination of non-food (consumer)

goods, containers and conveyances,

Radiation Protection Dosimetry,

164(1-2) (2015), pp. 160-164.

[2] Radiation Protection 101: Basis for the

definition of surface contamination

clearance levels for the recycling or

reuse of metals arising from

dismantling of nuclear installations,

European Commission, 1998.

[3] ICRP, Age-dependent Doses to the

Members of the Public from Intake of

Radionuclides - Part 5 Compilation of

Ingestion and Inhalation Coefficients,

ICRP Publication 72. Ann. ICRP 26 (1),

1995.

Authors

F. Russo

C. Mommaert

Bel V

Rue Walcourt, 148

1070 Brussels,

Belgium

T. van Dillen

National Institute for Public Health

and the Environment (RIVM)

P.O. Box 1

3720 BA Bilthoven,

The Netherlands

Operation and New Build

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


atw Vol. 63 (2018) | Issue 1 ı January

34

DECOMMISSIONING AND WASTE MANAGEMENT

atw-Special „Eurosafe

2017“. In cooperation

with the EUROSAFE

2017 partners,

Bel V (Belgium),

CSN (Spain), CV REZ

(Czech Republic),

MTA EK (Hungary),

GRS (Germany), ANVS

(The Netherlands),

INRNE BAS (Bulgaria),

IRSN (France),

NRA (Japan),

JSI (Slovenia),

LEI (Lithuania),

PSI (Switzerland),

SSM (Sweden),

SEC NRS (Russia),

SSTC NRS (Ukraine),

VTT (Finland),

VUJE (Slovakia),

Wood (United

Kingdom).

Revised version

of a paper presented

at the Eurosafe,

Paris, France, 6 and

7 November 2017.

Carbon-14 Speciation During

Anoxic Corrosion of Activated Steel

in a Repository Environment

E. Wieland, B. Z. Cvetković, D. Kunz, G. Salazar and S. Szidat

1 Introduction Carbon-14 is an important radionuclide in the inventory of radioactive waste [1,2] and, due its

long half-life (5730 y), it has been identified a key radionuclide in safety assessments [3,4]. 14 C is of specific concern due

to its potential presence as either dissolved or gaseous species in the disposal facility and the host rock, the high mobility

of dissolved carbon compounds in the geosphere caused by weak interaction with mineral surfaces in near neutral

conditions, and eventually because it can be incorporated in the human food chain. Current safety assessments are

based on specific assumptions regarding the rate of 14 C release from potential sources, the 14 C speciation upon release

and the mobility of the different chemical forms of 14 C in the cementitious near field and the host rock [1].

The main source of 14 C in L/ILW in

Switzerland are activated metallic

nuclear fuel components and reactor

core components as well as spent

­filters and ion exchange resins used in

light water reactors (LWR) for the

removal of radioactive contaminants

in a number of liquid processes and

waste streams. Compilations of the

activity inventories revealed that

in the already existing and future

arisings of radioactive waste in

Switzerland, the 14 C inventory is

mainly associated with activated

(or irradiated, respectively) steel

(~85 %) while the 14 C inventories

associated with nuclear fuel components

(e.g. Zircaloy) and wastes

from the treatment of reactor coolants

(e.g. spent ion exchange resins)

are much less. 14 C in activated steel

results mainly from 14 N activation

( 14 N(n,p) 14 C) [2]. Release of 14 C

occurs during anoxic corrosion of

activated steel in the cementitious

near field of the L/ILW repository.

Recent reviews of corrosion rates

suggest that steel corrosion in these

conditions is a very slow process [5,6].

Carbon-14 can be released in a

variety of organic and inorganic

chemical forms. 14 C will decay within

a disposal facility if the 14 C-bearing

compounds are retained by interaction

with the materials of the

engineered barrier. For example,

inorganic carbon, i.e. 14 CO 2 and its

bases, is expected to precipitate as

calcium carbonate within a cementbased

repository or undergo 14 CO 3

2-

isotopic exchange with carbonate

minerals. For this reason inorganic 14 C

has only a negligible impact on the

14 C-based dose release. By contrast,

gaseous species containing 14 C, such

as 14 CH 4 , 14 CO etc., could form and

migrate with bulk gas from the near

field into the host rock. It is indicated

from previous studies that a limited

number of small organic molecules

| | Fig. 1.

Schematic presentation of the design of the corrosion experiment. Reactor set-up for the corrosion experiment with activated steel

(top); analytical procedures for the detection of 14C-bearing dissolved organic compounds (bottom) and gaseous species (right).

are likely to be formed in the course

of the anoxic corrosion of activated

steel in alkaline conditions, in particular

reduced hydrocarbons, such

as methane, ethane etc., and oxidized

hydrocarbons, such as alcohols,

aldehydes and carboxylic acids [7].

It is to be noted that both oxidized

and reduced hydrocarbons have been

observed in anoxic iron-water systems

in anoxic (near neutral to alkaline)

conditions which seems to be inconsistent

with a view to the negative

redox potential associated with the

systems [8].

Although the 14 C inventory associated

with activated steel is well

known, our understanding of the

chemical form of the 14 C-bearing

compounds produced in the course of

the anoxic corrosion of activated steel

is limited. The present study is aimed

to fill this knowledge gap.

2 Corrosion study

with activated steel

The schematic presentation of the

experimental design is displayed in

Figure 1 which includes a reactor

system to perform the corrosion

experiment with activated steel and

analytical methods for the identification

and quantification of the

14 C-bearing compounds in the liquid

and gas phases. The corrosion study

was supposed to be carried out using

steel components exposed to neutron

flux in a Swiss nuclear power plant

(NPP). To this end five irradiated steel

guide-tube nuts were retrieved from

the storage pool of NPP Gösgen ­during

the annual maintenance work in

2012 and transferred to the PSI

hotlaboratory. The nuts had been

positioned at the bottom end of

fuel rods and exposed to a thermal

neutron flux for ~2 years. Each nut

weighed ~5 g and had a contact dose

Decommissioning and Waste Management

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


atw Vol. 63 (2018) | Issue 1 ı January

rate ~150 mSv/h (predominantly

caused by 60 Co). When planning

the corrosion experiment several

constraints had to be taken into

consideration: The low 14 C inventory

of the activated steel samples

(~18 kBq/g) [10] in combination

with the fact that only a small amount

of activated steel could be used in a

corrosion experiment outside a hot

cell due to the high dose rate of the

material and the very slow corrosion

of stainless steel in alkaline conditions

(typically


atw Vol. 63 (2018) | Issue 1 ı January

DECOMMISSIONING AND WASTE MANAGEMENT 36

A B C

| | Fig. 3.

A) Sketch of the reactor, B) picture of the lead shielding with door and

C) the sampling system for liquid and gaseous samples placed outside the lead shielding.

a naturally occurring radionuclide

produced in the upper atmosphere

and present in the chemical form

14 CO 2 (activity of 1 m 3 air ~53 mBq).

Furthermore, alkaline solutions are

commonly known as a sink for CO 2

and therefore for 14 CO 2 . Hence, the

14 C background concentration accumulated

in the course of the corrosion

experiment with activated steel could

be affected by an undesirable uptake

of 14 C from the atmosphere in any

stage of sample preparation and

handling. The average 14 C background

was determined to be F 14 C =

0.06 ± 0.02 (F 14 C = fraction modern)

in samples collected after high performance

ion exchange chromatography

(HPIEC) and using pre-cleaned plastic

vials for injection and collection. This

value is about an order of a magnitude

higher than background values

achieved in radio carbon dating.

Sample preparation for compoundspecific

14 C AMS method involves

va rious dilution processes during

chromatographic separation of single

compounds that had to be considered

adequately in order to reach the target

dynamic range of the AMS (Figure 2).

The analytical protocol required

that dilution of the samples by a

factor 1:25 and 1:50 occurred in the

course of the separation by HPIEC.

Tests measurements carried out at

increasing concentrations of 14 C-

labelled carboxylic acids standards

allowed the dynamic range of the

AMS-based analytical method to be

determined (~0.06 - ~50 F 14 C).

Recovery of the compound-specific

14 C AMS method was determined

using four different 14 C-labelled

carboxylic acids ( 14 C-acetic acid,

14 C-formic acid, 14 C-malonic acid and

14 C-oxalic acid) dissolved in either

deionized, decarbonated water (ultrapure

water generated by Millipore

Gradient A10 water purification

­system) or in ACW (pH 12.5). The

samples were sequentially injected

into the HPIEC system as single compounds.

The corresponding fractions

of the 14 C-labelled carboxylic acids

were collected and analyzed by AMS

[11, 12]. Recoveries (%) were determined

using single compounds and

mixtures of the compounds. In all

cases recovery was found to be close

to 100 % (97 ±17 %) [12].

Corrosion studies with unirradiated

iron powders revealed that

volatile organic compounds, such as

alkanes, alkenes, alcohols, aldehydes,

are also formed during iron corrosion

[9] which requires the development of

a compound-specific 14 C AMS

analytical method for 14 C-bearing

v olatile species. The analytical

approach is currently being developed

in a way similar to that previously

elaborated for dissolved organic

compounds and is based on gas

chromatographic (GC) separation of

single compounds in combination

with 14 C detection by AMS. To this

end, the GC system has to be coupled

directly to a combustion reactor and a

fraction sampling system for 14 CO 2

(Figure 1). Coupling of the three

devices, i.e. GC, com-bustion reactor

and fraction collector, is still under

development.

2.4 Development of the

corrosion reactor

The experimental set-up for the longterm

corrosion experiment with the

activated steel nut specimens consists

of a custom-made gas-tight over pressure

reactor placed within a 10 cm

thick lead shielding (Figure 3). For the

experiments two activated steel nut

segments of ~1 g each were immersed

in 300 mL ACW (pH 12.5) under a N 2

atmosphere (200 mL). The reactor is

equipped with a digital pressure transmitter,

a temperature sensor and a

sensor to detect dissolved oxygen

( Visiferm DO Arc, Hamilton, USA).

The overpressure reactor is designed in

such a way that all mani pulations

necessary for regular sampling can be

carried out outside the lead shielding

to minimize exposure of the experimentalist

to radiation. Leak tests

­confirmed gas-­tightness of the reactor.

2.5 Start of the corrosion

experiment

The corrosion experiment with the

activated steel nut segments was

started in May 2016. Results from the

first few samplings are exemplarily

listed in Table 1. They show an

increase in the activity of total organic

14 C (TO 14 C) with time, thus indicating

progressing corrosion. At present,

how­ever, identification and quantification

of the individual 14 C-bearing

organic compounds by compoundspecific

14 C AMS is not yet possible

because their concentration is still

below the detection limit of the

­compound-specific 14 C AMS method.

As a consequence, the analytical

methodology is currently further

improved by developing a procedure

that allows pre-concentration of

the liquid samples collected by the

fraction collector.

Time TO 14 C TOC Hydrocarbons [µM] Carboxylic acids [µM]

[d] [F 14 C] [Bq/L] [ppm] Methane Ethane Ethene Foramte Acetate Oxalate Gycolate Lactate

0 0.00 0.00 - - - < 5 n.d. < 0.1 n.d. n.d.

1 0.10 0.04 - n.d. n.d. n.d. 7 n.d. 0.3 0.4 n.d.

15 0.99 0.45 2.44 n.d. n.d. n.d. 8 n.d. 0.5 1.3 1.6

29 1.56 0.70 2.60 n.d. n.d. n.d. 7 n.d. 0.5 1.4 1.2

93 3.53 1.60 4.67 0.42 n.d. n.d. 13 n.d. 0.7 1.7 2.8

| | Tab. 1.

Compilation of the first results from the corrosion study with activated steel.

Decommissioning and Waste Management

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


atw Vol. 63 (2018) | Issue 1 ı January

The total concentration of organic

carbon ( 12 C + 14 C), i.e. TOC, also

tends to increase with time. Note that

TOC accounts for the total concentration

of organic compounds, that

is, 12 C-bearing and 14 C-bearing compounds.

However, the concentration

of the 14 C-bearing compounds is

orders of magnitudes lower than that

of the corresponding 12 C-bearing

counter parts. The concentration of

the analysed 12 C-bearing individual

compounds, i.e. hydrocarbons and

carboxylic acids (Table 1), is still

below (n.d.) or close to the detection

limits of the analytical techniques

(GC-MS and HPIEC-MS, respectively).

Note that the latter analytical techniques,

i.e. GC-MS and HPIEC-MS,

can be used to detect separately

both 12 C-bearing and 14C-bearing

species of the same kind based on

their differences in the mass of carbon.

Again, the concentration of the 14 C-

bearing compounds is orders of

magnitudes lower than that of the

corresponding

12 C-bearing counterparts.

Thus, the concentrations of the

hydrocarbons and carboxylic acids

listed in Table 1 correspond to those

of the respective 12 C-bearing organic

compounds. The first results clearly

support the need of a very sensitive

AMS-based analytical method for

the detection of both volatile and

dissolved 14 C-bearing carbon species,

i.e. compound-specific 14 C AMS.

3 Summary

Our current understanding of the type

of 14 C-bearing species produced

during anoxic corrosion of activated

metals is very limited. This information,

however, is required in conjunction

with safety assessment of

nuclear waste repositories containing

activated metals (e.g. activated steel,

Zircaloy) as waste materials. A unique

corrosion experiment with activated

steel from NPP Gösgen, Switzerland,

is currently being carried out with the

aim of identifying and quantifying the

14 C-bearing carbon species produced

in the course of the corrosion process

under hyper-alkaline, anoxic conditions.

A specific analytical technique

was developed by combining chro matographic

separation of 14 C-bearing

individual compounds with 14 C detection

by AMS (compound-specific 14 C

AMS). This approach was chosen

because the concentrations of these

compounds was expected to be extremly

low due to low amount of

activated steel that could be used in

the experiment, the low corrosion rate

of steel in hyper-alkaline conditions

and the low 14 C inventory determined

for activated steel. The compoundspecific

14 C AMS method is characterized

by a low 14 C detection limit

and a large dynamic range (~3 orders

of a magnitude) and therefore it is

well suited for application in the corrosion

experiment with activated

steel. The method was developed for

selected, potentially 14 C-bearing compounds

of interest as previous studies

with unirradiated iron have shown

that only a limited number of carbon

species are formed during corrosion.

The specific set-up developed for

the corrosion experiment with activated

steel allows continuous monitoring

of important physico-chemical

parameters (pressure, temperature,

dissolved oxygen) and further allows

sampling of liquid and gas phase from

the reactor to be conducted outside

the lead shielding. Analysis of the

­liquid and gas phases from the first

sampling campaigns show that the

concentrations of the individual

organic compounds ( 12 C- and 14 C-

bearing) are still very low, i.e. below

or close to the detection limit of the

analytical methods used in this study.

Nevertheless, the total organic 14 C

content increases with time, indicating

progressing corrosion. This

increase in TO 14 C is slow in line with

the very slow corrosion of steel in

alkaline media. The analytical method

will be developed further to identify

and quantify the 14 C-bearing single

compounds in future samplings.

Acknowledgement

We thank NPP Gösgen for providing

the irradiated steel nuts and

Ines Günther- Leopold (PSI), Matthias

Martin (PSI) and Robin Grabherr (PSI)

for sample preparation. Partial

funding for this project was provided

by swissnuclear and the National

Cooperative for the Disposal of

Radioactive Waste (Nagra), Switzerland.

The project has received funding

from the European Union's European

Atomic Energy Community's ( Euratom)

Seventh Framework Programme FP7/

2007-2013 under grant agreement

no. 604779, the CAST project.

References

[1] L. Johnson and B. Schwyn, 2008.

Proceedings of a Nagra/RWMC workshop

on the release and transport of

C-14 in repository environments, Nagra

Working Report NAB 08-22, Nagra,

Wettingen, Switzerland.

[2] M.-S. Yim and F. Caron, 2006. Life cycle

and management of carbon-14 from

nuclear power generation, Prog. Nucl.

Energ. 48, 2-36.

[3] Nagra, 2002. Project Opalinus Clay:

Safety Report. Demonstration of

Disposal Feasibility for Spent fuel,

Vitrified High-level Waste and Longlived

Intermediate-level Waste

(Entsorgungsnachweis), Nagra

Technical Report NTB 02-05, Nagra,

Wettingen, Switzerland.

[4] Nuclear Decommissioning Authority,

2012. Geological Disposal. Carbon-14

Project - Phase 1 Report,

NDA/RWMD/092, United Kingdom.

[5] N.R. Smart et al., 2004. The Anaerobic

Corrosion of Carbon and Stainless Steel

in Simulated Cementitious Repository

Environments: A Summary Review of

Nirex Research. AEAT/ERRA-0313, AEA

Technology, Harwell, United Kingdom.

[6] N. Diomidis, 2014. Scientific Basis for

the Production of Gas due to Corrosion

in a Deep Geological Repository, Nagra

Working Report NAB 14-21, Nagra,

Wettingen, Switzerland.

[7] E. Wieland and W. Hummel, 2015.

Formation and stability of carbon-14

containing organic compounds in

alkaline iron-water-systems: Preliminary

assessment based on a literature survey

and thermodynamic modelling,

Mineral. Mag. 79, 1275-1286.

[8] D. B. Vance, 1996. Redox reactions in

remediation, Environ. Technol. 6, 24-25.

[9] B. Cvetković et al., 2017. Formation of

low molecular weight organic

compounds during anoxic corrosion of

zero-valent iron in alkaline conditions.

Environm. Eng. Sci. (accepted).

[10] D. Schumann et al., 2014.

Determination of the 14 C content in

activated steel components from a

neutron spallation source and a nuclear

power plant. Anal. Chem. 86,

5448-5454.

[11] S. Szidat et al., 2014. 14 C analysis and

sample preparation at the new Bern

laboratory for the analysis of

radiocarbon with AMS (LARA).

Radiocarbon 56, 561-566.

[12] B. Cvetković et al., 2017. Analysis of

carbon-14 containing corrosion

products released from activated steel

by accelerator mass spectrometry.

Analyst (in prep.).

Authors

E. Wieland

B.Z. Cvetković

D. Kunz

Paul Scherrer Institut

Laboratory for Waste Management

5232 Villigen PSI, Switzerland

G. Salazar

S. Szidat

University of Bern

Department of Chemistry and

Biochemistry & Oeschger Centre

for Climate Change Research

3012 Bern, Switzerland

DECOMMISSIONING AND WASTE MANAGEMENT 37

Decommissioning and Waste Management

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


atw Vol. 63 (2018) | Issue 1 ı January

38

FUEL

atw-Special „Eurosafe

2017“. In cooperation

with the EUROSAFE

2017 partners,

Bel V (Belgium),

CSN (Spain), CV REZ

(Czech Republic),

MTA EK (Hungary),

GRS (Germany), ANVS

(The Netherlands),

INRNE BAS (Bulgaria),

IRSN (France),

NRA (Japan),

JSI (Slovenia),

LEI (Lithuania),

PSI (Switzerland),

SSM (Sweden),

SEC NRS (Russia),

SSTC NRS (Ukraine),

VTT (Finland),

VUJE (Slovakia),

Wood (United

Kingdom).

Revised version

of a paper presented

at the Eurosafe,

Paris, France, 6 and

7 November 2017.

1) Reactivity control

is ensured notably

by the motion of

rod cluster control

assemblies

requiring not to

exceed a limited

fuel assembly

deformation.

2) Core coolability

requires not to

exceed a limited

deformation of the

fuel rods geometry.

3) Fission products

containment is

primarily ensured

by the first barrier

integrity.

4) M5 is the reference

alloy designed by

AREVA while ZIRLO

and Optimized

ZIRLO are Westinghouse’s

alloys (the

historical Zircaloy-4

cladding is no

longer loaded in

EDF’s reactors since

the end of 2016).

5) Zr + 2H 2 O → ZrO 2 +

2H 2

Review of Fuel Safety Criteria in France

Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois

1 Background Depending on design basis condition of Pressurized Water Reactors (PWRs), the safety

­objective is either preventing or mitigating the release of fission products and other contaminants to the environment.

Fuel is involved in each of the three reactor safety functions: reactivity control 1 , core coolability 2 and fission products

containment 3 . A main issue in the safety demonstration for the French PWRs is to respect the objectives related to the

barriers behavior, depending on Plant Condition Category (PCC) divided into four categories: normal operation

( PCC-1), incident transients (PCC-2), moderate frequency accident transients (PCC-3) and hypothetical accident

transients (PCC-4).

The objectives associated with the

first barrier are the following:

• for PCC-1 and PCC-2, the fuel rods

must remain intact;

• for PCC-3 and PCC-4, although

fuel rod integrity may be lost, the

number of damaged fuel rods

must be limited (in PCC-4, a more

extensive number of damaged

fuel rods is allowed than in PCC-3)

and the geometrical structure of

the core must not be damaged in

order to ensure an adequate core

coolability.

For each category and transients type,

these objectives are then expressed as

requirements associated to the limitative

physical phenomena occurring

during PCC. Afterwards, the requirements

are supported by fuel safety

criteria that are limit values on computable

metrics representative of the

relevant physical phenomena. These

limit values are determined by experiments

intended to be representative of

situations encountered in PCC.

In France, the fuel safety criteria

(and notably their limit values) came

in the 1970s from Westinghouse’s

license. At that time state-of-the-art

and computing capacities lead to

establish decoupling criteria enabling

to implement simplified and robust

approaches to analyze the more complex

and severe accidental conditions.

For instance, to maintain core coolability,

requirements may be based on

either fuel rod cladding integrity or

the absence of fuel dispersal in

the primary coolant. Indeed, such

requirements avoid notably studying

the impact of hot or melt fuel interaction

with water on core coolability.

Since the French nuclear program

was initiated, both operating experience,

experiments carried out by

operators and research institutes as

well as international R&D programs,

which aim at improving computation

methodologies, have allowed continuous

progress in knowledge and

in collecting experimental results,

especially in RIA (Reactivity-Initiated

accident) and LOCA (Loss-of-Coolant

Accident) conditions. Moreover, new

cladding alloys characterized by

enhanced performances, especially

regarding cladding corrosion during

operating conditions, have been

introduced in French PWRs (such as

M5, ZIRLO and Optimized ZIRLO 4 ).

Besides, although some operating

conditions have changed, notably

with strech-out operating conditions

and with the increase of maximum

allowed fuel burn-up, most of fuel

safety criteria have not been reviewed

since EDF’s Nuclear Power Plants

(NPPs) were designed, except those

concerning LOCA, which have

changed as a result of rulemaking

occured between 2008 and 2016 (see

Eurosafe 2016) and those concerning

Pellet-Cladding Interaction assisted

by Stress Corrosion Cracking (PCI-

SCC) in PCC-2 which have been introduced

since the 90’s.

In this context, the fuel safety

criteria were reviewed from 2011 to

2017 in order to assess, on the one

hand the sufficiency and validity of

current requirements and fuel safety

criteria relating to all fuel degradation

modes in the light of state-of-the-art

and operating conditions. The consistency

of the fuel rod behavior under

the reference PCCs with the assumptions

used in radiological consequences

studies was also assessed.

Thus, the review concerned the

following limitative physical phenomena:

• cladding embrittlement due to

corrosion. In PWRs, fuel rod

cladding in Zirconium alloy is

oxidized by the primary coolant 5 ,

which leads to the development of

an oxide layer at the clad outer

surface and to the absorption of a

portion of the hydrogen in the

cladding, leading to precipitated

hydrides. As a consequence, cladding

strength decreases [2, 1]. The

kinetics of oxidation depends on

clad temperature, which is about

350 °C in normal operations. If a

PCC-2 may lead to a rise in clad

temperature to a value in the range

of 450 to 480 °C, clad temperature

under PCC-3 and PCC-4 is higher

(> 700 °C) due to boiling crisis;

• clad failure due to Pellet-Cladding

Mechanical Interaction (PCMI)

and PCI-SCC. During transients

characterized by an increase of the

reactor power, the heating of fuel

pellets induces their thermal

­expansion and potentially fission-­

gas-induced fuel swelling, resulting

in a thermomechanical loading

(stress and strain) on the cladding

and potentially to clad failure.

Depending on power increase

during the transient and on

the level of clad embrittlement,

two clad failures types may be

observed. On the one hand, the

clad loading may be purely

mechanical (PCMI) under the

effect of the stress exerted by

pellets on clad. Hydride precipitation,

in particular in high burnup

fuel rods, plays an important role

in the incipient cracking initiated

at the cladding outer surface which

penetrates inwards, resulting in

though-wall cracking (with the

risk of fuel dispersal in the primary

coolant) [3, 4]. This phenomenon

is associated with power pulses

characterized by a rapid power

increase. On the other hand, in

conjunction with some corrosive

fission products, such as iodine,

expelled from pellets, the clad

loading may be assisted by SCC,

clad failure may be initiated at the

cladding inner surface leading to

clad perforation (without the risk

of fuel dispersal in the primary

coolant) [5] (see PCI workshop at

Luca in 2016). This phenomenon

is associated with power ramps

characterized by a lower power

rate than for pulses and followed

by an holding time at the ramp

terminal level;

• consequences of Departure from

Nucleate Boiling (DNB). Due to

boiling crisis occurrence, clad temperature

can increase suddenly,

reaching also high value (>700 °C)

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for several seconds. This may lead

either to clad ballooning up to

burst if the rod internal pressure

due to fission gas releases (during

normal operation and transient)

is higher than the external one

( especially for medium burnup

fuel rods), or to clad collapse on

the fuel pellets in the opposite

case. Moreover, the overheated

clads, embrittled by high temperature

oxidation, may lead to their

failure due to the application

of thermal stress during the

rewetting phase [6, 7, 8];

• consequences of fuel melting. In the

extreme case of an excessive temperature

rise of fuel rods due to a

major reactivity insertion or boiling

crisis, fuel rods may melt at least

partially (especially for fresh or

very low burn-up fuel). Indeed,

since the fissile content becomes

low at high burnup, the possibility

of pellet melting is very weak even

taking into account the reduction of

the melting point due to burn-up.

Fuel pellets melting generally leads

to clad fragmentation and clad

failure mode depends on fuel type

(UO 2 versus MOX) [6, 9, 10, 11].

Moreover, the current EDF’s NPPs

operating conditions which are

allowed must be taken into account

in the safety demonstration. Two

phenomena need to be dealt with:

• fuel assemblies may undergo bow

in PWRs due to hydraulic loads

exerted by the water, mechanical

loads applied by the top nozzle,

irradiation and temperature. The

design of fuel assemblies, particularly

the thickness and material of

the guide thimble, their position in

the core, as well as the duration of

their irra diation, also play a role in

the assembly bow. The magnitude

of the bow measured during refueling

outages for some PWRs 6 is

in the order of a few millimetres

and can be as much as 20 mm in

case of excessive assembly bow

[12, 13]. This potentially has an

impact on the in-core power distribution

(at the pin scale) and on

the safety analyses supporting the

plant operations which rely on the

hypothesis of a uniform water gap

between fuel assemblies ;

• leaking fuel rods [14]. Even if

it is an infrequent event, in EDF’s

reactors, some fuel rods may lose

their integrity, for example as the

result of cladding wear due to the

vibration of a loose part 7 stuck in a

grid cell or due to design or manufacture

defects. The presence of a

| | Fig. 1.

Topics reviewed in the frame of French rulemaking on fuel safety criteria.

primary defect (original loss of fuel

rod integrity) allows water to enter

into rods, which frequently leads to

a fairly well explained physicochemical

mechanism linked to

steam oxidation at the inside cladding

surface, and to the occurrence

of a secondary defect. In this area,

which is typically located at about

two or three meters from the original

defect, the cladding becomes

very brittle and can fail inducing

a fuel dissemination in the reactor

coolant system, even in normal

operating conditions [15, 16]. The

impact of this dissemination is

taken into account by the radiochemical

specifications in the

­Operating Technical Specifications.

Due to some leaking fuel

rods in reactor, Rod Ejection Accident

(REA) may lead to sudden

fuel rods failures near the ejected

control rod and to the dispersal

of fuel pellets fragments in the

primary coolant, and thus to a violent

thermal interaction between

fuel pellets fragments and the

coolant. This interaction would

lead to a strong primary coolant

pressure increase and to a production

of a steam zone, which could

dry out the neighbouring rods

(near the ejected control rod) up to

their failure. In addition, the

primary coolant pressure would

propagate to neighbouring rods

and to the reactor vessel, potentially

damaging them.

In the French regulatory framework,

new fuel safety criteria are suggested

by the French utility EDF on request

of the French Nuclear Safety Autho rity

(ASN) and submitted to it for approval.

The safety assessment of EDF’s proposals

(based on test results, studies,

operating experience feedback, examinations

of irradiated fuel rods…)

is made by Institute of Radiological

Protection and Nuclear Safety (IRSN).

Based on IRSN’s technical assessment,

the Advisory Committee for

Reactors Safety of the Nuclear Safety

Authority (ASN) meeting about the

French rulemaking on fuel safety

criteria related to PCC-1, PCC-2,

PCC-3 and PCC-4 (except for LOCA)

was held in June 2017. The new

criteria are then assumed to be applied

for EDF’s French PWR (except for EPR)

and for claddings loaded in these

reactors (except for Zircaloy-4 which

is not used anymore in fresh fuel).

In this way, the paper describes the

main conclusions of IRSN’s assessment

about the evolutions of fuel

safety criteria for each PCC and each

limitative physical phenomena. The

following Figure 1 gives an overwiew

of French rulemaking.

2 Fuel safety criteria

before the french

rulemaking

2.1 In PCC-1 and PCC-2

At the reactor design stage, two

requirements associated with physical

phenomena likely to affect the fuel

rod integrity were used to design

reactor protection systems: the

­absence of DNB and the absence of

6) The fuel assembly

bow is not measurable

in core but out

of core during

refueling outages

for some EDF PWRs.

7) A loose part is a

fragment, usually

metal and very

small (less than

three millimetres),

which has generally

come off a larger

part during operating,

e.g. when

fuel assemblies are

being handled.

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8) The CFHR criterion

adopted in the

French safety

demonstration

results from the

interpretation of

critical flux tests

performed for a

given fuel assembly.

For this reason,

the CHFR criterion

is likely to undergo

changes in case of

modification to the

fuel materials and

design.

9) A RIA is caused by

a control REA,

which is defined as

the mechanical

failure of a Rod

Cluster Control Assembly

(RCCA)

drive mechanism

casing, located on

top of the reactor

pressure vessel

which is ejected

vertically from the

reactor core due to

the high coolant

pressure. Such a

RIA is characterized

by a very rapid increase

of reactivity

and power in some

rods of the reactor.

10) EDF’s safety

domain for REA:

Oxide thickness,

enthalpy variation,

pulse width, clad

temperature.

11) ECR : Equivalent

Cladding Reacted.

12) Expansion due to

compression using

various PWR cladding

alloys and

performed at

350°C and 10 -4 s -1 .

13) Uni-axial tensile

tests using

transverse samples

and carried out

from 280°C to

400°C at 10 -2 s -1 .

fuel melting. In the 1990s, the absence

of clad failure due to PCI-SCC was

added. Criteria were thus defined:

• in order to avoid DNB, the Critical

Heat Flux Ratio (CHFR) must

remain above a critical value

d epending on the fuel assembly 8 ;

• in order to avoid fuel melting,

the maximum Linear Power

Density (LPD) must remain below

590 W/cm;

• in order to avoid clad failure due to

PCI-SCC, some thermo- mechanical

limits must be verified.

In addition, fuel rod design criteria

were used to check that fuel rods

behave correctly during transients as

regards to:

• cladding corrosion. In PCC-1, oxide

thickness shall not exceed 100 µm.

In PCC-2, clad temperature at the

interface between the metal and

the oxide shall not exceed 425 °C;

• PCMI. In PCC-1 and PCC-2, the

circumferential clad strain shall

not exceed 1 %.

2.2 In PCC-3 and PCC-4

In France, at the start of the industrial

exploitation of NPPs, specific requirements

and empirical criteria were

­defined to demonstrate core coolability,

especially for Rod Ejection

Accident (REA) 9 :

• to ensure that there is no hot or

molten fuel dispersal in the

primary coolant during REA, the

maximum fuel enthalpy is limited

to 200 cal/g, the limit coming from

Westinghouse’s extrapolation of

fuel behavior established on the

basis of RIA full-scale SPERT-CDC

tests carried out at zero-power on

fresh and very low irradiated UO 2

fuel. This criterion is applicable for

mean fuel assembly burn-up up to

33 GWd/tU;

• regarding PCMI, the progressive

increase of fuel assembly discharge

burn-up led ASN to ask EDF to

demonstrate that the previous

criteria were still applicable for

REA. Thus, some full-scale tests

carried out in the French CABRI

test reactor and in the Japanese

NSRR test reactor using high

burn-up fuel rods led to fuel

dispersal in the primary coolant for

fuel enthalpy far below 200 cal/g.

These tests clearly showed that this

criterion was no longer relevant.

Based on the results of full-scale

tests, EDF established an empirical

safety domain defined by four

parameters 10

which intends precluding

PCMI clad failure and

burst during boiling crisis for high

mean fuel assembly burn-up

(> 47 GWd/tU);

• the maximum peak clad temperature

must remain below 1,482 °C

(2,700 °F). This limit was taken

from fuel failure boundary for

LOCA conditions. The rational for

retaining a higher temperature

limit for non-LOCA transients,

such as REA, is that film boiling

­occurs briefly during those

transients, so that fuel rods could

withstand this brief dry-out

without suffering serious damage.

In addition, the number of fuel rod

failures must be calculated so that the

radiological doses to the public can be

estimated. A requirement is defined to

limit the number of fuel rods affected

by DNB. The conservative assumption

is that all fuel rods entering into

boiling crisis are assumed to fail.

Thus, the percentage of fuel rods

­likely to suffer DNB is limited to 5 % in

PCC-3 and to 10 % in PCC-4. Besides,

all fuel rods that experience fuel

melting, especially for REA, are

assumed to be failed for radiological

doses calculations. Nevertheless, only

a limited amount of fuel melting is

accepted, less than 10 % of pellet

volume.

3 Evolution of fuel safety

criteria

3.1 Clad embrittlement

due to corrosion

During operating conditions, it is no

longer necessary, for cladding alloys

loaded in EDF’s reactors (M5, and

Optimized ZIRLO), to verify the oxide

thickness criterion limited to 100 µm

because of their improved corrosion

resistance. However, as in-reactor

hydrogen content has a major impact

on clad behavior under PCMI during

incidental and accidental conditions,

the validity of the various criteria

ensuring clad non-failure under PCMI

conditions relies on compliance with

limits of hydrogen content (see

§ 4.2.1).

During incidental conditions, the

absence of corrosion acceleration is

not likely to occur for cladding alloys

loaded in EDF’s reactors because of

their corrosion resistance and the

temperatures likely to be reached

­during PCC-2. Verification that the

clad temperature at the interface

between the metal and the oxide

­remains below 425°C is therefore no

longer necessary.

In accidental conditions, the

current clad temperature criterion

limited to 1482°C does not take into

account the time spent at high

temperature during boiling crisis,

even though cladding oxidation rate

is dependent on this. By analysing

experimental results available in the

literature, EDF plans to complete this

criterion by defining a new oxidation

rate (ECR 11 ) limit, which is expressed

as a function of maximum clad

­temperature and based on DNB tests

carried out in PBF reactor [7]. IRSN

considered that, although this

approach is acceptable, EDF hasn’t

taken into account all physical

phenomena that are likely to induce

clad embrittlement nor measurement

uncertainties to define the ECR limit.

EDF will complete its approach and

review this new criterion.

3.2 Clad failure due to PCMI

and PCI-SCC

3.2.1 PCMI clad failure

For PCC-2 power ramps likely to

induce PCMI clad failure, the clad

strain limit of 2 % is raised instead of

1 % until the in-reactor hydrogen

­content is below 250 ppm, based on

representative analytical tests 12 . In

addition, the uncontrolled with drawal

of control rod assembly bank(s) at

zero power is a particular PCC-2

transient leading to a rapid power

excursion, which may also induce

PCMI clad failure. Up to now, no

criterion was established for this

­transient. That is why, a specific limit

of 1 % of plastic clad strain has been

defined to ensure clad non-failure until

the in-reactor hydrogen content is

below 805 ppm. This criterion is based

on appropriate analytical tests 13 . IRSN

concludes that these evolutions, based

on a cautious interpretation of tests

results, are acceptable.

No requirement and fuel safety

criterion ensuring core coolability

were defined for mean fuel assembly

burn-up between 33 and 47 GWd/tU

in REA transients. Moreover, SPERT,

CABRI and NSRR tests were carried

out at zero-power while French safety

demonstration requires REA studies

for all initial power levels. That is why,

EDF has revised existing criteria and

completed the safety demonstration

for fuel assembly burn-up higher than

33 GWd/tU. The new acceptance

criteria, expressed by enthalpy rise

and pulse width, aim at precluding

PCMI clad failure. Their limits depend

on cladding corrosion performances,

more specifically on in-reactor hydrogen

content which is of interest to

cope with PCMI behavior. More precisely,

EDF’s approach to define the

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new REA criteria depends on the fuel

rods types:

• for UO 2 fuel rods with ZIRLO, Optimized

ZIRLO and M5 claddings,

the approach has been based on

the interpretation with SCANAIR

code [17] of some full-scale RIA

tests carried out in CABRI and

­NSRR reactors and associated with

PCMI issue. But, the threshold

values of enthalpy rise and pulse

width are different for M5 than for

ZIRLO and Optimized ZIRLO due

to specific cladding corrosion performances.

Regarding M5, IRSN

considers acceptable the 150 cal/g

of enthalpy rise criterion (the pulse

width limit definition being in

progress and the hydrogen content

limit is 160 ppm). However, concerning

ZIRLO and Optimized

­ZIRLO, IRSN identifies that no

uncertainty about experimental

data has been taken into account

by EDF to calculate the enthalpy

rise limit from the restrictive test,

CABRI CIP0-1 14 , which will lead

EDF to review the definition of the

associated criterion;

• for MOX fuel rods with M5 cladding,

EDF has used SCANAIR code

to reproduce PCMI behavior for

MOX fuel based on a specific RIA

test carried out on UO 2 fuel and

­related to ballooning. IRSN considers

that the approach is complicated

and unsupported. Eventually,

EDF plans to define fuel

safety criteria for MOX fuel rods

with M5 on the basis of the analysis

of specific integral RIA tests devoted

to MOX, as it has been done

for UO 2 fuel rods.

For REA initiated at non-zero power

levels, EDF has developed an approach

which aims at demonstrating that the

REA initiated at zero-power is the

most limiting compared to transients

initiated at higher power levels. IRSN

estimates that EDF’s approach, based

on the comparison of thermo- mechanical

parameters calculated with

­SCANAIR code for the PCMI behavior,

is acceptable. EDF will apply this

­approach for each NPPs series.

As in-reactor hydrogen content

plays an important role in the definition

of criteria related to PCMI,

IRSN will assess EDF’s correlations

giving hydrogen content as a function

of oxide thickness.

3.2.2 PCI-SCC clad failure

The risk of PCI-SCC clad failure is

currently taken into account in PCC-2

studies for which fuel rods integrity

must be demonstrated. However,

some PCC-3 or PCC-4 transients lead

to PCI-SCC. If the corresponding clad

failure mode is not likely to lead to a

loss of core coolability, the risk still

needs to be assessed for PCC-3 and

PCC-4 transients in order to ensure

that the radiological consequences of

the concerned accidents are conservatively

assessed. Thus, EDF has

developed an approach to verify the

absence of any risk of clad failure in

case of Uncontrolled Control Rod

Withdrawal accident at non-zero

­power level (PCC-3). IRSN considers

this approach to be acceptable.

Another transient, the Steam Line

Break accident initiated at non-zero

power level (PCC-4) is also likely to

lead to PCI-SCC clad failure. EDF

has provided justification concerning

some reactors concluding that the

PCI-SCC clad failure risk is no greater

than for PCC-2 transients. For IRSN,

the justification still needs to be confirmed

and extended to all reactors.

3.3 Consequences of DNB

In order to demonstrate the absence

of fuel dispersal in the primary coolant

after clads ballooning and burst

during boiling crisis, EDF has proposed

two approaches depending on

transients:

• for REA, the approach is based

on the comparison between the

restrictive PCMI criterion and

results of various full-scale tests

associated with ballooning and

burst (IGR, BIGR, NSRR, PBF – [18,

19, 20]). In the available experimental

database, no fuel dispersal

is observed up to EDF fuel rods

burn-up discharge limit (57 GWd/

tU) and up to the enthalpy rise

­limit of 150 cal/g (see § 4.2.1);

• for Uncontrolled Control Rod

With drawal at non-zero power

level (PCC-3) and Locked Rotor

(PCC-4) accidents, EDF has compared

the maximum fuel rod

burn-up calculated beyond which

boiling crisis is avoided and the

current non-dispersal threshold 15 .

However, as the absence of fuel

dispersal has been demonstrated

with a very small margin, IRSN

considers that EDF will have to

update its safety demonstration for

each ten-yearly outage review or in

case of modifications deemed to

impact this conclusion.

Besides, questionning the conservative

assumption is that all fuel

rods entering into boiling crisis are

assumed to fail, EDF foresees to limit

(up to 5 % for PCC-3 or 10 % for

PCC-4) the number of broken rods

due to ballooning during boiling

crisis. From EDF’s point of view, the

current criterion related to radiological

doses calculations is based on a

very conservative assumption considering

that all fuel rod entering into

boiling crisis is supposed to be failed

[21]. By applying a fuel rod burn-up

threshold calculated with SCANAIR

code [17] depending on fuel rod

design and irradiation, some fraction

of fuel rods can be excluded from the

counting of failed rods. IRSN considers

acceptable this method. However,

in case of plant operating conditions

modifications (for the future), EDF’s

evolution could lead to increase

radiological consequences, which is

not acceptable for IRSN.

Finally, regarding on-going RIA

investigations and research programs,

IRSN considers namely that Cabri

International Project (CIP 16 ) tests

planned in the CABRI-water loop

facility may be used to analyse clad

behavior during boiling crisis notably

for high fuel burn-up and will improve

knowledge on the MOX fuel behavior.

3.4 Consequences

of fuel melting

In the current safety demonstration,

no requirement associated with fuel

safety criterion was defined concerning

fuel melting risk during PCC-3. In

order to adress this gap, EDF plans to

verify the limit of 10 % molten fuel at

the pellet centre for the Uncontrolled

Control Rod Withdrawal accident

initiated at non-zero power level. For

IRSN, this evolution is acceptable, but

the radiological doses calculations

related to this transient will have to be

assessed consistently with the new

criterion.

Moreover, like the NRC’s requirement,

a limited amount of fuel melting

is acceptable provided it is restricted

to the fuel centerline region and is

less than 10% of pellet volume [22].

Indeed, during REA (PCC-4), due to

the effects of edge peaked power and

lower solidus temperature, fuel rods

may undergo fuel melting in the pellet

periphery. Thus, fuel melting outside

the centerline region is precluded to

avoid molten fuel coolant interaction.

Therefore, EDF will demonstrate that

this requirement is satisfied based on

appropriate analysis rules.

Besides, with regard to the

200 cal/g of maximum fuel enthalpy

criterion for REA (applied to fuel

assemblies with burn-ups up to

33 GWd/tU), EDF confirmed its validity

for MOX fuel on the basis of the

CABRI REP-Na9 test 17 . However, IRSN

14) For CIP0-1, the

measured

hydrogen content

is 1000 ppm.

15) Established at

55,2 GWd/tU in

mean fuel rod

burn-up, based on

Halden and

Studsvik LOCA

tests.

16) CABRI CIP: Tests

with water coolant

loop plan to start

in 2018.

17) CABRI REP-Na9

was carried out on

MOX fuel with a

low clad corrosion

and a fuel rod

burn-up of 28

GWd/tU. The

tested fuel rod was

not failed for a

maximum fuel

enthalpy of

200 cal/g.

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underlines that further tests on MOX

fuel would improve knowledge about

its sensibility as regard to fuel melting,

especially for high burn-up and high

plutonium levels characteristics of

MOX fuel loaded in EDF’s reactors.

4 Taking into account

rod failures and

assembly bow

4.1 Impact of fuel assembly

bowing on safety

demonstration

Fuel assemblies distort in-core and the

gap between fuel assemblies can vary

away from the design value. Using

several ex-core fuel assembly bow

measurements (from different reactors

and cycles), EDF has developed

a mechanical model to estimate interassembly

gap size distributions in

cores. IRSN assessed assumptions and

considered that the predictive model

is consistent with the current state- ofthe-art.

In addition to slower drop

times of RCCA due to friction in guide

tubes, fuel assembly distortion potentially

leads to neutronic, thermohydraulic

and mechanical impacts on

safety demonstration:

• the presence of larger inter-assembly

gaps causes a local variation in

the fuel-to-moderator ratio and

hence the local neutron moderation.

Basically, as fuel assemblies

move apart, the concentration of

thermal neutrons in the gap increases

and so does the power in

peripheral pins. EDF has developed

a new methodology for quantifying

and taking into account this effect

in the safety demonstration. IRSN

estimates this methodology satisfactory;

• for the same reasons, the Critical

Heat Flux Ratio (CHFR) decreases

at periphery, but also the hydraulic

diameter of the corresponding flow

channel increases, so does the

CHFR. Because of these antagonistic

effects, the flow channel in

which the minimum CHFR is

reached could become a peripheral

one (instead of a channel within

the fuel assembly in the current

safety demonstration). In such

flow channel, grid straps do not

have mixing vanes, which significantly

reduces CHFR. For EDF,

the minimum CHFR remains in the

center of the fuel assembly. However,

to evaluate the global effect,

EDF realized sensitivity studies

notably using unappropriate CHF

correlations. Because of the large

number of justifications still to be

provided, IRSN can’t conclude on

EDF evaluation;

• the presence of smaller water gaps,

and particularly the existence of

contact between grids, is likely to

increase the maximum impact

forces on fuel assembly grids under

seismic and LOCA loads. EDF not

yet assessed the effect of variable

inter-assembly gaps, repre sentative

of the in-reactor situation, on the

assembly grids buckling risk. In

addition, for IRSN, the validation

of EDF’s model to calculate the

impact force on grids during

accidentel conditions needs to be

completed, particularly because

it doesn’t include a comparison

with sufficiently representative

tests results. Thus, the safety

demonstration will be updated.

4.2 Leaking fuel rods

during normal operating

conditions

The behavior of defective fuel rods,

especially under REA, is an important

aspect of safe reactor operation, since

some EDF’s reactors (7 out of the 58

operating reactors currently) contain

a very small percentage of leaking fuel

rods (only 0,11% leaking fuel assemblies).

This issue has been assessed for

several years. The complexity of the

physical phenomena to be taken into

account and the lack of available

experimental data on waterlogged

fuel rods under this transient explain

the difficulty to conclude on the

potential unwanted effects: surrounding

fuel rods failures due to

mechanical and thermal effects or

even potential vessel damage [23,

24]. IRSN considered that EDF’s

demonstration takes into account

satisfactorily the state-of-the-art.

Finally, the large pressure pulse does

not lead to additional fuel rods failures

nor to vessel damage. However, for

IRSN, EDF should still justify that the

models used for assessing thermal

interaction and its consequences are

appropriate.

Considering other PCC-2 and

PCC-4 transients, IRSN estimates that

it is likely that in many cases, application

of stress would lead to the fuel

rods failure in the secondary defect

area and to fuel dispersal in the

primary coolant. However, these

phenomena are unlikely to affect the

core coolability or to have any significant

impact on the the radiological

doses calculations, except for steam

generator tube rupture accidents.

Indeed, these transients are characterized

by a break in the second

barrier, containment bypass and the

possibility that some contaminated

reactor coolant will be released into

the environment. EDF will study

the potential consequences of this

scenario.

References

1. A.M. Garde et al., Hydrogen Pick-Up

Fraction for ZIRLO Cladding Corrosion

and Resulting Impact on the Cladding

Integrity, Proceedings of Top Fuel 2009

Paris, France, September 6-10 (2009)

2. S. K. Yagnik, R-C Kuo, Y.R. Rashid et al.,

Effect of hydrides on the mechanical

properties of Zircaloy-4, Proceedings of

the 2004 International Meeting on

LWR Fuel Performance, Orlando,

Florida, September (2004)

3. R.L. Yang, R.O. Montgomery,

N. Waeckel, EPRI TR #1002865, Topical

report on reactivity-initiated accident:

bases for RIA fuel and core coolability

criteria (2002)

4. T. Sugiyama, High burnup fuel behavior

under high temperature RIA conditions,

FSRM 2010, Tokai, Japan, May (2010)

5. B. Julien et al., Performance of

advanced fuel product under PCI

conditions, Proceedings of the 2004

International Meeting on LWR Fuel

Performance, Orlando, Florida,

September 19-22 (2004)

6. P.E. Macdonald, W.J. Quapp et al.,

Response of unirratiated and irratiated

PWR fuel rods tested under Powercooling

mismatch conditions, Nuclear

Safety, vol.19, n°4, (1978)

7. F. M. Haggag, Zircaloy-cladding

embrittlement criteria : comparison of

in-pile and out-of-pile results, NUREG/

CR-2757 (1982)

8. T. Fuketa, Transient response of LWR

fuels (RIA), Compr. Nucl. Mater.

579-593 (2012)

9. W.G. Lussie, The response of mixed

oxide fuel rods to power bursts,

IN-ITR-114, Idaho Nuclear Corporation

(1970)

10. W.G. Lussie, The response of UO2 fuel

rods to power bursts, IN-ITR-112, Idaho

Nuclear Corporation (1970)

11. M.D. Freshley, Behavior of discret

plutonium dioxide particles in mixedoxide

fuel during rapid power transient,

Nuclear technology, Vol.15 (1972)

12. N. Waeckel, Fuel Assembly distortion in

EDF NPPs, Oral communication on

OECD WGFS, Paris (2014)

13. C. Durand, Fuel bowing performances,

EDF oral communication at OECD NEA

Workshop Advanced fuel modelling for

safety and performance enhancement

(2017)

14. Report OECD NEA/CSNI/R(2014)10,

Leaking Fuel Impacts and Practices

(2014)

15. Y. KIM, S. KIM, Kinetic studies on

massive hydriding of commercial

zirconium alloy tubing, Journal of

nuclear materials, 270, pp. 147-153

(1999)

Fuel

Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois


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atw Vol. 63 (2018) | Issue 1 ı January

44

AMNT 2017

16. D. H. Locke, The behavior of defective

reactor fuel, Nuclear Engineering and

Design (1972)

17. A. Moal, V. Georgenthum,

O. Marchand, SCANAIR: A transient fuel

performance code Part One: General

modelling description, Nuclear

Engineering and Design, Vol. 280,

pp. 150-171 (2014)

18. NUREG/IA-0213, Experimental study of

narrow pulse effects on the behavior of

high burn-up fuel rods with Zr 1 % Nb

cladding and UO2 fuel (VVER type)

under reactivity-initiated accident

conditions: program approach and

analysis of results, (2006)

19. M. Ishikawa, A study of fuel behavior

under reactivity initiated accident

conditions – Review, Journal of nuclear

materials, Vol. 95, pp. 1-30 (1980)

20. OCDE NEA Report N°6847, Nuclear fuel

behavior under Reactivity-initiated

accident (RIA) conditions (2010)

21. C. Bernaudat, J. Guion, N. Waeckel,

IAEA Technical meeting on fuel behaviour

and modelling under severe transient

and LOCA conditions, Mito

(Japon) (2011)

22. Draft regulatory guide DG-1327,

Pressurized water reactor control rod

ejection and boiling water reactor

control rod drop accidents, U.S, Nuclear

Regulatory Commission (NRC),

Washington DC (2016)

23. S. Tanzawa and T. Fujishiro, Effects of

waterlogged fuel rod rupture on

adjacent fuel rods and channel box

under RIA conditions, Nucl. Sci. and

Tech., 24(1):23-32 (1987)

24. T. Sugiyama and T. Fuketa, Mechanical

energy generation during high burnup

fuel failure under reactivity initiated

accident conditions, Nucl. Sci. and Tech.,

37(10):877-886 (2000)

Authors

Sandrine Boutin

Stephanie Graff

Aude Foucher-Taisne

Olivier Dubois

Institut de radioprotection

et de sûreté nucléaire

B.P. 17

92262 Fontenay-aux-Roses,

France

48 th Annual Meeting on Nuclear Technology (AMNT 2017)

Key Topic | Outstanding Know-How &

Sustainable Innovations

Technical Session: Reactor Physics,

Thermo and Fluid Dynamics

Neutron Flux Oscillations Phenomena

Joachim Herb

The Technical Session about Neutron Flux Oscillation Phenomena was chaired by Joachim Herb (Gesellschaft für

Anlagen und Reaktorsicherheit (GRS) GmbH) and well attended by approx. 50 listeners. It comprised of three keynotes

and two technical presentations. The main topics were the significant changes of the neutron flux noise levels in

different German and foreign pressurized water reactors (PWRs). For about ten years an increase in neutron noise

­levels has been observed in German PWRs. During the following five years the noise levels have been decreasing again.

In principle, a correlation of the neutron noise levels to the use of certain fuel element types was observed and the

­phenomenon of neutron flux oscillations had been known since decades. Nevertheless, no self-consistent physical

­theory exists so far, which can explain the observed changes and the absolute levels of the observed neutron flux noise

levels. Therefore, safety authorities, technical support organizations (TSO), utilities as well as research organizations

showed increased interest in this topic during the last years. The results of the corresponding work as well as an outlook

into soon-starting research projects were given in this session.

The first keynote of the session about

Neutron Flux Oscillations in PWR:

Safety Relevance was presented

by Kai-Martin Haendel (TÜV Nord

EnSys GmbH & Co. KG, Germany).

Mr. Haendel reported that the source

of the low frequency neutron flux

noise (< 2 Hz) had unexpectedly

changed which led to sporadic erroneous

activations of surveillance

signals (rod drop, reactor power

limitation) in the reactor limitation

system despite the existing filtering

of the neutron flux signal. A review of

the limitation and protection systems

was necessary to demonstrate that

safety functions were not compromised

by the higher levels of neutron

noise and that the actions of the

limitation system comply with the

given safety criteria, i.e. the safetyrelated

parameters adhere to all safety

limits under all design accidental

conditions. For the purpose of the

rod drop detection and the short-time

corrected thermal reactor power it

was shown that, as long as the delay

time of the filters stayed below certain

limits, all safety key parameters were

met. A reduction of the reactor power

results also in a decrease of the

neutron noise level and hence in the

absence of any erroneous activation of

the rod drop signal and a strongly

reduced occurrence of erroneous activations

of the reactor power signal.

Marcus Seidl (PreussenElektra

GmbH, Germany) presented the second

keynote with the title Neutron Flux

Oscillations in PWR: Operational

Experience. While neutron noise so

far has mainly been explained empirically

the existing theoretical frameworks

are unable to describe all its

observed properties in Konvoi and

Vor-Konvoi reactors in a consistent

manner. This is likely due to the fact

that a suitable (and not jet existing)

AMNT 2017

Technical Session: Neutron Flux Oscillations Phenomena ı Joachim Herb


atw Vol. 63 (2018) | Issue 1 ı January

theory needs to couple the neutronics,

thermal-hydraulics and mechanical

properties of the core. The goal is

­difficult to achieve because on the one

hand almost no suitable coupling

schemes exist at the moment for this

purpose. It is also difficult on the other

hand, because neutron noise mostly

has been treated as an unwanted signal

(besides being used for some recurrent

component oscillation checks) in commercial

power reactors in the sense

that is has been suitably filtered or

reduced by means of power reductions.

As other fields of science have

shown in the past the analysis of noisy

signals can often lead to better instruments

and in turn to the detection

of hitherto unrecognized phenomena.

So the main motivation to continue

current efforts to consistently explain

the neutron noise signals is to get a

better understanding of the mechanical

and thermal-hydraulic behaviour

of fuel assemblies under operating

conditions. For this purpose, it might

be necessary to corroborate upcoming

theoretical explanations by means of

better in-core temperature and mechanical

oscillation measurements. In

practice this can pay off in an improvement

of reactor performance by

leading to better fuel assembly designs

and an improved thermal margin

determination.

The third keynote about Neutron

Flux Oscillations in PWR: Clarification

of Possible Causes was

presented by Christophe Demazière

(Chalmers University of Technology,

Sweden). He gave a brief account of

the capabilities of core monitoring

using noise analysis, including a

historic overview starting 1949 with

early development in noise analysis at

the Clinton Pile at ORNL, USA, later

works on the detection of excessive

vibrations of control rods, core-barrel

vibrations, estimations of in-core

coolant velocities, detector tube

impacting and the analyses of BWR

instabilities. To generalize the use of

noise analysis, it is necessary to invert

the reactor transfer function, which

describes the effect of local disturbances

on the measured neutron flux

noise. Then Christophe Demazière

introduced the Horizon 2020 EUproject

CORTEX (CORe monitoring

Techniques and EXperimental validation

and demonstration) which was

expected to start on September 1 st ,

2017. The project aims are the

­development of high fidelity tools

for simulating stationary fluctuations,

the validation of those tools against

experiments to be performed at

research reactors, the development

of advanced signal processing techniques

(to be combined with the simulation

tools), the demonstration of the

proposed methods for both on-line

and off-line core diagnostics and

monitoring and the dissemination of

the knowledge gathered from within

the project to stakeholders in the

nuclear sector. The project will be led

and coordinated by Chalmers University

of Technology. 17 European

organizations (from eight countries)

and two non-European organizations

will be involved in the project.

Additionally, there will be an Advisory

End-User Group for the project.

Gaëtan Girardin (Kernkraftwerk

Gösgen-Däniken AG, Switzerland)

sum marized the recent investigation

on Neutron Flux Oscillation Phenomena

at Kernkraftwerk Gösgen

(KKG), which is a 3-loop pre-KONVOI

type PWR. It was observed that the

global amplitudes of the power oscillations

had slowly and monotonously

increased during the last seven operating

cycles. Moreover, no modification

of importance had been done on

the primary circuit and the reactor

core over the last years that could

possibly explain the amplitude increase

of the neutron noise. In order

to determine the possible reason

of the neutron noise increase, the

­already existing neutron flux measurements

were completed during the last

cycle by two extensive measurement

campaigns: one mid of cycle and the

second one end of cycle. Based on

these new measurements, it was

­obtained and confirmed that the

largest noise amplitudes are located in

one quadrant of the core between

Loop 1 and 3, and the simultaneous

measurements revealed that the noise

signals at two opposite sides of the

core had strong negative correlations.

Moreover, no time shifts were found

in the axial measurements between

the top and bottom neutron signals. It

was also found that the highest amplitudes

had not increased over last cycle

compared to previous increase in the

previous cycles. The observed saturation

of the noise amplitudes at quite

high amplitudes were correlated to a

core fully loaded with HTP design

fuel assemblies. The ex-core filters

were calibrated in a way so that few

activations of the power limitation

system were observed. It was also

observed that there existed a relationship

between fuel assembly bowing

and noise amplitudes. Based on the

analyses a stabilization of neutron

noise amplitudes was expected.

The final presentation was given

by Joachim Herb (Gesellschaft für

Anlagen- und Reaktorsicherheit, (GRS)

gGmbH, Germany) about the Analyses

of Possible Explanations for the

Neutron Flux Fluctuations in

German PWR. He reported, that no

comprehensive theory existed yet

which could explain the observed

­neutron flux fluctuation levels based

on first physical principles. Therefore,

GRS has started investigations on

which combination of thermal hy draulics,

structural mechanics and neutron

physics models were able to explain

the observed neutron flux fluctuation

and the change in the observed levels.

The analyses based on the evaluation

of measurements in German PWRs.

Using simple models, parts of the

observations could be explained: A

basic coupled thermal hydraulics/

point neutron kinetics model could

reproduce the shape of the neutron

flux noise spectrum as well as the

linear dependency between the noise

level and the moderator temperature

coefficient, but it could not explain

the spatial correlations between the

signals of different detectors. A point

source model for the neutron flux was

used to consistently explain the observations

at the different neutron flux

detector locations, but it could not

explain the shape of the noise spectrum.

A model based on the modification

of the cross sections of the

neutron reflector was able to produce

flux changes of about 10 %, but it had

to be shown what could cause the

assumed changes of the cross sections.

Also, different mechanical explanations

were discussed based on the assumption

of core-wide motions of fuel

assemblies and further core internals.

These motions might be produced by

excitations at the natural frequency,

forced excitations and/or self-excitation

due to fluid-structure interaction

with the coolant. Overall, it was concluded

that the phenomena is very

likely caused by a combination of

different physical effects which

requires further work on the combination

of different physical models

and coupled simulations.

Author

Joachim Herb

Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH

Abteilung Kühlkreislauf /

Cooling Circuit Department

Bereich Reaktorsicherheitsforschung

/ Reactor Safety

Research Division

Forschungszentrum

Boltzmannstraße 14

85748 Garching b. München,

Germany

45

AMNT 2017

AMNT 2017

Technical Session: Neutron Flux Oscillations Phenomena ı Joachim Herb


atw Vol. 63 (2018) | Issue 1 ı January

46

AMNT 2017

Key Topic | Enhanced Safety & Operation

Excellence

Focus Session: Radiation Protection

Erik Baumann and Angelika Bohnstedt

The objectives of radiation protection are to minimize the negative health effects due to radiation. Over many past

decades, the regulatory environment, i.e. the various international and national codes and standards but also

recommendations issued by IAEA and IRCP, was always subject to continuous development reflecting up to date

knowledge and experience. Currently, discussions focus on “new” areas of human activities like decommissioning and

on “new radioactive substances” and potential threats associated with their handling (handling and treatment of

substances containing naturally occurring radioactive materials). Latter already became subject to regulations issued

by EURATOM. For topics like decommissioning, statement can be found doubting that the existing regulations address

radiation protection in a sufficient manner.

Authors

Erik Bauman

New NP GmbH

Paul-Gossen-Str. 100

91052 Erlangen,

Germany

Dr. Angelika

Bohnstedt

Karlsruhe Institute

of Technology (KIT)

Programm Nukleare

Entsorgung, Sicherheit

und Strahlenforschung

(NUSAFE)

Hermann-von-

Helmholtz-Platz 1

76344 Eggenstein-

Leopoldshafen,

Germany

In the Focus Session Radiation

Protection – What about the basic

principles and objectives in the

current regulatory environment?

three presentations (a forth one had

to be cancelled on short term for

personal reasons) directed the view

on different aspects. This gave the

­occasion to the 25 to 30 participants

for interesting questions and a fruitful

exchange of opinion not only with the

lecturer but also among each other.

Especially the presentation dealing

in a somewhat provocative way

with the subject ‘Hormesis’ led to a

motivated discussion in the audience

about different point of views.

In the first presentation Hormesis

– a Miracle in reality? Discussion Required

Jan-Christian Lewitz (LTZ-

Consulting GmbH) started with from

literature compiled controversial conclusions

about the amount of harmed

people by the Chernobyl accident. This

was followed by a provocative statement

about the hormesis principle in

the way “when unhealthy things

become healthy” and “it is just

depending on the right dose”. He quoted

the explanation for hormesis as “biopositive

reaction of biological systems”

but also restricted that there are

“no general mechanism known for the

different hormetic effects” and indicated

that hormesis is not con sidered

for risk assessment. Then Mr. Lewitz

showed curves about the dose/effect

relation and the LNT ­model and

­remarked that little ­scientific evidence

of any measurable adverse health

effects at radiation doses below about

100 mSv is at the moment available.

As a discussible example for another

effect he shortly gave an overview of an

incident in Taiwan in the eighties of

the last century where buildings, used

by about 10,000 people, were constructed

with Co-60 contaminated

steel. Higher-than-normal radiation

levels were discovered after 9 years and

therefor surveys for cancer and birth

defects in this group of persons, some

lived up to 20 years in the building,

where executed. Mr. Lewitz presented

the result of the survey with a lower

mortality in the examined group than

in the normal average public.

He ended his presentation with the

questions “What should be looked

­after and be obeyed?” and “Is Optimization

below 100 mSv/y justified in

regard to limited resources?” and

encouraged the audience to discuss

with him his challenging point of view.

The second lecture Radiation

Instrumentation and Measurement

Technologies for High Radiation

Fields was given by Dr. Marina Sokcic-

Kostic (NUKEM Technologies Engineering

Services GmbH) who talked about

the possibility to monitor radioactive

materials in high dose-rate environments

where common types of gamma

detectors reach their limits. The first

instrument she presented was a

Geiger- Mueller-Counter where by

switching on and off the counting

tube dead-times can be avoided.

Next Ms. Sokcic-Kostic remarked that

measurement of particle radiation in

the presence of high gamma fields is

quite challenging. She explained a

­fission chamber, operable for gamma

radiation up to 50 to 100 Sv/h, where

ionization efficiency is set very low, so

that mainly the fission products

are measured and additionally by

adjusting the pulse heights neutrons

can be separated from gammas. Afterwards

she presented some applications

of this chamber. One device

with several chambers is used to

characterize irradiated fuel assemblies

in a storage pond by passive neutron

and passive gamma counting.

Another one she explained where the

chamber is combined with other

measurement instruments works with

active neutron detection monitors

using external neutron or ion sources.

Ms. Sokcic-Kostic conclude her pre sentation

with a Gamma camera which is

able to localize hot spots in waste

packages and some information about

Cherenkov detectors.

With the final talk Predictions of

Expected Dose Rates by validated

Activation Calculations as Input for

a step-wise Decommissioning and

Dismantling of a Nuclear Power

Plant Dr. L. Schlömer (together with

Dr. S. Tittelbach and Prof. P.-W.

Phlippen; all WTI Wissenschaftlich-

Tech nische Ingenieurberatung GmbH)

changed the subject to Monte-Carlo

modelling. He listed the specific

requirements for decommissioning

like licensing, planning of packaging,

probing and of course cost estimation.

Then he showed that Monte-Carlo

code coupled with modern variance

reduced techniques (ADVANTG) is a

good solution for radiological characterization

while reducing number

of samples and related costs. Mr.

Schlömer commented that even more

detailed calculations are able with

an activation and decay module

(­ORIGEN-S). With an example for a

BWR (same for a PWR) he explained

the steps which have to be performed

for the calculation procedure to get

from a technical drawing of a reactor

to a detailed MCNP-model. To validate

the method a comparison of

measured and calculated dose rates is

necessary. Therefore, Mr. Schlömer

continued, dose rate measurements

have to be executed on defined places

between RPV and biological shield. He

concluded his presentation with the

outcome that the methods of validation

show good results for the BWR

and the PWR.

AMNT 2017

Focus Session: Radiation Protection ı Erik Baumann and Angelika Bohnstedt


The International Expert Conference on Nuclear Technology

Estrel Convention

Center Berlin

29 – 30 May

2018

Germany

AMNT 2018

Key Topics

Outstanding Know-How &

Sustainable Innovations

Enhanced Safety &

Operation Excellence

Decommissioning Experience &

Waste Management Solutions

Preliminary Programme

December 15, 2017

Subject to change.

www.nucleartech-meeting.com


atw Vol. 63 (2018) | Issue 1 ı January

48

Plenary Session

Tuesday ı May 29 th 2018

All contributions translated simultaneously

in English/German.

Key Topic

Outstanding Know-How &

Sustainable Innovations

AMNT 2018

Welcome and Opening Address

| | Dr. Ralf Güldner, President of DAtF, Germany

33

Policy

Continuity or Disruption, What Future

for EU-UK Nuclear Partnership

| | Greg Clark MP, Secretary of State for Business,

Energy and Industrial Strategy, United Kingdom

(TBC)

Decommissiong and Interim Storage

after Assignment of Responsibilities

Rückbau und Zwischenlagerung

nach der Neuordnung

| | Dr. Dr. Jan Backmann, Head of Reactor Safety and

Radiation Protection, Ministry of Energy,

Agriculture, the Environment and Digitalization

of Schleswig-Holstein, Germany

33

Economy

The NDA's Current Strategy and

its Long term Objectives

| | David Peattie, CEO, Nuclear Decommissioning

Authority (NDA), United Kingdom

Nuclear Power under Current Market Conditions

in Switzerland

| | Dr. Willibald Kohlpaintner, Head of Nuclear Energy

Division, Axpo Holding AG, Switzerland

33

Competence

How Does Nuclear Phase-Out Affect

the International Business of German

Technical and Scientific Support

Organisations?

| | Dr. Dirk Stenkamp, CEO, TÜV Nord Group,

Germany

Phase-Out in Germany –

We Are International!

| | Carsten Haferkamp, Managing Director,

New NP GmbH

33

Communications

Trust Building by Participation – National

Societal Advisory Committee's Challenging

Objective

Vertrauen schaffen durch Partizipation –

Die große Aufgabe des Nationalen Begleitgremiums

bei der Endlagersuche in Deutschland

| | Prof. Dr. Klaus Töpfer, Former Federal Minister,

Member of the National Societal Advisory

Committee, Germany (TBC)

33

Waste Management

Site Selection in Practice:

Challenges at the Start of the Process

Standortauswahl in der Praxis: Herausforderungen

am Neubeginn des Verfahrens

Panel

| | Ursula Heinen-Esser, Managing Director,

Bundesgesellschaft für Endlagerung (BGE),

Germany

| | N.N.

| | N.N.

| | N.N.

Moderator

| | Johannes Pennekamp,

Frankfurter Allgemeine Zeitung, Germany

Award Ceremony

Award of the Honorary Membership of KTG

| | Presented by Frank Apel, Chairperson of KTG,

Germany

Outside the Box

Black Holes, Multidimensionality and Entropy

– Limits of Reality

| | Dr. Maria J. Rodriguez, Research Group Leader,

Gravitational and Black Hole Theory, Max Planck

Institute for Gravitational Physics, Germany

Focus Sessions

Tuesday ı 29 th May 2018

International Regulation | Radiation

Protection: The Implementation of the

EU Basic Safety Standards Directive

2013/59 and the Release of Radioactive

Material from Regulatory Control

Coordinator:

| | Dr. Christian Raetzke, CONLAR Consulting on

Nuclear Law, Licensing and Regulation, Germany

The EU Basic Safety Standards Directive has to be

implemented in national law by 6 February 2018. In

Germany a new Act on Radiation Protection has

been created. The changes present many challenges

to regulators and industry alike in EU countries. The

session will particularly focus on the release of radioactive

material from regu latory control and will put it

in the context of the new Directive.

The Implementation of the New EU BSS

in France

| | Sidonie Royer-Maucotel, Commissariat

á l'Énergie Atomique et aux Énergies Alternatives

(CEA), France (TBC)

The Implementation of the New EU BSS

in Germany

| | Dr. Goli-Schabnam Akbarian, Federal Ministry for

the Environment, Nature Conservation, Building

and Nuclear Safety (BMUB), Germany

Comparative Overview of Regulations

for Clearance in NEA Member States

| | Edward Lazo, OECD Nuclear Energy Agency (NEA),

France (TBC)

Necessary Modifications on Clearance

Regulations in Germany and Switzerland –

Comparative Analysis

| | Dr. Jörg Feinhals, DMT GmbH & Co. KG, Secretary

of the Working Group Disposal/Directory of Fachverband

für Strahlenschutz e. V. (Radiation

Protection Association)

Safety of Advanced

Nuclear Power Plants

Tuesday ı 29 th May 2018

Plenary Closing Remarks

| | Frank Apel, Chairperson of KTG, Germany

Coordinators:

| | Dr. Andreas Schaffrath, Gesellschaft

für Anlagen- und Reaktorsicherheit (GRS) gGmbH,

Dr. Thomas Mull, New NP GmbH

Social Evening

DAtF-Reception and

Meet-and-Greet in the Exhibiton Area

New Builds in UK

| | N.N.

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Current Developments in China

| | Prof. Xu Cheng, Karlsruhe Institute of Technology

(KIT), Germany

Status on Thermal-Hydraulic Passive Safety

Systems Design and Safety Assessment

| | Prof. Francesco D'Auria, University of PISA, Italy

Reactor Safety Research in Germany

| | Dr. Thomas Nunnemann, Federal Ministry for

Economic Affairs and Energy (BMWi), Germany

Einzeleffekt- und Integralexperimente

zur Untersuchung des Anlauf- und

Betriebsverhalten passiver Systeme

| | Dr. Thomas Mull, New NP GmbH, Germany;

Prof. S. Leyer, Université du Luxembourg,

Luxembourg; Prof. Uwe Hampel,

Dr. Christoph Schuster, Tech nische Universität

Dresden (TUD), Germany

Modellierung passiver Systeme mit

der nuklearen Rechenkette der GRS

| | Dr. Andreas Schaffrath, S. Buchholz,

Dr. A. Krüsssenberg, Gesellschaft für Anlagenund

Reaktorsicherheit (GRS) gGmbH, Germany

Technical Sessions

Wednesday ı 30 th May 2018

Outstanding Know-How &

Sustainable Innovations

Chair & Keynote Coordinator:

| | Dr. Matthias Lamm

Know-How, New Build and Innovations

Keynote

Can Nuclear Energy Thrive in a Carbon-

Constrained World? – Findings From a

New MIT Study

| | Jacopo Buongiorno, TEPCO Professor and

Associate Department Head, Director, Center for

Advanced Nuclear Energy Systems (CANES),

Massachusetts Institute of Technology (MIT), USA

Keynote

AP1000 –

On the Way to Commercial Operation

| | Tba

Operational Readiness

of the Barakah Nuclear Power Plant

| | Dr. Rolf Janke, Nawah Energy Company, Licensing

& Regulatory Affairs, United Arab Emirates

Russian Reactor Technologies:

Basic Development Trend and “Waiting List”

| | Dr. Andrey Gagarinskiy, NRC Kurchatov Institute,

Russia

Advanced Load Following Control

with Predictive Reactivity Management

(ALFC-PREDICTOR)

| | Andreas Kuhn, New NP GmbH, Germany

Improving Knowledge Transfer Through

Interactive Learning Strategies

| | Jeanne Bargsten, TÜV SÜD Energietechnik GmbH

BW, Germany

Digital Transformation in Nuclear Industry –

Focus: Backoffice Applications

| | Dr. Jan Leilich, New NP GmbH, Germany

Keynote

Co-Generation – A Game Changer

in Polands New Build Plans?

| | Prof. Dr. hab. Grzegorz Wrochna, National Centre

for Nuclear Research, Poland

The Future of Nuclear Power

Chair:

| | Dr. Thomas Mull & Fabian Weyermann

Keynote

DEMO – The Remaining Crucial Step T owards

the Exploitation of Fusion Power After ITER

| | Dr. Gianfranco Federici, EUROfusion, Spain

Application of Variance Reduction Techniques

in Neutronics Shielding Calculations of the

Stellarator Power Reactor HELIAS

| | André Häußler, Karlsruhe Institute of Technology

(KIT), Germany

Synergistic Effect of H and He on W Grain

Boundaries: A First-Principles Study

| | Litong Yang, Forschungszentrum Jülich GmbH,

Germany

Neutronics Analyses on the IFMIF- DONES Test

Cell Bio-Shield and Liner

| | Dr. Yuefeng Qiu, Karlsruhe Institute of Technology

(KIT), Germany

CFD Analysis of Centrifugal Liquid Metal

Pumps

| | Moritz Schenk, Karlsruhe Institute of Technology

(KIT), Germany

New Products, New Services

Chair:

| | Prof. Andreas Class and Ralf Schneider- Eickhoff

Steam Generator Segmentation Innovation

Project

| | Niklas Bergh, Westinghouse Electric Germany

GmbH, Germany

ASME Nuclear Certification and

Other Certification Programs

| | Dr. Dirk Kölbl, CIS GmbH Consulting Inspection

Services, TÜV Thüringen Group, Managing

Director, Germany

Equipment Qualification for Nuclear Power

Plants – Ensuring the Compliance

of Safety-Critical Nuclear Equipment

| | Dr. Ailine Trometer, TÜV SÜD Energietechnik

GmbH, Germany

SISTec: Mathematical Calibration

of Large Clearance Monitors

| | Tim Thomas, Safetec Entsorgungs- und Sicherheitstechnik

GmbH, Germany

Perimeter Security System Peri-D-Fence-L1

| | Steffen Christmann, Westinghouse Electric

Germany GmbH, Germany

A Multipurpose Inertial Electrostatic

Confinement Fusion for Medical Isotopes

Production

| | Dr. Yasser Shaban, Southern Medical University,

School of Biomedical Engineering, Expert

Committee Member, China

Neutronic Analysis of a Nuclear- Chicago NH3

Neutron Howitzer

| | Ahmet Ilker Topuz, Istanbul Technical University,

Turkey

Reactor Physics, Thermo and

Fluid Dynamics

Chair:

| | Dr. Andreas Schaffrath

Keynote Coordinator:

| | Dr. Tatiana Salnikova

Investigation of the Operation Mode

of Passive Safety System 1

PANAS: Experimental and Theoretical

Investigation of Generic Thermal Hydraulic

Issues of Passive Safety Systems

| | Dr. Christoph Schuster, Technische Universität

Dresden (TUD), Germany

EASY – Evidence of Design Basis Accidents

Mitigation Solely with Passive Safety Systems

| | Sebastian Buchholz, Gesellschaft für Anlagenund

Reaktorsicherheit (GRS) gGmbH, Germany

Modelling of Condensation Inside

an Inclined Pipe

| | Amirhosein Moonesi Shabestary, Helmholtz-

Zentrum Dresden-Rossendorf, Germany

Performance of the Passive Flooding System

in the Integral Tests of the Easy Project

| | Nadine Kaczmarkiewicz, Deggendorf Institute of

Technology, Mechanical Engineering, Germany

Investigation of the Operation Mode of

Passive Safety System 2

Chair:

| | Dr. Thomas Mull

Investigation of Thermal Coupling Model for

Evaporation Process in a Slightly Inclined Tube

and Tube Bundles

| | Yu Zhang, Technische Hochschule Deggendorf,

Germany

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Experimental and Theoretical Investigation

of Boiling in the Slightly Inclined Tubes of the

Containment Cooling Condenser

| | Frances Viereckl, TU Dresden, Chair of Hydrogen

and Nuclear Energy, Germany

Model Order Reduction of Low Pressure

Natural Circulation System

| | René Manthey, TU Dresden, Institute of Power

Engineering, Germany

Model Order Reduction of a High Pressure

Natural Circulation System

| | Alexander Knospe, TU Dresden, Institut für

Energietechnik, Germany

New Neutron Kinetic Developments

and Findings

Chair:

| | Dr. Tatiana Salnikova

Keynote

Insights of End-of-Life Core Design

from Utility Point of View

| | Dr. Marcus Seidl, PreussenElektra GmbH,

Germany

Frequency-Domain Investigation

on the Neutron Noise in KWU PWRs

| | Marco Viebach, TU Dresden, Institut für

Energietechnik, Germany

Nuclear Energy Campus

The Nuclear Energy CAMPUS leads through the

world of radioactivity, nuclear technology and

radiation protection with individual stations. There

will be contact persons available at all of the themed

stands to offer information in form of short talks,

movies, demonstrations or experiments. Besides,

information on study options and career perspectives

within nuclear industry are provided. The CAMPUS

language will be German..

Welcome and Introduction

| | Florian Gremme, Young Generation Network,

KTG, Germany

Post-Test Analysis of the RPV Lower Head Leak

Experiment at the INKA Test Facility Using

ATHLET

| | Michael Sporn, TU Dresden, Institute of Power

Engineering, Germany

New Thermal Hydraulic Development

and Findings

Chair:

| | Dr. Sanjeev Gupta

Keynote

International Cooperation in the Experimental

Field of Nuclear Thermohydraulics: Primary

Coolant Loop Test Facility (PKL)

| | N.N., OECD, France

Keynote

International Cooperation

on Pool Scrubbing Research:

Examples of NUGENIA/IPRESCA Project

| | Dr. Sanjeev Gupta, Becker Technologies GmbH,

Germany

Application of an Eulerian/Eulerian

CFD Approach to Simulate the

Thermohydraulics of Rod Bundles

| | Dr. Wei Ding, Helmholtz Zentrum Dresden

Rossendorf, Germany

Analysis of the Fatigue of the Bolts in the

Flange of a Reactor Pressure Vessel

| | Fabian Gottlieb, Kraftanlagen Heidelberg GmbH,

Technical Analysis, Germany

Preliminary Results of Water Hammer

Simulation in Two-Phase Flow Regimes

Using the Code ATHLET 3.1A

| | Christoph Bratfisch, Ruhr-Universität Bochum,

Germany

Design of Simplified and Optimized Heavy

Liquid Metal Loop for Future Applications

| | Dr.-Ing. Nader Ben Said, Westinghouse Electric

Germany GmbH, Germany

Investigation on Variation of Nodelized

Macroscopic Cross Sections Driven by

Deflection of Fuel assemblies with Serpent

| | Nico Bernt, Technische Universität Dresden (TUD),

Germany

PWR Cycle Analysis With the GRS Core

Simulator KMACS

| | Dr. Matías Zilly, Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH, Germany

Application of a Full-Core Statistical Approach

in LB-LOCA Analysis

| | Dr. Andreas Wensauer, PreussenElektra GmbH,

Germany

Nuclear Data Uncertainty Analyses

with XSUSA and MCNP

| | Dr. Winfried Zwermann, Gesellschaft für Anlagenund

Reaktorsicherheit (GRS) gGmbH, Germany

Workshop

Young Scientists' Workshop

Tuesday ı 29 th May 2018

Wednesday ı 30 th May 2018

Coordinator:

| | Prof. Dr.-Ing. Jörg Starflinger,

University of Stuttgart, Germany

Jury:

| | Prof. Dr. Marco K. Koch,

Ruhr-Universität Bochum

| | Prof. Dr. Jörg Starflinger, University of Stuttgart,

Institut für Kernenergetik und Energiesysteme

(IKE)

| | Dr. Wolfgang Steinwarz

| | Dr. Katharina Stummeyer, Gesellschaft

für Anlagen- und Reaktorsicherheit (GRS) gGmbH

Prize awarded by:

| | GNS Gesellschaft für Nuklear-Service mbH and

Forschungsinstitut für Kerntechnik und Energiewandlung

e. V.

Introducing of the

Young Generation Network

| | Yvonne Schmidt-Wohlfarth, Young Generation

Network, KTG, Germany

Nuclear Technology in and

Beyond our Daily Lifes

| | N.N.

Working in NPPs

| | Sebastian Hahn, Young Generation Network, KTG,

Germany

Radioactivity and Radiation Protection

| | Sven Jansen, VKTA – Strahlenschutz, Analytik &

Entsorgung Rossendorf e. V., Germany

Final Disposal of Radioactive Waste

| | Dr. Thilo von Berlepsch (BGE), Germany

Nuclear Fusion

| | André Häußler, Elena Nunnemann, Karlsruhe

Institute of Technology (KIT), Germany

Stations of Nuclear Campus

1 NPPs & Decommissioning

2 Electricity Market – Composition

of the Electricity Price

3 Packaging, Casks & Conditioning of Waste

4 Nuclear Medicine Applications

Modeling of Post-Dryout Heat Transfer

| | Dali Yu, Karlsruhe Institute of Technology (KIT),

Germany

Detailed session programme to be announced.

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Key Topic

Enhanced Safety &

Operation Excellence

Focus Sessions

Tuesday ı 29 th May 2018

Radiation Protection during

decommissioning – are there special

needs?

Coordinators:

| | Eric Baumann, New NP GmbH, Germany

Dr. Angelika Bohnstedt, Karlsruhe Institute of

Technology (KIT), Germany

The first area of the session addresses protection of

personnel engaged in decommissioning activities.

Decommissioning activities are associated with

large changes in the NPP configuration. The overall

remaining radioactive inventory will shift. Systems

are taken out of service. Originally confined radioactive

sources become open sources due to the fact

that systems are decommissioned. Demolition of

civil structures and other unit elements require

additional technical systems to cope with air

contamination.

The second area of discussion deals with clearance of

radioactive material and public acceptance. A large

amount of medium and low level waste has to be

examined and removed from the site. This topic

also addresses the question of “conditional” and

“ unconditional” release of components and material.

It is not only a question about clearance levels but

also about public acceptance of receiving cleared, i.e.

non-nuclear, waste at normal landfill sites.

To all decommissioning activities, the ALARA

approach applies. German rules and regulations,

e.g. Radiation Protection Ordinance, various KTA

rules, and international rules (BSS) and recommendations

issued e.g. by IAEA or EU provide an appropriate

framework for workers protection. Is there a

need for specific “German decommissioning rules”?

The final closure of a site requires the removal of all

material. The largest amount originating from the

demolition of buildings is non-nuclear waste. Some

amount of waste has gone through the clearance.

Some amount of waste was never subject to nuclear

regulatory surveillance because it originates from

office buildings, cooling towers, turbine buildings

(in PWR plants), pumping station structures and

others. Beside construction waste, valuable raw

materials are extracted – e.g. copper from electrical

cables. How is the public acceptance of “evil stuff”

from a nuclear power plant?

This session tries addressing some of these questions

and tries providing some answers. Some of the

presentation will give an interesting introduction

and the answer might be gained during lively

discussions between the session participants.

Detailed session programme to be announced.

International Operational Experience

Coordinator:

| | Dr.-Ing. L. Mohrbach, VGB PowerTech e.V.,

Germany

The operation of nuclear power plants involves a

wide scope of specialized areas of expertise, from

materials to human factors. Beyond daily business,

some background information from different fields

of operational activities might not only be regarded

as personally worthwhile but may also be well suited

to complement the general knowledge base for

nuclear.

This session addresses some of these questions

and tries providing some answers. Some of the

presen tations will give an introduction and produce

questions. The answer might be gained during lively

discussions between the session participants.

Summary of the QUENCH LOCA

Experimental Program

| | Dr. Andreas Wensauer, PreussenElektra GmbH,

Germany

Practical Safeguards in Nuclear Power Plants

| | Dr. Irmgard Niemeyer, Dipl.-Ing. Katharina Aymanns,

Forschungszentrum Jülich GmbH, Germany (TBC)

Comparison of Employment Effects

of Low-Carbon Generation Technologies

| | Geoffrey Rothwell, OECD Nuclear Energy Agency

(NEA), France

Application of Lubricants and

other Consumables in Nuclear Power Plants

| | Dr. Fred Böttcher, EnBW Kernkraft GmbH;

Dr. Dittmar Rutschow, VGB PowerTech e. V.,

Germany

New Developments in Radiation Protection

| | N.N.

Benefits of Simulator Training

| | N.N., KSG Kraftwerks-Simulator- Gesellschaft mbH,

GfS Gesellschaft für Simulatorforschung mbH,

Germany

Technical Sessions

Wednesday ı 30 th May 2018

Operation and Safety

of Nuclear Installations

Chair:

| | Dr. Thorsten Hollands

Keynote Coordinator:

| | Dr. Erwin Fischer

Chair:

| | Dr. Thorsten Hollands

Keynote

Safe to the Last Day – A Challenge for Operators

| | Christoph Heil, EnBW Kernkraft GmbH, Executive

Director, Germany

Keynote

Is Safety Culture Perceptible and Measurable?

| | Uwe Stoll, Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH, Scientific and

Technical Director, Germany

Keynote

Preserving and Ensuring Competence

and Motivation

| | Dr. Frank Sommer, PreussenElektra GmbH,

Head of CoC Operations, Germany

Digital Transformation in Nuclear Industry –

Focus: Site Applications

| | Dr. Jan Leilich, New NP, IBGM Product Management,

Germany

Save to the Last Day – How to Manage the

Complexity in a Multi-Year End of Life Process

| | Prof. Dr. Rüdiger von Der Weth, Hochschule für

Wirtschaft und Technik Dresden, Faculty of

Business Administration, Germany

Loca Scenario-Related Zinc Borate

Precipitation Studies at Lab Scale

| | Dr. Ulrich Harm, Technische Universität Dresden

(TUD), Germany

Simulation of Asymmetric Severe Accidents

Using the Code System AC2

| | Liviusz Lovasz, Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH, Germany

Simulation of the Bundle Test QUENCH-07

with the Severe Accident Analysis Codes

ASTEC V2.1 and AC^2 – ATHLET CD

| | Florian Gremme, Ruhr-Universität Bochum, Chair

of Energy Systems and Energy Economics,

Germany

Sensitivity and Uncertainty Analysis

of the MCCI Model Results in AC2/COCOSYS

for the OECD-CCI3 Experiment

| | Dr. Claus Spengler, Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH, Germany

Dry Filter Method DFM 2.0 – The Newest

Generation of Filtered Containment Venting

System

| | Dr. Peter Hausch, Caverion Deutschland GmbH,

Business Unit Krantz, Germany

3D Surface Radiation Dosimetry of

a Nuclear-Chicago NH3 Neutron Howitzer

| | Ahmet Ilker Topuz, Istanbul Technical University,

Nuclear Energy, Turkey

Chair:

Dr. Jürgen Sydow

TESPA-ROD Code Prediction of the Fuel Rod

Behaviour During Long-Term Storage

| | Dr. Heinz-Günther Sonnenburg, Gesellschaft für

Anlagen- und Reaktorsicherheit (GRS), Germany

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Summary of Experimental Investigations

at the ALADIN Test Facility

for the Thermal Hydraulic Analysis

of Accident Scenarios in Spent Fuel Pools

| | Christine Partmann, Technische Universität

Dresden (TUD), Germany

Considerations for Multi Unit Effects

in Probabilistic Risk Assessment

| | Dr. Felix Philipp Sassen, Westinghouse Electric

Germany GmbH, Germany

Model-Based Vulnerability Analysis

of Complex Infrastructures

| | Mathias Lange, Hochschule Magdeburg- Stendal,

Germany

Canadian Nuclear Fire PRA

| | Hossam Shalabi, Canadian Nuclear Safety

Commission, Canada

Key Topic

Decommissioning

Experience & Waste

Management Solutions

Focus Sessions

Tuesday ı 29 th May 2018

Post-operation and Decommissioning

in Germany

Coordinator:

| | Dr. Erich Gerhards, PreussenElektra GmbH,

Germany

Preparing for Decommissioning – Meeting the

Changing Requirements for Decommissioning

| | N.N., OECD Nuclear Energy Agency (NEA)

The Paradigm Shift in Nuclear Waste

Management in Germany

Coordinator:

| | Michael Köbl, GNS Gesellschaft

für Nuklear-Service mbH, Germany

In summer 2017 the responsibilities for nuclear

waste management in Germany have been fundamentally

reorganized. While the operators remain

responsible for the decommissioning and dismantling

of their NPPs as well as the packaging of the

nuclear waste, the German government assumes

responsibility not only for final disposal, but additionally

already for interim storage. This means that

the waste pro ducers, who used to be obliged to store

their HLW/ILW until the future availability of the federal

repository “Konrad”, from now on can directly

hand over their suitably packaged waste to the state

owned interim storage facilities. This essentially new

procedure poses huge challenges to the waste producers

as well as to the authorities. It is the aim of

this Focus Session to outline the new regulations and

discuss the consequences for all the parties involved.

TBA

| | Responsible Authorities and Federal Corporations:

BMUB, BfE, BGE, BGZ

TBA

| | Independent Experts

TBA

| | Waste Producers

TBA

| | Suppliers/Vendors

Panel Discussion

| | All Participants

This session will be held in German

with simultaneous English translation.

Keynote

Decommissioning and Waste Management of

Obsolete Nuclear Research Facilities

| | Dr. Vincenzo V. Rondinella, Joint Research Center

(JRC) of the European Commission, Germany

Keynote

Global Status of Decommissioning

| | Patrick J. O’Sullivan, International Atomic Energy

Agency (IAEA), Austria

Ventilation Concepts for Different Phases

During Decommissioning of Nuclear Facilities

| | Dirk Thybussek, Caverion Deutschland GmbH,

Business Unit Krantz, Germany

Bladecutter: A Novel Technology

for Removing Nuclear Sludge

| | Shuai Wang, The University of Manchester, School

of Electrical and Electronic Engineering, United

Kingdom

Untersuchungen Zum Abtrag Asbesthaltiger

Spachtelmasse Mittels Feuchtsandstrahlen

| | Simone Müller, KIT – Rückbau konventioneller &

kerntechnischer Bauwerke, Germany

Das Ausschreibungsverfahren für

den Abbau des Reaktordruckgefäßes und der

RDG-Einbauten im Kernkraftwerk Lingen

| | Stefan Lindemann, RWE Power AG, Germany

How to Improve Decommissioning

by Virtual Engineering Tools

| | Prof. Dr. Ulrich W. Scherer, Hochschule Mannheim,

Faculty Chemical Process Technology, Germany

Characterizing the Radioactivity

of the Concrete Shielding During

Decommissioning of the LFR

| | Perry Young, NRG, Research & Innovation – CP4S

–, Netherlands

Requirements on Operation and Decommissioning

| | Dr. Heinz Drotleff, German Waste Management

Commission, Germany

The Role of Service Operation for Decommissioning

– A Practitioner’s Experience

| | Dr. Thomas Volmar, RWE Power AG, Germany

Competencies and Ressources Required

to Assure Safe Service Operation and

Decommissioning

| | A. Dinter, PreussenElektra GmbH, Germany

Decommissioning and Service Operation

in Sweden

| | M. Bächler, UNIPER Technology, Germany

Full Scope Approach – Hand over of

Operations, Decommissioning, Dismantling

and Waste Management

| | Robert Bonner, AECOM, United Kingdom

This session will be held in German

with simultaneous English translation.

Detailed session programme to be announced.

Technical Sessions

Wednesday ı 30 th May 2018

Decommissioning Experience &

Waste Management Solutions

Chair:

| | Martin Brandauer

Keynote Coordinator:

| | Thomas Seipolt

Keynote

Evaluation of Approaches to Automate

Reactor Internals Segmentation/Evaluation

of New or Enhanced Techniques

for Concrete Decontamination

| | PhD Richard Reid/Richard McGrath, The Electric

Power Research Institute (EPRI), USA

Application of the System FREMES

to Characterize and Sort Soil During

the Remediation of FBFCi Dessel

Fuel Element Factory

| | Felix Langer, NUKEM Technologies Engineering

Services, O-P, Germany

Design 3D, Laser Scanning and Radiological

Data Visualization

| | Sergi Milà, Westinghouse Electric Spain, Spain

Full System Decontamination Project at Bohunice

| | Randall Duncan, Westinghouse Electric Company,

USA

Under Water Cutting Technologies

| | John Hubball, Westinghouse Electric Company,

DDR&WM, USA

Vorstellung eines Magnetfiltersystems

zur Behandlung von Sekundärabfällen der

Wasser-Abrasiv-Suspensions-Schneidtechnik

| | Carla Krauß, Karlsruhe Institute of Technology

(KIT), Germany

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Decommissioning Characterisation Through

Compressive Gamma-Ray Imaging

| | Dr. David Boardman, ANSTO,

Nuclear Stewardship, Australia

AuDeKa: A BMBF Funded Project to Develop

an Automated Deconta mination Cabin with

Documentation Based on Industry 4.0 Features

| | Franz Borrmann, Institut für Umwelt technologien

& Strahlenschutz GmbH, Germany

Lean Documentation Approach

in Decommissioning

| | Franz Borrmann, Institut für Umwelt technologien

& Strahlenschutz GmbH, Germany

Activation Analysis, Validation and Component-

Wise Packaging Concept for the Decommission

Planning of the Gundremmingen NPP

| | Dr. Ben Volmert, Nagra, Inventory & Logistics,

Switzerland

Primary Circuit Decontamination

at Biblis Unit A NPP

| | Markus Thoma, Siempelkamp NIS

Ingenieurgesellschaft mbH, Germany

Rückbau und Entsorgung der Reaktordruckbehälter-Einbauten

und der RDBs

der Kernkraftwerke Philippsburg 1 (KKP 1)

und Neckarwestheim I (GKN I)

| | Dr. Bernhard Wiechers, Westinghouse Electric

Germany GmbH, Decommis sioning, Dismantlling

& Remediation, Germany

This session can be held in German/English

with simultaneous translation.

Radioactive Waste Management,

Storage and Disposal

Chair:

| | Dr. Alexander Zulauf

Keynote Coordinator:

| | Iris Graffunder

Keynote

Challenges in the Management of Concrete

Waste from the Dismantling of Nuclear

Facilities – Case Study Rheinsberg NPP

| | Jörg Möller, EWN Entsorgungswerk

für Nuklearanlagen GmbH, Germany

Keynote

Managing Waste at the Remote- handled

Dismantling of Activated Concrete and Steel

Structures of the Biological Shield of KNK

| | Johannes Rausch, KTE Kerntechnische Entsorgung

Karlsruhe GmbH, Germany

Keynote

Clearance Measurement of Demolition Waste:

Measurement Process with High Operational

Throughput

| | Stefan Thierfeldt, Brenk Systemplanung GmbH,

Germany

Nuclear Energy and Society

Engaging with Society – Past, Present and

Future. Results From the Collabo rative

Interdisciplinary Project HoNESt – History of

Nuclear Energy and Society

| | Dr. Jan-Henrik Meyer, University of Copenhagen,

Saxo Institute, Denmark

Waste Treatment

Fortum NURES®-BORES Concept of Treating

Liquid Radioactive Waste Containing Boron

| | Dr. Jussi-Matti Mäki, Fortum Power and Heat Oy,

Nuclear Services, Finland

Investigations of Process Parameters Using

Microwave Technology for the Treatment of

Radioactive Waste

| | Prof. Dr. Ulrich W. Scherer, Hochschule Mannheim,

Faculty Chemical Process Technology, Germany

Characterization

Advantages and Limits of Spectroscopic

Measurement for the Classification

of Radioactive Wastes

| | Dr. Marina Sokcic-Kostic, NUKEM Technologies

Engineering Services, Engineering, Germany

Waste Management

Development of a Calculation Tool

for Optimal Holistic Disposal Planning

| | Dr. Anton Anthofer, VPC GmbH, Germany

Endlagerdokumentation Neu Gedacht

| | Dr. Anton Anthofer, VPC GmbH, Germany

Development of a Monitoring Concept

for Transport and Storage Containers

for Spent Fuel and Heat-Generating

High-Level Radioactive Waste on Prolonged

Intermediate Storage

| | Daniel Fiß, Hochschule Zittau/Görlitz, Germany

Use of a Statistical Toolset for Risk Aware

Package Planning of Activated Core Internals

| | Dr. Maarten Becker, Institut für Umwelttechnologien

& Strahlenschutz GmbH, Germany

Use of Flexible Packaging and Real Time Assay

Techniques to Divert Low Activity Waste LLW

from the UK LLWR Facility

| | Ian Wigginton, Nuvia Ltd, Waste & Environment,

United Kingdom

Packaging

MOSAIK Casks for Transport, Storage and Final

Disposal of All Kinds of Intermediate Level

Waste – A Success Story Spanning More Than

Three Decades

| | Dr. Jörn Becker, GNS Gesellschaft für

Nuklear-Service mbH, Technik, Germany

One Cask Fits All – The New MOSAIK® II-S

for All Kinds of Intermediate Level Waste

| | David Bergandt, GNS Gesellschaft

für Nuklear-Service mbH, TP2 Project

Management, Germany

GNS SBoX® A New Family of Robust,

Self-Shielded Containers

| | Martin Beverungen, GNS Gesellschaft

für Nuklear-Service mbH, Germany

Automated Ultrasonic Testing of CASTOR®

Cask Bodiesin Serial Production –

A Progress Report

| | Jörg Frank, GNS Gesellschaft für Nuklear-Service

mbH, Cask Manufacturing (Orders), Germany

Quivers for Non Standard Fuel Rods –

Advances and First Utilizations

| | Olga Di Paola, GNS Gesellschaft für Nuklear-

Service mbH, Germany

Experiences in the Assessment

of a Dual Purpose Transport Cask Loaded

with Damaged Spent Nuclear Fuel

| | Dr. Thorsten Schönfelder, Bundesanstalt für

Materialforschung und -prüfung (BAM), Germany

Preliminary Experimental Study on Reduction

of Hydrogen Concentration in a Small- Scale

Radioactive Waste Long-Term Storage

Container with Catalysts

| | Prof. Dr. Kazuyuki Takase, Nagaoka University

of Technology, Japan

Simulation-Based Investigation

of Suitability of Thermography and Muon Flux

Measurements for Non-Invasive Monitoring

of Transport and Storage Containers

for Spent Fuel

| | Michael Wagner, Technische Universität Dresden

(TUD), Germany

Repository

Entsorgung von Wärme Entwickelnden

Radioaktiven Abfällen – Herausforderungen

und Lösungsansätze

| | Matthias Bode, Leibniz Universität Hannover,

Germany

This session can be held in German/English

with simultaneous translation.

53

AMNT 2018

AMNT 2018

Preliminary Programme


atw Vol. 63 (2018) | Issue 1 ı January

54

KTG INSIDE

Fachgruppe Reaktorsicherheit:

Vorstand neu aufgestellt

Inside

Dr. Tatiana Salnikova hat den Vorsitz der KTG Fachgruppe

„Reaktorsicherheit“ von Uwe Stoll erfolgreich übernommen.

Am Moskauer Energetischen Institut studierte Dr. Salnikova

zunächst Umwelttechnik. Im Anschluss daran wechselte sie

zum Kerntechnikstudium an die Hochschule Zittau/Görlitz.

Ihre Promotion im Bereich der thermohydraulischen

Modellierung von Brennelementen mithilfe numerischer

Methoden erfolgte in Kooperation zwischen der TU Dresden

und AREVA NP. Im Jahr 2007 startete Tatiana Salnikova als

Projektleiterin bei der AREVA GmbH. Ihre Arbeitsschwerpunkte

liegen heute auf dem Gebiet der nuklearen Sicherheit.

Dazu gehören die Erstellung von Sicherheitsanalysen

für KKW, die Mitarbeit in nationalen und internationalen

Gremien wie Reaktor-Sicherheitskommission (RSK), International

Atomic Energy Agency (IAEA), und Electric Power

Research Institute (EPRI). Derzeit beschäftigt sie sich unter

anderem mit Fragestellungen zur Lastwechselfahrweise

von KKW. Ebenfalls seit diesem Jahr hat es bei der Position

des Kassenwartes einen Wechsel von Dr. Walter Tromm zu

Dr. Frank Sommer gegeben. Dr. Frank Sommer ist seit

2013 Bereichsleiter für das Kompetenzcenter Betrieb der

PreussenElektra GmbH in Hannover. Er studierte Maschinenbau

an der Ruhr-Universität in Bochum und promovierte

dort im Anschluss am Lehrstuhl für Strömungstechnik. Seit

1992 ist Frank Sommer in verschiedenen Funktionen

bei PreussenElektra bzw. ihren Vorgängerunternehmen beschäftigt.

Für die geleistete Arbeit bedanken wir uns herzlich bei

den Amtsvorgängern.

Dr. Angelika Bohnstedt (KIT) als stellvertretende Fachgruppensprecherin

gewählt. Herzlichen Glückwunsch.

Erik Baumann

Sprecher der Fachgruppe Strahlenschutz

6. Bilaterales Treffen WiN Schweden

und WiN Germany

26./27. Oktober 2017 – Informationszentrum

Kernkraftwerk Biblis

Bereits zum sechsten Mal trafen sich schwedische und

deutsche Women in Nuclear (WiN) zum Erfahrungsaustausch.

Nach Oskarshamn im April 2016 lud Deutschland

am 26./27.Oktober 2017 nach Biblis ein – das Kernkraftwerk

Biblis war neben dem bilateralen Treffen auch Gastgeber

für die Mitgliederversammlung von WiN Germany

2017.

Dr. Tatiana Salnikova

(Sprecherin der KTG Fachgruppe Reaktorsicherheit)

und Dr. Frank Sommer

(Kassenwart der KTG Fachgruppe Reaktorsicherheit)

| | „Insgesamt sind wir gut aufgestellt, um das Rückbauprojekt erfolgreich

durchzuführen – es ist gut, dass der Rückbau jetzt begonnen hat“,

resümiert Kemmeter am Ende seines Vortrags.

Fachgruppe Strahlenschutz:

Jahresrückblick 2017

Der Schwerpunkt der Tätigkeit der KTG Fachgruppe

Strahlenschutz lag 2017 in der Vorbereitung und Durchführung

der Focus Session Radiation Protection im Rahmen

des gemeinsam von der KTG e.V: und dem DAtF e.V. veranstalteten

48. Annual Meeting on Nuclear Technology

(AMNT 2018, Jahrestagung Kerntechnik).

Die Focus Session ist seit 2015 fester Bestandteil im Programm

der AMNT. Durch die gemeinsame Anstrengung

der Mitglieder der Fachgruppe gelang es auch 2017 eine

interessante Focus Session mit dem Thema „Radiation

Protection – What about the basic principles and objectives

in the current regulatory environment?“ zu gestalten. Die

Berichterstattung dazu findet sich in der Ausgabe 1 (2018)

der atw.

Am Rande der Jahrestagung Kerntechnik fand eine

Versammlung der Fachgruppe Strahlenschutz statt, zu der

alle Mitglieder vorab per E-Mail eingeladen worden waren.

Ein wesentlicher Tagesordnungspunkt bestand in der Wahl

eines neuen Stellvertreters, da der bisherige Stellvertreter,

Herr Sinisa Simic nicht mehr zur Verfügung steht. Einstimmig

wurde von den anwesenden Mitgliedern

Der große Dank für die Einladung und finanzielle

Unterstützung wurde seitens der WiNner persönlich dem

Gastgeber Horst Kemmeter – Leiter des Kernkraftwerkes

Biblis – überbracht, der in seinem Einführungsvortrag den

Stand der Rückbauaktivitäten des KKW Biblis vorstellte

und das Motto des WiN-Treffens The long way to green field

durchaus passend für den Standort Biblis fand.

Nach Besichtigung des Standortzwischenlager (SZL), in

dem Castor®- und Mosaik-Behälter lagern, sowie der Baustelle

des neu entstehenden LAW-II-Lagers (Low Active

Waste-Lager) fasste Martina Etzmuß (Preussen Elektra) im

Rahmen des ­offiziellen Vortragsprogrammes die politische

Situation in Deutschland insbeson dere nach dem Erd beben

und Tsunami in Japan und der sofortigen Still legung von 8

Kraftwerksblöcken zusammen.

Maria Taranger (Barsebäck AB) stellt die politischen

Rahmenbe dingungen in Schweden vor: Das National

Energy Agreement vom Juni 2015 hat zumindest für eine

mittelfristige Sicherheit gesorgt, denn eine Stilllegung von

KKWs aus politischen Gründen ist danach nicht mehr

möglich.

Das schützt jedoch nicht vor wirtschaftlichen Entscheidungen,

so wie sie in Ringhals 1 und 2 von Vattenfall

im letzten Jahr mit vorzeitiger Abschaltung getroffen

wurden. Anna Collin (Ringhals AB) berichtete vom Projekt

KTG Inside


atw Vol. 63 (2018) | Issue 1 ı January

STURE, mit dem die Stilllegung der beiden Ringhals-

Blöcke geregelt ist. Gleichzeitig sollen die Blöcke 3 und 4

sicher bis 2045 weiter betrieben werden. Dies bedeute ein

starker Fokus auf den sogenannten Human Factor, wie

Mitarbeiterqualifikationen und Flexibilität, so Collin.

Katarina Andersson (OKG) und Maria Taranger (BKAB)

stellten in einer gemeinsamen Präsentation die Rückbauaktivitäten

von Barsebäck 1 und 2 sowie Rückbauplanungen

von Oskarsham 1 und 2 vor. An vielen Stellen

profitiert man von der guten Zusammenarbeit, trotzdem

gäbe es standortspezifische Anforderungen.

Strategische Aspekte des Abfallmanagements wurden

von Sofia Eliasson (OKG) vorgestellt.

Katrin Hertkorn-Kiefer (RWE) trug Einzelheiten zu den

Rückbauprojekten der RWE vor und stellte fest, dass das

Konzept des sicheren Einschlusses für Biblis keine Option

gewesen wäre, es sei der Öffentlichkeit nicht mehr zu

vermitteln.

In die abschließende Diskussionsrunde Are we well

prepared for dismantling? führte Martina Sturek (SKB) mit

ihrem Vortrag zum schwedischen Entsorgungskonzept

ein. In Schweden sind die Betreiber der Kernkraftwerke

für die Entsorgung und Endlagerung verantwortlich.

Sie haben hierfür die gemeinsame Gesellschaft Svensk

Kärnbränslehantering AB (SKB) gegründet, die auch für

Transporte und Zwischenlagerung zuständig ist. Der

hochradioaktive Abfall soll im Wirtsgestein Granit im

Endlager Forsmark gelagert werden.

Dr. Christiane Vieh (BGE) vermittelte Eindrücke von der

Verantwortung der BGE für die Endlager in Deutschland.

Mit der Neugründung von zwei bundeseigenen Gesellschaften,

der Bundesgesellschaft für Endlagerung (BGE)

und der Bundesgesellschaft für Zwischenlagerung (BGZ)

übernimmt die Bundesrepublik Deutschland die Verantwortung

für die Zwischen- und Endlagerung von

radioaktiven Abfällen. Hingegen sind die Betreiber der

Kernkraftwerke weiterhin für den Rückbau Ihrer Anlagen

nach der Stilllegung zuständig.

In Deutschland ist die Suche nach einem Endlager für

wärmeentwickelnde radioaktive Abfälle in vollem Gange

und die Entscheidung für einen Standort wird im Jahr

2031 erwartet. Der Schacht Konrad, ein stillgelegtes

Eisenerz-Bergwerk, wird derzeit zum Endlager für radioaktive

Abfälle mit vernachlässigbarer Wärmeentwicklung

umgerüstet.

Gabi Voigt – ja, die aktuelle WiN Global Präsidentin ist

Mitglied von WiN Germany – berichtete unter anderem

stolz, dass WiN-Global seit dem 20. August 2017 als NGO

registriert wurde. Dies war eine notwendige Formalie für

| | Martina Sturek, WiN Präsidentin von WiN Schweden – im Bild links mit

WiN Germany Präsidentin Jutta Jené rechts) sowie WiN Global Präsidentin

Gabi Voigt (Mitte) – hat ein Treffen voraussichtlich im November 2018 im

Kernkraftwerk Ringhals angekündigt.

| | Gabi Voigt, aktuelle WiN Global Präsidentin

die WiN-Global Konferenz Ende August 2017 in China und

wurde mit hohem Aufwand noch fristgerecht erreicht.

Gabi blickt auf viele Aktivitäten im Jahr 2017 zurück und

stellte fest, dass es sehr viel mehr Arbeit gewesen sei, als sie

erwartet habe. Ziele für das kommende Jahr sind u.a. eine

stärkere Präsenz in den sozialen Medien und dass verschiedene

Konzepte der Zusammenarbeit (Memorandum

of Understanding) mit anderen Organisationen wie WNA,

ICRP, IRPA oder INYG mit Leben gefüllt würden.

Verleihung des WiN Germany Preises 2017

im Rahmen des bilateralen Treffens

Zum ersten Mal in der Geschichte von WiN Germany e.V.

fand die Präsentation der eingereichten wissenschaft lichen

Arbeit für den WiN Germany Preis im Rahmen des bilateralen

Treffens statt. Larissa Klaß, die zurzeit ihre Doktorarbeit

am Forschungszentrum Jülich schreibt, trug aus ihrer

Masterarbeit zum Thema Modified diglycolamides for a

selective separation of Am (III): complexation, structural investigations

and possible application vor. Das Fachwissen und

die Eloquenz von Larissa beim Vortrag einschließlich ihrer

55

KTG INSIDE

| | 19 WiNners aus Schweden und 24 WiNners aus Deutschland trafen sich beim 6. bilateralen Treffen der

beiden WiN-Chapter am Standort Biblis

| | WiN-Präsidentin Jutta Jené gratuliert Larissa Klaß – eine würdige

WiN-Preisträgerin, die sich über die für sie einmalige Gelegenheit freute,

vor einem ausschließlich weiblichen Publikum vortragen zu können.

KTG Inside


atw Vol. 63 (2018) | Issue 1 ı January

56

KTG INSIDE

KTG Inside

Verantwortlich

für den Inhalt:

Die Autoren.

Lektorat:

Sibille Wingens,

Kerntechnische

Gesellschaft e. V.

(KTG)

Robert-Koch-Platz 4

10115 Berlin

T: +49 30 498555-50

F: +49 30 498555-51

E-Mail: s.wingens@

ktg.org

www.ktg.org

souveränen Antworten auf fachliche Detailnachfragen

überzeugten das gesamte deutsch-schwedische Auditorium.

Kurz vor Ende der diesjährigen Veranstaltung traf eine

äußerst erfreuliche Nachricht von URENCO Deutschland

GmbH ein: URENCO sponsort den WiN Germany Award mit

1.500 Euro, da dem Unternehmen die Förderung von Frauen

in der Kerntechnik am Herzen liegt und Bestandteil

Herzlichen

Glückwunsch

Januar 2018

91 Jahre wird

1. Prof. Dr. Werner Oldekop,

Braunschweig

89 Jahre wird

20. Dr. Devana Lavrencic-Cannata, Rom/I

88 Jahre wird

10. Dipl.-Ing. Hans-Peter Schmidt,

Weinheim

87 Jahre wird

12. Dr. Rolf Hueper, Karlsruhe

86 Jahre wird

3. Dipl.-Ing. Fritz Kohlhaas, Kahl/Main

85 Jahre wird

9. Prof. Dr. Hellmut Wagner, Karlsruhe

83 Jahre werden

10. Dipl.-Ing. Walter Diefenbacher,

Karlsruhe

17. Dipl.-Ing. Helge Dyroff, Alzenau

24. Theodor Himmel, Bad Honnef

82 Jahre werden

5. Obering. Peter Vetterlein, Oberursel

23. Prof. Dr. Hartmut Schmoock,

Norderstedt

30. Dipl.-Phys. Wolfgang Borkowetz,

Rüsselsheim

30. Dipl.-Ing. Friedrich Morgenstern,

Essen

81 Jahre werden

7. Dipl.-Ing. Albrecht Müller,

Niederrodenbach

9. Dipl.-Ing. Werner Rossbach,

Bergisch Gladbach

25. Dipl.-Ing. (FH) Heinz Wolf,

Philippsburg

80 Jahre werden

7. Dipl.-Ing. Manfred Schirra, Stutensee

8. Dipl.-Ing. Wolfgang Repke, Waldshut

10. Dr. Dieter Türck, Dieburg

12. Dipl.-Ing. Hans Dieter Adami, Rösrath

18. Dr. Werner Katscher, Jülich

22. Dr. Frank Müller, Erlangen

79 Jahre werden

11. Dipl.-Ing. Gerwin H. Rasche, Hasloch

13. Dr. Udo Wehmann, Hildesheim

16. Dr. Wolfgang Kersting, Blieskastel

21. Prof. Dr. Detlef Filges, Langerwehe

28. Dr. Sigwart Hiller, Lauf

78 Jahre wird

4. Dipl.-Ing. Wolfgang Semenau,

Laudenbach

77 Jahre werden

3. Dipl.-Ing. Ferdinand Wind,

Tettnang-Burgermoos

12. Dr. Hand-G. Bogensberger,

Anthem/USA

15. Dipl.-Ing. Ulf Rösser,

Heiligkreuzsteinach

26. Dr. Heinrich Pierer von Esch, Erlangen

76 Jahre werden

6. Dipl.-Ing. Günter Höfer, Mainhausen

31. Dipl.-Phys. Werner Scholtyssek,

Stutensee

75 Jahre werden

19. Dr. Gerd Habedank,

Seeheim-Jugenheim

24. Dr. Günter Bäro Weinheim

70 Jahre wird

20. Dipl.-Ing. Edgar Bogusch, Erlangen

60 Jahre werden

7. Rüdiger König, Essen

19. Dipl.-Ing. Erwin Neukäter, Sugiez/CH

50 Jahre werden

12. Dipl.-Phys. Karl Froschauer,

Freigericht-Somborn

19. Dipl.-Ing. Sönke Holländer, Essen

21. Dipl.-Ing. Torsten Fricke, Hohnstorf

Februar 2018

90 Jahre wird

10. Dipl.-Ing. Hans-Peter Schabert,

Erlangen

89 Jahre wird

20. Dr. Helmut Hübel, Bensberg

88 Jahre wird

5. Dr. Eberhard Teuchert,

Leverkusen

ihres Nachhaltigkeitsprogrammes ist. Damit ist nicht nur

der diesjährige Preis finanziert, sondern auch die Vergabe

des WiN-Preises 2018 ist gesichert. Entsprechend groß viel

der Beifall aus. WiN Germany sagt herzlichen Dank an

URENCO Deutschland für die großzügige Spende und hofft

auf Nachahmer!

87 Jahre wird

14. Dipl.-Ing. Heinrich Kahlow,

Rheinsberg

85 Jahre wird

11. Dr. Rudolf Büchner, Dresden

Yvonne Broy

84 Jahre werden

9. Dr. Horst Keese, Rodenbach

12. Dipl.-Ing-. Horst Krause, Radebeul

23. Prof. Dr. Dr.-Ing. E.h. Adolf Birkhofer,

Grünwald

82 Jahre werden

6. Dr. Ashu-T. Bhattacharyya, Erkelenz

17. Dr. Helfrid Lahr, Wedemark

81 Jahre werden

5. Prof. Dr. Arnulf Hübner, Berlin

6. Dipl.-Ing. Heinrich Moers,

Maitland/USA

11. Dr. Günter Keil, Sankt Augustin

18. Dipl.-Ing. Hans Wölfel, Heidelberg

21. Dipl.-Ing. Hubert Andrae, Rösrath

80 Jahre werden

15. Dr. Hans-Heinrich Krug, Saarbrücken

27. Dr. Klaus Wolfert, Ottobrunn

79 Jahre werden

3. Dr. Roland Bieselt, Kürten

8. Dr. Joachim Madel, St. Ingbert

8. Dr. Herbert Spierling, Dietzenbach

22. Dr. Manfred Schwarz, Dresden


28. November 2017

Dipl.-Phys.

Erich Neuburger

Karlsruhe

Die KTG verliert in ihm ein langjähriges

aktives Mitglied, dem sie ein

ehrendes Andenken bewahren wird.

Ihren Familien gilt unsere Anteilnahme.

KTG Inside


atw Vol. 63 (2018) | Issue 1 ı January

78 Jahre werden

9. Dr. Gerhard Preusche,

Herzogenaurach

13. Dr. Hans-Ulrich Fabian, Gehrden

14. Dipl.-Ing. Kurt Ebbinghaus,

Bergisch Gladbach

21. Dr. Jürgen Langeheine, Gauting

23. Dr. Gerhard Heusener, Bruchsal

25. Prof. Dr. Sigmar Wittig, Karlsruhe

77 Jahre wird

16. Dr. Jürgen Lockau, Erlangen

76 Jahre werden

6. Dr. Michael Schneeberger, Linz/A

22. Cornelis Broeders, Linkenheim

75 Jahre werden

5. Dr. Joachim Banck, Heusenstamm

9. Dr. Friedrich-Karl Boese, Leonberg

13. Dr. Ingo-Armin Brestrich, Plankstadt

20. Ing. Leonhard Irion, Rückersdorf

28. Dr. Klaus Tägder, Sankt Augustin

70 Jahre werden

7. Dr. Hans-Hermann, Remagen

8. Dr. Max Hillerbrand, Erlangen

14. Reinhold Rothenbücher, Erlangen

23. Dr. Rudolf Görtz, Salzgitter

29. Dr. Anton von Gunten,

Oberdiessbach

65 Jahre werden

3. Dr. Reinhard Knappik, Dresen

20. Dipl.-Ing. Berthold Racky, Nidderau

60 Jahre werden

3. Prof. Dr. Sabine Prys, Offenburg

3. Dipl.-Ing. Siegfried Wegerer,

Tiefenbach

10. Dipl.-Ing. (FH) Anton Hums,

Essenbach

50 Jahre werden

5. Dr. Volker Wunder, Ottensoos

20. Dr. Josef Engering, Jülich

22. Toralf Wolf, Plauen

28. Dipl.-Ing. Jörg Schneider, Radebeul

Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag

und wünscht ihnen weiterhin alles Gute!

Wenn Sie keine

Erwähnung Ihres

Geburtstages in

der atw wünschen,

teilen Sie dies bitte

rechtzeitig der KTG-

Geschäftsstelle mit.

57

NEWS

Top

Back to the Future –

MIT resurrects 1940s-era

nuclear experiment

(nei) On 2 December 2017, the Massachusetts

Institute of Techno logy’s

­Nuclear Science and Engineering Department

restarted a “subcritical

graphite exponential pile” dating back

to the earliest years of the atomic age.

The pile is similar in design to the

famous Chicago Pile-1 (CP-1) built by

Enrico Fermi under the bleachers of

the University of Chicago football

stadium, which in 1942 initiated the

world’s first man-made, self-sustaining

fission chain reaction.

“Those were the good old days,

when scientists were more important

than university football coaches…,”

mused MIT Nuclear Reactor Laboratory

Director David Moncton at a

Dec. 2 ceremony attended by nearly

50 faculty, students and guests marking

the restart of the pile at 4:25 p.m.

EST, 75 years to the minute since

CP-1’s first criticality – and 60 years

since MIT’s pile was first assembled by

the university’s students.

The university intends to use the

unique assembly to teach students

about “real world” reactor physics

measurements and to conduct new

experiments – including some relevant

to advanced reactor designs, says

MIT’s Professor of the Practice of

­Nuclear Science and Engineering

Kord Smith.

“This is an extremely important

facility for teaching students about

measuring reactor parameters,” Smith

says. “This will give our students the

rare opportunity to handle and load

uranium fuel themselves.”

The MIT graphite reactor consists

of a 2.5-meter cubical pile of 30 metric

tons of stacked graphite rectangular

bars, with holes drilled at regular

­intervals to allow 2.5 metric tons of

natural uranium metal fuel rods to be

inserted. (Fermi chose the term “pile”

from the word “pila,” which means

stack in Italian.) With no moving

parts, the only other components of

the pile will be a plutonium-beryllium

or californium-252 neutron “source”

to drive the subcritical flux distribution,

a neutron-absorbing cadmium

rod to adjust subcritical reactivity, and

indium activation foils to measure the

spatial distribution of neutrons within

the pile.

Smith says Fermi’s original subcritical

experiments were built to

verify early nuclear physics theories

about the size and spacing of fuel

rods and the neutron slowing-down

or “moderating” material needed to

allow a neutron chain reaction to

become self-sustaining.

He explained that Fermi’s design

was “brilliantly simple,” allowing the

measurement of a single parameter –

the axial profile of neutrons in the pile

– to return information about how

close the assembly was to a selfsustaining

chain reaction and what

scale-up of pile dimensions was

needed in order for CP-1 to become

an actual critical reactor.

The simplicity of Fermi’s design

allowed MIT’s pile, like many others at

universities and laboratories around

the country, to be built in a month’s

time in 1957. However, with the

advent of more powerful water-cooled

reactors soon after, these teaching

tools soon fell into disuse and were

gradually forgotten. In fact, MIT’s

| | MIT Nuclear Reactor Laboratory Director David Moncton (L) with Associate

Department Head Jacopo Buongiorno, Professor of the Practice Kord Smith

and Professor Emeritus Neil Todreas in front of the graphite exponential

pile. (Photo: NEI, 4572)

graphite pile was “rediscovered” last

year, more or less hidden for decades

under its aluminum metal covers.

“We couldn’t believe the pile was

still here,” Smith says.

With the help of Moncton, departmental

colleagues and staff, Smith

restored the facility to working order

in time for the Dec. 2 restart.

Moncton, who operates the MIT

Reactor – the second-largest universitybased

research reactor in the country

– says both the U.S. Nuclear Regulatory

Commission and the U.S. Department

of Energy have been very helpful and

accommodating of MIT’s plans to

restart the subcritical graphite pile.

The university is awaiting an NRC

operating license, which hopefully will

be issued by the end of this year.

Once that happens, Smith expects

to use the pile for undergraduate and

graduate courses in the fundamentals

of reactor physics starting next year.

Among the activities in which the

students will be involved include

“testing of physics kernels of neutron

interactions within reactor-grade

graphite,” he says.

“Modeling and simulation are often

oversold by those who have never

done reactor measurements, and

students are beginning to believe that

News


atw Vol. 63 (2018) | Issue 1 ı January

58

NEWS

*)

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

1)

Refueling

2)

Inspection

3)

Repair

4)

Stretch-out-operation

5)

Stretch-in-operation

6)

Hereof traction supply

7)

Incl. steam supply

8)

New nominal

capacity since

January 2016

9)

Data for the Leibstadt

(CH) NPP will

be published in a

further issue of atw

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB

everything can be computed accurately,”

Smith explains. “How ever,

calculations are no better than the

[underlying] physics models. Graphite

piles are a great place to study the

physics of how neutrons interact in a

graphite-moderated system.”

Another application will be to

­model neutron fields in solid media

with large voids, with possible research

applications for graphite- moderated

advanced reactors and in test reactors

like Idaho National ­Laboratory’s Transient

Reactor Test Facility, also known

as the TREAT reactor – which was restarted

Nov. 15 after a 23-year operational

hiatus. The TREAT reactor in

turn will be used for tests that will support

the development of advanced

accident tolerant fuels for the U.S.

commercial reactor fleet.

In another echo of the past, Associate

Department Head Jacopo

Buongiorno sent away to Italy for a

bottle of Chianti, which was duly

signed by the 49 attendees at the

ceremony – just as Fermi and his

49 colleagues did 75 years ago.

| | www.nei.org, 4572

ONR (United Kingdom):

Regulators approve new

nuclear power station design

(onr) The UK Advanced Boiling Water

Reactor (UK ABWR), designed by

Hitachi-GE link to external website, is

suitable for construction in the UK,

the regulators confirmed following

completion of an in-depth assessment

of the nuclear reactor design.

The Office for Nuclear Regulation

(ONR), the Environment Agency link

to external website and Natural Resources

Wales link to external website,

the regulators who undertake the Generic

Design Assessment of new reactor

designs, are satisfied that this reactor

meets regulatory expectations on

safety, security and environmental

protection at this stage of the regulatory

process.

ONR has issued a Design Acceptance

Confirmation (DAC) and the

environment agencies have issued a

Statement of Design Acceptability

(SoDA) to Hitachi-GE.

Horizon Nuclear Power link to

external website is proposing to build

and operate two of these reactors in

Wylfa Newydd on Anglesey and

Oldbury- on-Severn near Thornbury

in South Gloucestershire.

Operating Results September 2017

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated. gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 720 649 662 5 631 282 252 863 137 100.00 96.19 99.80 94.83 99.16 94.46

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

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

KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0

KKB 2 Beznau 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

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

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

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

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

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

Dukovany B3 PWR CZ 500 473 0 0 2 309 273 101 934 129 0 82.69 0 71.14 0 70.50

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

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

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

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

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

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

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

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

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

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

Operating Results October 2017

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy utilisation

Energy generated, gross Time availability Energy availability

[MWh]

[%]

[%] *) [%] *)

Month Year Since Month Year Month Year Month Year

commissioning

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

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

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

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

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

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

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

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

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atw Vol. 63 (2018) | Issue 1 ı January

Mark Foy, ONR’s Chief Nuclear

Inspector said: “The completion of the

generic design assessment of the UK

ABWR is a significant step in our

regulation of the overall process to

construct this type of reactor in the

UK, ensuring that the generic design

meets the highest standards of safety

that we expect in this country. We’re

already working on our assessment

of Horizon’s site licence application

and on the development of the site

specific safety case to progress, in

due course, the construction and

operation of these reactors at Wylfa

Newydd.”

Dr Jo Nettleton, Deputy Director

for Radioactive Substances and Installations

Regulation at the Environment

Agency said: “We’ve concluded that

the generic design of the UK ABWR

should be capable of meeting the high

standards of environment protection

and waste management that we

require in the UK. We only came

to this conclusion after carefully reviewing

the submissions provided by

Hitachi- GE and their responses to the

questions and issues we raised. We’ve

also carefully considered all the comments

we received from people during

our public consultation and we’re

grateful for all who took part for

taking time to respond.”

Tim Jones, Natural Resources

Wales’s Executive Director for North

and Mid Wales, said: “It is our job to

ensure that any new nuclear power

station will meet high standards of environmental

protection and waste

management, ensuring that our communities

and environment are kept

safe.

“Following a public consultation

on our initial findings, we have

concluded that the UK ABWR design is

acceptable. We will now work on the

detailed assessments of the permits,

licences and consents that Horizon

Nuclear Power will need to have in

place to build Wylfa Newydd.”

The regulators have documented

progress of each stage of their assessment

through a series of reports on its

joint website.

| | www.onr.org.uk, 8874

IAEA conference says more

nuclear power needed to meet

global goals on climate change

(iaea) Nuclear power remains an

important option for many countries

to strengthen energy security and mitigate

the effects of global warming and

air pollution, but substantial growth in

its use is needed for the world to meet

its climate goals, according to an IAEA

international conference that concluded

in the United Arab Emirates.

The some 700 participants from 67

IAEA Member States and five international

organizations who attended

the event in Abu Dhabi this week

enjoyed a wide convergence of views,

Ambassador Hamad Alkaabi, president

of the International Ministerial Conference

on Nuclear Power in the 21 st Century,

said in his concluding statement.

“While respecting the right of each

State to define its national energy

policy, the Conference recognized that

nuclear power remains an important

option for many countries to improve

energy security, reduce the impact of

volatile fossil fuel prices and mitigate

the effects of climate change and air

pollution, including by backing up

intermittent energy sources,” Alkaabi,

the UAE’s Permanent Representative

to the IAEA in Vienna, said at the conference’s

closing session, attended by

IAEA Director General Yukiya Amano.

The three-day conference provided

a forum for high-level dialogue on

the role of nuclear power in the coming

decades. Nuclear power emits

virtually no greenhouse gases during

operation. It produces 11 percent of

the world’s electricity, which amounts

to one-third of all electricity generated

from low-carbon sources. Participants

noted that some 6.5 million

deaths a year are linked to air pollution,

with that number set to increase

significantly in the coming decades

in the absence of greater action to

curb emissions and expand access to

low-carbon energy.

To meet targets set out in the

Paris Agreement on climate change,

“substantial growth in nuclear

­electricity generation by 2050 will be

required,” Alkaabi said, citing the

International Energy Agency.

While nuclear power will play a key

role for many countries in achieving

the Sustainable Development goals

and reducing greenhouse-gas emissions,

“nuclear is not currently attracting

the necessary global investment” to

limit the average global temperature

increase to 2° C as required by the Paris

Agreement, he said. “In addition, a

number of plants are being shut down

in some countries before the end of

their safe operational lifetimes for both

political and economic reasons.”

The conference was the fourth

such ministerial event following previous

gatherings in Paris in 2005,

Beijing in 2009 and St. Petersburg in

2013. Organized in cooperation with

the Nuclear Energy Agency (NEA) of

the Organisation for Economic

| | Panellists at the International Ministerial Conference on Nuclear Power

in the 21 st Century, with the conference president, Ambassador Hamad

Alkaabi of the UAE, second from right. (Photo: D. Calma/IAEA, 8345)

Co- operation and Development, the

conference was hosted by the UAE

Government through the Ministry of

Energy and the Federal Authority for

Nuclear Regulation.

Ministers and senior officials from

IAEA Member States engaged in

discussions on issues including their

countries’ energy strategy and vision

for the role of nuclear power and challenges

to its introduction, continued

operation and expansion. In addition,

four panel sessions with selected

speakers from diverse backgrounds

discussed nuclear power and sustainable

development; challenges to

nuclear-power infrastructure development;

nuclear safety and reliability;

and innovations and advanced

nuclear technologies.

Alkaabi said participants widely

agreed on other key areas, including

the need to create an enabling environment

to facilitate the introduction

of nuclear power and ensure its safety

and sustainability; that nuclear power

is a safe, reliable and clean energy

option; and that “innovations in technology

design – including reactor size

– as well as in investment and ownership

models could facilitate the introduction

of nuclear power in more

countries.”

Small modular reactors currently

under development “may allow for

expanded use of nuclear power – including

on smaller grids and in remote

settings, as well as for non-electrical

applications – and improve access to

nuclear energy,” the ambassador said.

The conference repeatedly highlighted

the importance of public

­con­fidence for the future of nuclear

power. “Open and transparent decision

making involving all stakeholders

can improve the public perception of

nuclear power and lead to broader

public acceptance,” Alkaabi said.

In conclusion, participants recognized

the IAEA’s leading role in

promoting peaceful uses of nuclear

energy and supporting efforts to

strengthen global nuclear safety,

nuclear security and safeguards.

| | www.iaea.org, 8345

59

NEWS

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atw Vol. 63 (2018) | Issue 1 ı January

60

NEWS

World

France postpones plans to

reduce nuclear share after

warning of shortages

(nucnet) The French government has

postponed a target to reduce the share

of nuclear energy in the country’s

energy mix after grid operator RTE

warned it risked supply shortages after

2020 and could miss a goal to lower

carbon emissions. In 2015 the previous

government of Francois Hollande

established an energy transition law

which set a target of reducing the

share of nuclear in the energy mix to

50% by 2025 from the current 75%.

But environment minister Nicolas

Hulot said on 8 ­November 2017 this

would not be realistic. He said reducing

the nuclear share in a hurry

would increase France’s CO 2 emissions,

endanger the security of power

supply and put jobs at risk. Mr Hulot

said president Emmanuel Macron’s

government remains committed to

reducing nuclear energy and ordered

his ministry to produce a new timetable.

He later said in a television

interview that the government would

be working towards a 2030 to 2035

timeframe. RTE said in its 2017-2035

Electricity Outlook that if France went

ahead with plans to simultaneously

shut down four 40-year-old nuclear

­reactors and all its coal-fired plants,

there could be risks of power supply

shortages. State-controlled utility

EDF, which operates France’s 58

commercial nuclear power plants, has

argued instead to extend the operation

of its nuclear fleet from 40 to at least

50 years. France is the second largest

generator of nuclear electricity behind

the US. According to the International

Atomic Energy Agency, France’s

­nuclear fleet produced almost 28% of

the country’s electricity in 2016.

| | www.gouvernement.fr, 7763

Bill Gates’ TerraPower forms

new company with China to

develop twr technology

(nucnet) TerraPower, the company

founded in 2008 to develop advanced

nuclear technology and backed by

Microsoft founder Bill Gates, has

signed a joint venture with China

­National Nuclear Corporation (­CNNC)

to form a company that will work to

complete the Travelling Wave Reactor

(TWR) design and commercialise TWR

technology. TerraPower said on its

website that the formation of the new

company, Global Innovation ­Nuclear

Energy Technology Company Ltd, was

made possible under policies and

agreements signed by the governments

of the US and China. Terra Power said

the collaboration with ­CNNC aims to

pioneer new options in civilian nuclear

energy that address safety, environmental

and cost concerns. Unlike traditional

nuclear reactors, TWR technology

will be capable of using fuel made

from depleted uranium, which is currently

a waste byproduct of the

uranium enrichment process. Its

unique design gradually converts the

fuel through a nuclear reaction without

removing it from the reactor’s core,

eliminating the need for reprocessing.

This means the reactor can generate

heat and produce electricity over a

much longer period of continuous

operation. Additionally, eliminating

reprocessing reduces proliferation

concerns, lowers the overall cost of the

nuclear energy process, and helps to

protect the environment by making use

of a waste by-product and reducing the

production of greenhouse gases. On

3 November 2017 in Beijing, Mr Gates

met the premier of China’s state council,

Li Keqiang, to discuss increased

cooperation between China and the

US in the development of the next

generation of reactor technologies.

| | terrapower.com, 8832

Barakah project brought $ 3.3 bn

of economic benefit to UAE

(nucnet) More than 1,400 local companies

have been contracted in the

development of the United Arab

­Emirates’ first nuclear power station

project at Barakah, Mohamed Al-

Hammadi, chief executive officer of

the Emirates Nuclear Energy Corporation

(Enec), told an International

Atomic Energy Agency conference in

Abu Dhabi. Mr Al-Hammadi told the

International Ministerial Conference

on Nuclear Power in the 21st Century

that the construction of Barakah

brought over $3.3bn (€2.8bn) worth of

contracts to UAE-based companies,

| | Barakah project brought $ 3.3 bn of economic

benefit to UAE. View of the Barakah

construction site in September 2017.

(Courtesy: ENEC, 8877)

providing economic benefits to the

Gulf country. Enec signed a contract

with Korea Electric Power Corporation

in 2009 for building four APR-1400

units at the Barakah station. Construction

of the units began in 2012. Enec

said yesterday that Unit 1 at Barakah is

now more than 96% complete, Unit 2

more than 87%, Unit 3 more than 78%

and Unit 4 more than 58%. Overall,

construction of the four units is more

than 84% complete.

| | www.enec.gov.ae, 8877

Dominion to apply for second

life extension at North Anna

Nuclear Station – 80 operation

years advised

(nucnet) Dominion Energy Virginia has

notified the US Nuclear ­Regulatory

Commission that it intends to apply for

a second 20-year life extension for the

twin-reactor North Anna nuclear

power station in Virginia. The company

said it would file a licence renewal application

with the NRC in 2020, following

a similar application to extend the

operating lifetime of two reactors at

the Surry nuclear station, also in

Virginia, to 80 years. Dominion said it

expects to invest up to $4bn (€3.3bn)

in upgrades to the two North Anna

units and the two Surry units as

part of the relicensing process. The

Washington-­based Nuclear Energy

Institute said that of the 99 commercial

nuclear power reactors operating in

the US, 84 have had their original

40-year operating licences extended to

60 years. Three others that were issued

licence renewals have since shut down.

Another seven applications are under

NRC review, and the remaining four

are expected to apply between 2020

and 2022. By 2040, half of the nation’s

nuclear plants will have been operating

for 60 years. Under its second

licence renewal programme, the

industry is planning for a second round

of licence renewals to allow operation

out to 80 years.

| | www.dominion.com, 3882

Household energy prices

in the EU down compared

with 2016

(eurostat) In the European Union

(EU), household electricity prices

slightly decreased (-0.5%) on average

between the first half of 2016 and the

first half of 2017 to stand at €20.4 per

100 kWh. Across the EU Member

States, household electricity prices in

the first half of 2017 ranged from

­below €10 per 100 kWh in Bulgaria to

more than €30 per 100 kWh in

Denmark and Germany.

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atw Vol. 63 (2018) | Issue 1 ı January

| | First concrete poured for unit 1 at Bangladesh’s

Rooppur. Artist’s view of the site

with two reactors. (Courtesy: Rosatom, 7745)

Highest increases in electricity

prices in Cyprus, Greece and

Belgium, largest falls in Italy,

Croatia and Lithuania

Across the EU Member States, the

highest increase in household electricity

prices in national currency

­between the first half of 2016 and

the first half of 2017 was registered

by far in Cyprus (+22.0%), followed

by Greece (+12.8%), Belgium

(+10.0%), Poland (+6.9%), Sweden

(+5.5%) and Spain (+5.1%). In contrast,

the most noticeable decreases

were observed in Italy (-11.2%),

Croatia (-10.2%) and Lithuania

(-9.3%), well ahead of Luxembourg

(-4.9%), Austria (-4.1%), Romania

(-4.0%) and the Netherlands (-3.6%).

Expressed in euro, average household

electricity prices in the first half

of 2017 were lowest in Bulgaria (€9.6

per 100 kWh), Lithuania (€11.2) and

Hungary (€11.3) and highest in

­Denmark and Germany (both €30.5)

followed by Belgium (€28.0). The

average electricity price in the EU

was €20.4 per 100 kWh.

When expressed in purchasing

power standards (PPS), an artificial

common reference currency that

eliminates general price level differences

between countries, it can be seen

that, relative to the cost of other goods

and services, the lowest household

electricity prices were found in Finland

(12.8 PPS per 100 kWh), Luxembourg

(13.5) and the Netherlands (14.2),

and the highest in Germany (28.7),

­Portugal (28.6), Poland (25.9),

­Belgium (25.6) and Spain (25.4).

Half or more of the electricity

price is made up of taxes and

levies in Denmark, Germany and

Portugal

The share of taxes and levies in total

household electricity prices varied

significantly between Member States,

ranging from two-thirds in Denmark

(67% of household electricity price is

made up of taxes and levies) and over

half in Germany (54%) and Portugal

(52%) to 5% in Malta in the first half

of 2017. On average in the EU, taxes

and levies accounted for more than a

third (37%) of household electricity

prices.

| | ec.europa.eu, 8921

Reactors

Argentina to start construction

of two new reactors

(nucnet) Argentina plans to start construction

of two new nuclear reactor

units in the second half of 2018,

Argentina’s undersecretary for nuclear

energy Julian Gadano told Reuters.

Mr Gadano said Argentina is in the process

of finalising negotiation of the

commercial and financial contracts to

build the two plants. In May 2017,

­Argentina signed a $12.5bn (€10.7bn)

agreement with China for the construction

and financing of two nuclear power

units. According to the agreement,

China’s National Nuclear Corporation

and Nucleoeléctrica Argentina will begin

construction of Atucha-3, a 700-

MW Candu-6 pressurised heavy water

reactor (PHWR), in 2018 and will start

building a 1,000-MW Hualong One, or

HPR1000, pressurised- water reactor

unit in 2020. Argentina has three operating

commercial power reactors – a

Candu unit at the Embalse nuclear

station and two PHWRs at Atucha.

Under the May 2017 contract, China

agreed to provide a long term-loan for

85% of the required financing, which

will be repaid when the plants begin

generating electricity, according to

comments at the time by Mr Gadano.

| | www.na-sa.com.ar, 3345

First concrete poured for unit

1 at Bangladesh’s Rooppur

(nucnet) First concrete was poured on

30 November 2017 for the nuclear

island basemat of Unit 1 at the planned

Rooppur nuclear power station in

Bangladesh, Russian state-owned

nuclear corporation Rosatom said

in a statement. The ceremony was attended

by Rosatom’s director- general

Alexey Likhachev and the prime minister

of Bangladesh Sheikh Hasina, the

statement said. In October 2013, Russia

signed an agreement with Bangladesh

for design work on Rooppur, on

the banks of the Ganges river about

160 km from the Bangladeshi capital

Dhaka. In 2014, Rosatom said the

Rooppur units – the first nuclear power

reactors in Bangladesh – would both

be 1,200-MW V-392M pressurised water

reactors. According to Rosatom,

the first unit at Rooppur is scheduled

to begin commercial operation in 2023

with the second unit following in

2024. In July 2017, Russia agreed to

release a state loan to finance the construction

of the bulk of the Rooppur

project. No ­mention was made of the

amount of the loan, but earlier media

reports put it at $12.6bn (€10.6bn).

According to earlier reports, first concrete

for Unit 1 at Rooppur was expected

to be laid in December 2017.

| | www.rosatom.ru, www.baec.gov.bd,

7745

Bulgaria extends Kozloduy-5

operating licence by 10 years

(nucnet) The operating licence for

Unit 5 at the Kozloduy nuclear power

station in Bulgaria has been extended

by 10 years until 2027, the country’s

energy ministry said. The 963-MW

VVER V-320 unit, which began commercial

operation in December 1988,

could operate until 2047, the ministry

said, but a 10-year extension is the

longest allowed under Bulgarian law.

Its existing operating licence was due

to expire this month. Bulgaria has two

nuclear units in commercial operation,

Kozloduy-5 and Kozloduy-6.

They are both Russian-designed

VVERs and produce about 33% of the

country’s electricity. The operating

licence for Kozloduy-6 expires in

August 2019. Extending the life of the

two units is a priority for Bulgaria’s

government, energy minister Temenuzhka

Petkova said. Lachezar Kostov,

the head of the Bulgarian Nuclear

Regulatory Agency, said last year that

the main tasks for Bulgaria’s nuclear

energy sector are lifetime extensions

at Kozloduy-5 and -6, modernisation

of the two units by increasing

their capacity, construction of a new

unit at Kozloduy, and development of

a national repository for low- and

medium-level radioactive waste.

| | www.kznpp.org, 8834

Excavation of foundation pit

begins at Iran’s Bushehr-2

(nucnet) Excavation of the foundation

pit for Iran’s Bushehr-2 nuclear power

plant began on 31 October 2017,

Russian state-owned nuclear corporation

Rosatom said in a statement.

The start of work was given in a

ground-breaking ceremony attended

by Rosatom’s director-general Alexey

Likhachev and Ali Akbar Salehi, head

of the Atomic Energy Organisation of

Iran, the statement said. In March

2017, construction work formally

began at Bushehr-2, a pressurised

water reactor unit of the Russian

61

NEWS

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atw Vol. 63 (2018) | Issue 1 ı January

62

NEWS

VVER-1000 design. In September 2017,

Rosatom said site preparation works

had begun for Bushehr-2 and -3, both

of the same Russian design. Rosatom

said at the time that first concrete

for Bushehr-2 was planned for the

third quarter of 2019. Construction of

Bushehr-2 is expected to be completed

in 2024 and of Bushehr-3 in 2026. Iran

and Russia signed an agreement to

build two additional units at Bushehr

in November 2014. Official media in

Iran said the construction of Bushehr-2

and -3 would cost about $10bn

(€ 8.6bn). Bushehr-1 is Iran’s only

commercial nuclear unit. It is a 915-

MW pressurised-water reactor which

was supplied by Russia and began commercial

operation in September 2013.

| | www.rosatom.ru, 8872

Governor approves restart

of Japan’s Ohi-3 and -4

(nucnet) The governor of Fukui

Prefecture in southwest Japan has

approved the restart of the Ohi-3 and

-4 nuclear reactor units, operator

Kansai Electric Power Company said

on 27 November 2017. His decision

clears the final regulatory hurdle for

the restarts of both units early next

year. Ohi-3 is a 1,127-MW pressurised

water reactor that began commercial

operation in 1991. Ohi-4, also a

1,127-MW PWR, began commercial

operation in 1993. All of Japan’s 48

reactors were shut between 2011 and

2012 after the March 2011 Fukushima-

Daiichi accident. Five units have resumed

commercial operation after

meeting revised regulatory standards.

They are: Takahama-3 and -4, Ikata-3

and Sendai-1 and -2. According to the

Japan Atomic Industrial Forum, 12

nuclear units at six sites have now

been approved as meeting new regulatory

standards introduced following

the accident. Ohi-3 and -4 were the

first two reactors to resume operation

in Japan following the Fukushima-

Daiichi accident, but were both taken

offline in September 2013 for scheduled

refuelling and maintenance. But

| | Finland’s Posiva makes progress with

final repository excavation works. Artist’s

view of the encapsulation plant.

(Courtesy: Posiva, 8871)

restarts where delayed when, in May

2014, the Fukui district court ruled

that it would not allow Ohi-3 and -4 to

return to operation. A lawsuit filed by

a group of almost 200 people living

within a 250km radius of the Ohi station

claimed that the plant was sited

near several active seismic faults and

was not adequately protected against

earthquakes. Kansai Electric appealed

the decision and it was overturned by

a higher court in March 2017.

| | www.kepco.co.jp, 8871

Energoatom and Toshiba to

cooperate on modernisation

of Ukraine nuclear plants

(nucnet) Ukraine’s state-owned

nuclear operator Energoatom and

Japan-based Toshiba have signed an

agreement to cooperate on the modernisation

of turbine island equipment

at Ukrainian nuclear power

stations. Energoatom said the modernisation

aims to increase the power

output and efficiency, and improve

the safety of Ukraine’s plants. The

agreement will increase cooperation

in the long-term servicing of existing

plant equipment, a statement said.

Energoatom said a committee will be

formed to ensure the implementation

of the agreement. According to the

International Atomic Energy Agency,

Ukraine has 15 reactors in commercial

operation which produced 52% of the

country’s electricity in 2016.

| | www.energoatom.kiev.ua, 8834

Completion of Vogtle units

is best economic choice

(nucnet) Completing the Vogtle-3 and

-4 AP1000 nuclear reactor units represents

the best economic choice for

customers and preserves the benefits

of carbon-free, baseload generation

for the state of Georgia, Georgia

Power chairman, president and chief

executive officer Paul Bowers told a

Georgia Public Service Commission

(PSC) hearing into the project on

7 November 2017. Mr Bowers said. All

the project owners – Georgia Power,

Oglethorpe Power, MEAG Power and

Dalton Utilities – have agreed to continue

with the project. This decision

was based on the results of a schedule,

cost and cancellation assessment that

was prompted by the bankruptcy of

Westinghouse, supplier of the AP1000

technology being used for the plants.

Mr Bowers said assessments of

the project have included economic

analysis, evaluation of various alternatives

including abandoning one or

both units, and assumptions related to

potential risks. The Georgia PSC will

hear from owners and partners in the

project as well as public witnesses.

The PSC will issue its final recommendation

on 6 February 2018. Mr Bowers

said construction has continued

uninterrupted at the Vogtle site over

the past six months. Southern ­Nuclear,

the nuclear operating subsidiary

which operates the existing units

at the Georgia station, is now the

project manager at the site. Bechtel is

managing daily construction efforts.

| | www.georgiapower.com, 8432

Waste Management

Finland’s Posiva makes

progress with final repository

excavation works

(nucnet) Finnish nuclear waste

manage ment company Posiva has

completed the excavations for the

­encapsulation plant at the final deep

geologic disposal facility under construction

at Olkiluoto, Posiva’s owner

Teollisuuden Voima Oyj (TVO) said in

a statement. Excavation works began

in October 2016. TVO said Posiva has

also made progress with excavation

work for the vehicle access tunnels

leading to the final disposal facility

­itself. TVO said the first phase of excavations

for the final disposal facility is

estimated to take two and a half years.

In December 2016, Posiva was given

regulatory approval to begin construction

of a deep geologic repository at

Olkiluoto on the country’s southwest

coast – the first final repository in the

world to enter the construction phase.

| | www.posiva.fi, 8871

Research

Wendelstein 7-X now ready

for virtual tours!

(ipp-mpg) The new 360-degree panorama

featured on the internet pages of

Max Planck Institute for Plasma

Physics (IPP) leads right into the

plasma vessel of the Wendelstein 7-X

fusion research device at Greifswald.

The address www.ipp.mpg.de/

panoramaw7x takes observers on an

extraordinary tour to the core of the

device, otherwise accessible only to

experts; they can stroll through the

experimentation hall and view the

facilities that heat the plasma to many

millions of degrees.

By way of PC, tablet or smartphone

they can cast an eye at every angle and

zoom in on even tiniest details. Short

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atw Vol. 63 (2018) | Issue 1 ı January

videos in which IPP scientists describe

their workplaces are started and

stopped by mouse click; info panels

can be slotted in to explain important

components. The panorama was

recorded by Munich photographer

Volker Steger, who had already done

the panorama of IPP Garching’s

ASDEX Upgrade fusion device

(www.ipp.mpg.de/panorama).

The objective of IPP’s research is a

fusion power plant to derive energy

from fusion of light atomic nuclei, just

as the sun does. At Garching IPP also

operates the ASDEX Upgrade experiment,

a large-scale device of the tokamak

type. IPP’s branch institute at

Greifswald is conducting research on

the large Wendelstein 7-X stellarator.

As of now both devices are accessible

at any time for a virtual tour.

| | www.ipp.mpg.de, 8892

High performance computing

for energies with EoCoE

(cea) Computer simulation being an

amazing driver of innovation, it is

strategic for Europe to develop supercomputing

resources at the most

advanced level. The UE provides

support to supercomputing infrastructures

(PRACE1), hardware (ETP4H-

PC2) as well as software technologies.

The support to application software

development is spread over nine thematic

centers of excellence – EoCoE

being one of them.

Predicting wind and sunlight intensity,

designing innovative materials to

store electricity, optimizing the management

of water reservoirs, predicting

the performance of geothermal plants

or even stabilizing plasma in a nuclear

fusion reactor are essential tasks to

master in order to accomplish a successful

diversification of the energy mix.

Dedicated to low-carbon energies,

EoCoE, which stands for ‘European

Energy-Oriented Center of Excellence’

(and can be pronounced as “Echo”),

targets the fields of weather forecast,

materials, water management and

nuclear fusion—all of which require

high calculation capacities. The center

brings together twenty-two partners

from eight European countries,

involved in both HPC and energies,

committed to tackling the challenges

in these fields.

“Computer simulation is driven by

the constant upgrades of high-performance

computers,” said Edouard

Audit, the CEA Director of Maison de

la simulation3 and coordinator of

EoCoE. “Yet the challenge is not so

much to gain time than to achieve

things that were previously

inacces sible. In materials science, for

instance, it is now possible to digitally

test a very large number of materials.”

Exascaling the future

The mission of a laboratory such as

Maison de la simulation is to develop

cutting-edge digital tools in close

collaboration with scientists from the

related disciplines, as well as transversal

tools such as linear algebra, input/

output data management, and result

visualization. “We provide support to

researchers as they develop their

own code to help them achieve the

expected result. The help we offer

ranges from applied mathematics to

algorithms and HPC” Mr. Audit

explained. “Meanwhile, we are also

preparing for the future, that is to say

the development of exascale architectures

(1018 operations per second),

that are massively multi-core. They

differ from previous architectures by

the fact that now, not all their processors

are of the same nature. This is

why we must change the way we compute—and

how we manage memory

storage in particular.”

First concrete achievements

Several significant advances have

already been achieved thanks to

EoCoE. During the working sessions,

the scientists learn to “instrument”

their simulation code to monitor the

results step by step, and optimize them.

For nuclear fusion, the Gysela code

developed at CEA (IRFM4) describes

ion transport in plasma inside the

reactor’s toric chamber (tokamak). In

addition to being necessary for the

R&D activities of tokamaks WEST

(CEA) and ITER in Cadarache, this

code also deepens the fundamental understanding

that physicists have of fusion

plasma turbulence. It is now suitable

for hundreds of thousands of computing

cores. The meticulous audit

work accomplished within EoCoE has

saved 10 % in computing time and has

helped prepare for the future upgrade

to the exascale.

| | www.cea.fr, 9983

Company News

MATRIX by Areva TN: a game

changer in used fuel dry storage

(areva) AREVA TN, the ­nuclear

­logistics affiliate of New ­AREVA, is

launching an advanced used nuclear

fuel storage overpack, ­NUHOMS®*

MATRIX. With its improved capacity

and performance, NUHOMS® ­MATRIX

addresses the challenges faced by our

customers when it comes to storing

used fuel ­safely, ­efficiently and competitively.

The unique 2-level horizontal and

modular set-up reduces the inde pendent

spent fuel storage installation

(­ISFSI) footprint by 45% which in turn

reduces pad construction costs. This

makes NUHOMS® MATRIX the smallest

storage pad on the market for the

same capacity, in a context where space

is at a premium on nuclear sites. Its

design accommodates canisters of

different sizes and it can store high burnup

short cooled fuel, which is of particular

interest for shutdown nuclear

reactors. New features and devices

allow for the complete inspection of the

canister without removing it from the

module, as aging management and

retrieval of the canister for future transport

to a consolidated storage site have

become a challenge for utilities.

NUHOMS® storage systems

securely store the dry fuel storage

containers in a horizontal position

within a sturdy, low-profile, reinforced

concrete structure. This fortress-like

structure serves as a robust barrier.

“As more communities, policymakers

and utilities across the world

discuss securely storing used nuclear

fuel, our NUHOMS® MATRIX system is

a competitive, safe and timely solution

for those needs and concerns,” said

Greg Vesey, president, TN Americas.

With more than 1,250 dry storage

systems loaded worldwide, AREVA TN

offers its customers an unrivaled

experience for the management of

used fuel.

| | www.new.areva.com, 4532

People

Camilla Hoflund new

President and CEO of Studsvik

(studsvik) The current President and

CEO Michael Mononen and the Board

of Directors have together concluded

that a changeover in the chief executive

post is appropriate after the major

changes in the Group that have been

made in recent years. Studsvik’s Board

of Directors has therefore appointed

Camilla Hoflund as new President and

CEO from January 1, 2018.

Camilla Hoflund is a mining

engineer from the Royal Institute of

Technology (KTH) and has been head

of Studsvik’s Fuel and Materials

Technology business area since 2014.

She has worked at Studsvik since

1994, with a short break in 2000-2003

* NUHOMS: Nuclear

Horizontal Modular

Storage

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when she was a business developer

and consultant for risk management

services at Det Norske Veritas.

| | www.studsvik.com, 983

Westinghouse appoints

Ken Canavan Chief

Technology Officer

(westinghouse) Westinghouse Electric

Company announced that Ken

Canavan has been appointed chief

technology officer (CTO), effective

January 2, 2018.

Westinghouse’s CTO role has strategic

responsibility to drive nextgeneration

technology and innovation

solutions that align with the com pany’s

global business strategy. Canavan will

lead these efforts, as well as strengthen

Westinghouse with regard to technology

leadership development.

Canavan, 53, previously was director

of engineering for the Electric Power

Research Institute (EPRI). There he

was responsible for turning industry

needs into compelling research and

development plans. These plans improved

safety and performance of the

global nuclear fleet. He has more than

30 years of experience in key engineering

and risk management roles. Prior

to his work at EPRI, Canavan was

responsible for risk applications at

­Data Systems and Solutions, ERIN

Engineering and Research and GPU

Nuclear. He also was a safety analysis

engineer with Davis-Besse Nuclear

Power Station in Ohio (USA).

Canavan has a bachelor’s degree in

chemical engineering, with a nuclear

engineering minor, from Manhattan

College, New York. Ken, his wife,

Paula and his two children will relocate

to the Pittsburgh area.

| | www.westinghousenuclear.com,

8831

WANO Nuclear Excellence

Awards 2017

(wano) At the closure of its fourteenth

Biennial General Meeting held in

Gyeongju, the World Association of

Nuclear Operators (WANO) tonight

acknowledged the outstanding contribution

made by nine nuclear professionals

to promote excellence in the

safe operation of commercial nuclear

power.

| | WANO Nuclear Excellence Awards 2017 (873)

The honorary awards were established

in 2003 to recognise individuals

who have made extraordinary contributions

to excellence in the operation

of nuclear power plants, or the infrastructure

that supports the nuclear

power enterprise, or through WANO.

Potential award recipients undergo

a rigorous nomination and selection

process before being approved. The

awards are presented during each

WANO Biennial General Meeting.

This year’s award recipients are:

Brian Cowell, EDF Energy; Bum-nyun

Kim, Korea Hydro & Nuclear Power

Company (KHNP); Pavlo Pavlyshyn,

Rivne Nuclear Power Plant, NNEGC

Energoatom; Pierre Pilon, Bruce

­Power; Philippe Sasseigne, Électricité

de France; Debbie Sims, WANO ­Atlanta

Centre; Jouko Turpeinen, Fortum

Power and Heat Oy; Jean Van Vyve,

ENGIE Electrabel; Makoto Yagi, The

Kansai Electric Power Company, Inc.

| | www.wano.info, 873

Publications

Nuclear Energy Data – 2017

(nea) Nuclear Energy Data is the

­Nuclear Energy Agency’s annual compilation

of statistics and country

reports documenting nuclear power

status in NEA member countries and in

the OECD area. Information provided

by governments includes statistics on

total electricity produced by all sources

and by nuclear power, fuel cycle capacities

and requirements, and projections

to 2035, where available. Country

reports summarise energy policies,

updates of the status in nuclear energy

programmes and fuel cycle developments.

In 2016, nuclear power continued

to supply significant amounts

of low-carbon baseload electricity,

despite strong competition from lowcost

fossil fuels and subsidised renewable

energy sources. Three new units

were connected to the grid in 2016, in

Korea, Russia and the United States. In

Japan, an additional three reactors

returned to operation in 2016, bringing

the total to five under the new regulatory

regime. Three reactors were

­officially shut down in 2016 – one in

Japan, one in Russia and one in the

United States. Governments committed

to having nuclear power in the energy

mix advanced plans for developing or

increasing nuclear generating capacity,

with the preparation of new build projects

making progress in Finland,

Hungary, Turkey and the United Kingdom.

Further details on these and

other developments are provided in

the publication’s numerous tables,

graphs and country reports. Download

the report at oe.cd/nea-data-2017

| | www.oecd-nea.org, 3342

Market data

(All information is supplied without guarantee.)

Nuclear Fuel Supply

Market Data

Information in current (nominal)

­U.S.-$. No inflation adjustment of

prices on a base year. Separative work

data for the formerly “secondary

market”. Uranium prices [US-$/lb

U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =

0.385 kg U]. Conversion prices [US-$/

kg U], Separative work [US-$/SWU

(Separative work unit)].

January to December 2013

• Uranium: 34.00–43.50

• Conversion: 9.25–11.50

• Separative work: 98.00–127.00

January to December 2014

• Uranium: 28.10–42.00

• Conversion: 7.25–11.00

• Separative work: 86.00–98.00

January to June 2015

• Uranium: 35.00–39.75

• Conversion: 7.00–9.50

• Separative work: 70.00–92.00

June to December 2015

• Uranium: 35.00–37.45

• Conversion: 6.25–8.00

• Separative work: 58.00–76.00

2016

January to June 2016

• Uranium: 26.50–35.25

• Conversion: 6.25–6.75

• Separative work: 58.00–62.00

July to December 2016

• Uranium: 18.75–27.80

• Conversion: 5.50–6.50

• Separative work: 47.00–62.00

2017

January 2017

• Uranium: 20.25–25.50

• Conversion: 5.50–6.75

• Separative work: 47.00–50.00

February 2017

• Uranium: 23.50–26.50

• Conversion: 5.50–6.75

• Separative work: 48.00–50.00

March 2017

• Uranium: 24.00–26.00

• Conversion: 5.50–6.75

• Separative work: 47.00–50.00

April 2017

• Uranium: 22.50–23.50

• Conversion: 5.00–5.50

• Separative work: 45.50–48.50

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atw Vol. 63 (2018) | Issue 1 ı January

May 2017

• Uranium: 19.25–22.75

• Conversion: 5.00–5.50

• Separative work: 42.00–45.00

June 2017

• Uranium: 19.25–20.50

• Conversion: 5.55–5.50

• Separative work: 42.00–43.00

July 2017

• Uranium: 19.75–20.50

• Conversion: 4.75–5.25

• Separative work: 42.00–43.00

August 2017

• Uranium: 19.50–21.00

• Conversion: 4.75–5.25

• Separative work: 41.00–43.00

September 2017

• Uranium: 19.75–20.75

• Conversion: 4.60–5.10

• Separative work: 40.50–42.00

October 2017

• Uranium: 19.90–20.50

• Conversion: 4.50–5.25

• Separative work: 40.00–43.00

November 2017

• Uranium: 19.90–20.50

• Conversion: 4.50–5.25

• Separative work: 40.00–43.00

| | Source: Energy Intelligence

www.energyintel.com

Cross-border Price for Hard Coal

Cross-border price for hard coal in

[€/t TCE] and orders in [t TCE] for

use in power plants (TCE: tonnes of

coal equivalent, German border):

2012: 93.02; 27,453,635

2013: 79.12, 31,637,166

2014: 72.94, 30,591,663

2015: 67.90; 28,919,230

2016: 67.07; 29,787,178

I. quarter: 56.87; 8,627,347

II. quarter: 56.12; 5,970,240

III. quarter: 65.03, 7.257.041

IV. quarter: 88.28; 7,932,550

2017:

I. quarter: 95.75; 8,385,071

II. quarter: 86.40; 5,094,233

III. quarter: 88.07; 5,504,908

| | Source: BAFA, some data provisional

www.bafa.de

EEX Trading Results

November 2017

(eex) In November 2017, the ­European

Energy Exchange (EEX) achieved a

total volume of 276.6 TWh on its

­power derivatives markets (November

2016: 423.2 TWh). The November

volume comprised 163.8 TWh traded

at EEX via Trade Registration with

subsequent clearing. Clearing and

settlement of all exchange transactions

was executed by European

Commodity Clearing (ECC).

| | Uranium spot market prices from 1980 to 2017 and from 2007 to 2017. The price range is shown.

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.

| | Separative work and conversion market price ranges from 2007 to 2017. The price range is shown.

)1

In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.

The shift of liquidity from the

German- Austrian power futures into

the German Phelix DE future continued,

resulting in a new record

­volume of 72,651,053 MWh traded in

the German contract.

The Settlement Price for base load

contract (Phelix Futures) with delivery

in 2018 amounted to 37.60 €/MWh.

The Settlement Price for peak load contract

(Phelix Futures) with delivery

in 2018 amounted to 45.70 €/MWh.

On the EEX Market for emission

allowances, 144.3 million tonnes of

CO 2 (November 2016: 79.2 million

tonnes of CO 2 ) were traded in

­November. The total volume increased

by 82%. Primary market auctions contributed

81.9 million tonnes of CO 2 to

the total volume. On the emission

­derivatives market 57.9 million tonnes

of CO 2 were traded which is more than

three times the volume of the same

month of the previous year (November

2016: 18.5 million tonnes of CO 2 ).

The E-Carbix amounted to

7.57 ­€/EUA, the EUA price with

delivery in December 2017 amounted

to 7.35/7.92 €/ EUA (min./max.).

| | www.eex.com

MWV Crude Oil/Product Prices

October 2017

(mwv) According to information and

calculations by the Association of the

German Petroleum Industry MWV e.V.

in October 2017 the prices for super

fuel, fuel oil and heating oil noted

inconsistent compared with the previous

month September 2017. The

average gas station prices for Euro

­super consisted of 134.72 €Cent

(­September 2017: 137.12 €Cent,

­approx. -1.75 % in brackets: each

information for pre vious month or

rather previous month comparison),

for diesel fuel of 116.19 €Cent (114.36;

+1.60 %) and for heating oil (HEL)

of 57.07 €Cent (55.84, +2.20 %).

The tax share for super with

a ­consumer price of 134.72 €Cent

(137.12 €Cent) consisted of

65.45 €Cent (48.58 %, 65.45 €Cent)

for the current constant mineral oil

tax share and 21.51 €Cent (current

rate: 19.0 % = const., 21.89 €Cent)

for the value added tax. The product

price (notation Rotterdam) consisted

of 36.20 €Cent (26.87 %, 37.79 €Cent)

and the gross margin consisted of

11.74 €Cent (8.74 %; 11.99 €Cent).

Thus the overall tax share for super

results of 67.58 % (66.73 %).

Worldwide crude oil prices

(monthly average price OPEC/Brent/

WTI, Source: U.S. EIA) were again

approx. +3.27 % (+6.68 %) higher in

September compared to September

2017.

The market showed a stable development

with higher prices; each in

US-$/bbl: OPEC basket: 55.5 (53.44);

UK-Brent: 57.51 (56.15); West Texas

Inter­mediate (WTI): 51.58 (49.82).

| | www.mwv.de

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66

NUCLEAR TODAY

Links to

reference sources:

Bangladesh

new nuclear project:

http://bit.ly/2BxD8z7

UK nuclear skills

warning:

http://on.ft.com/

2iIIML8

Author

John Shepherd

nuclear 24

41a Beoley Road West

St George’s

Redditch B98 8LR,

United Kingdom

‘Newcomer’ Nuclear Nation

Leads Way into New Nuclear Year

John Shepherd

At the start of a new year, it is appropriate that a ‘newcomer’ nuclear nation has launched work on building its first

nuclear power plant. First nuclear safety-related concrete has been poured for the plant at Rooppur in Bangladesh –

­making the South Asia nation the first in 30 years to start building its first commercial reactor unit following the United

Arab Emirates in 2012 and Belarus in 2013.

In Bangladesh, it is Russia’s Atomstroyexport that has been

selected to build two VVER type (AES-2006) pressurised

water reactors, each with a 1,200 MW(e) gross electricity

generating capacity. The units are expected to be commissioned

in 2023 and 2024 respectively.

In addition to supporting the country’s increasing

electricity needs, the reactors will “transform Bangladesh

into a middle income country” and a developed one by

2041, said Prime Minister Sheikh Hasina.

Despite setbacks that nuclear has endured in recent

years, there are nearly 60 reactors under construction

around the world, mostly in Asia, according to the International

Atomic Energy Agency (IAEA). Some 447 commercial

reactor units are in operation in 30 countries.

IAEA director-general Yukiya Amano told the recent

fourth International Ministerial Conference on Nuclear

Power in the 21 st Century in the United Arab Emirates that

the agency’s latest projections showed the global potential

for nuclear energy up to 2050 continues to be high,

­although figures show expansion is likely to slow.

Amano warned: “It is difficult to see other low-carbon

energy sources growing sufficiently to take up the slack if

nuclear power use fails to grow.”

But there is cause for optimism, beyond Bangladesh, as

a new nuclear year gets under way. Key developments to

look forward to include a review of the role of nuclear in

France, following a long-overdue acceptance, of sorts, that

the obsession of former president François Hollande to

­reduce the national nuclear share to 50 % by 2025 from

the current 75 % was flawed.

France’s grid operator RTE had warned that the country

faced potential supply shortages beyond 2020 – in addition

to increasing CO 2 emissions – if nuclear power were rolled

back. The new administration of President Emmanuel

Macron has chosen to fudge the issue, by saying it remains

committed to reducing nuclear’s role. A new “timetable” to

reduce the nuclear share is being drawn up and environment

minister Nicolas Hulot has indicated that the government

is now considering a period of 2030 to 2035. Therefore,

it will be for a future leader of France to potentially

revisit the issue.

Another highlight of this new nuclear year will be in

Pakistan, which is set to see construction start on a Chinese

Generation III HPR1000 Hualong One reactor at the

country’s Chashma nuclear power plant. This follows a

cooperation agreement signed recently by the China

National Nuclear Corporation and the Pakistan Atomic

Energy Commission.

China is also making strides in the U