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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
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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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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
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
| | 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
Environment and Safety
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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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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
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).
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
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
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
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
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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
OPERATION AND NEW BUILD
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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.
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
| | 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)
Fuel
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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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FUEL 40
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
News
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
News
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.
News
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
News
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
63
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NEWS
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 governme
