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nucmag.com
2018
8/9
437
Akademik Lomonosov:
Preparations for Premiere
in Full Swing
442 ı Fuel
Westinghouse EnCore Accident Tolerant Fuel
446 ı Operation and New Build
Neutron Flux Fluctuations in PWR
ISSN · 1431-5254
24.– €
457 ı Research and Innovation
Coated Ceramic Honeycomb Type Passive
Autocatalytic Recombiner
463 ı AMNT 2018
Young Scientists Workshop
Call for Papers
Inside
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atw Vol. 63 (2018) | Issue 8/9 ı August/September
Nuclear Energy: The Dead Live Longer or
the Summer of 2018
Dear Reader, Although nuclear energy offers both comprehensive technical potential with further development
prospects for use in power generation and attractive economic conditions, both for existing plants and for new plants –
assuming a reliable regulatory and political environment – there was no visible impetus for this for a long time.
Nuclear energy has also been or is facing serious market
challenges. There are two reasons why it cannot exploit its
economic advantages: On the one hand, there are hardly
any free electricity markets left; regulated markets with
subsidy systems, some of which are excessive and barely
manageable, prevent any market development towards
efficient systems as a whole. On the other hand, plants
with long depreciation periods, as is the case with nuclear
energy at around 20 years, are not very attractive.
Remarkable developments in spring/summer 2018 set
clear signals for future impulses, especially with their
technical accents:
1. At the end of April 2018, the Akademik Lomonosov was
launched in St. Petersburg, Russia. The lighter is
equipped with two KLT-40S type nuclear reactors,
which have been successfully used in icebreakers for
many decades. Each reactor can supply up to 35 MW of
electricity and 200 GJ/h of district heating, sufficient to
supply around 100,000 people in polar regions. After
the launch, the lighter was towed through the Baltic
and North Sea to Murmansk, where it is loaded with
nuclear fuel. Next year, the Akademik Lomonosov will
be towed to the Chukchi region in eastern Russia to its
final location.
2. On 6 June 2018, the Taishan 1 nuclear power plant unit
in the province of Guangdong in southern China
achieved first criticality. This is the first active EPR type
plant in the world and thus the second Generation III+
reactor type to go into operation after the Russian
VVER-1200 in Novovoronezh, which went into operation
in 2016. With a gross nominal output of 1750 MW, it is
the world's most powerful type of nuclear power plant.
Construction of the plant began in 2009. 2 blocks of the
same type have been under construction in Europe
since 2005 (Olkiluoto 3, Finland) and 2007 ( Flamanville
3, France). Originally, EPR reactors were developed
for a Western European expansion program and are
supplied by Framatome. A second unit is currently being
commissioned at the Taishan site in China. French
President Emmanuel Macron and Indian President
Narenda Modi signed a contract in March 2018 to build
six EPRs in India.
3. On 21 June 2018, the Sanmen 1 nuclear power plant
unit in the Chinese province of Zhejiang achieved first
criticality. This is the first AP1000 plant worldwide
and thus the third Generation III+ reactor type in
operation. Construction of the plant began in 2009 and
on 8 August 2018 the identical Haiyang 1 block in the
Chinese province of Shandong also achieved first
criticality. A further block is under construction at each
of the two sites. The AP1000 with a gross output of
around 1250 MW is a development of Westinghouse. In
the USA, four units are under construction at the Vogtle
and Summer sites; construction of the two Summer
units was suspended in August 2017, partly because the
Westinghouse Electric Company, as the manufacturer,
had to initiate Chapter 11 insolvency procedure.
Meanwhile, the Canadian Brookfield Business Partners
has taken over the nuclear technology company. Among
others, the Indian government is confident of signing a
contract for the construction of 6 AP1000s in India in
the near future.
These start-ups not only mark the fact that, despite all the
challenges and the associated delays, new technical
ground can be successfully broken in nuclear technology.
EPR, AP1000 or VVER-1200 can now provide impetus for
the marketing of nuclear energy in the new markets
available - even if these markets are not necessarily located
in Europe at present.
Oh yes, Europe ... two sentences about the Old World:
1. Nuclear energy, and thus the reactors at the Belgian
sites of Tihange and Doel, which are almost prayer- milllike
in some media, have so far this year covered around
60 % of the country's electricity requirements. In April
2018, the current Belgian government had confirmed
an “energy pact” for the country's nuclear power plants,
which intends for the plants to be decommissioned
between 2022 and 2025. This is about the seventh exit
announcement by a Belgian government.
2. The UK government is promoting the development
and construction of small modular reactors (SMR). A
£ 200 million investment programme as part of the
country's long-term industrial strategy is to accelerate
the construction of a pilot plant at Trawsfynydd in
northern Wales.
So it is not only exciting with regard to the future of nuclear
energy worldwide, there are now also future prospects for
expansion worldwide with currently 454 commercial units
in operation, as many as never before.
Christopher Weßelmann
– Editor in Chief –
427
EDITORIAL
Editorial
Nuclear Energy: The Dead Live Longer or the Summer of 2018
atw Vol. 63 (2018) | Issue 8/9 ı August/September
EDITORIAL 428
Kernenergie: Totgesagte leben länger
oder der Sommer 2018
Liebe Leserin, lieber Leser, obgleich die Kernenergie ein sowohl umfassendes technisches Potenzial mit
weiteren Entwicklungsperspektiven für den Einsatz in der Energieerzeugung als auch attraktive betriebswirtschaftliche
Rahmenbedingungen, sowohl für bestehende Anlagen als auch für Neuanlagen – ein verlässliches regulatorisches und
politisches Umfeld vorausgesetzt – bietet, fehlten hierzu lange sichtbare Impulse.
Die Kernenergie wurde bzw. wird zudem mit ernsten
Herausforderungen der Märkte konfrontiert. So kann sie
ihre wirtschaftlichen Vorteile aus zwei Gründen nicht
ausspielen: Zum einen existieren kaum noch freie Strommärkte;
regulierte Märkte mit teils überbordenden und
kaum noch überschaubaren Subventionssystemen verhindern
jegliche Marktentwicklung in Richtung effizienter
Systeme überhaupt. Zum anderen sind Anlagen mit langen
Abschreibungszeiten, wie es bei der Kernenergie mit rund
20 Jahren der Fall ist, wenig attraktiv – langer Atem ist für
Kernkraftwerksbetreiber erforderlich.
Bemerkenswerte Entwicklungen im Frühjahr/Sommer
2018 setzen insbesondere mit ihren technischen Akzenten
deutliche Zeichen für Zukunftsimpulse:
1. Ende April 2018 lief in St. Petersburg, Russland, die
Akademik Lomonosov vom Stapel. Der Leichter ist ausgerüstet
mit zwei Kernreaktoren vom Typ KLT-40S, wie sie
erfolgreich seit vielen Jahrzehnten in Eisbrechern zum
Einsatz kommen. Jeder Reaktor kann bis zu 35 MW Strom
liefern sowie 200 GJ/h Fernwärme, ausreichend für die
Versorgung von rund 100.000 Menschen in polaren
Regionen. Der Leichter wurde nach dem Stapellauf durch
Ost- und Nordsee nach Murmansk geschleppt, wo die
Kernbrennstoffbeladung erfolgt. Im kommenden Jahr
wird die Akademik Lomonosov in die Tschuktschen-Region
im Osten Russlands zu ihrem endgültigen Einsatzort
geschleppt.
2. Am 6. Juni 2018 erreichte der Kernkraftwerksblock
Taishan 1 in der im Süden Chinas gelegenen Provinz
Guangdong Erstkritikalität. Es ist dies die erste Anlage
weltweit vom Typ EPR und damit nach dem 2016 in Betrieb
gegangenen russischen WWER-1200 in Nowoworonesch
der zweite Reaktortyp der Generation III+ in Betrieb. Mit
einer Nennleistung von 1750 MW brutto ist es der weltweit
leistungsstärkste Kernkraftwerkstyp. Der Bau der Anlage
begann im Jahr 2009. In Europa sind 2 typgleiche Blöcke
seit 2005 (Olkiluoto 3, Finnland) bzw. 2007 (Flamanville 3,
Frankreich) in Bau. Ursprünglich waren Reaktoren des
Typs EPR für ein westeuropäisches Zubauprogramm entwickelt
worden und werden von Framatome geliefert. Am
chinesischen Standort Taishan befindet sich ein zweiter
Block in der Inbetriebnahme. Der französische Staatspräsident
Emmanuel Macron und der indische Präsident
Narenda Modi unterzeichneten im März 2018 einen
Vertrag, der zum Bau von sechs EPR in Indien führen soll.
3. Am 21. Juni 2018 erreichte der Kernkraftwerksblock
Sanmen 1 in der chinesischen Provinz Zhejiang Erstkritikalität.
Es ist dies die erste Anlage weltweit vom Typ
AP1000 und damit der dritte Reaktortyp der Generation
III+ in Betrieb. Der Bau der Anlage begann im Jahr 2009.
Am 8. August 2018 erreichte der baugleiche Block Haiyang
1 in der chinesischen Provinz Shandong ebenfalls Erstkritikalität.
An beiden Standorten ist jeweils ein weiterer
Block in Bau. Der AP1000 mit einer Bruttoleistung von rd.
1250 MW ist eine Entwicklung von Westinghouse. Der Bau
begann im Jahr 2009. In den USA sind an den Standorten
Vogtle und Summer vier Blöcke in Bau; für die beiden
Blöcke Summer wurde im August 2017 ein Baustopp
beschlossen, u.a. da die Westinghouse Electric Company
als Hersteller ein sog. „Chapter 11-Insolvenzverfahren“
ein leiten musste. Inzwischen hat die kanadische Brookfield
Business Partners das Kerntechnikunternehmen übernommen.
Unter anderem die indische Regierung ist
zuversichtlich, einen Vertrag über den Bau von 6 AP1000
in Indien in der nächsten Zukunft unterzeichnen zu
können.
Diese Inbetriebnahmen kennzeichnen nicht nur, dass bei
allen Herausforderungen und auch damit verbundenen
Verzögerungen, technisches Neuland in der Kerntechnik
erfolgreich beschritten werden kann. EPR, AP1000 oder
auch WWER-1200 können jetzt Impulse mit sich bringen,
die der Vermarktung auf den bereit stehenden neuen
Märkten für die Kernenergie Schwung liefern – auch wenn
diese Märkte derzeit nicht unbedingt in Europa liegen.
Ach ja Europa ... zwei Sätze zur Alten Welt:
1. Die Kernenergie und damit die in manchen Medien fast
gebetsmühlenartig gescholtenen Reaktoren an den
belgischen Standorten Tihange und Doel haben im
bisherigen Jahresverlauf rund 60 % des Strombedarfs des
Landes gedeckt. Die derzeitige belgische Regierung
hatte im April 2018 für alle Kernkraftwerke des Landes
einen „Energiepakt“ bestätigt, der eine Stilllegung der
Anlagen in den Jahren 2022 bis 2025 vorsieht. Es ist
dies ungefähr die siebte Ausstiegsankündigung einer
belgischen Regierung.
2. Die Regierung Großbritanniens fördert die Entwicklung
und den Bau von modularen Kernreaktoren kleiner
Leistung (SMR: small modular reactor). Ein 200-Mio.-
Pfund Investitionsprogramm im Rahmen der langfristigen
Industriestrategie des Landes soll den Bau einer Pilotanlage
am Standort Trawsfynydd im Norden Wales
forcieren.
Es bleibt also nicht nur spannend, was die Zukunft der
Kernenergie weltweit betrifft, es gibt jetzt auch Zukunftsperspektiven
sogar für einen Ausbau weltweit– mit derzeit
454 Kernkraftwerken weltweit in Betrieb...so viele wie
noch nie zuvor.
Christopher Weßelmann
– Chefredakteur –
Editorial
Nuclear Energy: the Dead Live Longer or the Summer of 2018
Kommunikation und
Training für Kerntechnik
Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?
Wählen Sie aus folgenden Themen: Dozent/in Termin/e Ort
3 Atom-, Vertrags- und Exportrecht
Ihr Weg durch Genehmigungs- und Aufsichtsverfahren RA Dr. Christian Raetzke 18.09.2018
02.04.2019
22.10.2019
Das Recht der radioaktiven Abfälle RA Dr. Christian Raetzke 23.10.2018
05.03.2019
17.09.2019
Atomrecht – Navigation im internationalen nuklearen Vertragsrecht Akos Frank LL. M. 03.04.2019 Berlin
Atomrecht – Was Sie wissen müssen
Export kerntechnischer Produkte und Dienstleistungen –
Chancen und Regularien
3 Kommunikation und Politik
RA Dr. Christian Raetzke
Akos Frank LL. M.
RA Kay Höft M. A.
RA Olaf Kreuzer
Dr. Ing. Wolfgang Steinwarz
Berlin
Berlin
04.06.2019 Berlin
12.06. - 13.06.2019 Berlin
Schlüsselfaktor Interkulturelle Kompetenz –
International verstehen und verstanden werden
Public Hearing Workshop –
Öffentliche Anhörungen erfolgreich meistern
Kerntechnik und Energiepolitik im gesellschaftlichen Diskurs
– Themen und Formate
Angela Lloyd 26.09.2018 Berlin
Dr. Nikolai A. Behr 16.10. - 17.10.2018
05.11. - 06.11.2019
Berlin
N.N. 12.11. - 13.11.2018 Gronau/
Lingen
3 Rückbau und Strahlenschutz
In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:
Stilllegung und Rückbau in Recht und Praxis
Das neue Strahlenschutzgesetz –
Folgen für Recht und Praxis
Dr. Matthias Bauerfeind
RA Dr. Christian Raetzke
Maria Poetsch
RA Dr. Christian Raetzke
24.09. - 25.09.2018 Berlin
05.11. - 06.11.2018
12.02. - 13.02.2019
25.06. - 26.06.2019
Berlin
3 Nuclear English
Enhancing Your Nuclear English Devika Kataja 22.05. - 23.05.2019 Berlin
Advancing Your Nuclear English (Aufbaukurs) Devika Kataja 10.10. - 11.10.2018
10.04. - 11.04.2019
18.09. - 19.09.2019
3 Wissenstransfer und Veränderungsmanagement
Berlin
Veränderungsprozesse gestalten – Heraus forderungen
meistern, Beteiligte gewinnen
Erfolgreicher Wissenstransfer in der Kern technik –
Methoden und praktische Anwendung
Dr. Tanja-Vera Herking
Dr. Christien Zedler
Dr. Tanja-Vera Herking
Dr. Christien Zedler
28.11. - 29.11.2018
26.11. - 27.11.2019
Berlin
26.03. - 27.03.2019 Berlin
Haben wir Ihr Interesse geweckt? 3 Rufen Sie uns an: +49 30 498555-30
Kontakt
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Die INFORUM-Seminare können je nach
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der Fachkunde geeignet sein.
atw Vol. 63 (2018) | Issue 8/9 ı August/September
430
Issue 8/9
August/September
CONTENTS
437
Akademik Lomonosov:
Preparations for Premiere
in Full Swing
| | The world’s only floating power unit ‘Akademik Lomonosov’ takes the sea. On 28 April 2018, the floating nuclear power unit (FPU)
‘Akademik Lomonosov’ has left the territory of Baltiyskiy Zavod in St. Petersburg, Russia, where its construction had been carried out
since 2009, and headed to its base in Chukotka.
Editorial
Nuclear Energy: the Dead Live Longer
or the Summer of 2018 427
Kernenergie: Totgesagte leben länger
oder der Sommer 2018 428
Abstracts | English 432
Abstracts | German 433
Inside Nuclear with NucNet
A Stark Warning to Trump on China, Russia
and the ‘Crisis’ Facing US Nuclear Industry 434
NucNet, David Dalton
Calendar 436
442
| | Neutron radiographs of U3Si2 pins from ATR.
Energy Policy, Economy and Law
Akademik Lomonosov:
Preparations for Premiere in Full Swing 437
Roman Martinek
440
Spotlight on Nuclear Law
Nuclear Phase-out Last Act?
Are the New Compensation Regulations for
Frustrated Expenses in Accordance
with the Constitution? 440
Atomausstieg letzter Akt?
Sind die neuen Entschädigungs regelungen
für frustrierte Aufwendungen und nicht mehr
verstrombare Elektrizitätsmengen im Atomgesetz
verfassungsgemäß? 440
| | Upper part of a pressurized reactor vessel during maintenance.
Tobias Leidinger
Contents
atw Vol. 63 (2018) | Issue 8/9 ı August/September
431
Fuel
Innovations for the Future
Westinghouse EnCore® Accident Tolerant Fuel 442
Gilda Bocock, Robert Oelrich, and Sumit Ray
Operation and New Build
Analyses of Possible Explanations for the
Neutron Flux Fluctuations in German PWR 446
457
CONTENTS
Joachim Herb, Christoph Bläsius, Yann Perin,
Jürgen Sievers and Kiril Velkov
| | Coated Ceramic Honeycomb type Passive Autocatalytic Recombiner.
Detailed Measurements and Analyses of the
Neutron Flux Oscillation Phenomenology
at Kernkraftwerk Gösgen 452
A Preliminary Conservative Criticality Assessment
of Fukushima Unit 1 Debris Bed 473
G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff
María Freiría López, Michael Buck and Jörg Starflinger
452
AMNT 2018
Key Topic | Outstanding Know-How
& Sustainable Innovations
Focus Session International Regulation:
Radiation Protection: The Implementation
of the EU Basic Safety Standards Directive 2013/59
and the Release of Radioactive Material
from Regulatory Control 477
Christian Raetzke
| | Schematic representation of the 3002 MW 3-Loop KKG core.
DAtF Notes 456
Research and Innovation
Effects of Airborne Volatile Organic Compounds on
the Performance of Pi/TiO 2 Coated Ceramic Honeycomb
Type Passive Autocatalytic Recombiner 457
Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
AMNT 2018
Young Scientists' Workshop 463
Report
Report: GRS Workshop “Safety of Extended
Dry Storage of Spent Nuclear Fuel” 480
Klemens Hummelsheim, Florian Rowold and Maik Stuke
KTG Inside 483
News 484
Nuclear Today
Why do We Allow Nuclear to Take
the ‘Silly Season’ Media Heat? 490
Jörg Starflinger
John Shepherd
Heuristic Methods in Modelling Research
Reactors for Deterministic Safety Analysis 464
Imprint 485
Vera Koppers and Marco K. Koch
Development and Validation of a CFD
Wash-Off Model for Fission Products
on Containment Walls 469
Katharina Amend and Markus Klein
Aachen Institute for Nuclear Training
AMNT 2019: Call for Papers
Inforum: Seminar Programme 2018/2019
Insert
Insert
Insert
Contents
atw Vol. 63 (2018) | Issue 8/9 ı August/September
432
ABSTRACTS | ENGLISH
A Stark Warning to Trump on China, Russia
and the ‘Crisis’ Facing US Nuclear Industry
NucNet, David Dalton | Page 434
The US has the largest number of nuclear plants in
the world – 99 in commercial operation at the time
of writing – but its global leadership position is said
to be declining as efforts to build a new generation
of reactors have been plagued by problems, and
aging plants have been retired or closed in the face
of economic, market, and financial pressures. A
recent report by the Atlantic Council issued a
stark warning, arguing that the US nuclear energy
industry is facing a crisis that the Trump administration
must immediately address as a core part of
its “all of the above” energy strategy.
Akademik Lomonosov:
Preparations for Premiere in Full Swing
Roman Martinek | Page 437
At the end of July 2018, the loading of the floating
power unit Akademik Lomonosov with nuclear fuel
started in Murmansk. This is one of the key stages of
the project, which as of today has no analogues in the
world. In 2019, the power unit will begin to supply
local population and industrial facilities in North-
Eastern Siberia with heat and electricity. The project is
expected to open up opportunities for the mass production
of floating nuclear power plants – a number
of countries have already voiced their interest.
The Akademik Lomonosov is intended for providing
energy to remote industrial facilities, port cities, as
well as gas and oil platforms located on the high seas.
Nuclear Phase-out Last Act?
Are the New Compensation Regulations for
Frustrated Expenses in Accordance with the
Constitution?
Tobias Leidinger | Page 440
Shortly before it was passed, the legislature reacted
to the constitutional deficiencies which the Federal
Constitutional Court (BVerfG) objected to in its
judgment of 6 December 2016 on the nuclear
phase-out (BVerfGE 143, 246) and for which a
constitutional situation had to be established by
30 June 2018. However, the newly created compensation
regulations in the 16 th amendment to the
Atomic Energy Act raise new legal questions,
especially those relating to their constitutionality.
Westinghouse EnCore ® Accident
Tolerant Fuel
Gilda Bocock, Robert Oelrich
and Sumit Ray | Page 442
The development and implementation of accident
tolerant fuel (ATF) products, such as Westinghouse’s
EnCore® Fuel, can support the long-term
viability of nuclear energy by enhancing operational
safety and decreasing energy costs. The first introduction
of Westinghouse EnCore Fuel into a commercial
reactor is planned for 2019 as segmented
lead test rods (LTRs) utilizing chromium-coated
zirconium cladding with uranium silicide (U 3 Si 2 )
pellets. The EnCore Fuel lead test assembly (LTA)
program, with LTAs planned for 2022 insertion, will
introduce silicon carbide/silicon carbide composite
cladding with U 3 Si 2 pellets.
Analyses of Possible Explanations for the
Neutron Flux Fluctuations in German PWR
Joachim Herb, Christoph Bläsius, Yann Perin,
Jürgen Sievers and Kiril Velkov | Page 446
During the last 15 years the neutron flux fluctuation
levels in some of the German PWR changed
significantly. During a period of about ten years, the
fluctuation levels increased, followed by about five
years with decreasing levels after taking actions like
changing the design of the fuel elements. The
increase in the neutron flux fluctuations resulted in
an increased number of triggering the reactor
limitation system and in one case in a SCRAM.
Several models based on single physical effects are
used to simulate the neutron flux. Each of these
simple models can reproduce some of the characteristics
of the observed neutron flux fluctuations.
Detailed Measurements and Analyses of the
Neutron Flux Oscillation Phenomenology at
Kernkraftwerk Gösgen
G. Girardin, R. Meier, L. Meyer,
A. Ålander and F. Jatuff | Page 452
Recent investigations on measured neutron flux
noise at the Kernkraftwerk Gösgen-Däniken are
summarised. The NPP in operation since 1979 is a
German KWU pre-KONVOI, 3-Loop PWR with a
thermal power of 3,002 MWth (1,060 MWe). In a
period of approx. 7 cycles from 2010 to 2016, an
increase of the measured neutron noise amplitudes
in the in- and out-core neutron detectors has been
observed, although no significant variations have
being detected in global core, thermohydraulic
circuits or instrumentation parameters. Verifications
of the instrumentation were performed and it was
confirmed that the neutron flux instabilities
increased from cycle to cycle in this period. In the last
two years, the level of neutron flux noise remains
high but seems to have achieved a saturation state.
Effects of Airborne Volatile Organic
Compounds on the Performance of Pi/TiO 2
Coated Ceramic Honeycomb Type Passive
Autocatalytic Recombiner
Chang Hyun Kim, Je Joong Sung,
Sang Jun Ha and Phil Won Seo | Page 457
Ensuring the containment integrity during a severe
accident in nuclear power reactor by maintaining the
hydrogen concentration below an acceptable level
has been recognized to be of critical importance after
Fukushima Daiichi accidents. Although there exist
various hydrogen mitigation measures, a passive
autocatalytic recombiner (PAR) has been considered
as a viable option for the mitigation of hydrogen risk
under the extended station blackout conditions
because of its passive operation char acteristics for
the hydrogen removal. As a post- Fukushima action
item, all Korean nuclear power plants were equipped
with PARs of various suppliers. The capacity and
locations of PAR as a hydrogen mitigation system
were determined through an extensive analysis for
various severe accident scenarios.
49 th Annual Meeting on Nuclear Technology
(AMNT 2018): Young Scientists Workshop
Jörg Starflinger | Page 463
During the Young Scientists Workshop of the 49 th
Annual Meeting on Nuclear Technology (AMNT
2018), 29 to 30 May 2018, Berlin, 13 young
scientists presented results of their scientific
research as part of their Master or Doctorate theses
covering a broad spectrum of technical areas. Vera
Koppers, Katharina Amend and Maria Freiria were
awarded for their presentations by the jury.
Heuristic Methods in Modelling Research
Reactors for Deterministic Safety Analysis
Vera Koppers and Marco K. Koch | Page 464
A new method for rapid and reliable modelling of
research reactors for deterministic safety analysis is
presented. A rule-based software system is being
developed to support the modelling process in
ATHLET for selected research reactor types in the
light of limited available data. The fundamental
elements of the input deck are generated automatically
by few input data necessary.
Development and Validation of a
CFD Wash-Off Model for Fission Products
on Containment Walls
Katharina Amend and Markus Klein | Page 469
The research project aims to develop a CFD model
to describe the run down behavior of liquids and the
resulting wash-down of fission products on surfaces
in the reactor containment. The paper presents a
three-dimensional numerical simulation for water
running down inclined surfaces coupled with an
aerosol wash-off model and particle transport using
OpenFOAM. The wash-off model is based on Shields
criterion. A parameter variation is conducted and
the simulation results are compared to experiments.
A Preliminary Conservative Criticality
Assessment of Fukushima Unit 1 Debris Bed
María Freiría López, Michael Buck and
Jörg Starflinger | Page 473
A conservative criticality evaluation of Fukushima
Unit 1 debris bed has been carried out. In order to
obtain a multi-dimensional criticality map, parameters,
such as debris size, porosity, particle size, fuel
burnup, water density and boration were varied. As
a result, safety parameter ranges where recriticality
can be excluded have been identified. It was found
that most of the possible debris would be inherently
subcritical because of its porosity and 1600 ppm B
would ensure subcriticality under any conditions.
49 th Annual Meeting on Nuclear Technology
(AMNT) Key Topic | Outstanding Know-How
& Sustainable Innovations
Christian Raetzke | Page 477
The report summarises the presentations of the
Focus Session International Regulation | Radiation
Protection: The Implementation of the EU Basic
Safety Standards Directive 2013/59 and the Release
of Radioactive Material from Regulatory Control
presented at the 49 th AMNT 2018, Berlin, 29 to 30
May 2018.
Report: GRS Workshop “Safety of Extended
Dry Storage of Spent Nuclear Fuel”
Klemens Hummelsheim, Florian Rowold
and Maik Stuke | Page 480
Conference report on the GRS Workshop “Safety of
Extended Dry Storage of Spent Nuclear Fuel”, 6 to 8
June 2018.
Why do We Allow Nuclear to Take the
‘Silly Season’ Media Heat?
John Shepherd | Page 490
The time of year always means all manner of weird
and wonderful stories finding their way into the
news. For the nuclear industry, the hot spell fanned
the media flames of an old anti-nuclear favourite, as
it became clear operations at some nuclear power
plants were being halted temporarily to comply
with restrictions that prevent cooling water further
heating local rivers and waterways. It’s a question
why the nuclear community does not use the time
of year to communicate their important and
interesting topics.
Abstracts | English
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Eine deutliche Warnung für die
US-Nuklearindustrie – auch vor der
Konkurrenz aus China und Russland
NucNet, David Dalton | Seite 434
In den USA ist die weltweit größte Anzahl von
Kernkraftwerken in kommerziellem Betrieb – 99
Anlagen; aber die globale Führungsposition der
USA schwindet, da die Bemühungen zum Bau einer
neuen Generation von Reaktoren mit Problemen
behaftet ist und ältere Anlagen angesichts wirtschaftlichen
Drucks stillgelegt werden. Ein kürzlich
veröffentlichter Bericht des Atlantic Council warnt
die US-Nuklearindustrie vor einer Krise, der die
Trump-Regierung als Kernstück ihrer „All of the
above“-Energiestrategie begegnen muss.
Akademik Lomonosov: Vorbereitungen
für die Inbetriebnahme in vollem Gange
Roman Martinek | Seite 437
Ende Juli 2018 begann in Murmansk die Kernbrennstoffbeladung
des schwimmenden Kraftwerks
Akademik Lomonosov. Dies ist eine der bedeutenden
Phasen des Projekts, das bis heute weltweit
einzigartig ist. Das Kraftwerk wird ab 2019 eine
ganze Region in Nordostsibirien mit Wärme und
Strom versorgen. Das Projekt soll Möglichkeiten für
die Serienproduktion von schwimmenden Kernkraftwerken
eröffnen – einige Länder haben dafür
bereits ihr Interesse bekundet. Die Akademik
Lomonosov ist für die Energieversorgung abgelegener
Industrieanlagen, Hafenstädte sowie von Gasund
Ölplattformen auf hoher See konzipiert.
Atomausstieg letzter Akt? Sind die neuen
Entschädigungsregelungen für frustrierte
Aufwendungen und nicht mehr verstrombare
Elektrizitätsmengen im Atomgesetz
verfassungsgemäß?
Tobias Leidinger | Seite 440
Kurz vor knapp hat der Gesetzgeber auf die
verfassungs rechtlichen Mängel reagiert, die das
Bundesverfassungsgericht (BVerfG) in seinem Urteil
vom 6. Dezember 2016 zum Atomausstieg (BVerfGE
143, 246) höchstrichterlich beanstandet hat und für
die bis zum 30. Juni 2018 ein verfassungsgemäßer
Zustand herzustellen war. Doch die neu geschaffenen
Entschädigungsregelungen in der 16.
AtG-Novelle werfen neue Rechtsfragen auf, insbesondere
die nach ihrer Verfassungsgemäßheit.
Westinghouse EnCore® Accident
Tolerant Fuel
Gilda Bocock, Robert Oelrich und
Sumit Ray | Seite 442
Entwicklung und Einsatz von „störfalltolerantem
Kernbrennstoff“ wie z.B. EnCore® von Westinghouse,
kann der Kernenergie weitere Zukunftsperspektiven
durch Erhöhung der Betriebssicherheit und
Senkung der Kosten eröffnen. Der erste Einsatz von
Westinghouse EnCore Fuel in einem kommerziellen
Reaktor ist für 2019 geplant. Testbrennstäbe mit
verchromtem Zirkoniummantel und Uransilicid
(U3Si2)-Pellets sind dafür vorgesehen. Das für
2022 geplante EnCore Fuel Lead Test Assembly
(LTA)-Programm sieht ein Siliziumkarbid/Siliziumkarbid-Verbundhüllrohr
mit U3Si2-Pellets vor.
verändert. Während eines Zeitraums von etwa zehn
Jahren nahmen die Schwankungsbreiten zu, gefolgt
von etwa fünf Jahren mit abnehmender Tendenz
nach z.B. einer Änderung der Auslegung der Brennelemente.
Die Zunahme der Neutronenflussschwankungen
führte zu einer erhöhten Anzahl von
Auslösungen des Reaktorbegrenzungssystems und in
einem Fall zu einem SCRAM. Zur Simulation des Neutronenflusses
werden mehrere Modelle verwendet,
die auf einzelnen physikalischen Effekten basieren.
Detaillierte Messungen und Analysen
der Neutronenflussschwingungen
im Kernkraftwerk Gösgen
G. Girardin, R. Meier, L. Meyer, A. Ålander
und F. Jatuff | Seite 452
Aktuelle Untersuchungen zum Neutronenflussrauschen
im Kernkraftwerk Gösgen-Däniken
werden zusammengefasst. Das seit 1979 in Betrieb
befindliche Kernkraftwerk In einem Zeitraum von
ca. 7 Zyklen von 2010 bis 2016 wurde ein Anstieg
der gemessenen Neutronenrauschamplituden beobachtet,
obwohl keine signifikanten Schwankungen
der globalen physikalischen und thermohydraulischen
sowie Instrumentierungsparametern
festgestellt wurden. Überprüfungen der Instrumentierung
wurden durchgeführt und es wurde bestätigt,
dass die Neutronenflussinstabilitäten in diesem
Zeitraum von Zyklus zu Zyklus zunahmen. In den
letzten zwei Jahren blieb das Neutronenflussrauschen
hoch, scheint aber einen Sättigungszustand
erreicht zu haben.
Einfluss von flüchtigen organischen
Verbindungen auf Pi/TiO 2 -beschichtete
keramische Wabenkörpern von passiven
autokatalytischen Rekombinatoren
Chang Hyun Kim, Je Joong Sung, Sang Jun Ha
und Phil Won Seo | Seite 457
Nach den Unfällen von Fukushima Daiichi wurde
festgestellt, dass der Integrität des Sicherheitsbehälters
bei einem schweren Unfall in einem
Kernkraftwerk höchste Priorität gilt, indem die
Wasserstoffkonzentration unterhalb akzeptabler
Werte gehalten wird. Obwohl es verschiedene
Maßnahmen zur Wasserstoffminderung gibt, wird
ein passiver autokatalytischer Rekombinator (PAR)
wegen seiner Betriebseigenschaften als praktikable
Option angesehen. Als Post-Fukushima-Maßnahme
wurden alle koreanischen Kernkraftwerke mit PARs
verschiedener Anbieter ausgestattet. Die Kapazitäten
und optimalen Einbauorte von PARs als
Wasserstoffminderungssystem wurden durch eine
umfangreiche Analyse für verschiedene schwere
Unfallszenarien ermittelt.
49. Jahrestagung Kerntechnik (AMNT 2018):
Young Scientists Workshop
Jörg Starflinger | Seite 463
Im Rahmen des Young Scientists Workshop der
49. Jahrestagung Kerntechnik (AMNT 2018) vom
29. bis 30. Mai 2018 in Berlin stellten 13 Nachwuchswissenschaftlerinnen
und -wissenschaftler im
Rahmen ihrer Master- oder Doktorarbeiten ein breites
Spektrum von Fachthemen vor. Vera Koppers,
Katharina Amend und Maria Freiria wurden für ihre
Präsentationen von der Jury ausgezeichnet.
für die Durchführung von deterministischen
Sicherheitsanalysen vorgestellt. Für ausgewählte
Forschungsreaktor-Typen wird ein regelbasiertes
Softwaresystem konzipiert, das den Modellierungsprozess
für ATHLET unterstützt. Die Entwicklung
wird unter dem Aspekt limitierter verfügbarer
Daten vorgenommen. Die fundamentalen Elemente
des Datensatz werden unter Verwendung weniger
Eingabedaten automatisch generiert.
Entwicklung und Validierung eines
CFD-Modells für das Auswaschen von Spaltprodukten
auf Containment-Oberflächen
Katharina Amend und Markus Klein | Seite 469
Ziel des Forschungsvorhabens ist ein CFD-Modell
für das Ablaufverhalten von Wasser und den
resultierenden Abwasch von Spaltprodukten auf
Oberflächen im Reaktorsicherheitsbehälter. Das
Paper präsentiert eine dreidimensionale numerische
OpenFOAM Simulation von Wasser auf geneigten
Oberflächen gekoppelt mit einem Aerosol-
Abwaschmodell und dem Partikeltransport. Das
Abwaschmodell basiert auf dem Shields Kriterium.
Es wird eine Parametervariation durchgeführt und
die Simulationsergebnisse mit Experimenten verglichen.
Eine vorläufige konservative
Kritikalitätsbeurteilung des Schüttbetts
des Reaktors Fukushima-1
María Freiría López, Michael Buck und
Jörg Starflinger | Seite 473
Eine konservative Kritikalitätsanalyse des Fukushima
Unit 1 Schüttbetts wurde durchgeführt. Um eine
mehrdimensionale Kritikalitätskarte zu erstellen,
wurden Parameter wie Schüttbettgröße, Porosität,
Partikelgröße, Brennstoffabbrand, Wasserdichte und
Boranteil variiert. Als Resultat, wurden Bereiche
identifiziert, in denen Rekritikalität ausgeschlossen
werden kann. Es stellt sich heraus, dass die meisten
entstehenden Schüttbetten aufgrund seiner Porosität
inhärent unterkritisch sind, und dass auch
1600 ppm B Unterkritikalität sicherstellen.
49. Jahrestagung Kerntechnik (AMNT 2018)
Key Topic | Outstanding Know-How &
Sustainable Innovations
Christian Raetzke | Seite 477
Der Bericht fasst die Vorträge der Focus Session
International Regulation | Radiation Protection:
The Implementation of the EU Basic Safety
Standards Directive 2013/59 and the Release of
Radioactive Material from Regulatory Control
zusammen, die auf der 49. Jahrestagung Kerntechnik
(AMNT 2018) präsentiert wurden.
Report: GRS Workshop “Safety of Extended
Dry Storage of Spent Nuclear Fuel”
Klemens Hummelsheim, Florian Rowold und
Maik Stuke | Seite 480
Tagungsbericht zum Workshop “Sicherheit einer
zeitlich längeren trockenen Lagerung abgebrannter
Brennelemente”, 6 bis 8 Juni 2018.
Warum lassen wir zu, dass die Kernenergie
in der “Saure Gurken Zeit“ Thema wird
433
ABSTRACTS | GERMAN
Analysen zu Neutronenflussschwankungen
in deutschen DWR
Joachim Herb, Christoph Bläsius, Yann Perin,
Jürgen Sievers und Kiril Velkov | Seite 446
In den letzten 15 Jahren haben sich die Neutronenflussschwankungen
in einigen der deutschen DWR
Heuristische Methoden in der Modellierung
deterministischen Sicherheitsanalysen von
Forschungsreaktoren
Vera Koppers and Marco K. Koch | Seite 464
Es wird eine neue Methode zur schnellen und zuverlässigen
Modellierung von Forschungsreaktoren
John Shepherd | Seite 490
In der „Saure Gurken Zeit“ des Jahres werden von
der Presse teils seltsame und teils wunderbare
Geschichten aufgenommen. Immer wieder trifft
dies auch die Kernenergie – warum lassen wir dies
zu, mit den wichtigen positiven Botschaften, die wir
mit der Kernenergie haben?
Abstracts | German
atw Vol. 63 (2018) | Issue 8/9 ı August/September
434
INSIDE NUCLEAR WITH NUCNET
A Stark Warning to Trump on China, Russia
and the ‘Crisis’ Facing US Nuclear Industry
NucNet, David Dalton
The US has the largest number of nuclear plants in the world – 99 in commercial operation at the time of
writing – but its global leadership position is said to be declining as efforts to build a new generation of
reactors have been plagued by problems, and aging plants have been retired or closed in the face of economic,
market, and financial pressures.
A recent report by the Washington-based think-tank the
Atlantic Council issued a stark warning, arguing that the
US nuclear energy industry is facing a crisis that the Trump
administration must immediately address as a core part of
its “all of the above” energy strategy that is intended to
herald an era of American energy dominance, with tens of
billions of dollars to be spent on drilling and construction
of pipelines, processing plants and liquefied natural gas
export terminals. The administration might be bullish on
energy policy, but the nuclear industry is worried.
Six US nuclear plants have been shut down permanently
since 2013 and 12 more are slated to retire over the
next seven years. The Washington-based Nuclear Energy
Institute, which represents the nuclear industry in the US,
says the US electricity grid is enduring “unprecedented
tumult and challenge” because of the loss of thousands
and thousands of megawatts of carbon-free, fuel-secure
generation that nuclear plants represent. Closing nuclear
plants makes electricity prices go up and is putting
emissions reduction targets hopelessly out of reach, NEI
president and chief executive officer Maria Korsnick said.
The Atlantic Council says the decline of the nuclear
power industry in the US is “an important policy problem”
that is not receiving the attention it deserves. The report
was made public in the same week that Ohio-based utility
FirstEnergy announced plans to permanently shut down
its three nuclear power stations – Davis-Besse, Perry and
Beaver Valley – within the next three years without some
kind of state or federal relief.
The nuclear industry has long argued that electricity
markets should be reformed to recognise the ability of
traditional baseload generation with onsite fuel supplies –
including nuclear power plants – to provide grid resiliency
during extreme events like hurricanes or extreme winter
weather.
To save financially-ailing nuclear plants, state legislatures
in Illinois and New York last year approved subsidies to keep
nuclear plants operating after utilities made appeals about
protecting consumers and jobs. But other proposed bailouts
of nuclear plants have stalled in New Jersey, Connecticut,
Massachusetts, Ohio and Pennsylvania. In Minnesota, the
state legislature is considering a bill that would help Xcel
Energy, owner and operator of the Monticello and Prairie
Island nuclear stations, plan for the high costs of maintaining
old nuclear power plants. The proposed legislation would
give utilities earlier notice about how much money they could
recover for costly work, Minnesota Public Radio reported.
The Atlantic Council report says nuclear power should be
elevated in the Trump administration’s national security
strategy because nuclear is an important strategic sector, and
US global leadership and engagement in nuclear power are
“vital to US national security and foreign-policy interests”.
It also argues that nuclear power is an important
component of a diversified US energy mix, but notes that in
sharp contrast to developments in the US, China and Russia
are pushing to expand their nuclear industries, develop
complete fuel cycles, and build and commercialise new
reactors for both domestic and international markets. The
results of these efforts are striking – nearly two-thirds of the
new reactors under construction worldwide are estimated to
be using designs from China and Russia, countries that have
the advantage of using “state- monopoly and authoritarian
systems” to advance nuclear energy for geopolitical means.
China has the largest nuclear construction programme
in the world by far, with 20 of the 53 total reactors under
construction worldwide. The 13 th Five-Year Plan (2016 to
2020) calls for 58 GW of nuclear capacity online by
2020 to 2021, and an additional 30 GW under construction
at that time.
But what is really worrying the US nuclear industry is
the success of China’s nuclear strategy to establish joint
ventures with Western companies (Toshiba-Westinghouse,
Framatome-Areva, SNC-Lavalin, Energoatom) to build and
evaluate different technologies (AP-1000, EPR, Candu,
VVER-1000), and to incorporate this experience into its
own indigenous designs. Although cost estimates are
difficult to obtain, China has seemingly been able to build
reactors quicker, and at lower cost, than the US, Europe,
and even South Korea, the report says.
China brings a complete package of design, construction,
labour, technology, and financing, which improves
the economics compared to industries in the West.
Both China and Russia offer attractive financing
packages to fund these projects. China goes into markets
abroad with financing options from its Export-Import
Bank, while Russia uses resources from both the Russian
state budget and the Russia Wealth Fund.
In contrast, says the NEI, the US Export-Import Bank’s
board of directors remains without a quorum and as a
result cannot consider medium- and long-term transactions
exceeding $ 10 m. Typically, commercial nuclear
deals are measured in billions of dollars, not millions:
Turkish President Tayyip Erdoğan said that the investment
in the country’s first nuclear power plant, being built by
Russia’s Rosatom, will exceed $ 20 bn.
While China’s relationship with nuclear power is
relatively new – with its first nuclear plant completed in
1991 – Russia’s long history with nuclear power dates to
1954, when the first reactor was commissioned in Obninsk.
The industry has since grown to 37 reactors in commercial
operation and five under construction. Nuclear generation
reached a record of 196.3 TWh in 2016, accounting
for 17 % of domestic electricity generation, and further
increased to 202.868 TWh and 19.9 % in 2017.
Inside Nuclear with NucNet
A Stark Warning to Trump on China, Russia and the ‘Crisis’ Facing US Nuclear Industry ı NucNet, David Dalton
atw Vol. 63 (2018) | Issue 8/9 ı August/September
The Chinese and Russian use of nuclear-power financing
and technology as a means of expanding their overseas
physical presence, and their foreign-policy influence in
key countries, has important implications for the US the
Atlantic Council report says.
On one hand, US companies are collaborating with
China on building, developing, and demonstrating new
reactors; GE has won tenders for the supply of turbine
generators for new Russian-supplied units in Hungary and
Turkey. On the other hand, Russia and China are vying for
expanded influence in countries critical to US diplomacy,
namely Iran, Saudi Arabia, Turkey, Jordan, Egypt, and
Pakistan.
The Middle East is emerging as an arena of intense
nuclear competition and positioning, with the first South
Korean nuclear unit recently completed at Barakah in the
United Arab Emirates, Jordan continuing to negotiate
on financing for two Russian nuclear reactors, Egypt
beginning construction of a nuclear station at Akkuyu
with Russia, and Saudi Arabia announcing its intention
to proceed with two reactors after years of delay. The
Chinese, French, Russians, and South Koreans have
submitted initial bids in Saudi Arabia, and a US has also
submitted a bid on this first phase of the process of
short listing companies. The bid was approved by the US
Department of Energy (DOE), even though the US has not
yet concluded a 123 nuclear framework agreement with
the Saudis, which would be necessary before a US export
deal could be finalised.
Drew Bond, a senior fellow and director of energy
innovation programmes at the American Council for
Capital Formation Centre for Policy Research, agrees
that this is a critical time for the Trump administration,
energy secretary Rick Perry and US domestic nuclear
infrastructure. He says the country’s 30-year hiatus in
building new reactors coupled with the rise of state-owned
competitors abroad has taken “a significant toll on the US
nuclear industry and has seriously undermined America’s
global influence over nonproliferation and other matters”.
The US used to be the overwhelming leader in
designing, building, and fuelling nuclear reactors around
the world, but no longer, said Mr Bond. “Unfortunately, in
recent years we have ceded this role – along with our
influence – to other nations, particularly Russia, China,
and South Korea. More than a dozen countries have
planned or proposed to build new reactors in the coming
years. Whether those reactors are designed and built up to
US or Russia safety standards is critical, not to mention the
geopolitical implications for the world.”
President Trump and his administration have been
calling for an “all of the above” energy strategy that
achieves US energy dominance. But advanced fossil fuels
and renewables can’t do it alone. According to Mr Bond,
nuclear energy and the supply chain that comes with it
must be a part of the picture.
The Atlantic Council report, said the NEI, shows the
need for the administration and Congress to support
American commercial nuclear exports through concrete
action.
“It’s critical for our industry that, given aggressive
overseas, state-owned competitors, we work with the
White House and Congress to give American companies
the tools they need to compete and win abroad,” NEI
vice-president Dan Lipman said.
“That means reestablishing a quorum at Ex-Im Bank,
ensuring US expot controls for nuclear technology are
more efficient, ensuring Section 123 bilateral nuclear
cooperation agreements are concluded, and fully funding
commercial nuclear energy research and development in
the federal budget.
“It’s not only American jobs that are at stake, but our
influence on safety, security and nonproliferation norms
across the world.”
Author
NucNet
The Independent Global Nuclear News Agency
David Dalton
Editor in Chief, NucNet
Avenue des Arts 56
1000 Brussels, Belgium
www.nucnet.org
The Atlantic Council
report is online:
https://bit.ly/
2GmNx3k
INSIDE NUCLEAR WITH NUCNET 435
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A Stark Warning to Trump on China, Russia and the ‘Crisis’ Facing US Nuclear Industry ı NucNet, David Dalton
atw Vol. 63 (2018) | Issue 8/9 ı August/September
436
CALENDAR
Calendar
2018
02.09.-06.09.2018
19 th International Nuclear Graphite Specialists
Meeting (INGSM-19). Shanghai Institute of Applied
Physics, Shanghai, China, ingsm.csp.escience.cn
03.09.-06.09.2018
Jahrestagung des Fachverbandes Strahlenschutz.
Dresden, Germany, Fachverband für
Strahlenschutz e.V., www.fs-ev.org
04.09.-05.09.2018.
8. Symposium Lagerung und Transport
radioaktiver Stoffe. Hannover, Germany,
TÜV NORD Akademie, www.tuev-nord.de
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.
San Francisco, CA, USA, EPRI – Electric Power
Research Institute, www.epri.com
12.09.-14.09.2018
SaltMech IX – 9 th Conference on the Mechanical
Behavior of Salt. Hannover, Germany, Federal
Institute for Geosciences and Natural Resources
(BGR) Hannover, the Institute of Geomechanics (IfG)
Leipzig and the Technical University of Clausthal
(TUC), www.saltmech.com
16.09.-20.09.2018
55 th Annual Meeting on Hot Laboratories and
Remote Handling – HOTLAB 2018. Helsinki,
Finland, VTT and International Atomic Energy
Agency (IAEA), www.vtt.fi/sites/hotlab2018/
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
19.09.-21.09.2018
Workshop Sicherheitskonzepte Endlagerung.
Grimsel, Switzerland. Fachverband für Strahlenschutz
e.V., www.fs-ev.org
26.09.-28.09.2018
44 th Annual Meeting of the Spanish Nuclear
Society. Avila, Spain, Sociedad Nuclear Española,
www.sne.es
30.09.-05.10.2018
14 th Pacific Basin Nuclear Conference (PBNC).
San Francisco, CA, USA, pbnc.ans.org
30.09.-03.10.2018
Fifteenth NEA Information Exchange Meeting on
ctinide and Fission Product Partitioning and
Transmutation. Manchester Hall, Manchester, UK,
OECD Nuclear Energy Agency (NEA), National
Nuclear Laboratory (NNL) in co‐operation with the
International Atomic Energy Agency (IAEA),
www.oecd-nea.org
30.09.-04.10.2018
TopFuel 2018. Prague, Czech Republic, European
Nuclear Society (ENS), American Nuclear Society
(ANS). Atomic Energy Society of Japan, Chinese
Nuclear Society and Korean Nuclear Society,
www.euronuclear.org
01.10.-05.10.2018
3 rd European Radiological Protection Research
Week ERPW. Rovinj, Croatia, ALLIANCE, EURADOS,
EURAMED, MELODI and NERIS, www.erpw2018.com
02.10.-04.10.2018
7 th EU Nuclear Power Plant Simulation ENPPS
Forum. Birmingham, United Kingdom, Nuclear
Training & Simulation Group, www.enpps.tech
08.10.-11.10.2018
World Energy Week. World Energy Council Council’s
Italian Member Committee, www.worldenergy.org
09.10.-11.10.2018
8 th International Conference on Simulation
Methods in Nuclear Science and Engineering.
Ottawa, Ontario, Canada, Canadian Nuclear Society
(CNS), www.cns-snc.ca
10.10.-11.10.2018
IGSC Symposium 2018 – Integrated Group for the
Safety Case; Current Understanding and Future
Direction for the Geological Disposal of Radioactive
Waste. Rotterdam, The Netherlands, OECD
Nuclear Energy Agency (NEA), www.oecd-nea.org
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
15.10.-18.10.2018
International Conference on Challenges Faced by
Technical and Scientific Support Organizations
(TSOs) in Enhancing Nuclear Safety and Security:
Ensuring Effective and Sustainable Expertise.
Brussels, Belgium, International Atomic Energy
Agency (IAEA), www.iaea.org
16.10.2018
The next steps for nuclear energy projects in the
UK. London, United Kingdom, Westminster Energy,
Environment & Transport Forum,
www.westminsterforumprojects.co.uk
16.10.-17.10.2018
4 th GIF Symposium at the 8th 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
24.10.-26.10.2018
NUWCEM 2018 Cement-based Materials for
Nuclear Waste. 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 Plants. Magdeburg, Germany,
VGB PowerTech e.V., www.vgb.org
05.11.-08.11.2018
International Conference on Nuclear
Decom missioning – ICOND 2018. Aachen,
Eurogress, Germany, Aachen Institute for Nuclear
Training GmbH, www.icond.de
06.11-08.11.2018
G4SR-1 1 st International Conference on
Generation IV and Small Reactors. Ottawa,
Ontario, Canada. Canadian Nuclear Society (CNS),
and Canadian Nuclear Laboratories (CNL),
www.g4sr.org
12.11.-13.11.2018
15. Deutsche Atomrechtssymposium. Berlin,
Germany, Bundesministerium für Umwelt,
Naturschutz und nukleare Sicherheit, Wiss. Ltg. Prof.
Dr. Martin Burgi, www.grs.de/ars_anmeldung
13.11.-15.11.2018
24 th QUENCH Workshop 2018. Karlsruhe, Germany,
Karlsruhe Institute of Technology in cooperation with
the International Atomic Energy Agency (IAEA),
quench.forschung.kit.edu
22.11.2018
Weiterbildungskurs 2018 – IT-Sicherheit im Alltag
– Praxiswissen für Mitarbeiter in der Nukleartechnik.
Baden, Switzerland, Nuklearforum Schweiz,
www.nuklearforum.ch
03.12.-14.12.2018
United Nations, Conference of the Parties –
COP24. Katowice, Poland, United Nations
Framework Convention on Climate Change –
UNFCCC, www.cop24.katowice.eu
06.12.2018
Nuclear 2018. London, United Kingdom, Nuclear
Industry Association (NIA), www.niauk.org
2019
25.02.-26.02.2019
Symposium Anlagensicherung. Hamburg,
Germany, TÜV NORD Akademie, www.tuev-nord.de
10.03.-15.03.2019
83. Annual Meeting of DPG and DPG Spring
Meeting of the Atomic, Molecular, Plasma Physics
and Quantum Optics Section (SAMOP), incl.
Working Group on Energy. Rostock, Germany,
Deutsche Physikalische Gesellschaft e.V.,
www.dpg-physik.de
10.03.-14.03.2019
The 9 th International Symposium On
Supercritical- Water-Cooled Reactors (ISSCWR-9).
Vancouver Marriott Hotel, Vancouver, British
Columbia, Canada, Canadian Nuclear Society (CNS),
www.cns-snc.ca
09.04.-11.04.2019
World Nuclear Fuel Cycle 2019. Shanghai, China,
World Nuclear Association (WNA),
www.world-nuclear.org
07.05.-08.05.2019
50 th Annual Meeting on Nuclear Technology
AMNT 2019 | 50. Jahrestagung Kerntechnik.
Berlin, Germany, DAtF and KTG,
www.nucleartech-meeting.com – Save the Date!
27.10.-30.10.2019
FSEP CNS International Meeting on Fire Safety
and Emergency Preparedness for the Nuclear
Industry. Ottawa, Canada, Canadian Nuclear Society
(CNS), www.cns-snc.ca
Calendar
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Akademik Lomonosov:
Preparations for Premiere in Full Swing
Roman Martinek
At the end of July, the loading of the floating power unit Akademik Lomonosov with nuclear fuel started in Murmansk.
This is one of the key stages of the project, which as of today has no analogues in the world. In 2019, the power unit will
begin to supply local population and industrial facilities in North-Eastern Siberia with heat and electricity. The project
is expected to open up opportunities for the mass production of floating nuclear power plants – a number of countries
have already voiced their interest.
On July 25, the Russian city of Murmansk, the largest
non-freezing seaport in the world and the largest city
above the Arctic Circle, saw the start of the loading of
nuclear fuel into the reactors of the world’s only floating
nuclear power unit (FPU) Akademik Lomonosov. The
project, named after the outstanding Russian scientist and
laid down back in 2006, is the first in a series of mobile
transportable small-capacity power units. It is designed to
operate as part of a floating nuclear thermal power plant
(FNPP) and represents a new class of energy sources based
on Russian technologies of nuclear shipbuilding.
The Akademik Lomonosov is intended for the regions in
the High North and the Far East. Its main goal is to provide
energy to remote industrial facilities, port cities, as well as
gas and oil platforms located on the high seas. The permanent
mooring site of the floating NPP will be the Siberian
city of Pevek on the Chukchi Peninsula in the northeastern
extremity of Eurasia. The new plant will replace there two
technologically obsolete generation facilities: Bilibino NPP
and Chaunskaya CHPP. After being brought into operation,
the Akademik Lomonosov will become the northernmost
nuclear power plant in the world.
In the spring of this year, the floating power unit was
towed from the territory of the Baltic Shipyard, where its
construction was carried out from 2009, to the base of
Atomflot in Murmansk. During its transportation, the ship
144 meters long and 30 meters wide travelled the 4,000 km
route through the waters of four seas – the Baltic Sea, the
North Sea, the Norwegian Sea and the Barents Sea –
around the Scandinavian Peninsula and along the coasts
of Estonia, Sweden, Denmark and Norway. On May 19,
the Akademik Lomonosov was successfully moored in
Murmansk, where it was presented to the public in a
ceremonial atmosphere.
Vitaliy Trutnev, Head of Rosenergoatom’s Directorate for
the Construction and Operation of FNPPs, commented on
the current status of the project development: “Here in
Murmansk, we finalize the remaining technological
operations. Specialists have begun to implement one of the
most important tasks – the stage-by-stage loading of
nuclear fuel into the reactor plants. The next key stages
that are planned to be implemented before the end of
this year will be the physical launch of the reactors and the
beginning of complex mooring tests – after obtaining the
appropriate Rostekhnadzor permits (Federal Service for
Environmental, Technological and Nuclear Supervision –
author's note).
The FNPP project is based on the technology of
small modular reactors (SMRs) – this category, according
to IAEA classification, typically includes reactors with
electrical power up to 300 MW. A characteristic feature
of the majority of such designs is the integrated layout
of the reactor plant, in which the active zone, the steam
generator, the pressure compensator and a number of
other types of equipment are assembled in a single unit – a
factory-finished monoblock delivered ready-made to the
437
ENERGY POLICY, ECONOMY AND LAW
Energy Policy, Economy and Law
Akademik Lomonosov: Preparations for Premiere in Full Swing ı Roman Martinek
atw Vol. 63 (2018) | Issue 8/9 ı August/September
ENERGY POLICY, ECONOMY AND LAW 438
site. This technology has been known since the 1960s: for
instance, the U.S. floating nuclear power plant Sturgis was
used for ten years to provide energy to the Panama Canal
in case of a threat of an intentional failure of the groundbased
power supply system, but it was decommissioned in
1976. To date, despite the existence of many similar developments
in the world, the Akademik Lomonosov is the only
floating power unit in the world, which gives uniqueness
to the Russian project.
The FPU is equipped with two KLT-40S icebreaker-type
reactors with a capacity of 35 MW each – together they are
able to produce up to 70 MW of electricity and 50 Gcal/h
of heat energy in the nominal operating mode, which is
enough to support the life of a city with a population of
about 100 thousand people. In addition to the floating
power unit itself, the structure of the FNPP project 20870
includes hydrotechnical facilities that provide installation
and detachment of the FPU and transfer of generated
electricity and heat to the shore, as well as onshore
facilities for transmitting this energy to external networks
for distribution to consumers. Currently, specialists are
working on the creation of this infrastructure in Pevek.
One of the main features of the project being implemented
is the placement of two reactor units in a small
hull of the vessel while preserving all the functional
characteristics of the ground-based nuclear power plant
with fewer maintenance personnel. At the same time, the
highest reliability and safety of operation is provided with
no environmental impact.
The floating power unit is supposed to have a lifespan
of from 35 to 40 years. For its operation, low-enriched
uranium will be used, and spent fuel will be accumulated
on the platform itself. Once every three years, fuel will be
reloaded, with the average annual duration of the reactor
refuelling not exceeding 60 days. In addition, on an annual
basis, scheduled shutdowns will be carried out at the plant
for routine maintenance, the average annual duration of
which will be no more than 20 days.
In designing the Akademik Lomonosov, priority was
given to such aspect as the safety of its operation. The
technological solution for the design components of the
FNPP is based on the tried and tested reference technology
used on nuclear icebreakers since 1988. The icebreakers
Taimyr and Vaigach were used as prototypes – their reactor
units have operated without fail for several decades in
the most difficult conditions of the Arctic. At the same
time, it should be noted that the technologies of the reactor
facilities for the icebreaking fleet are constantly being
improved and have made a qualitative step forward since.
This development is taking into account the fact that
increasingly high demands are being placed on nuclear
safety in the world.
Energy Policy, Economy and Law
Akademik Lomonosov: Preparations for Premiere in Full Swing ı Roman Martinek
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Thanks to the use of this experience, the Akademik
Lomonosov is today equipped with advanced icebreaker
reactors, and the FPU vessel is designed to withstand a
collision with an iceberg, the pressure exerted by a tsunami
wave as well as hurricanes – this safety margin makes the
ship virtually unsinkable and invulnerable to natural
disasters. From the outside environment, the FPU premises
are insulated with a double hull of the vessel, and reactor
facilities are equipped with special biological barriers that
do not allow radiation to spread beyond the compartments
where these facilities are located.
The FPU vessel design has also taken into account the
climatic conditions in which the FNPP will be operated.
The main body and load-bearing structures are made of
steel, resistant to brittle fracture under low temperature
conditions. In addition, the FPU is equipped with ice
strengthening – additional structural elements that ensure
the vessel’s strength during navigation in ice-covered waters,
as well as all the means necessary for towing with the
help of an icebreaker.
The primary importance of safety in the operation of
small modular reactors is emphasized by Professor Marco
K. Koch, head of the working group Plant Simulation and
Safety at the Ruhr University Bochum, who is also a board
member of the German Nuclear Society (KTG): “Compliance
with all safety standards, including safe nuclear fuel
management, is absolutely imperative”. The expert also
highlighted the advantages of SMRs in this aspect:
“ Depending on the design chosen, it is possible to increase
the safety of small modular reactors by combining active
and passive safety systems. Due to the smaller size
and thus the lower capacity compared to today's power
reactors, in the event of a hypothetical accident, SMRs
have greater capabilities in terms of external cooling, as
well as a higher dynamics of reactor start-up and shutdown.
In addition, due to the lower inventory, absolutely less
fission products are produced”.
Another important feature of the FPU, which determines
the critical importance of technology for energy
supply to hard-to-reach areas, is its environmental
friendliness. Every day of the FPU operation, either directly
or indirectly due to gas savings, reduces annual consumption
to 200,000 tons of coal and 120,000 tons of fuel oil.
This seems particularly relevant in the light of the global
goals of the Paris Climate Agreement. As part of the fight
against climate change, the Russian side plans to reduce
greenhouse gas emissions by 2030 to 70 percent of the
1990 baseline. At the same time, the only way to achieve
these goals, in terms of the energy sector, is to implement a
program for the development of carbon-free energy.
“ Provided safety aspects are taken into account, small
modular reactor technologies are an environmentally
friendly alternative to energy supply due to the use of
smaller areas and the absence of CO 2 emissions”, agrees
Prof. Marco K. Koch.
The floating power unit Akademik Lomonosov is the first
representative in a series of plants, whose production is
planned to be established in the future, not least for
exports to other countries. “SMR concepts can really be of
interest for countries with decentralized energy supply”,
says Prof. Thomas Schulenberg, director of the Institute
of Nuclear and Energy Technologies at the Karlsruhe Institute
of Technology. “Decentralized energy supply should be
understood as an energy grid that is not interconnected, as
in Europe, but limited to small areas – for example, in
island regions such as Indonesia, or in sparsely populated
regions on land”, the professor explained.
The expert's words have been confirmed by real
experience: Director General of Rosatom Alexei Likhachev
noted interest in the new Russian development coming
from island states, including in South-East Asia. “In the
near future, we plan to move to negotiations on specific
deliveries, and if the result is achieved, sufficiently large
capacities of Russian shipbuilding will be loaded with
orders”, he added.
Prof. Marco K. Koch notes that small modular reactors
can be used both in countries that already have nuclear
infrastructure on their territory and in the countries that
are new to the industry. Another significant argument
in favor of the development of these technologies is
significantly lower financial costs compared to large energy
facilities. In Prof. Schulenberg’s view, a developing country
is very difficult to find an amount of 10 billion euros for the
construction of a large nuclear power plant – it is much
easier to get a loan for the amount of an order of magnitude
less. These circumstances lead to the conclusion that the
use of small modular reactors in floating power plants is
able to open a wide potential not only for energy supply to
remote regions, but also for expanding the club of states
using atomic energy for peaceful purposes.
Author
Roman Martinek
Expert for Communication
Czech Republic
ENERGY POLICY, ECONOMY AND LAW 439
Energy Policy, Economy and Law
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440
SPOTLIGHT ON NUCLEAR LAW
Atomausstieg letzter Akt?
Sind die neuen Entschädigungs regelungen für frustrierte
Aufwendungen und nicht mehr verstrombare Elektrizitätsmengen
im Atomgesetz verfassungsgemäß?
Tobias Leidinger
Kurz vor knapp hat der Gesetzgeber auf die verfassungsrechtlichen Mängel reagiert, die das Bundesverfassungsgericht
(BVerfG) in seinem Urteil vom 6. Dezember 2016 zum Atomausstieg (BVerfGE 143, 246) höchstrichterlich
beanstandet hat. Doch die neu geschaffenen Entschädigungsregelungen in der 16. AtG-Novelle werfen neue
Rechtsfragen auf, insbesondere die nach ihrer Verfassungsgemäßheit.
I. Die Vorgaben des Bundesverfassungsgerichts
Nach dem BVerfG-Urteil vom 6. Dezember 2016 musste
der Gesetzgeber bis zum 30. Juni 2018 in Bezug auf
den Atomausstieg einen verfassungsmäßigen Zustand
herstellen (vgl. dazu Leidinger, atw 2017, S. 26 ff.). Dies
erfolgt jetzt durch Entschädigungsregelungen, die durch
das Sechzehnte Gesetz zur Änderung des Atomgesetzes
(16. AtGÄndG), in das Atomgesetz eingefügt werden (vgl.
BT-Drs. 19/2508). Da das Änderungsgesetz im Hinblick
auf seine beihilferechtlichen Auswirkungen noch der
Überprüfung durch die EU-Kommission bedarf, kann das
Gesetz, das vom Bundestag am 28. Juni 2018 beschlossen
wurde, nicht sofort in Kraft treten.
Das Bundesverfassungsgericht hatte eine Kompensation
in zweifacher Hinsicht gefordert: Zum einen bedarf es
eines angemessenen Ausgleichs für frustrierte Aufwendungen,
die die Betreiber im Vertrauen auf den
Bestand der Ende 2010 zusätzlich gewährten Elektrizitätsmengen
getroffen hatten. Zum anderen ist eine Kompensationsregelung
für die Strommengen erforderlich, die
den Betreibern 2002 im Rahmen des „Energiekonsens“
(Atomausstieg I) zugestanden worden waren, die aber
nunmehr – infolge des endgültigen Atomausstiegs II bis
Ende 2022 – nicht mehr konzernintern verstromt werden
können. Letzteres betrifft allein die Betreiber Vattenfall
und RWE. E.ON verfügt noch über freie Kapazitäten, auch
wenn sämtliche eigenen Mengen verstromt sind. EnBW ist
nach eigenen Angaben nicht betroffen.
Neben dem Deutschen Bundestag hat sich auch der
Bundesrat mit den Regelungen befasst (BR-Drs. 205/18).
Auch eine Sachverständigenanhörung hat es dazu am
13. Juni 2018 im Umweltausschuss des Bundestages
gegeben. Die vom Bundesrat erhobene Forderung, im
Rahmen der gesetzlichen Neuregelung sicherzustellen,
dass Rest strommengen nicht auf norddeutsche Kernkraftwerke
(z.B. Emsland, Brokdorf) im Netzausbaugebiet
übertragen werden dürfen – weil dann die Einspeisung
regenerativer Energien eingeschränkt werde –, hat die
Bundesregierung – zu Recht – zurückgewiesen (BT- Drs.
19/2705). Eine solche Einschränkung von Übertragungsmöglichkeiten
müsste zu weiteren, nicht mehr
erzeugbaren Elektrizitätsmengen führen. Das wirft
erneut verfassungsrechtliche Fragen auf, insbesondere
nach einem finanziellen Ausgleich. Im Ergebnis käme es
zu einer noch größeren Belastung für den öffentlichen
Haushalt.
II. „Angemessenheit“ der Kompensation
von zentraler Bedeutung
Von entscheidender Bedeutung ist, ob durch die
neuen Entschädigungsregelungen die verfassungsrechtlich
ge botene Angemessenheit in Bezug auf frustrierte
Auf wendungen und nicht mehr verstrombare Strommengen
hergestellt wird. Denn die „Angemessenheit“
des Ausgleichs ist vom Bundesverfassungsgericht als
zentrales Kriterium einer verfassungskonformen Regelung
bestimmt worden. Fehlt es daran, wären die vom BVerfG
aufgestellten Maßgaben verletzt. Fraglich ist also, ob der
Gesetzgeber das ihm insoweit zukommende Gestaltungsermessen
verfassungskonform ausgeübt hat.
Für den Ausgleich nicht verstrombarer Strommengen
hatte das Gericht drei verschiedene Optionen eröffnet:
Zunächst wäre eine zeitlich auskömmliche Laufzeitverlängerung
bis zu dem Zeitpunkt denkbar, in dem die
ausgleichspflichtigen Strommengen tatsächlich konzernintern
verstromt sind. Das wäre – aus Sicht des Steuerzahlers
– der mit Abstand kostengünstigste Weg. Er wurde
indes nicht beschritten. Es bleibt vielmehr dabei, dass
die Nutzung der Kernenergie „zum frühestmöglichen
Zeitpunkt beendet werden soll“, d.h. es wird am Enddatum
31. Dezember 2022 unverändert festgehalten. Dieses
Datum beruht indes auf einer rein politischen Festlegung,
die bereits in der 13. AtG-Novelle im Jahr 2011 („Atomausstiegsgesetz“)
vorgenommen wurde. Sodann besteht die
Option, eine Weitergabemöglichkeit von Reststrommengen
zu ökonomisch zumutbaren Bedingungen gesetzlich
sicherzustellen oder – als dritte Möglichkeit – einen
angemessenen finanziellen Ausgleich für konzernintern
nicht verstrombare Reststrommengen zu gewähren.
III. Ausgleich für nicht mehr verstrombare
Elektrizitätsmengen
Das neue Gesetz bestimmt mit § 7f AtG (neu) einen
lediglich „konditionierten“ Geldausgleich für nicht mehr
verstrombare Elektrizitätsmengen. Danach müssen sich
die Kraftwerksbetreiber mit nicht verstrombaren Elektrizitätsmengen
zunächst, d.h. primär „ernsthaft darum
bemühen“, diese Mengen an andere Kraftwerksbetreiber
„zu angemessenen Bedingungen zu übertragen“, die zwar
noch über Kernkraftwerke, aber nicht mehr über Elektrizitätskontingente
zur Verstromung verfügen. Nur wenn und
soweit Strommengen zu diesen Bedingungen nicht mehr
übertragen werden konnten, greift dann – sozusagen
subsidiär – eine finanzielle Kompensation.
Es ist mehr als fraglich, ob das Gesetz mit dieser
Regelung den höchstrichterlichen Vorgaben gerecht wird:
Der vom Bundesverfassungsgericht festgestellte Verstoß
gegen Art. 14 Abs. 1 (Eigentum) und das Gleichheitsgebot
aus Art. 3 Abs. 1 GG resultiert doch gerade daraus, dass es
aufgrund des Ausstiegsgesetzes (13. AtG-Novelle) zu
einem Nachfragemonopol hinsichtlich der nicht mehr
verstrombaren Mengen kommt, also einer Situation, die
per se keine „angemessenen Bedingungen“ für eine
konzernübergreifende Veräußerung der Strommengen
zulässt (vgl. BVerfGE 143, 246 (361)).
Spotlight on Nuclear Law
Nuclear Phase-out Last Act? Are the New Compensation Regulations for Frustrated Expenses in Accordance with the Constitution? ı Tobias Leidinger
atw Vol. 63 (2018) | Issue 8/9 ı August/September
| | Blick auf den oberen Teil des Reaktordruckbehälters eines Kernkraftwerks in Deutschland während der Revision mit Brennelementwechsel.
SPOTLIGHT ON NUCLEAR LAW 441
Der Gesetzgeber hat sich damit für ein Regelungsmodell
entschieden, das die verfassungsgerichtliche Kritik
am Atomausstieg im Kern ignoriert: Die in ihren Grundrechten
verletzten Konzerne werden nicht etwa entschädigt,
sondern sollen ihre Reststrommengen zu
Bedingungen verkaufen, die das BVerfG als unzumutbar
und gleichheitswidrig qualifiziert hat.
Hinzu kommt, dass das Gesetz keine Regelungen trifft,
die Angemessenheit des Ausgleichs auf der Ebene der
Anteilseigner zu schaffen, sondern es stellt insofern allein
auf die Genehmigungsinhaber ab. Die Feststellungen des
BVerfG bezogen sich indes auf die beschwerdeführenden
Konzerngesellschaften RWE und Vattenfall, die an vorzeitig
abgeschalteten Anlagen wie Krümmel oder in ihren
Laufzeiten verkürzten Anlagen wie Gundremmingen
beteiligt sind. Diese Regelung führt dazu, dass Ansprüche
der Genehmigungsinhaber auf Ausgleich bei den Gemeinschaftsunternehmen,
an denen Vattenfall beteiligt ist, in
Höhe dieser Beteiligungsquote gekürzt werden. Es ist
fraglich, ob die so konzipierte Regelung den Vorgaben des
Urteils entspricht. Das BVerfG hatte es dem Gesetzgeber an
sich leicht gemacht, indem es die verfassungswidrige
Benachteiligung von RWE und Vattenfall in Bezug auf die
Reststrommengen konkret beziffert hatte: Für RWE waren
40 TWh und für Vattenfall 46 TWh bestimmt worden. Die
Gesetzesregelung bleibt hinter diesen höchstrichterlichen
Vorgaben zurück.
Schließlich führt die Entschädigungsregelung in § 7f
dazu, dass die genaue und endgültige Festsetzung des
Ausgleichs erst nach der Abschaltung des letzten deutschen
Kernkraftwerks mit Ablauf des 31. Dezember 2022
erfolgen kann. Das bedeutet weitere Rechtsunsicherheit
für die Ausgleichsberechtigten, denn die behördliche
Entscheidung darüber, ob die Übertragungsangebote
„ angemessen“ sind bzw. waren, ergeht erst nach dem
31. Dezember 2022 – im Zusammenhang mit der Entscheidung
darüber, ob und in welcher Höhe ein Ausgleich
gewährt wird. Wenn sich dann herausstellt, dass ein
Ausgleichsberechtigter die Übertragung zu für den Übernehmenden
günstigeren Konditionen hätte anbieten
müssen, ist sein Ausgleichsanspruch insoweit ausgeschlossen.
IV. Ausgleich für frustrierte Aufwendungen
§ 7e AtG (neu) sieht einen angemessenen Ausgleich für
Investitionen vor, die Kraftwerksbetreiber im Vertrauen
auf die Ende 2010 zusätzlich gewährten Elektrizitätsmengen
getroffen haben. Das Bundesverfassungsgericht
hat das für eine Kompensation relevante „berechtigte
Vertrauen“ auf die Zeit vom 28. Oktober 2010 bis zum
16. März 2011 beschränkt. Dabei kommt es nicht auf den
Zeitpunkt der Leistungserbringung, sondern den der
Vermögensdisposition an, z.B. die Eingehung einer vertraglichen
Verpflichtung. Das im Gesetz formulierte
Kausalitätserfordernis zwischen dem Entzug der 2010
gewährten Zusatzmengen und der Frustration von
Investitionen ist dem Wortlaut nach zu eng gefasst.
Investitionen sind zu berücksichtigen, wenn die Zusatzmengen
dafür ein tragender, nicht aber der alleinige
Grund waren. Auch der jetzt normierte Verweis im Atomgesetz
auf den Rechtsgedanken des § 254 BGB (Mitverschulden)
wirft für die Rechtsanwendung praktisch
schwierige Abgrenzungs-, Bewertungs- und Beweisfragen
auf. Erschwert wird die Problematik dadurch, dass für den
auf die Kompensation gerichteten Ausgleichsantrag eine
Ausschlussfrist von nur einem Jahr ab Inkrafttreten der
neuen Regelung gilt.
V. Rechtsunsicherheit verbleibt
Mit der Neuregelung der §§ 7e-g AtG verbleiben mithin
erhebliche Unsicherheiten: Sie resultieren nicht nur aus
einer Reihe neuer Begriffe, sondern vor allem aus dem
vom Gesetzgeber für die Entschädigung der nicht mehr
verstrombaren Reststrommengen gewählten „konditionierten
Entschädigungsmodell“; das so keiner der vom
Bundesverfassungsgericht eröffneten Regelungsoptionen
entspricht. Damit ist weiterer Streit über den Atomausstieg
vorprogrammiert.
Author
Prof. Dr. Tobias Leidinger
Rechtsanwalt und Fachanwalt für Verwaltungsrecht
Luther Rechtsanwaltsgesellschaft
Graf-Adolf-Platz 15
40213 Düsseldorf
Spotlight on Nuclear Law
Nuclear Phase-out Last Act? Are the New Compensation Regulations for Frustrated Expenses in Accordance with the Constitution? ı Tobias Leidinger
atw Vol. 63 (2018) | Issue 8/9 ı August/September
442
FUEL
Innovations for the Future
Westinghouse EnCore® Accident
Tolerant Fuel
Gilda Bocock, Robert Oelrich, and Sumit Ray
EnCore® and
ADOPTTM are trademarks
and registered
trademarks of Westinghouse
Electric
Company LLC, its
affiliates and/or its
subsidiaries in the
United States of
America and may be
registered in other
countries throughout
the world. All rights
reserved. Unauthorized
use is strictly prohibited.
Other names
may be trademarks of
their respective owners
The development and implementation of accident tolerant fuel (ATF) products, such as Westinghouse’s EnCore® Fuel,
can support the long-term viability of nuclear energy by enhancing operational safety and decreasing energy costs. The
first introduction of Westinghouse EnCore Fuel into a commercial reactor is planned for 2019 as segmented lead test
rods (LTRs) utilizing chromium-coated zirconium cladding with uranium silicide (U 3 Si 2 ) pellets. The EnCore Fuel lead
test assembly (LTA) program, with LTAs planned for 2022 insertion, will introduce silicon carbide/silicon carbide
composite cladding with U 3 Si 2 pellets.
Over the past several years, the
Westinghouse EnCore Fuel features
have been tested in autoclaves, in
research reactors, at national laboratories
and in the Westinghouse Ultrahigh
Temperature Test Facility to
confirm and fully understand the
science behind ATF materials. Based
on the positive results to date, fuel rod
and assembly design in preparation
for the LTR and LTA programs is
underway, as well as licensing efforts
with the U.S. Nuclear Regulatory
Commission (NRC). Accident analyses,
coupled with economic evaluations,
have been continuing to define the
value of ATF to utilities.
These new designs will offer
design- basis-altering safety, greater
uranium efficiency and significant
economic benefits. Adoption of the
Westinghouse ATF, in conjunction
with a transition to 24-month cycle
operation, is the recommended path
forward for implementation of the
Westinghouse EnCore Fuel.
1 Introduction
Nuclear energy remains a fundamental
component of many industrialized
nations’ energy supply mixes due to its
demonstrated reliability in baseload
electrical supply, as well as inherent
carbon-free energy production. Two
factors are critical to maintaining this
capability: (a) enhancing safety to
help safeguard the plant and public
from highly impacting events such as
that which occurred at the Fukushima
Daiichi Nuclear Power Plant and (b)
decreasing operating costs to compete
with other sources of energy. The
development and implementation
of Accident Tolerant Fuel (ATF) products,
such as Westinghouse’s EnCore®
Fuel features, can support both of
these critical factors for long-term
operation.
Development of nuclear fuels with
enhanced accident tolerance is being
accelerated to support implementation
into commercial reactors as soon
as possible. The major objectives for
ATF designs include: 1) improved
cladding reaction to high-temperature
steam; 2) reduced hydrogen generation;
and 3) reduced beyond design
basis accident source term. In addition
to improving safety margins
for light water reactors (LWRs), fuel
designs using advanced, ATF materials
can improve fuel efficiency, enhance
debris resistance and extend fuel
management capability. Encore Fuel,
being developed by Westinghouse
Electric Company LLC (Westinghouse),
includes two unique accident tolerant
or fault tolerant fuel designs: chromium
(Cr)-coated zirconium (Zr)
alloy cladding with uranium silicide
(U 3 Si 2 ) fuel pellets, and silicon
carbide (SiC) cladding with U 3 Si 2 fuel
pellets.
The first introduction of Westinghouse
EnCore Fuel into a commercial
reactor is planned for 2019 as segmented
lead test rods (LTRs). The
LTRs will utilize chromium-coated
zirconium cladding with U 3 Si 2 highdensity,
high-thermal conductivity
pellets. The EnCore Fuel lead test
assembly (LTA) program, planned
for 2022 insertion, will introduce
SiC/SiC composite cladding along
with chromium- coated zirconium
cladding and the high-density, /highthermal
conductivity U 3 Si 2 pellets
modified to achieve higher oxidation
resistance.
Over the past several years,
Westinghouse’s ATF test program
has tested the chromium-coated
zirconium and SiC claddings in
autoclaves and in the Massachusetts
Institute of Technology’s (MIT) reactor
and U 3 Si 2 pellets in Idaho National
Laboratory’s (INL) Advanced Test
Reactor (ATR). Tests in the Ultrahigh
Temperature Test Facility at
Westinghouse’s U.S. Materials Center of
Excellence Hot Cell Facility in Churchill,
Pennsylvania, have been carried out to
confirm the time and temperature
limits for the SiC and chromiumcoated
zirconium claddings. Additionally,
an extensive research program to
fully understand the science behind
ATF materials continues with the
Westinghouse-led International Collaboration
for Advanced Research on
Accident Tolerant Fuel (CARAT) group
and at United States (US) and United
Kingdom (UK) national laboratories.
Based on the positive results to date,
fuel rod and assembly design in preparation
for the LTR and LTA programs is
underway, as well as licensing efforts
with the U.S. Nuclear Regulatory Commission
(NRC), and accident analyses
coupled with economic evaluations
for both operating savings and fuel
savings have been continuing to define
the value of ATF to utilities.
2 Lead test rod program
LTR programs are an essential step in
the introduction of new nuclear fuel
technologies into commercial energyproducing
reactors. In the EnCore LTR
program, two Westinghouse 17x17
optimized fuel assemblies (OFA) will
contain up to 20 ATF rods with
Cr-coated Zirconium alloy cladding,
and U 3 Si 2 and enhanced ADOPT fuel
pellets in Exelon’s Byron Unit 2 in
Cycle 22. Coated tubes and U 3 Si 2 and
ADOPT pellets will be delivered to the
Westinghouse Columbia Fuel Fabrication
Facility for manufacturing of
the assemblies. The shipping date for
the assemblies containing the LTRs is
February, 2019.
Westinghouse is continuing development
work with the University of
Wisconsin-Madison to continue the
optimization of coating performance,
and also working with commercial
vendors and the U.S. Army Research
Lab (ARL) to scale-up production
to full-length tubes. The U 3 Si 2 fuel
Fuel
Westinghouse EnCore® Accident Tolerant Fuel ı Gilda Bocock, Robert Oelrich, and Sumit Ray
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Material Process Vendor Maximum
Days
Titanium Nitride/ Titanium
Aluminum Nitride
Average
Corrosion Rate
(mg/dm 2 /day)
Std. Dev.
Corrosion Rate
(mg/dm 2 /day)
Average
Zr Corrosion
(mg/dm/day)
Corrosion
Rate
(microns/year)
PVD* PSU** 169 1.07 0.80 2.22 7.67
Chromium Cold spray UW*** 20 0.03 0.06 3.27 0.14
*Physical Vapor Deposition **Pennsylvania State University ***University of Wisconsin
| | Tab. 1.
Autoclave Corrosion Performance for the Top Zirconium Alloy Coatings.
FUEL 443
pellets are being fabricated at INL. The
fuel rod and fuel assembly designs are
progressing and the manufacturing
plan is being refined.
3 Recent testing
3.1 Autoclave testing of ATF
claddings
A primary benefit of ATF coating is to
enhance survivability in high-temperature
steam or water conditions, as
may occur in postulated accident
scenarios. To demonstrate this improved
survivability, Westinghouse
has performed corrosion testing using
the autoclave facility at the Churchill,
Pennsylvania site to screen various
coatings and SiC preparation methods
for corrosion resistance. As part of a
multi-year program, more than 12
types of coatings on zirconium alloys
and approximately 10 versions of SiC
have been tested in autoclaves. As a
result of this testing, two coatings,
titanium-nitride/titanium-aluminumnitride
and chromium (Table 1), were
identified for testing in the MIT
reactor.
Testing in the MIT reactor further
narrowed the options to the chromium
coating (Figure 1). The chromiumcoated
zirconium showed no signs of
peeling and had minimal weight gain
after taking into account the uncoated
inner surface of the tube. The very
positive results from these tests helped
validate the viability of the Cr coating
for use in LTRs being inserted in a
commercial pressurized water reactor
(PWR).
Initial autoclave and reactor testing
resulted in relatively high levels of
SiC corrosion. Autoclave testing with
hydrogen peroxide was used to simulate
the more aggressive oxidation
conditions of the reactor and to
explore coolant conditions that would
minimize SiC corrosion rates. This
testing has been used to refine the
manufacturing parameters of the SiC
composites such that, along with
hydrogen addition to the primary
coolant, above 40 cc/kg [2], the
current corrosion rates for SiC meet or
exceed the target 7of microns/year
recession rate. For a full core of
SiC cladding, this would result in a
maximum of 150 kg of silicon dioxide
(SiO 2 ) or about 350 ppm over an
18-month cycle. This is well below the
solubility limit of ~700 ppm SiO 2 at
the coldest steam generator conditions.
Note also that commercially
available resins to remove SiO 2 could
be added to the current resins used
to maintain water chemistry on a
continuous basis.
In addition to corrosion resistance,
crud buildup on the outside surface
of fuel rod claddings has long been
identified as a potential factor in fuel
rod operation, especially at higher
operating temperatures. Westinghouse
continues to assess the potential for
crud buildup on advanced ATF claddings.
Limits on crud buildup on SiC
claddings are likely to be different than
for coated claddings because the SiC
surface may be corroding underneath
any potential crud buildup. Therefore,
testing in the high heat transfer rate
and crud deposition test loop (WALT
loop) at the Westinghouse facility in
Churchill, Pennsylvania, has been
carried out from mid-2017 and will
continue until 2019 to study heat
transfer rates and crud buildup on the
SiC and chromium-coated cladding
surfaces. Preliminary results indicate a
somewhat higher crud deposition rate
for chromium-coated cladding than for
uncoated zirconium cladding. Surface
treatments are being explored to
reduce the crud deposition rate.
3.2 High-temperature testing
of ATF claddings
One goal of the ATF program is to
develop fuels that can withstand
post-accident temperatures greater
than 1,200 °C without the cladding
igniting in steam or air. Therefore, a
crucial part of the testing carried out
by Westinghouse during the previous
year was aimed at quantifying the
maximum temperature at which the
ATF claddings could operate without
excessive corrosion. The test apparatus
(Figure 2) currently uses a
graphite rod which is inserted into insulation
and then into the test piece.
| | Fig. 1.
Chromium-coated zirconium alloy tubes before and after testing in the MIT
reactor [1].
| | Fig. 2.
SiC rodlet undergoing testing in the ultra-high
temperature apparatus in steam at 1600°C at
Churchill. The sample is mounted inside the
shield tube that is glowing white in the photograph.
The SiC tube is inside the shield tube
with steam injected both above and below
the sample. The steam exits through the hole
that is visible in the shield tube.
This results in a very stable heating of
the test pieces.
Chromium-coated zirconium has
now been tested at up to 1,500 °C. This
is above the chromium- zirconium low
melting eutectic point of 1,333 °C. At
1,400 °C, there was noticeable reaction
between the Cr and the Zr. However,
there was not the rapid oxidation that
Fuel
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FUEL 444
uncoated zirconium experiences at
1,200 °C. At temperatures of 1,300 °C,
however, the Chromium- coated zirconium
alloy was stable for reasonable
lengths of time. Combined with the
lowering of zirconium oxidation at
normal operating temperatures, which
vastly reduced the formation of zirconium
hydrides, and therefore embrittlement,
the chromium- coated zirconium
provides significant performance
improvements during normal operation,
transients, design basis accidents
and beyond design basis accidents, as
compared to uncoated zirconium.
Similar tests were run with SiC at
temperatures from 1,600 °C up to
1,700 °C. These tests were terminated
only because of excessive corrosion of
the heater element. At 1,600 °C, the
SiC cladding was visually untouched.
At 1,700 °C, there were indications of
small beads on the surface, presumably
SiO 2 from the reaction of SiC
with steam, but on the whole, no
significant deterioration of the SiC.
Changes are being made to the heating
rod to increase the flow of Helium
cover gas and to allow accurate weight
changes to be made on the SiC rodlets
so that kinetic data can be obtained.
3.3 Testing of Westinghouse
U 3 Si 2 ATF high-density fuel
U 3 Si 2 is a revolutionary material for
LWR fuel service because its inherent
thermal conductivity is much greater
than existing UO 2 -based fuel, resulting
in significantly lower pellet temperatures.
U 3 Si 2 -based fuel can also
have up to 17 percent greater uranium
density than UO 2 -based fuel, so considerably
more energy can be economically
realized from each individual
fuel assembly. However, due to
these differences, considerable data
is required on the behavior of U 3 Si 2
at LWR operating temperatures
(estimated to be from 600 °C and up to
1,200 °C during transients).
To obtain the necessary data,
U3Si2 fuel pellets were manufactured
at INL and put into rodlets in the ATR
in 2015. The first rodlets came out of
the ATR at the end of 2016 (Figure 3)
and post-irradiation examination
(PIE) was performed in the summer
of 2017 at INL [3]. The PIE results
indicate some small amount of
cracking that may have been due to
impurities within the U 3 Si 2 . Fission
gas release and swelling were both
essentially zero with an exit burnup
of 20 MWd/kgU. Considering the
ATR high heat generation rates (12 to
15 kW/ft), which are significantly
above the average of 5 kW/ft and peak
of 9 kW/ft normally found in LWRs,
this was exceptionally good behavior.
The next set of U 3 Si 2 pins is due out in
2018 and will have achieved a burnup
of 40 MWd/kgU.
U 3 Si 2 was tested for air and steam
oxidation and compared to UO 2 using
digital scanning calorimeters at both
the Westinghouse Fuel Fabrication
Facility in Columbia, South Carolina
(USA) [4] and at Los Alamos National
Laboratory (LANL) [5]. The Westinghouse
test results indicate that the
ignition temperatures for UO 2 and
U 3 Si 2 are between 400 °C and 450 °C.
The LANL results indicate an ignition
temperature of about 400 °C. The
reasons for this difference are being
studied. The heat and mass generated
by the oxidation of the U 3 Si 2 is considerably
higher than for UO 2 . The
effect of this difference in heat release
and mass on the stability of the rods
was investigated in rodlet tests in the
autoclaves in the Churchill facility
during the summer of 2017. Unacceptable
tube bulging was found and programs
are now underway to increase
the oxidation resistance of the U 3 Si 2 .
| | Fig. 3.
Neutron radiographs of 20 MWd/kgU U3Si2 pins from ATR. Note the lack of pellet cracking and
distortion.(Ref. 4).
It is noted, however, that ATF cladding
surfaces are much harder than zirconium
alloy cladding and grids, so it is
expected that the likelihood of grid to
rod fretting leakages will be greatly
reduced from the current ppm levels.
4 Accident scenario
evaluations
To assess and demonstrate the performance
of ATF materials in postulated
accident scenarios, Modular Accident
Analysis Program, Version 5 (MAAP5),
calculations were performed for chromium-coated
zirconium and SiC claddings
along with high-density fuels
for the station blackout scenario and
the Three Mile Island Unit 2 (TMI2)
small-break loss-of-coolant (LOCA)
scenario with replenishment of the
primary coolant [6].
The chromium-coated zirconium
option offers modest ATF gains
(~200 °C) before large-scale melting
of the core begins in beyond design basis
events, such as a long-term station
blackout. Though it would not prevent
the contamination of the PWR primary
loop due to ballooning and bursting at
about 800 °C to 900 °C, the chromiumcoated
zirconium option could prevent
a TMI-2 type of accident from extending
into the fuel meltdown phase and
prevent extensive contamination of
the containment and perhaps preserve
the nuclear plant. This is because,
although the Cr-coated Zr may begin
to fail as the temperature exceeds
1,400 °C due to eutectic formation, it
does not rapidly oxidize as uncoated
zirconium alloys do, and does not provide
a rapid energy input spike into the
core (Figure 4). Note that, in this case,
Iron-chromium-aluminum (FeCrAl)
was used to model the performance
of chromium-coated zirconium since
the temperature and oxidation performance
is about the same. The
results for the station blackout
scenario (Figure 5) indicate that
fission products can be contained
within SiC cladding for up to two
hours longer than current Zr-based
cladding due to its higher temperature
capability (~2,000 °C decomposition
temperature). These two hours can
be used to implement additional
responses by the operators. The lower
pressure in the system due to minimal
hydrogen production (Figure 6.) increases
the chances that alternate
means to feed cooling water to the
core at about 40 gpm can result in
avoidance of fuel melting, indefinitely
extending the coping time as long as
the water flow continues. The SiC
cladding, of course, prevents any
Fuel
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FUEL 445
| | Fig. 4.
Hottest core node for TMI-2 accident where coolant is restored
at ~9,900 seconds.
| | Fig. 5.
Hottest core node for PWR station blackout.
leakage of fission products into the
primary loop since it will not balloon
and burst. Due to the short timespan
before coolant was re-introduced to
the system, the SiC cladding would
have had no adverse consequences
from a TMI-2 type accident (Figure 4).
5 Transition cycle analysis
for optimum ATF
implementation in
current PWRs
5.1 U 3 Si 2 fuel
As previously noted, one of the
primary benefits of U 3 Si 2 is that it
increases the uranium density by up
to 17 percent as compared to UO 2 .
This yields an effective enrichment
of 0.84 weight percent U-235 as
compared to 0.71 weight percent
U-235 found in natural uranium. This
increase in density will support
improved fuel cycle economics and
reduce the total number of fuel
bundles that need to be inserted into
a reactor, resulting in significant
savings. Because of the increased
density, the use of U 3 Si 2 also extends
the energy output and cycle length
capability for PWR fuel assemblies,
while remaining below the 5 weight
percent enrichment limit for commercial
fuel. The Westinghouse ATF can
thus either decrease the fuel cycle cost
of 18-month cycles by reducing the
number of feed assemblies and increasing
fuel utilization, or it can
make 24-month cycles economical for
today’s uprated, high-power density
PWRs.
Economic analysis shows that the
Westinghouse EnCore Fuel has very
favorable economics, not only at the
ATF equilibrium cycle, but also during
the transition cycles from UO 2 to ATF.
This is especially applicable when
transitioning to a 24-month cycle
operational regime, which thus represents
the recommended path forward
for implementation. The higher
thermal conductivity of the U 3 Si 2 also
provides a very high tolerance for
transients while operating at higher
linear heat generation rates than is
possible for UO 2 – which will increase
plant operating margin. In addition,
the higher uranium density can
extend the core operating capability
compared to current fuels, while
maintaining the current 5 weight
percent 235U enrichment limit for
commercial fuel; yet enable economically
competitive fuel management
schemes for the longer cycles.
In particular, the introduction of
ATF in a current 18-month cycle
high-power density PWR to accomplish
a transition from UO 2 to ATF by
either maintaining the currently predominant
18-month cycle operational
regime, or extending it to a 24-month
cycle has been analyzed. Implementing
the Westinghouse ATF to achieve a
more cost effective 18-month cycle
will deliver fuel cost savings due to
fewer fresh assemblies per reload
and improved fuel utilization. Implementing
the Westinghouse ATF in
conjunction with a transition to
24-month cycle will yield economic
benefits due to the resulting reduced
number of outages and related
savings, which offset the slightly
higher fuel costs (as compared to
18-month cycle fuel costs). Analyses
have shown that the economic impact
of the transition cycles to implement a
24-month cycle operation with ATF is
significantly better than the economic
impact of transition cycles which implement
ATF and maintain an
18-month cycle operation.
It is anticipated that the fabrication
costs to make the U 3 Si 2 powder could
increase as compared to existing
UO 2 fabrication. However, after the
powder is made, only minor cost increases
are expected to occur in the
rest of the fuel manufacturing process.
Therefore, the overall cost increase
is anticipated to be offset by
the safety, economic and operational
benefits.
| | Fig. 6.
Total hydrogen generated for PWR station blackout.
5.2 Chromium-coated
zirconium cladding
Chromium-coated zirconium alloy
offers a higher accident temperature
capability, compared to uncoated
zirconium alloy cladding, of between
1,300 ˚C and 1,400 ˚C. The coated
cladding also reduces corrosion and
hydrogen pickup. Resistance to rod
wear is another benefit of this cladding
type. The potential for exothermic
reactions is greatly reduced
during LOCA or transient events
that lead to high-temperature fuel
transients. These attributes provide
both safety and economic benefits
that support licenseability and economically
viable transition scenarios.
5.3 SiC cladding
SiC cladding provides 25 percent lower
thermal neutron cross-sections
than current Zr cladding. This would
afford even greater neutron economy.
Additionally, the fuel and cladding
would be able to withstand temperatures
~2,000 ˚C in the event of a
beyond design basis accident. This
temperature increase could result in a
rise in design basis operating margins.
6 Licensing
To get EnCore Fuel licensed and
loaded into commercial reactor cores
in region quantities by 2027, Westinghouse
has initiated a program to
significantly compress the licensing
timeframe from initial testing to
Fuel
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446
OPERATION AND NEW BUILD
commercial delivery, while improving
the quality of the data and resulting
design models used to describe the
fuel. This new approach would be a
significant improvement compared
to the current, largely empirical
approach, which requires years to
obtain limited data from a very expensive
test reactor(s), as well as for the
fabricating, testing, cooling, transportation
and post-irradiation examination
of samples. To reduce the
licensing timeframe for EnCore Fuel,
Westinghouse plans to utilize:
• Atomic scale modeling:
• By utilizing first principles to
determine physical properties
of irradiated materials
• By leveraging Westinghouse
involvement in the Nuclear
Energy Advanced Modeling &
Simulation (NEAMS) Department
of Energy (DOE) program
on basic property prediction
• By leveraging Westinghouse
involvement in the Consortium
for Advanced Simulation of
Light Water Reactors (CASL) –
Virtual reactor design
• By continuing to utilize MedeA
and Thermo-Calc software
• Real-time data generation to verify
the atomic scale modeling:
• Poolside data generation
PP
Gamma emission tomography
based on gamma-ray spectroscopy
and tomographic reconstruction
can be used for
rod-wise characterization of
nuclear fuel assemblies without
dismantling the fuel to detect
pellet swelling, pellet- cladding
interaction and pellet cracking
PP
Potential use of a spectroscopic
detection system to select
different gamma-ray emitting
isotopes for analysis, enabling
nondestructive fuel characterization
with respect to a variety
of fuel parameters (fission gas
release)
• Wired or wireless transmission
technology for measuring
PP
Centerline temperature
PP
Fuel rod gas pressure
PP
Swelling of fuel
In addition to saving time and cost, with
this approach Westinghouse hopes to
achieve, an increased confidence by the
U.S. NRC due to the predictability of
performance that can be obtained since
the performance models will have a
theoretical basis in addition to an
empirical basis. There should also be
reduced time and effort due to the reduction
in the number of submissionreview-revision-
submission cycles. This
should remove the review process from
the critical path to commercialization.
Communication with the U.S. NRC
Commissioners, and coordination
between the DOE, NRC and industry
for licensing of ATF, are in progress
and continuing.
7 Conclusion
Westinghouse and its partners are
continuing to make good progress on
U 3 Si 2 fuel, SiC cladding, and chromium-coated
zirconium cladding. These
new designs will offer design-basisaltering
safety, greater uranium efficiency,
and significant economic
benefits per reactor per year for PWRs.
While all testing and development to
date has been engineered for LWR
designs, Westinghouse believes the
technology could provide some of the
same safety and economic benefits to
CANDU and other reactor designs.
Fuel and accident modeling with
other types of reactor systems will be
required to evaluate the actual potential
for these benefits. This, together
with more beneficial power peaks,
lower impact of the transition cycles
and reduced dependence on uranium
price assumptions, make adoption of
the Westinghouse ATF, in conjunction
with a transition to 24-month cycle
operation, the recommended path
forward for implementation of the
Westinghouse ATF, EnCore Fuel.
References
[1] Gordon Kohse, MIT, 2016.
[2] Ed Lahoda, Sumit Ray, Frank Boylan,
Peng Xu and Richard Jacko, SiC Cladding
Corrosion and Mitigation, Top Fuel 2016,
Boise, ID, September 11, 2016, Paper
Number 17450, ANS, (2016).
[3] Jason Harp, Idaho National Laboratory
preliminary photographs.
[4] Lu Cai, Peng Xu, Andrew Atwood,
Frank Boylan and Edward J. Lahoda,
Thermal Analysis of ATF Fuel Materials
at Westinghouse, ICACC 2017, Daytona
Beach, FL, January 26, 2017.
[5] E. Sooby Wood, J.T. White and A.T.
Nelson, Oxidation behavior of U-Si
compounds in air from 25 to 1000 C,
Journal of Nuclear Materials, 484,
pages 245-257 (2017).
[6] Eugene van Heerden, Chan Y. Paik,
Sung Jin Lee and Martin G. Plys,
Modeling Of Accident Tolerant Fuel
for PWR and BWR Using MAAP5,
Proceedings of ICAPP 2017, Fukui and
Kyoto ,Japan, April 24-28, 2017.
Authors
Gilda Bocock
Robert Oelrich
Sumit Ray
Westinghouse Electric Company
5801 Bluff Road
Hopkins, SC 29061, USA
Analyses of Possible Explanations for the
Neutron Flux Fluctuations in German PWR
Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
Revised version of a
paper presented at
the Annual Meeting
of Nuclear Technology
(AMNT 2017), Berlin.
During the last 15 years the neutron flux fluctuation levels in some of the German PWR changed significantly. During
a period of about ten years, the fluctuation levels increased, followed by about five years with decreasing levels after
taking actions like changing the design of the fuel elements [1, 2]. The increase in the neutron flux fluctuations resulted
in an increased number of triggering the reactor limitation system and in one case in a SCRAM [3].
There exist different possible explanations
how neutron flux oscillations are
caused by physical phenomena inside
a PWR. Possible explanations can be
based on complicated interactions
between thermo-hydraulical (TH),
structural-mechanical and neutron
physical processes (see Figure 1).
Yet, no comprehensive theory
exists, which can explain the neutron
flux fluctuation histories observed
in German PWR based on first
physical principles. Therefore, GRS
has started investigations to
explain the observed neutron flux
Operation and New Build
Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
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| | Fig. 1.
Possible causes for neutron flux oscillations.
fluctuations and amplitude changes
[4].
Characteristics of neutron flux
fluctuations in German PWR
The neutron flux level during full
power operation is measured by inand
ex-core detectors sensitive to
thermal neutrons [5]. In German
Vorkonvoi and Konvoi type PWRs in
total 16 ex-core detectors (ionization
chambers) are located at four azimuthal
positions (see Figure 2) and at
four different axial heights outside the
RPV wall within the biological shield.
The signals of the upper two and lower
two are combined. They measure
the neutrons released from the core.
Inside the reactor core eight measurement
rods are located (see Figure 2).
Each measurement rod consists of six
self-powered neutron detectors (SP-
NDs) located at different elevations.
The neutron flux measurements used
in the following analyses have been
provided by an operator of a German
Vorkonvoi PWR. The data were sampled
with 250 Hz after they had been
low-pass filtered with a cutoff frequency
of 100 Hz.
Figure 3 (left) shows the power
spectral densities (per Hz) of the neutron
flux signals of ex-core detectors
measured in a Vorkonvoi PWR. The
power levels measured at the four
different azimuthal positions show no
significant differences. The highest
power spectral density of the neutron
flux is measured at low frequencies up
to 1 Hz.
Figure 3 (right) shows the distribution
of the time-dependent spectral
power density. For each time step,
a Fast-Fourier-Transformation was
calculated, using the following
parameters: sampling frequency =
250 Hz, numbers of samples = 4096,
Hanning window function. The time
steps in Figure 3 (right) are separated
by 14.3 s, which is 7/8 of the length of
a single FFT window (16.4 s). For each
point of time and frequency the
spectral power level is color coded.
The spectral power density changes
over time in a “chaotic” way. This
means that the frequency of the
maximal power density changes over
time. This observation does not
change if the number of samples used
for the FFT or the time resolution used
for the calculation spectrogram is
reduced or increased.
The top row of Figure 4 shows the
measured coherence and the phase
angles of two combinations of two different
ex-core detectors each. The coherence
was calculated by dividing
the absolute value squared of the cross
correlation of the corresponding two
detector signals by the autocorrelation
of the signals. Both detector
combinations show a strong coherence
at 1 Hz. The phase of the complex
valued frequency dependent
cross correlation was used to calculate
the frequency dependent phase shown
in Figure 4 (converted into units of
degree). The two detectors located at
perpendicular horizontal positions
relative to the core center (at 45° and
135°) exhibit constant phase difference
in the frequency range up to
1 Hz. In contrast, the two detectors
located at opposing sides of the core
center (at 45° and 225°) show a nearly
constant phase difference of 180° in
the frequency range up to 5 Hz. This
phase difference of 180° can be found
for all detector combinations calculated
by the cross correlation for two
detectors placed at opposing sides
relative to the reactor center.
The bottom row of Figure 4 shows
the relative signal strengths over time
of six in-core detectors located at different
axial heights on the Co4 measurement
rod (see Figure 2 for the positions
of the detectors). Even though
the amplitude at different elevations
shows some differences (higher amplitudes
at middle elevations than at
lower or higher ones) the temporal
OPERATION AND NEW BUILD 447
| | Fig. 2.
Horizontal positions of in-core (marked light
blue) and ex-core (marked red) detectors
within the core shroud and outside the reactor
pressure vessel.
| | Fig. 3.
Power spectral density (left) and spectrogram (right) of ex-core detector measurements in a Vorkonvoi PWR.
Operation and New Build
Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
atw Vol. 63 (2018) | Issue 8/9 ı August/September
OPERATION AND NEW BUILD 448
| | Fig. 4.
Coherence and phase angles between different ex-core detectors (top),
relative neutron flux measurements at six different elevations of the C04
in-core measurement rod (bottom); all measurements in a Vorkonvoi PWR.
progressions of the curves are identical
for all six elevations. It has to be
emphasized that no time lag can be
identified between measurements at
the bottom of the reactor core
compared with measurements at
the top. The same signal pattern can
be observed for all eight in-core
measurement positions.
All these observations are consistent
with different measurements
and analyses done during the last
decades [6, 7, 8]. Fiedler [8] compared
neutron flux fluctuation levels
in different plant types. He found that
the prominence of the 180° phase
difference between opposing detectors
(referred to as “beam mode”) is
special to KWU type PWRs.
Possible explanation based on
thermo-hydraulics effects
Already at the beginning of the
1970s, a model was published [9, 10]
coupling a point-kinetics neutron
physics model with a one-dimensional
TH model. It allows predicting neutron
flux fluctuation levels based
on coolant temperature or density
oscillations. Based on this model
it is already possible to understand
essential characteristics of the neutron
| | Fig. 5.
Simulated temperature fluctuations in frequency (top, left) and time (top, right) domain; layout of the coupled ATHLET-QUABOX/
CUBBOX model for a mini-core (bottom, left) and the resulting neutron flux fluctuations spectrum (bottom, right).
noise spectrum qualitatively, e. g. the
dependency of the neutron flux fluctuation
amplitude on the value of the
moderator temperature coefficient.
Following this approach and based
on some new simulations with the
CTF/PARCS codes [11, 12] a model of
the reactor core has been developed
using a coupled version of ATHLET
and QUABOX/CUBBOX [13]. In [12]
temperature fluctuations at the core
inlet were applied based on different
spectral properties. Temperature
oscil lations based on a white noise
spectrum resulted in much smaller
power/neutron flux oscillations than
temperature oscillations based on a
low-pass-filtered spectrum. A possible
explanation for that observation
might be alias-effects due to the limited
spatial and temporal resolution of
the coupled system. To avoid such
problems with the coupled system of
ATHLET and QUABOX/CUBBOX, a
Kolmogorov type spectrum [14] has
been applied for the temperature
fluctuations at the inlet of the reactor
core. Figure 5 (top row, left) shows
the power spectral density of the
temperature oscillations over the
frequency. Such spectra were observed
in different reactors [15, 16,
17].
Based on the assumption that the
temperature fluctuations follow such
a Kolmogorov type spectrum the time
dependent temperature fluctuations
are calculated (Figure 5, top right).
The temperature fluctuations have the
same variance as a sine-wave with an
amplitude corresponding to 1 K.
The TH model layout is shown in
Figure 5 (bottom, left). It consists of
nine interconnected core channels
with common inlet and outlet thermofluid
elements. The mini core has a
typical neutron-physics characteristic
of an end of fuel cycle (EOC).
Figure 5 (bottom, right) shows the
power spectral density of the resulting
fluctuations in the reactor power
production, which is proportional to
the neutron flux amplitude. For frequencies
smaller than 3 Hz the calculated
power spectral density fits the
measured ex-core detector signals of a
Vorkonvoi PWR quite well over several
orders of magnitude. This suggests
that temperature fluctuations at the
inlets of the core channels are part of
the explanation. This model can also
explain the correlation between the
amplitude of the fluctuations and the
moderator temperature coefficient.
However, it is not possible to explain,
why no phase differences could be
observed between measurements of
Operation and New Build
Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
atw Vol. 63 (2018) | Issue 8/9 ı August/September
SPNDs at the same horizontal but
different axial locations. The transport
of temperature fluctuations
through the core channels should
result in a delay of measurements
between detectors in lower regions
and the upper regions of the core.
Furthermore, this approach cannot
explain the strict 180° phase differences
between measurement positions
at opposing sides of the reactor
(neither for ex-core nor for in-core
detector combinations). If this
approach should be continued sensitivity
studies on the parameters of
the applied Kolmogorov typed spectrum
will be necessary.
Possible explanations based
on mechanical motions
For decades, the analyses of neutron
flux fluctuations have been used for
the detection of mechanical oscillations
inside the reactor pressure
vessel, see e.g. [6, 8, 18]. However, the
mechanical oscillations considered in
these analyses are harmonic oscillations
with resonance frequencies
exceeding 2 Hz. Nevertheless, the
simultaneousness of detector signals
of one measurement rod as well as the
location of the maximum of the neutron
noise level in the middle of the
core height indicate that mechanical
motions in the reactor core, which
behave synchronous and without
phase differences over the full core
height, also contribute to the observed
fluctuations at low frequencies.
Point Source Model
To check whether the observed fluctuations
are consistent with a core wide
mechanical motion a model based on
a point source for the neutron flux has
been developed. The model is based
on the assumption that the signal
| | Fig. 6.
Moving point source model (yellow circles:
detectors used for trilateration, black star:
idle position of point source, blue star: point
position derived by trilateration, red circle:
estimation for position uncertainty).
| | Fig. 7.
Different detector combinations used for trilateration (left), estimated horizontal point source locations over time (right).
strengths at the detectors depend linearly
on the distances between the
point source and the detectors (see
Figure 6). Based on this assumption
the position of the point source can
be calculated by trilateration using
different detector combinations (see
Figure 7 left). Additionally, an estimation
of the uncertainty of the
assumed position of the point source
can be derived. The three combinations
considered here are the four
ex-core detectors (marked red), three
in-core detectors located at the left
side of the reactor core (marked
green), and three in-core detectors
located at the right side (marked
blue).
Figure 7 (right) shows for different
time steps the pathways of the
assumed point source calculated
by a combination of the four ex-core
detectors (red), three left in-core
detectors (green) and three right incore
detectors (blue). The position
calculated by the ex-core detectors is
scaled by a factor of 1/3 relative to the
center of the reactor core. Also shown
are the estimated uncertainties of the
point source position for the different
detector combinations.
The model results in consistent
point source location estimations for
the three detector combinations. Also
the estimated uncertainties are small
compared with the pathways of the
point source. If instead of the detectors
marked in Figure 7 (left) the two
inner-most detectors (J06, G10) are
included in the calculation of the
trajectories, no consistent trajectories
can be derived.
This indicates that a phenomenon
involving the full reactor core plays a
significant role for explaining the
observed neutron flux fluctuations.
But it cannot explain the shape of the
measured power spectral density.
Structural-Mechanics
Considerations on Core-Wide
Motions of Fuel Assemblies
and further Core Internals
A synchronous excitation or synchronization
via mechanical coupling
can lead to core-wide correlated
mechanical motions of fuel assemblies,
which effect both in- and excore
neutron flux instrumentation.
This explanation is supported by
both the successful simulation of
the detector signals by an empirical
model of a moving point source and
the correlation between the neutron
fluctuation levels and the use of fuel
assemblies with reduced lateral
stiffness due to changes in the spacer
design. It also explains the simultaneity
of signals at different vertical
levels and the bow-shaped vertical
amplitude characteristic with a maximum
at or slightly below middle core
height.
Core barrel, grid plate and the
collective of fuel assemblies form
an enhanced system of coupled
mechanical oscillators. Core barrel
motions can have additional effects
on the neutron flux signal via
modulation of absorption and
reflection in the water gap between
core barrel and reactor pressure
vessel. The fuel assemblies within
this coupled oscillator differ in type
and service time and thus mechanical
parameters, which can lead to chaotic
motions and interaction effects and
thus oscillations in a broad frequency
band. In a low-leakage loading pattern
the fuel assemblies with the longest
service time and lowest remaining
stiffness are located at the core
periphery, which can evoke additional
effects on the ex-core and outer
in-coresensors, e.g. via water gap
modulation or motion in a strong
flux gradient.
OPERATION AND NEW BUILD 449
Operation and New Build
Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
atw Vol. 63 (2018) | Issue 8/9 ı August/September
OPERATION AND NEW BUILD 450
There are three possibilities for an
excitation in general:
• Stochastic fluid forces from turbulent
flow would lead to oscillations
of components at their natural
frequencies. The lowest natural
frequency of the fuel assemblies is
reported around 2.6 to 4 Hz [6, 18],
which would not explain the coherence
maximum at 1 Hz. Calculations
with simplified finite element
models show that depending on
design and operational behavior,
i.e. lateral stiffness decrease due to
radiation induces spring relaxation
in the spacers, the lowest natural
frequency can be shifted significantly
towards lower values. In [6,
16] an additional mode of the fuel
assemblies around 1 Hz in form of
synchronously moving cantilevered
beams (fixed at the bottom) is supposed.
Nevertheless, regarding the
fixture of the fuel assemblies in the
grid plate, the manifestation of this
mode is questionable. A further
explanation is the excitation of the
coupled system of core barrel, grid
plate and fuel assemblies, which
might have additional natural
system frequencies below the natural
frequencies of the single fuel
assemblies.
• A second possibility would be the
existence of an excitation force,
which is oscillating at around 1 Hz
and evokes a subsequent transient
deflection of the fuel assemblies.
Pressure fluctuations from residual
imbalances of the coolant pump,
standing waves, cavity resonances
in the pressurizer or vibrations of
other components of the loop are
known to induce core barrel motions
which could propagate to the
fuel assemblies. Fluid mechanical
oscillating forces with direct effect
on the fuel assemblies, e.g. pressure
differences, are also possible.
• A third possibility would be a selfexcitation
of fuel assemblies in a
constant axial flow. Research on
fuel assembly bow gives hints that
in fluid-structure-interaction (FSI)
simulations local forces can arise
leading to instability of the zero
position of the fuel assembly [19].
To investigate and prove the mentioned
hypotheses, a coupled FSI
model of core components and the
surrounding fluid is essential.
Simulations of reflector
influence
Further, the reflector influence has
been studied by means of a simplified
2D core model, in which the reflector
Case description
Maximum (relative)
increase on the left side
cross-sections are manipulated in
order to simulate the effect of varying
water gap between core barrel and
reactor pressure vessel, which corresponds
to the reflector region. These
variations could be caused by mechanical
motions, e.g. of core barrel or
fuel assemblies at the core periphery,
and their effect increases with decreasing
boron concentration. In this
model the TH parameters are homogeneous
and representative of the
hot full power state at zero burnup.
Further assumptions are: fuel temperature
= 900 K, moderator density =
702 kg/m 3 and boron concentration
= 1,300 ppm.
Table 1 summarizes the results
obtained for different variations of
the thermal absorption and fast-tothermal
scattering crosssection. The
reflector is modified only in one half of
the core (the left side) to reproduce
the spatial oscillations observed in the
PWR. The results show that the effects
of thermal absorption and scattering
are additive. The amplitude of the
power variation can reach the same
order of magnitude as observed in the
PWR.
Additional study is necessary to
determine if actual mechanical motions
can cause such changes leading
to increase/decrease of the moderator
volume (coolant water) in the reflector
zone and in that way changing
the homogenized assembly crosssections.
In addition, time-dependent
simulations are needed to check if
the frequency observed in the PWR
can be reproduced. Nevertheless, this
preliminary result shows that this
hypothesis is very promising. The
recently published study [20] showed,
that a variation of the gap size
between fuel elements of about one
centimeter can result in changes
of the neutron flux amplitudes at
the ex-core detectors of up to the
order of magnitude of 10 %. Therefore,
the influence of mechanical
motions of the fuel elements relative
to each other and as an ensemble
relative to the reflector cannot be
ruled out as explanation of the observed
neutron flux oscillations.
Summary and outlook
Several models based on single
physical effects (TH fluctuations at
the core inlet, movement of a point
source, coupled oscillations of core
internals, changes in the reflector
coefficients) are used to simulate the
neutron flux. Each of these simple
models can reproduce some of the
characteristics of the observed neutron
flux fluctuations but does not
encompass all features observed in a
real reactor. This suggests that further
work on the combination of models
is needed. Thereby, the biggest challenges
will lie in FSI simulations of
fuel assemblies including further core
internals, neutron physics simulations
using time-dependent geometries,
and possibly the coupling of all three
physical models.
Acknowledgment
This work has been performed in the
framework of the German Reactor
Safety Research and was funded by
the German Federal Ministry for
Economic Affairs and Energy (BMWi,
project no. RS1533). The authors
would like to thank the operators of
one German Vorkonvoi PWR and one
Konvoi PWR for providing data of
neutron flux measurements.
References
Maximum (relative)
decrease on the right side
-10 % thermal absorption 4 % -3 %
-10 % scattering 7 % -5 %
-10 % thermal absorption
-10 % scattering
10 % -8 %
-20 % thermal absorption 11 % -7 %
-20 % scattering 14 % -11 %
| | Tab. 1.
Summary of the reflector study results.
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GWU built PWRs, Progress in Nuclear
Energy 85, pp 668-675, 2015.
2. Reaktor-Sicherheitskommission,
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3. Bundesamt für Strahlenschutz,
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14.12.2012.
Operation and New Build
Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
atw Vol. 63 (2018) | Issue 8/9 ı August/September
4. C. Bläsius et al., Untersuchungen der
Ursachen für Neutronenflussschwankungen,
GRS-408, Gesellschaft
für Anlagen- und Reaktorsicherheit
(GRS) gGmbH, 2016.
5. G. Kaiser et al., Reaktorinstrumentierung.
Prozeßtechnik und Leistungsregelung im
Kernkraftwerk, VDE Verlag, 1983
6. J. Runkel, Rauschanalyse in
Druckwasserreaktoren, 1987.
7. L. J. Kostić, J. Runkel, D. Stegemann,
Thermohydraulics Surveillance of Pressurized
Water Reactors by Experimental
and Theoretical Investigations of the
Low Frequency Noise Field, Progress in
Nuclear Energy 21, pp. 421-430, 1988.
8. J. Fiedler, Schwingungsüberwachung
von Primärkreiskomponenten in Kernkraftwerken,
2002.
9. G. Kosaly, M. M. R. Williams, Point
theory of the neutron noise induced by
in-let temperature fluctuations and
random mechanical vibrations, Atomkernenergie
18(3) p. 203-208, 1971.
10. G. Kosaly., L. Mesko, Remarks on the
transfer function relating inlet temperature
fluctuations to neutron noise, Atomkernenergie
20(1), pp. 33-36, 1972.
11. A. Abarca et al., Analysis of Thermalhydraulic
Fluctuations in Trillo NPP with
CTF/PARCSv2.7 Coupled Code, 23 nd
International Conference Nuclear
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Slovenia, 2014.
12. G. Verdú et al., Study of the Noise
Propagation in PWR with Coupled
Codes, International Conference on
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and Engineering (M&C 2011), Rio de
Janeiro, Brazil, 2011.
13. S. Langenbuch, K. Velkov, Overview on
the Development and Application of
the Coupled Code System ATHLET-
QUABOX/CUBBOX, Proceedings of
Mathematics and Computation,
Supercomputing, Reactor Physics and
Nuclear and Biological Applications,
Avignon, France, 2005.
14. J. O. Hinze, Turbulence, McGraw-Hill,
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15. G. C. Van Uitert, H. Van Dam, Analysis
of Pool-Type Reactor Noise, Progress in
Nuclear Energy 1, pp. 73-84, 1977.
16. E. Türkcan, Review of Borssele PWR
noise experiments, analysis and
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17. Hashemian et al., Sensor Response
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18. R. Sunder, Sammlung von Signalmustern
zur DWR-Schwingungs überwachung
– Informationsgehalt der
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GRS-A-1074, Gesellschaft für Reaktorsicherheit
(GRS) mbH, 1985.
19. A. J. Petrarca, Y. Aleshin, Y. Xu, R. Corpa
Masa, J.M. Gómez Palomino, Effect of
lateral hydraulic forces on fuel assembly
bow, Proceedings of the TopFuel Conference
in Zurich, Switzerland, 2015.
20. J. Konheiser et al., Investigation of the
effects of a variation of fuel assembly
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detection in a PWR, Journal of Nuclear
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pp. 188-195, 2017.
Authors
Joachim Herb
Christoph Bläsius
Yann Perin
Jürgen Sievers
Kiril Velkov
Gesellschaft für Anlagen- und
Reaktorsicherheit (GRS) gGmbH
Boltzmannstr. 14
85748 Garching, Germany
OPERATION AND NEW BUILD 451
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Operation and New Build
Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov
atw Vol. 63 (2018) | Issue 8/9 ı August/September
OPERATION AND NEW BUILD 452
Revised version of a
paper presented at
the Annual Meeting
of Nuclear Technology
(AMNT 2017), Berlin.
Detailed Measurements and Analyses
of the Neutron Flux Oscillation
Phenomenology at Kernkraftwerk Gösgen
G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff
1 Introduction This paper summarises recent investigations [1], [2], [3] on measured neutron flux noise at
the Kernkraftwerk Gösgen-Däniken AG, who is operating since 1979 a German KWU pre-KONVOI, 3-Loop PWR with a
thermal power of 3,002 MWth (1,060 MWe). In a period of approx. 7 cycles from 2010 to 2016, an increase of the
measured neutron noise amplitudes in the in- and out-core neutron detectors has been observed, although no significant
variations have being detected in global core, thermo-hydraulic circuits or instrumentation parameters. Verifications of
the instrumentation were performed and it was confirmed that the neutron flux instabilities increased from cycle to
cycle in this period. In the last two years, the level of neutron flux noise remains high but seems to have achieved a
saturation state.
In a power reactor, neutron noise is
the result of random fluctuations of
many parameters, primarily neutronic
ones such as the number of neutrons
emitted per fission, thermal-hydraulic
parameters such as the fluctuations of
the primary water inlet temperature,
and mechanical parameters as for
example main circulation pump vibrations
or core internal vibrations. In
a KWU-PWR as KKG, the significant
neutron noise is observed at a frequency
in the range of 0.1 Hz to about
10 Hz, with a peak close to 1 Hz. Each
component has a typical spectral
response in the frequency domain,
and such a spectrum analysis can be
used as a diagnostic tool for surveillance
[4]. A significant variation of
the measured spectrum during a cycle
can be potentially interpreted as of
relevance for the plant performance
or safety. For that reason the Reactor
Pressure Vessel (RPV) and main
| | Fig. 1.
Schematic representation of the 3002 MW 3-Loop KKG core and the radial
positions of the in-core (left white on the map) and ex-core neutron flux
detectors. The colour map shows the relative power map (Fq) at the
assembly level. The inner axial flux distribution is monitored via six axially
and uniformly distributed in-core Self-Powered Neutron Detectors, while
the four radial ex-core channels contain two compensated ionisation
chambers, i.e. for the upper and lower core regions.
cir culation pumps at KKG are
equipped with acceleration and absolute
position sensors.
To deepen the understanding of
this behaviour, neutron flux signals at
different core locations and burnup
have been newly measured at a
sampling rate up to 100 Hz in order to
analyse possible spatial correlations
between the measured signals. The
measurements corresponded to
Middle- of-Cycle (MOC) and End-of-
Cycle (EOC) conditions, for two
successive cycles aiming at analysing
noise evolution, additionally to the
known linear increase during the
cycle. During the cycle itself, the noise
amplitude increase is linearly correlated
to the decrease of the negative
moderator temperature reactivity
coefficient (Γ T ), which is caused by
the decrease of the boron con centration
in the primary circuit; this
behaviour is well known and predictable.
The phenomena to be
investigated here is the variation from
cycle-to-cycle, which was unexpected.
Auto- and cross-correlations between
neutron signals in the time and
frequency domain were investigated
by means of signal analysis tools. In
this respect several hypotheses behind
the increase of neutron noise – e.g.
core loading pattern, fuel structure
design, variations of the core inlet
temperature, core asymmetry, etc. –
were identified and checked on
the measured high-frequency data.
Globally it was observed that the
highest neutron noise amplitudes
were to be found in one single core
quadrant, located between Loop 1 and
Loop 3 of the core. Radial correlations
were also identified between core
quadrants, but no measurable time
delays were found axially between
measurements from top and bottom
neutron signals.
Additional measurements of various
plant parameters were also performed,
in a second phase, to extend
the analysis not only to neutron flux
signals, but also temperature, pressure
or component vibrations. Correlations
between vibration signals and
neutron flux signals were analysed as
well.
A brief description of the KKG core
is provided in Section 2. The performed
measurements, neutron noise analysis
performed at KKG [3], along with the
results are described in Section 3.
Section 4 presents a summary of the
performed analysis and the current
model explaining its origin.
2 KKG Core design
The reactor is a Pressurized Water
Reactor (PWR) pre-KONVOI 3-Loop,
manufactured by KWU-Siemens with
a thermal power of 3002 MWth
(1060 MWe). The core contains 177
fuel assemblies with a 15 x 15 fuel
assembly layout and an active core
height of 352 cm.
Since 2014 (Cycle 36) the core is
for the first time fully loaded with
HTP fuel assemblies manufactured
by AREVA GmbH, whose fuel design
features Zircaloy/Duplex cladding
material, modern spacer grid geometries
and UO 2 fuel with 4.95%-wt
enrichment equivalent. The reactor is
typically operated at full power for
12-month cycles and has five different
radial burnup regions. The moderator
temperature coefficient of reactivity
Γ T is in the range of 30 pcm/K at BOC
to 70 pcm/K at EOC. The boron
concentration is typically 950 ppm at
BOC and is continuously decreasing
at a rate of ~ 3 ppm/day. The core
is operated at a maximal Linear
Heat Generation Rate (LHGR) of
525 W/cm, with an average power
density q’’’ of about 105 W/cm 3 [5].
Operation and New Build
Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff
atw Vol. 63 (2018) | Issue 8/9 ı August/September
| | Fig. 2.
Illustration of four ex-core neutron flux signals (S1 at 100 Hz measured during 10 s.
The core features 48 Rod Cluster
Control Assemblies (RCCA), the
absorber fingers being inserted within
20 guide tubes per fuel assembly. The
neutron flux within the reactor core is
monitored with six channels, each of
which contains six axial Self-Powered
Neutron Detectors (SPNDs) and a
3D aeroball system in 24 radial
positions. Four quadrants of the
core are equipped in the biological
shielding each with two ex-core
γ-compensated ionisation chambers
for the full power measurements:
one for the lower part of the core
and one for the upper part. The incore
and ex-core detectors allow a
detailed and continuous measurement
of the spatial distribution of
the neutron flux; an illustration of
the instrumentation's location is given
in Figure 1.
3 Neutron noise measurements
and analysis
In order to analyse the possible
reasons of the neutron noise increase
at KKG, the already existing neutron
flux measurements were complemented
in cycle 36 with two extensive
measurement campaigns using a
sampling rate of 100 Hz: one at
MOC and the other at EOC. Figure 2
depicts a typical ex-core neutron flux
measurement.
The in-core and ex-core neutron
signals, including signals from the
vibration monitoring system (“SÜS”)
of the RPV were measured for at least
two continuous hours. The large
amount of data were analysed with inhouse
MATLAB scripts in order to
determine and compare neutron noise
characteristics.
Figure 3 shows the Power Spectrum
Density (PSD) of two in-core
channels at core positions J14 and
G02. On the figure are depicted five
axial levels, the detector E01 is located
at the core top and E06 at the core
bottom. It can be observed that the
results have a non-white noise spectral
component and that position G02 has
lower neutron noise compared to J14,
although the core is symmetrically
loaded.
More specifically, auto- and cross-
correlations between the neutron
signals in the time and frequency
domain were carefully investigated;
Figure 4 describes these correlations
in a graphic form. The analysis of
these results led to the interesting
observation that no time shifts were
found for the axial measurements
between top and bottom neutron
signals; suggesting that the origin of
the increased neutron noise amplitudes
are not primarily associated
with inlet temperature variations
that would propagate vertically at
flow velocity and thus requiring ca.
1 second to propagate.
Figure 5 shows the Probability
Density Functions (PDF) calculated for
two in-core detectors J14 and G02. It is
interesting to notice that, although
the two detectors are symmetrically
located in the core, the shape of the
PDF is highly asym metrical for position
J14. The curve features an upper
OPERATION AND NEW BUILD 453
a) b)
| | Fig. 3.
Power Spectrum Density (PSD) of SPND- J14 (a) and symmetric core position G02 (b), calculated from a sample of 4096 points measured on 18.12.2014. The instrumentation channel contains
axially 5 detectors at different heights starting with detector E01 on the top of the fuel assembly to E06 to the bottom. Higher intensities are measured at low frequency (< 1 Hz). A second
small peak at about 1.8 Hz (J14) is typically identified and corresponds approximately to the first eigenfrequency of HTP fuel assemblies.
Operation and New Build
Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff
atw Vol. 63 (2018) | Issue 8/9 ı August/September
OPERATION AND NEW BUILD 454
tail distribution. High but short peaks
and low prob ability are mostly responsible
for the activation of the power
limitation function of the digital I&C
system. Probability density function
(e.g. Generalized Extreme Value GEV
a)
[6]) and fits of measured parameters
were calculated in an attempt to
predict maximal values and frequency
occurrences of the measured neutron
flux [3], which are of relevance for
operational core control.
Although ex-core raw signals
from the ionisation chambers are
electronically filtered in the signal
processing, high amplitude noise are
nevertheless registered with a certain
low residual probability of triggering
b)
| | Fig. 4.
Radial cross-correlations of the four ex-core detectors (a) and axial cross-correlations of in-core detector G02 (b) measured on 18.12.2014.
a)
| | Fig. 5.
Probability Density Functions (PDF) a) and probability distribution b) of in-core detectors at position J14 and position G02 axial level 5. The signals are fitted with
the Generalized Extreme Value (GEV) and Gauss functions (b). GEV fit is well suited for asymmetric distributions as observed at certain core positions in KKG.
b)
a)
| | Fig. 6.
Signal sample from vertical movement detector A1 located on the reactor pressure lid of the oscillation surveillance system (SÜS) recorded at MOC (a) and
probability distribution of two absolute position sensors at MOC (b).
b)
Operation and New Build
Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff
atw Vol. 63 (2018) | Issue 8/9 ı August/September
an alarm. If two channels out of
four are simultaneously measuring a
reactor power PPKG.2.Max._Signal >
103 %, an alarm will be activated in
the control room and RCCA insertion
will be activated in order to reduce the
neutron flux. For this reason, probability
density functions of the in-core
and ex-core detectors were specifically
analysed (Figure 5).
Additional to physical measurements
of neutron flux and vibration
signals (Figure 6), special care was
given to the signal analysis of digitallybuilt
signals, for example the corrected
reactor thermal power, used into the
digital I&C system. Useful insights,
among other the evolution of the
positions with high neutron noise,
were obtained by comparing statistical
distributions at MOC and EOC of
those signals.
The neutron noise evolutions of
the in-core and ex-core detectors are
presented in Figure 7. The withincycle
evolutions of the neutron noise
amplitude are to be seen mostly as
linear; local trends are observed
and well coincide with the average
neutron flux trend within the cycle,
whose distribution is a result of boron
acid concentration, burnup of hot
spots in the core, decrease of the
radial peaking factors and RCCA
positions.
The signal correlations given in
Figure 4 revealed that the noise
signals at two opposite sides of the
core had strong negative correlations;
detectors of instrumentation channels
1 and 3 are strongly correlated. This
means that the measured flux increase
in one quadrant is at the same time
compensated by a flux reduction in
the opposite core quadrant. The
analysis has also shown, as illustrated
in Figure 7, that the largest noise
| | Fig. 7.
In-cycle evolution of neutron noise (1-σ standard deviation) measured during Cycle 36: ex-core ionization chambers (S1 – S4) and
in-core SPNDs at axial position 5 (close to fuel assembly inlet). The peak observed at ~20 EFPD is the result of a conducted power
level change.
amplitudes are located primarily in
one quadrant of the core centred on
core position J14 between Loop 1 and
3. The reason for the high neutron
noise in this region was analysed.
It is to note here that the core fuel
loading is 90° symmetric whereas the
RPV with the three loops is 120°
symmetric, implying that there is no
simple core symmetry; in addition,
the individual symmetries show
deviations from theory. To illustrate
this assumption, it can be mentioned
that the thermal loops have different
thermal powers, and their layout is no
perfectly 120° from one another.
Further thermo-hydraulic investigations
would be required to check the
impact of these asymmetries on the
neutron noise amplitudes. It can also
be mentioned that the 48 RCCA are
not positioned with a 90° symmetry in
the core.
Finally, the within-cycle evolution
of neutron noise was compared, at a
macroscopic level, to plant-specific
parameters such as the reactor power,
calibrated ex-core and in-core LHGRs,
and the calculated core flowrate
deduced from the pressure sensors
in the three loops. For illustration
purposes, the neutron flux measured
by two different channels (Middle
range and SPNDs) and the primary
water temperature span are shown in
Figure 8.
4 Summary
The phenomena leading to an increase
of the neutron flux noise from
cycle to cycle since about 2010 have
been studied in detail through
detailed measurements performed in
the timeframe 2014 to 2015 over two
cycle at MOC and EOC states. The
results show that this increase can
OPERATION AND NEW BUILD 455
a) b)
| | Fig. 8.
Cycle Evolution during Cycle 36 at KKG of a) Measured neutron flux and b) Average core temperature difference (ΔT = T oulet – T inlet ).
Operation and New Build
Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff
atw Vol. 63 (2018) | Issue 8/9 ı August/September
456
DATF NOTES
hardly be attributed to the primary
water inlet temperature variations,
which remain relatively well known
since decades, because the noise has
essentially no time shift dependence
along the water flow through the
assembly channel. The high neutron
flux noise is concentrated essentially
in one quarter of the core, radial and
azimuthal correlations build a consistent
picture supporting this observation.
The model explaining the increase
of the neutron flux noise is at the
present time associated with the
replacement of FOCUS fuel assemblies
by the HTP assemblies, which took
place basically since 2010. The current
core configuration has no longer
FOCUS assemblies, and the (high)
neutron noise achieved seems to be
saturated, bracketing the period of
insertion of the HTP-assemblies well.
The reason for the neutron noise
increase is associated to the thermalhydraulics
pattern in the core, not fully
symmetric (3 loops with asymmetries),
probably promoting a more
intense cross flow towards one specific
loop that exercises a lateral dragging
force on the HTP assemblies. Since
these assemblies hold the fuel rods in a
less fixed way than the previous
FOCUS, with the purpose of minimising
the rod-to-grid fretting potential
further, the guiding tubes do not
count in HTP assemblies with the
stiffness of the fuel rods themselves to
give a combined, stronger assembly
stiffness, as it was the case of the
FOCUS assemblies. HTP are considered
to be mechanically more prone
to elastic lateral oscillations. The
increase of neutron flux noise would
be the result of larger variations of
the water gap thickness between
HTP assemblies, an effect that was
enhanced as the core was loaded
increasingly with HTP assemblies.
Further work is ongoing to
bring complementary information to
support or discard this assembly
behaviour model. In particular, KKG
participates in the CORTEX international
research programme within
the Horizon 2020 EU Framework
Programme for Research and Innovation,
and a different organisation
will take independent new measurements
to refine the analyses available.
Acknowledgments
The authors would like to thank
the Electrical Division at KKG for
their support and collaboration, in
particular R. Härry, K. Heydecker
and A. Ploner for performing several
additional measurements during last
cycle. We are also thankful to the
director of the Nuclear Fuel Division,
B. Zimmermann, for his support
during the course of this research.
References
[1] Neutronenflussrauschen, R. Meier,
ANO-D-41205, 2010. Restrictive.
[2] Noise Analysis of KKG’s neutron flux
detector signals, A. Alander, Studsvik
Scandpower, TN-04/2011, Document
Kernkraftwerk Gösgen-Däniken AG.
2011. Restrictive.
[3] Studie des Neutronenflussrauschens im
Zyklus 36, G. Girardin, Kernkraftwerk
Gösgen-Däniken, BER-F-78937, Internal
Document Kernkraftwerk Gösgen-
Däniken AG, 2015. Restrictive.
[4] Use of Neutron Noise for Diagnosis Of
In-Vessel Anomalies in Light-Water Reactors,
ORNL/TM-8774, 1984.
[5] KKGG – Reaktorphysikalische
Rechnungen für den 36. Zyklus; FS1-
0016977 v1, Endgültiger Umsetz plan
für den 35. BE-Wechsel (Stand:
10.06.2014), Internal Document Kernkraftwerk
Gösgen-Däniken AG, 2014.
Restrictive.
[6] Handbook of statistical Distributions
with Applications (Statistics: A Series of
Textbooks and Monographs),
K. Krishnamoorthy,
ISBN-978-1584886358.
Authors
Dr. Gaëtan Girardin
Fuel Assembly Design
Dr. Rudolf Meier
Nuclear Technic
Phys. Lukas Meyer
Core Surveillance
Phys. Alexandra Ålander
Transport and Storage
Dr.-Ing. Fabian Jatuff
Projects and Processes
Kernkraftwerk Gösgen-Däniken AG
Kraftwerksstrasse
4658 Däniken, Switzerland
Notes
For further details
please contact:
Nicolas Wendler
DAtF
Robert-Koch-Platz 4
10115 Berlin
Germany
E-mail: presse@
kernenergie.de
www.kernenergie.de
First half of 2018:
Electricity production
in Germany
For the first half of 2018, the seven nuclear
power plants in Germany produced about
34.8 billion kWh (net) electricity and had
therefore a share of 12.9 % of the whole
production.
Although five power plants were
tem porarily shut down due to scheduled
inspections, the nuclear energy shows a
rise of 9 % relating to its electricity
pro duction of the first half of 2017.
Net electricity production (269.5 billion kWh)
for first half of 2018 in percent
12.9
Nuclear
energy
41.4
Renewable
energy
among:
20.4 Wind power
8.5 Biomass
8.3 Photovoltaics
4.2 Hydro power
24.7
Lignite
7.6
Gas
13.4
Hard coal
Quelle: VGB; AG Energiebilanzen; Fraunhofer ISE
DAtF Notes
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Effects of Airborne Volatile Organic
Compounds on the Performance of
Pi/TiO 2 Coated Ceramic Honeycomb
Type Passive Autocatalytic Recombiner
Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
1 Introduction Ensuring the containment integrity during a severe accident in nuclear power reactor by
maintaining the hydrogen concentration below an acceptable level has been recognized to be of critical importance
after Fukushima Daiichi accidents. Although there exist various hydrogen mitigation measures, a passive autocatalytic
recombiner (PAR) has been considered as a viable option for the mitigation of hydrogen risk under the extended station
blackout conditions because of its passive operation characteristics for the hydrogen removal [1]. As a post-Fukushima
action item, all Korean nuclear power plants were equipped with PARs of various suppliers. The capacity and locations
of PAR as a hydrogen mitigation system were determined through an extensive analysis for various severe accident
scenarios [2]. For some plants, dual hydrogen mitigation systems were equipped with a combination of newly installed
PARs and the existing igniters that each system has 100 % of full capacity for hydrogen control for postulated severe
accident conditions. Among a total of 24 operating units in Korea, a Pt/TiO 2 coated ceramic honeycomb type PAR
supplied by Ceracomb Co. Ltd. [10] has been installed in 18 operating plants and almost units have reached the second
or the third overhaul period since their first installation in 2013.
457
RESEARCH AND INNOVATION
The PAR makes use of a catalyst to
convert hydrogen (H 2 ) and oxygen
(O 2 ) into water vapor and heat. The
heat of reaction creates a natural
convective flow through the recombiner,
eliminating the need of pumps
or fans to transport new hydrogen to
the surface of the catalyst. In spite of
an advantage of its passive operation,
there have been concerns about
adverse effects on the performance of
PARs by potential deactivators (chemical
poisons and physical inhibitors)
[3, 4, 5]. PARs are required to perform
their safety function not only after
exposure to potential contaminants
during operation, but also in an accident
environment that may contain
various gases or aerosols that are
potentially poisonous to the PAR
catalyst elements [6]. The Ceracomb
also has performed various tests and
demonstrated that its performance
degradation of hydrogen removal
capacity is within 25 % in severe
accident conditions such as fission
product poisons, aerosols, cable
burns, carbon monoxide, etc. However,
its performance under the longterm
exposed condition to containment
air has not been fully investigated
because the Ceracomb PAR has
no operational experience in nuclear
power plants.
Under the long-term exposed condition
by airborne substances, it
has been known that the catalyst
shows a delayed response for hydrogen
removal [6]. These airborne substances
are known as volatile organic
compounds (VOC) that adsorb on
active sites of the catalyst surface thus
making them unavailable for catalytic
reaction to proceed. As a result, the
recombiner would require either a
higher hydrogen concentration, or a
higher temperature, or both, to start
the hydrogen recombination reaction,
compared with the catalyst in as-new
condition. The VOCs could be originated
from solvents, lubricants, oils,
insulations and paints, etc. which are
commonly used materials in the plant
maintenance. The key prameters of
catalyst performance under the longterm
exposed condition of VOCs
could be the start-up delay time for
catalyst reaction and its hydrogen
depletion (removal) rate because
these parameters directly affect the
results of hydrogen control analyses
in design basis and severe accident
conditions. The catalyst performance
should be verified up to sufficient
periods of plant operation and be
compared with the parameters on the
PAR performance used in the hydrogen
control analysis. Therefore, with
the exposure time to containment air,
the VOC effects will play a more
important role in PAR maintenance
during normal nuclear power plant
operation [7]. In comparison to the
performances under the accident conditions,
however, the performances
under the long term exposed condition
to containment air during normal
operation (i.e., effects of volatile organic
compounds) have not been fully
investigated becaue it requires long
time up to several overhaul periods in
the containment to obtain catalyst
samples and it includes the proprietary
information of PAR suppliers and
utilites.
This paper describes the test results
on the effect of airborne volatile
organic compounds in the containment
air on the performance of TiO 2
coated ceramic honeycomb type PAR
in Korean nuclear power plants
performed in 2014 ~ 2016 overhaul
periods. The test plants are extended
to seventeen (17) operating plants
compared to the previous eight (8)
operating plants [8]. A total of 152
tests have been performed with 680
catalyst samples to investigate the
effect of volatile organic compounds
(VOC) on the start-up performance on
the hydrogen removal. A total of 62
tests have been performed with 248
catalyst samples to identify the influence
on the hydrogen depletion rate
by the VOC effects. The analysis for
VOC components has been performed
for selected samples from seven (7)
plants to identify airborne substances
adsorbed on the surface of catalysts
using a qualitative GC/MS (gas
chromatograph/mass spectrometer)
method.
2 Test method
2.1 Pt/TiO 2 ceramic
honeycomb PAR
Figure 1 shows an illustrated view of
Pt/TiO 2 coated ceramic honeycomb
type PAR that has been installed in
eighteen (18) operating units. This
type of PAR has been developed and
Research and Innovation
Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
atw Vol. 63 (2018) | Issue 8/9 ı August/September
RESEARCH AND INNOVATION 458
| | Fig. 1.
Pt/TiO 2 Coated Ceramic Honeycomb type Passive Autocatalytic Recombiner.
| | Tab. 1.
Specifications of Ceracomb PAR.
Small-Size Medium-Size Large-Size
Weight (kg) 42.1 75.8 144.3
Width (cm) 37.8 72.5 142.8
Depth (cm) 34.3 36.5 36.5
Height (am) 100 100 100
No. of Catalysts 4 8 16
H 2 Depletion Rate (g/sec) a
(4 %-H 2 , 60 °C, 1.5 bar)
> 0.2 g/sec > 0.4 g/sec > 0.9 g/sec
a) Required hydrogen depletion rate in the technical specification for PAR purchase of Korean NPPs.
2.2 Test facility
The VOC effect tests have been performed
using the PAR performance
test facility (PPTF). The PPTF comprises
a carbon steel pressure vessel
with the internal volume of 12.5 m 3 (a
cylindrical shape with 3.3 m in height
and 2.2 m in diameter). It was constructed
to perform performance tests
in various conditions of pressure,
temperature, humidity, hydrogen
concentration and chemical water
spray. Figure 3 shows types and locations
of measurements in the pressure
vessel of PPTF. Inside of the vessel,
mixing fans, spray nozzles and electrical
heaters are installed to maintain
a desired test condition. At the center
of the vessel a test PAR is located. A
small-sized PAR with four (4) catalysts
is used as a test PAR. Gates are
equipped at the PAR entrance and exit
to prevent air and hydrogen from
being in contact with the catalyst
surface before the test starts. The
hydrogen concentration is measured
with an accuracy of 2 % of full scale
sampling rate. The time lag of the
hydrogen concentration signal due to
the length of the gas sampling line is
estimated as below 50 sec.
| | Fig. 2.
Ceramic Honeycomb Catalyst.
supplied by Ceracomb Co. Ltd. [9, 10].
The Ceracomb PAR consists of a
stainless steel housing equipped with
catalysts inside the lower part of the
housing. The PARs are installed with
floor mount type or wall mount type
in the containment and its structures
are designed to meet the seismic
requirements of each plant. Air and
hydrogen mixture flows from bottom
of the PAR to the exit openings at the
upper part of PAR. The housing is
designed to have chimney effects so
that the heat generated in the catalytic
reaction in lower part of the housing
can promote a strong driving force for
natural convective flow and to protect
the catalyst from the direct impinge of
containment spray. There are three
different sizes of PAR according to the
number of the catalyst. The specifications
of the Ceracomb PAR are
summarized in Table 1.
Different types of catalytic recombiners
have been supplied by various
PAR suppliers such as AREVA, CANDU
Energy, NIS (formerly NUKEM), KNT
and Ceracomb. AREVA, CANDU Energy
and NIS utilized plate type catalysts
while original NUKEM invented a
specialized cartridge containing pellet
type catalysts. KNT and Ceracomb PAR
utilized ceramic honeycomb type
catalysts. In the present Pt/TiO 2
coated ceramic honeycomb type PAR,
a cubical catalyst with a honeycomb
microstructure has been used to
increase the surface area for the
reaction. The catalyst is manufactured
by coating a mixture of TiO 2 and Pt on
the supporting structure of the ceramic
honeycomb of 35 CPSI (cell per
square inch). Figure 2 shows an
illustrated view of ceramic honeycomb
catalyst. The dimensions of the
standard honeycomb catalyst are
15 cm by 15 cm with the height of
5 cm. A protected metal frame is
used to protect the catalyst because
the ceramic catalyst is fragile and
vulnerable to impact.
2.3 Test methods
Key parameters of catalyst performance
are considered as the start-up
delay time and hydrogen removal rate
which are directly related to PAR
modeling in the hydrogen control
analysis to determine the capacity and
locations of PAR system [2]. Under
the VOC-affected conditions, its performance
is hard to identify through
the perioic inspection method because
the start-up delayed time and
the hydrogen removal rate are defined
under the natural convection conditions.
Therefore, a number of catalysts
are withdrawn out of containment
during an overhaul period of each
plant and their performance is tested
in the PAR performance test facility
(PPTF) under the natural convection
conditions. Table 2 shows the number
of catalysts taken from various plants
for VOC effect tests performed during
2014 ~ 2016 outage periods in seventeen
(17) plants. Further tests for
other plants are scheduled according
to their outage schedules. The VOC
effect tests are performed into three
groups; (a) the measurement of startup
delay time for hydrogen removal,
(b) the measurement of hydrogen
depletion (removal) rate and (c) VOC
component analysis to identify airborne
substances adsorbed on the
Research and Innovation
Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
atw Vol. 63 (2018) | Issue 8/9 ı August/September
| | Fig. 3.
PAR Performance Test Facility (PPTF) with Measurement Types and Locations.
Plant ID Plant Type Test Date
(yyyy/mm)
| | Tab. 2.
Number of Catalysts for VOC Effect Tests.
surface of catalysts. Four (4) catalysts
are withdrawn from one PAR considering
the installed location in the
containment so that the catalyst
samples be distributed uniformly
throughout containment area in order
to avoid local effects of the test results.
The exposure time of catalysts to the
containment air includes the normal
operation time of ~18 months and the
plant outage time that depends on the
outage schedule of each plant.
The start-up delay times for hydrogen
removal are measured in the PPTF
facility. Four (4) catalysts are mounted
in a small sized test PAR housing
No. of catalysts (No. of Tests)
Delay
Time
Depletion
Rate
VOC
Component
C1 PWR (W) a) 2014. 04 8 (2) 8 (2) 1
C2 PWR (W) 2015. 08 32 (8) 20 (5) -
D1 PWR (W) 2014. 11 28 (7) 12 (3) -
D2 PWR (W) 2016. 02 20 (5) 12 (3) -
F1 PWR (W) 2014. 11 28 (7) 20 (5) -
F2 PWR (W) 2016. 06 20 (5) 12 (3) -
G PWR (F) b) 2014. 12 84 (21) 20 (5) 1
H1 PWR (F) 2014. 11 84 (21) 20 (5) 4
H2 PWR (F) 2016.06 84 (21) 12 (3) 1
L PWR (O) c) 2014. 07 20 (5) 12 (3) -
M1 PWR (O) 2014. 07 40 (10) 12 (3) 1
M2 PWR (O) 2016. 04 20 (5) 12 (3) -
N PWR (O) 2015. 03 40 (10) 12 (3) -
O PWR (O) 2014. 12 20 (5) 12 (3) -
P PWR (O) 2014. 06 40 (10) 20 (5) 1
W PHWR d) 2015. 10 20 (5) 12 (3) -
Y PHWR 2014. 07 20 (5) 12 (3) 1
Total 608 (152) 248 (62) 10
Notes: a) PWR (W) : Westinghouse designed PWR b) PWR (F) : Framatome designed PWR
c) PWR (O) : Optimized Power Reactor (OPR) 1000 d) PHWR : CANDU6
* Each Data sets of C1/C2, D1/D2, F1/F2, H1/H2 and M1/M2 represent the same plants
but the tests are performed on different outage schedule.
that is the same model of the commercial
PAR so that four (4) catalyst samples
are used for a test. The test PAR is
installed at the center in the test vessel
of the PPTF. After the test vessel is
closed, mixing fans are turned on and
the hydrogen is injected to a desired
hydrogen concentration. Until desired
conditions are achieved, gates at the
PAR entrance and exit are closed in
order to prevent air and hydrogen
from being in contact with the catalyst
surface. The start-up delay tests are
performed at the initial conditions of
the hydrogen concentration of
3 vol. % and temperature of 60 °C
under the pressure of 1.5 bar (abs).
The start-up delay time is defined as
the required time for the hydrogen
concentration in the test vessel to start
to decrease by one percent (relative)
of the initial hydrogen concentration
after the hydrogen in the test vessel
starts to contact the catalysts in the
PAR (i.e., the gates at the PAR entrance
and outlet are opened).
The hydrogen depletion rates with
degraded catalysts under the normal
operation environments for an overhaul
period are measured using the
PPTF facility. The tests are performed
with the same procedure of the startup
delay time tests but with different
initial conditions. The hydrogen
depletion tests are performed with
the initial conditions with a hydrogen
concentration of 6.9 vol. % and temperature
of 60 °C under the pressure
of 1.5 bar (abs). The hydrogen
depletion rate is calculated from the
gradient of the hydrogen concentration
when the concentration at the
PAR entrance is 4 vol. %. The hydrogen
depletion rate from the present
tests are compared with the hydrogen
depletion rate required in the technical
specification of PAR purchase,
which is defined as above 0.2 g/s for
the small sized PAR at the conditions
of 4 vol. % of hydrogen, temperature
of 60 °C and pressure of 1.5 bar.
The composition adsorbed airborne
substances on the catalyst
surfaces is analyzed with GC/MS (gas
chromatograph/mass spectrometer)
method. Tests are performed by
Frontier Laboratories Co. Ltd. [11]
using Agilent 6890 GC/5973N MSD
and PT-2020D Pyrolyzer. Each catalyst
is heated up in an oven and the
temperature is raised up to 300 °C and
600 °C successively with a rate of
20 °C/min. The VOCs desorbed from
the catalyst surface were separated
continuously and their components
are analyzed qualitatively with GC/
MS method.
4 Results and Discussion
The performance of the catalyst
should be inspected periodically using
a specially designed device during
every plant outage period. In case of
the present ceramic honeycomb type
PAR, at least a quarter of the entire
catalysts are tested in every outage
period. The catalysts are tested in
single arrangement under the predetermined
flow and temperature
of air and hydrogen mixture by
measuring the temperature rise of
air-hydrogen mixture between inlet
and outlet of the test device. Figure 4
RESEARCH AND INNOVATION 459
Research and Innovation
Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
atw Vol. 63 (2018) | Issue 8/9 ı August/September
RESEARCH AND INNOVATION 460
| | Fig. 4.
Tempearature and Hydrogen Concetration at the Exit of Test Device of Periodic Inspection
(New Catalyst: 3 % hydrogen and air mixture at 60 °C and 1 bar).
shows temperature rise behavior of
new catylists, which shows a similar
trend with time. Therefore, the PAR
supplier suggested the accepatance
criteria of the periodic inspection as
the temperature rise at a given time
(The exact values of temperature rise
and time are not described in this
paper because that information is a
supplier’s proprietary). Figure 5
shows temperature rise bebavior of
catylists that were exposed to containment
air during one overhaul period.
The behavior of temperature rise is
affected by the existence of VOC.
Some catalysts showed delayed startup
of hydrogen recombination and
others showed further increase of
temperature by combustion of VOC
itself. Figure 5 also shows the hydrogen
volume faction of air-hyrogen
mixture at the outlet of the test device.
It showed that the hydrogen recombination
already started although
the temperature does not reach the
required value. Therefore, there is a
possibility of unneccesary failure of
plant inspection with the current
method by temperature rise. This
method requires relatively long test
time because of larger heat capacity of
ceramic structure. In addition, it is
difficult to correlate the hydrogen
recombination performance with the
amount of temperature rise and test
time. Threfore, we decided to change
the inspection method from the temperature
rise to the direct measurement
of hydrogen concentration with
new acceptance criterion.
Under the VOC-affected conditions,
the performance of PAR is hard
to identify through the current perioic
inspection method because the startup
delayed time and the hydrogen
removal rate are defined under the
| | Fig. 5.
Tempearature and Hydrogen Concetration at the Exit of Test Device of Periodic Inspection
(After the Exposue of One Overhaul Period to Containment Air, 3 % hydrogen and
air mixture at 60 °C and 1 bar).
natural convection conditions. Therefore,
a number of catalysts are withdrawn
out of containment during an
overhaul period of each plant and
their performance is tested in the PAR
performance test facility (PPTF) under
the natural convection conditions.
A total of 152 tests are performed
with 608 catalyst samples to investigate
the effect of volatile organic
compounds (VOC) on the startup
performance on the hydrogen
removal. The catalyst samples are
taken from seventeen (17) plants with
four (4) different reactor types. For
plants C, D, F, H and M, the tests are
performed twice in the first and
second outage period to compare test
resuts between the first and the
second outages in the same plant.
Figure 6 shows the measured start-up
delay times in conditions of hydrogen
of 3 vol. %, temperature of 60 °C and
pressure of 1.5 bar. These test conditions
are selected because a start-up
delay time is considered after the
hydrogen concentration and the
temperature reached at both 3 vol. %
and 60 °C in the analysis of hydrogen
control to determine the capacity
and locations of PARs as a hydrogen
mitigation system [2]. Fifteen (15)
minutes of the start-up delay time are
assumed in severe accident analyses
while 12 hours of the start-up delay
time is assumed in design basis accident
analysis [12]. For new catalysts a
certain time is required until the flow
is fully developed by naural convection.
This time has been measured as
about 404 sec with a standard deviation
of 66.9 sec. As shown in Fig. 6,
the start-up delay times are well
within 15 minutes except the plants G
and H. The start-up delay times for
plant G and H1 show an average time
of 1,006 sec and 893 sec with a
standard deviation of 160 sec and
215 sec, respectively. The total averaged
start-up delay time for all plants
is estimated as 660.6 sec with a standard
deviation of 237.8 sec. For plants
C, D, F, H and M, the second tests does
not show a noticeable difference
compared to its first tests.
In the design basis accident such as
a loss-of-coolant-accident (LOCA),
the hydrogen is generated gradually
and the hydrogen concentration could
be reached at 4 vol. % after several
days without a hydrogen mitigation
system after a LOCA takes places. In
the analysis of hydrogen concentration
in the LOCA, twelve (12) hours of
the start-up delay time were assumed
after the hydrogen concentration and
the catalysts temperature reach at
both 3 vol. % and 60 °C. Although the
start-up delays of 12 hours are considered,
there is a sufficient margin to
maintain the hydrogen concentration
below the regulatory limit of 4 vol. %.
However, in the severe accident conditions,
the hydrogen concentration in
the containment abruptly increases at
the timing of the reactor vessel failure
so that the margin for start-up delay
for hydrogen removal may not be
sufficient compared to the situation of
a design basis accident. The regulatory
position in Korea is that the startup
delay times should be verified and
compared to the assumptions used in
the analysis of hydrogen control in
DBA and severe accident conditions.
In the case of plant G, H and N, the
analysis of hydrogen control in severe
accident conditions has been re-evaluated
with a longer delay time of
30 minutes in consideration of the
results of the start-up delay time
measurement tests in 2014. For the
Research and Innovation
Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
atw Vol. 63 (2018) | Issue 8/9 ı August/September
| | Fig. 6.
Start-up Delay Times after One Overhaul Period Exposure to VOC.
Plant ID
Compounds
other plants, the re-evaluation has
been performed in 2017.
Figure 7 shows the hydrogen
depletion rates after an overhaul
period of exposure to VOCs in containment
air. A total of 62 tests are
performed with 248 catalyst samples
from seventeen (17) plants as
described in Table 2. The test results
show that the hydrogen depletion
rates are much higher than the
required depletion rate of 0.2 g/sec
that is specified in technical specification
of PAR purchase in Koran
nuclear power plants. A total averaged
value is estimated as 0.270 g/sec with
C1 G H1 H2 M1 P Y Estimated Sources
of VOCs
Benzene ! ! ! ! ! ! ! Paint, Insulation, Glue
Docosane ! ! ! ! Oil
Eicosane ! ! ! ! ! Oil
Heptadecane ! ! ! ! ! ! ! Oil
Heptane, 3-methylene- ! ! ! Oil
Hexadecane ! ! ! ! ! ! Oil
Octadecane ! ! ! ! ! ! Oil
1-Propene, 2-methyl- ! ! ! Paint
Dibutylformamide ! ! ! Insulation
Diethyl phtalate ! ! ! Paint, Insulation
Heneicosane ! ! ! ! Oil
Methylstyrene ! ! ! ! Paint, Insulation
Nonadecane ! ! ! Oil
Tridecane ! ! ! ! Oil
Nonaneitrile ! ! ! Oil, Resin
Tetradecane ! ! ! Oil
Toluene ! ! Paint, Sealing
| | Tab. 3.
Major Compounds Adsorbed on the Sample Catalyst Surface.
a standard deviation of 0.03 sec. The
measured hydrogen depletion rates of
catalysts exposed to VOCs have no
difference with those of new catalysts
that is estimated as 0.2687 g/sec with
a standard deviation of 0.0108 sec.
The recombination reaction takes
place on some active sites on the
degraded catalyst releasing the heat
of reaction. This causes the catalyst
surface temperature to increase
creating a driving force for convective
flow. Increase convective flow
accelerates the reaction rate leading
to further increase in the catalyst
temperature until all the adsorbed
| | Fig. 5.
Hydrogen Depletion Rates after One Overhaul Period Exposure to VOC.
VOCs desorb and all the active sites
are free, i.e., the catalyst is fully
regenerated. The same conclusion
about the hydrogen depletion rate
has been reported in reference [6].
The adsorbed airborne substances
on the catalyst surface are analyzed
qualitatively using GC/MS (gas
chromatograph/mass spectrometer)
method for selected samples from
seven (7) plants. Various VOCs are
detected and their major compounds
are summarized in Table 3. It is
estimated that these compounds are
originated from paints, oils, lubricant,
insulation, glues, etc., which are commonly
used in the plant maintenance.
Although benzene, heptadecane etc.
are commonly detected, the detected
volaticle organic compounds differ
from each plants. In the previous
results, the plant H1 showed a relatively
longer start-up delay time compared
to other plants [8]. There was a
steam generator replacement in plant
G and H when the PARs were installed
in 2013. Further tests are performed
in next overhaul for plant H. The test
results of H 2 represents test results in
the second overhaul (2016) in plant H.
The detected VOCs are different from
the results of the first overhaul (2014)
but the start-up delay time still
remained in relatively larger value
than other plants. The common VOCs
detected in plant G, H1 and H2 are
benzene, hetadecane, octadecane etc.
(the plant G and H are the same type
plants). However, these materials are
also detected in other plants having a
relatively shorter start-up delay time.
From the present results, it is considerd
that the detected materials are
plant-specific and strongly dependent
on the maintenance activities. The
VOC materials presented in Table 3
are at least not strongly related to the
RESEARCH AND INNOVATION 461
Research and Innovation
Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
atw Vol. 63 (2018) | Issue 8/9 ı August/September
RESEARCH AND INNOVATION 462
start-up performance of PARs. Therefore
we could not identify which
materials of VOC could affect the
start-up performance using the
present GC/MS method.
The regulatory position in Korea
on the PAR is that the start-up delay
time and the hydrogen depletion rates
should be verified periodically and
compared to those assumed in the
hydrogen control analysis for design
basis and severe accidents because the
long-term operational experience of
PAR in the nuclear power plants has
not been fully insvestigated. Therefore,
the present paper have been
mainly focused on the start-up delay
time and hydrogen depletion rate in
a given condition to validate the
assumptions used in the hydrogen
control analysis. It is considered that
there is a sufficient margin to control
hydrogen below the regulatory limit
of 4 vol. % of hydrogen concentration
in design basis accidents. However, in
the severe accident conditions, the
hydrogen in the containment abruptly
increases at the timing of the reactor
vessel failure. There may not be
sufficient margin for hydrogen control
in some severe accident scenarios if an
additional start-up delay time more
than 30 minutes is considered. However,
the capacity and locations of PAR
have been determed from very conservative
severe accident analyses [2]
and the temperature of containment
air is expected to be above or around
100 °C in severe accident conditions.
It could be postulated that the temperature
will be high enough to regenerate
the PAR catalyst that had resided
in the containment for a prolonged
time period so that the PAR will
promptly respond to hydrogen. Therefore,
it is important to identify in
which conditions the PAR will
promptly react with hydrogen in such
a long time exposed condition to
possible VOCs.
4 Conclusions
The hydrogen depletion rates and
the start-up delay time of a Pt/TiO 2
coated ceramic honeycomb PAR have
been measured using a total of 856
catalyst samples from seventeen (17)
operating nuclear power plants after
one overhaul period of normal operation
since its first installation in order
to investigate the effect of volatile
organic compounds (VOCs) on the
catalyst functionality. The measured
hydrogen depletion rate and start-up
delay time were compared to those
used in the hydrogen control analysis
because these are key parameters in
the determination of the capacity and
location of PARs. The tests showed
that the hydrogen depletion rates are
not affected by VOC accumulation on
the catalyst surface due to its volatile
nature at high temperature by exothermic
catalytic reaction. Through a
series of tests on the start-up delays
using VOC-affected catalysts, the VOC
delays the start-up for hydrogen
removal by poisoning or blocking of
the catalytic surface. Although the
measured delay times were well
within 30 minutes in the condition of
3 vol. % of hydrogen, 60 °C of temperature
and 1.5 bar of pressure, it is
expected that the delay time would
further increase in proportion to the
exposure time to containment air. The
type of airborne substances was
identified through qualitative GC/MS
(gas chromatograph/mass spectrometer)
method from selected samples
from seven (7) plants. The volatile
organic substances adsorbed on the
catalyst surface were estimated
mainly from paints, lubricants, glues,
insulations and oils etc. It is expected
that the reduction of VOC in the
containment air may be a challenging
work. Therefore, it is important to
identify in which conditions the PAR
will promptly react with hydrogen in
such a long exposed condition of
possible VOCs. To this end, further
extensive tests on the catalyst performances
in various hydrogen concentrations
and temperatures will be
performed with catalysts that had
resided in various reactor containments
and for various exposure times
to containment air.
References
1. Status Report on Hydrogen Management
and Related Computer Codes,
NEA/CSNI/R(2014)8 (2014).
2. Kim, C. H. et al., Analysis Method for the
Design of a Hydrogen Mitigation
System with Passive Autocatalytic
Recombiners in OPR-1000, The 19 th
Pacific Basin Nuclear Conference (PBNC
2014), Vancouver, Canada, August 24–
28, 2014, Paper No. PBNC2014-072
(2014).
3. Effects of Inhibitors and Poisons on the
Performance of Passive Autocatalytic
Recombiners (PARs) for Combustible
Gas Control in ALWRs, EPRI ALWR
Program Report, Palo Alto CA (1997).
4. Studer, E. et al., Assessment of Hydrogen
Risk in PWR, 1 st IPSN/GRS EURSAFE
Meeting, Paris (1999).
5. OCED/NEA THAI Project: Hydrogen and
Fission Product Issues Relavent for
Containment Safety Assessment under
Severe Accident Conditions, NEA/
CSNI/R(2010)3 (2010).
6. Kelm, S. et al., Ensuring the Long-Term
Functionality of Passive Auto-Catalytic
Recombiners under Operational
Containment Atmosphere Conditions –
An Interdisciplinary Investigation,
Nuclear Engineering and Design,
Vol.239, pp. 274-280 (2009).
7. Reinecke, E-A. et al., Open Issues in the
Applicability of Recombiner
Experiments and Modeling to Reactor
Simulations, Progress in Nuclear Energy,
Vol.52, pp. 136-147 (2010).
8. Kim, C. H. et al., Operational Experience
of Ceramic Honeycomb Passvie Autocatalytic
Recombiner as a Hydrogen
Mitigation System, The 16 th International
Topcical Meeting on Nucear
Reactor Thermal Hydraulics
(NURETH-16), Chicago, IL, USA,
August 30 – September 4 (2015)
9. Kang, Y. S. et al., Hydrogen Recombination
over Pt/TiO2 Coated Ceramic
Honeycomb Recombiner, Appl. Chem.
Eng., Vol.22, No.6, pp. 648-652 (2011).
10. Ceracomb Co. Ltd., http://
www.ceracomb.co.kr/en/ (homepage).
11. Frontier Laboratories, Co. Ltd.,
http://frontier-lab.com (homepage).
12. Final Safety Analysis Report of Ulchin
Nuclear Power Plants, Units 3 & 4,
Section 6.2, Korea Hydro and Nuclear
Power Co., Ltd. (revised in 2013).
Authors
Chang Hyun Kim
Je Joong Sung
Sang Jun Ha
Central Research Institute
Korea Hydro and Nuclear Power
Co., Ltd.
25-1 Jang-dong, Yuseong-gu,
Deajeon, 305-343, Rep. of Korea
Phil Won Seo
Department of Research &
Development,
Ceracomb Co., Ltd.
312-25 Deuksan-dong, Asan-si,
Chungcheongnam-do, 336-120,
Rep. of Korea
Research and Innovation
Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo
atw Vol. 63 (2018) | Issue 8/9 ı August/September
49 th Annual Meeting on Nuclear Technology (AMNT 2018)
Young Scientists' Workshop
Jörg Starflinger
During the Young Scientists' Workshop of the 49 th Annual Meeting on Nuclear Technology (AMNT 2018), 29 to 30
May 2018, Berlin, 13 young scientists presented results of their scientific research as part of their Master or Doctorate
theses covering a broad spectrum of technical areas.
This demonstrated again the strong
engagement of the younger generation
for nuclear technology and the
significant support of German institutions
involved.
Dr. Katharina Stummeyer (Gesellschaft
für Anlagen- und Reaktorsicherheit
gGmbH), Dr.-Ing. Wolfgang
Steinwarz (Founder and former jury
chairman of the Workshop „Preserving
Competence in Nuclear Technology”),
Prof. Dr.-Ing. Marco K. Koch (Ruhr-
Universität Bochum), and Prof. Dr.-Ing.
Jörg Starflinger (Universität Stuttgart)
as members of the jury assessed the
written compacts and the oral
presentations to award the prices
supported by GNS Gesellschaft für
Nuklear-Service mbH, Essen and
Forschungsinstitut für Kerntechnik und
Energiewandlung e.V., Stuttgart.
Vera Koppers (Gesellschaft für
Anlagen- und Reaktorsicherheit (GRS)
gGmbH, mentoring: Prof. Koch) as first
speaker reported on the present status
on Heuristic Methods in Modelling
Research Reactors for Deterministic
Safety Analysis. The goal is a
deeper understanding of modelling of
research reactors using the code
ATHLET. Good agreement of ATHLET
results with experiments from literature
has been achieved.
The presentation by Sebastian
Unger (Helmholtz-Zentrum Dresden-
Rossendorf, mentoring: Prof. Hampel)
described Experimental Investigation
on the Heat Transfer of Innovative
Finned Tubes for Passive
Cooling of Nuclear Spent Fuel Pool.
A single-phase cooling system for
spent fuel pools has been introduced.
The bottle neck in heat transfer lays
on the air-side heat exchange, which
is enhanced by innovative fins. The
potential of enhancement of heat
transfer has clearly been demonstrated
on small-scale.
Martin Arlit (Technische Universität
Dresden, mentoring: Prof. Hampel)
informed about Heat Transport from
Dried Surfaces of a Spent Fuel
Mock-up under Accident Conditions
with a Thermal Anemometry
Grid Sensor. A grid sensor has
been developed enabling the spatially
resolved measurement of fluid temperatures
and velocities within a rod
bundle. Small-scale experiments
showed that heat dissipation by convection
of the overall heating power is
below 10 %, but is of importance for
the cooling of the dried rod bundle
section above the water level.
Maria Freirìa López (Universität
Stuttgart, mentoring: Prof. Starflinger)
reported on Criticality Evaluation of
Debris Beds after a Severe Accident.
By means of Monte-Carlo-Code simulations,
a criticality map for debris
beds, forming during beyond-design
accidents, is currently being developed.
The first analyses indicates that
debris beds in fact might get critical,
but they also showed parameter
combinations (debris size, boration,
porosity, etc.), where criticality can be
intrinsically excluded.
Larissa Klaß (Forschungszentrum
Jülich GmbH, mentoring: Prof. Modolo)
described Modified Diglycolamides
for a Selective Separation of
Am(III): Complexation, Structural
Investigations and Process Applicability.
In her work the complexation
behaviour of new hydrophilic complexants
towards trivalent actinides
and lanthanides was investigated in
order to achieve a deeper understanding
of their coordination chemistry.
For the first time, the formation of
mixed complexes of hydrophilic and
lipophilic complexant in the organic
phase has been measured. Based on
this result, an innovative solvent
extraction procedure was developed,
which could simplify the existing
procedures.
Corbinian Nigbur (Universität
Stuttgart, mentoring: Prof. Starflinger)
introduced the Application of the
Integral Diffusion Approach to
the Modelling of the Oxidation of
Mixtures of Fuel and Zirconium.
The objective is to simulate the oxidation
process during accidents with
one integral approach replacing the
different numerical approaches within
thermal-hydraulic system codes.
Comparison of numerical simulations
with data from crucible experiments
showed good agreement.
Numerical implementation of
methods considering dynamic
soil-structure interaction was the
subject of the presentation given
Arthur Feldbusch, (Technische Universität
Kaiserslautern mentoring: Prof.
Sadegh-Azar). A dynamic model has
been developed for soil-structureinteraction.
Using the “Thin Layer
Method”, a tool is derived to evaluate
the soil-structure behaviour due to
mechanical loads. The model is
limited to linear calculations, but shall
be extended to non-linear capabilities.
Pascal Distler (Technische Universität
Kaiserslautern, mentoring: Prof.
Sadegh-Azar) reported about Airplane
Crash Analysis: Semi-hard and
hard Missile Impact on Reinforced
Concrete Structures, in which the
damage mechanisms were presented
and explained to determine the load
bearing capacity of the hard and the
soft impact of projectiles. A numerical
model has been set up and compared
with impact tests, which show reasonable
agreement. In the next step, the
actual model will be extended to
describe the interaction between the
reinforced concrete structure (target)
and the impacting projectile.
Danhong Shen (Karlsruhe Institute
of Technology (KIT), mentoring: Prof.
Cheng) gave an overview on An improved
turbulent mixing model in
sub-channel analysis code. Using
CFD simulations of two adjacent
sub-channels of a fuel assembly, an
improved numerical turbulent mixing
model has been derived to be used in
sub-channel codes. Three empirical
correlations are proposed to describe
the relationship between each turbulent
mixing coefficient and the
Reynolds number as well as the
geometry parameter. This investigation
will improve computation
capabilities of sub-channel codes.
The presentation of Dali Yu (Karlsruhe
Institute of Technology (KIT),
mentoring: Prof. Cheng) described
Modeling of post-Dryout Heat
Transfer. The aim of the work is to
predict the wall surface temperature
under Dryout conditions. The whole
post-dryout flow region is divided into
463
AMNT 2018 | YOUNG SCIENTISTS WORKSHOP
AMNT 2018 | Young Scientists' Workshop
Young Scientists' Workshop ı Jörg Starflinger
atw Vol. 63 (2018) | Issue 8/9 ı August/September
464
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Award winners, sponsors and jury of the 49 th Annual Meeting on Nuclear Technology (AMNT 2018)
Young Scientists Workshop: (from left): Dr. Wolfgang Steinwarz, Prof. Dr. Marco K. Koch (Ruhr-Universität
Bochum), Dr. Katharina Stummeyer (Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH),
Vera Koppers, Winner of the 2018 Young Scientists' Award, Dr. Jens Schröder (GNS Gesellschaft für
Nuklear- Service mbH), Katharina Amend (2 nd ranked young scientist), Prof. Dr. Jörg Starflinger
( Universität Stuttgart), María Freiría López 3 rd ranked young scientist.
several different sections, each of
them modelled separately with
different correlations. A comparison
with experimental data showed fairly
reasonable results which are subject
for improvement as a next step.
Tobias Jankowski (Ruhr-Universität
Bochum, mentoring: Prof. Koch) reported
about Development and
Validation of a Correlation for
Droplet Re-Entrainment Estimation
from Liquid Pools. The correlation is
based on a dimensional analysis and
therefore considers thermohydraulic
boundary conditions by dimensionless
quantities, which are quantified
by empirical constants. These constants
are obtained by four nearly
steady test phases, taken from two
experimental facilities of different
scale. The correlation results are in a
good agreement with experimental
data.
The presentation entitled Comparison
of Different Wash-off
Models for Fission Products on
Containment Walls was given by
Katharina Amend (Universität der
Bundeswehr München, mentoring:
Prof. Klein). A parameter variation
was conducted with in the setting of a
simplified geometry and with the
geometry of the laboratory tests. One
key influencing parameter for the
resulting washed off mass is the
percentage of area covered by water
in each case, which differs with
inclination and mass flow rate. First
simulations with the laboratory
geometry show satisfactory agreement,
when compared to the experiments.
Moritz Schenk (Karlsruhe Institute
of Technology (KIT), mentoring: Prof.
Cheng) gave a presentation about CFD
Analysis of centrifugal Liquid Metal
Pumps. Using the open-source software
OpenFOAM the influence of the
physical properties of liquid metals on
the performance of a pump impeller
and on the flow field is investigated.
In general, the simulations show a
relatively strong negative influence on
head and efficiency for much higher
viscosities and nearly no effect for
lower viscosities compared to water.
This qualitative behaviour is in good
agreement with the literature. The
optimization of the liquid metal pump
is ongoing, focussing on the corrosion
potential of the liquid metal.
Summarizing, the scientific quality
of papers presented by the young
scientists in this year reached again
a very high level. Therefore, all participants
of the workshop should get
honourable recognition.
The jury awarded Vera Koppers
(Gesellschaft für Anlagen- und Reaktorsicherheit
(GRS) gGmbH) the 1 st price
of the 2018 competition. Her compact
as well as those of both the 2 nd ranked
author, Katharina Amend (Universität
der Bundeswehr München) and the 3 rd
ranked author Maria Freiria (Universität
Stuttgart) are published in this
issue of atw – nucmag.
Author
Prof. Dr.-Ing. Jörg Starflinger
Institute of Nuclear Technology
and Energy Systems (IKE)
Pfaffenwaldring 31
70569 Stuttgart, Germany
Young
Scientists'
Workshop
WINNER
Vera Koppers was
awarded with the
1 st price of the 49 th
Annual Meeting on
Nuclear Technology
(AMNT 2018) Young
Scientists' Workshop.
Heuristic Methods in Modelling
Research Reactors for Deterministic
Safety Analysis
Vera Koppers and Marco K. Koch
1 Introduction The national and international fundamental nuclear safety objective is to protect the public
from ionising radiation [IAEA2016]. Although research reactors may have a smaller risk potential to the public than
nuclear power plants, operators and researchers are at a higher risk [IAEA2016]. Deterministic safety analyses using
thermal-hydraulic system codes are a prevalent and important instrument to evaluate the safety of nuclear power plants
and research reactors. A wide range of safety analysis codes that are used for simulations of nuclear power plants are
applicable to simulations of research reactors. The application range of the thermal-hydraulic system code ATHLET
(Analysis of thermal-hydraulics of leaks and transients) – developed by GRS (Gesellschaft für Anlagen- und Reaktorsicherheit
gGmbH) – was extended to simulated subcooled nucleate boiling processes at low pressure in 1994 [GRS2009].
AMNT 2018 | Young Scientists' Workshop
Heuristic Methods in Modelling Research Reactors for Deterministic Safety Analysis ı Vera Koppers and Marco K. Koch
atw Vol. 63 (2018) | Issue 8/9 ı August/September
After that, research reactor simulations using ATHLET were successfully performed at national and international
research institutes. ATHLET uses the finite volume method and solves the partial differential equations matrix at
discrete meshed volumes. In order to simulate a plant system, the user has to build up a network of thermal hydraulic
volumes. This approach allows a wide range of code application due to free thermal-hydraulic nodalisation, but it takes
large amount of human resources and requires detailed plant descriptions. In ATHLET, the main modules are thermal
fluid dynamics (TFD), heat transfer and heat conduction (HECU), neutron kinetics (NEUKIN) as well as plant control
(GCSM). The user has to choose adequate input options out of a wide range of possibilities for each module. Analysing
foreign research reactors, technical support organisations and research institutes might be confronted with limited
available information of plant data. In case of emerging safety related questions, the complex input data structure of
safety analysis codes impede a fast response.
The present paper describes the development
of a new method for rapid
input deck development in the light of
limited available data. Due to high
diversity of research reactor designs,
a rule-based software system is
engineered to support the modelling
process for deterministic safety analysis
utilising the system code ATHLET.
The use of heuristic rules allows
an adequate input deck generation
despite limited data. The fundamental
elements of the input deck are generated
automatically by few input data
necessary. In the case of unavailable
data and urgently safety related questions,
the user is supported by this
software. In the following, the applied
heuristic rules realising the new
strategy of modelling are described.
After that, first functionality of the
new modelling system is demonstrated.
2 Heuristic methods
in modelling research
reactors
In this paper, heuristic methods are
defined as an approach to achieve an
appropriate modelling quality of
research reactors despite incomplete
data. For this purpose a new software
is developed that is structured in the
following main modules:
• process of user input
• build the research reactor model
• transform to ATHLET input format
• export as input deck
The required key data, which the user
has to provide to run the software, are
constricted to publicly available data.
Detail technical documentations, such
as safety analysis report, operating
manual, system descriptions and
schematics as well as technical
drawings are assumed to be not
accessible. The next text section
describes the main steps of the
strategy that are implemented in the
modelling software.
medicine, research reactors have a
wide range of designs and operation
modes. Realising a heuristically process
for research reactor modelling,
the number of reactor types considered
in this study was restricted.
To date, 241 research reactors
are operated around the world
[RRDB2018]. The TRIGA (Training,
Research, Isotopes, General Atomic)
and MTR (Material Testing Reactors)
reactors represent the most widely
installed research reactor types. About
25 % of the research reactors are
of MTR type and 21 % are of TRIGA
design [RRDB2018]. Consequently,
these types are selected as a model
design basis. The considered reactor
designs are abstracted to open core
and tank-in pool reactors as pictured
in Figure 2-1. The TRIGA design is
currently limited to reactors with
natural convection cooling.
To structure the research reactor
types, a modularisation approach is
used. The first level of modularisation
is also shown in Figure 2-1. On the
second level, the reactor components
are decomposed into their further
elements. Focusing on the central
component, the “reactor core”, typical
MTR research reactors have a cluster
of multiple assemblies installed at the
lower part of the reactor pool. The
assembly consists of several parallel
arranged fuel plates and the assembly
feet. The basic TRIGA core design
(Mark I and II) consists of a cylindrical
geometry and uses fuel moderator
rods. The TRIGA core is also located at
the lower part of the reactor pool.
The fuel elements of both types (MTR
or TRIGA) are made of a fuel meat
section containing the fissile material
and outside cladding material. Within
this work, the fuel meat and cladding
material are the smallest units of
which a fuel element is made of. In
Figure 2-2 the modularization of an
MTR core is shown.
| | Fig. 2-1.
Generic design sense of TRIGA and MTR research reactors and modularisation of main components.
465
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
2.1 Abstraction and
modularisation of research
reactor designs
Due to their different applications in
the field of science, technology and
| | Fig. 2-2.
Modularisation of the reactor core (MTR example).
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AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Fig. 2-4.
Nodalisation of MTR Fuel Assembly.
| | Fig. 2-3.
Nodalisation of MTR Fuel Assembly.
The main system boundary to be
modelled in the input deck is defined
at the pool with the inlet and the
outlet pipe. The reactor pipework is
composed of different pipes that are
built up by pipe segments (horizontal,
vertical, etc.). The pipes may also
contain valves and pumps. The modularisation
process is used as the basis
for object-oriented software design.
2.2 Applied nodalisation
rules for selected MTR and
TRIGA types
To realise the transformation and
exportation of reactor data into
ATHLET- format, nodalisation schemes
have to be developed and their rules
have to be implemented in the software.
For different research reactor
types, different nodalisation rules
have to be applied. Within the system
code ATHLET, the thermal hydraulic
nodalisation is represented by
thermo-fluiddynamic objects (TFOs).
TFOs are classified into pipes, branches
and special objects. Pipe objects
simulate one-dimensional fluid flow,
branch objects represent major
branching, and special objects are
used for simulation of components
with special requirements, e.g. cross
connections.
Focusing on the core geometry of a
MTR research reactor, each assembly
has several separated cooling channels
between the fuel plates. To cover
different postulated initial events, e.g.
blockage of one cooling channel in a
fuel element, the reactor core is
considered in detail and for each
cooling channel one representative
pipe is used. To reduce calculation
time, it is possible to group assemblies,
if they have identical characteristics.
Otherwise, there are modelled
separately. In Figure 2-3, the applied
nodalisation scheme for MTR fuel
assemblies is presented. Every fuel
assembly is linked to a common
branch before entering and leaving
the reactor core. The fuel plates are
modelled as Heat Conduction Objects
(HCOs). Internal fuel plates are
coupled on both sides to corresponding
TFOs. External fuel plates are
coupled one-sided to a TFO representing
a core channel and the other
side is coupled to a common bypass
channel.
Focusing on the TRIGA research
reactor, the core is composed of
several fuel rods in one tank. In contrast
to the MTR core, the fuel rods
have no separated cooling channels.
Therefore, the determination of
nodalisation depends on the core
layout. Based on typical TRIGA core
grid structures (Mark I and II), heuristics
are derived and realised in a
simple algorithm to determine the
linkage of TFOs. This approach
reduces the required input data to the
number of grid positions n in the first
circle around the centre point and the
number of grid positions along the
radius r (starting at the centre point)
– see Figure 2-4. Further, the length
of r is required. In radial direction, the
cooling area is divided into rings starting
at the centre point. In tangential
direction, the cooling area is divided
into segments.
The number of segments depends
on the number of grid position in the
first circle. The algorithm also computes
the belonging cross connections
and geometrical data. In the pictured
nodalisation in Figure 2-4, there are
13 pipes connected by cross connection
objects (6 grid positions along
r-direction and 6 grid positions in the
first circle). As already applied for
MTR core design, the pipes are linked
to a common branch before entering
and leaving the reactor core. The fuel
rods are modelled as cylinders and
defined adiabatic at the inner side.
The outer side is coupled to the
corresponding TFO.
As default setting, the axial power
profile for both core designs (MTR
and TRIGA) follows a sinus curve.
While the geometry of guide boxes
and control plates/rods are not considered,
the external reactivity is
modelled by a signal in the general
control simulation module of ATHLET.
In the following Figure 2-5, the
generated core layouts by the software
for input deck generation is
presented. Only fuel assemblies with
fuel plates (MTR) and fuel rods
( TRIGA) are shown. Other components
or empty positions are not
AMNT 2018 | Young Scientists' Workshop
Heuristic Methods in Modelling Research Reactors for Deterministic Safety Analysis ı Vera Koppers and Marco K. Koch
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pictured. Typically, four control rods
are required for reactivity control in
TRIGA reactors with thermal power
levels of less than 1 MW [IAEA2016B].
Further, graphite elements are at the
outer positions. For MTR research
reactors, there are often empty places
at the centre of the core grid for
radiation samples. If the input number
of fuel assemblies or elements
does not match the number of grid
positions, the implemented algorithm
considers these typical core characteristics.
The assemblies or elements are
positioned in respect of this information.
Furthermore, the free flow path
is calculated as a function of total core
area and number of fuel assemblies,
elements and other components
inside the research reactor core.
3 Generated input decks
of exemplary MTR and
TRIGA reactors
In this part, first functionality of
the new modelling system is demonstrated
by generating an exemplary
MTR and TRIGA research reactor
model. For this purpose, two reference
research reactors were chosen.
Providing technical details in
[ABD2008A] and comparative data in
[ABD2008B], the ETRR-2 was identified
as a MTR reference facility. The
ETRR-2 is a multipurpose research
reactor located in Inshas, the Arab
Republic of Egypt. It corresponds to
the rightmost research reactor design
in Figure 2-1. The ETRR-2 reactor
consists of 29 fuel assemblies of MTR
type with 19 fuel plates each and has
22 MW nominal power. Further
description is presented in [ABD2008].
The main nodalisation of the generated
ETRR-2 model in ATHLET is
pictured in Figure 3-1. On the left
side, the coolant loop is presented
in bright blue. The reactor pool is
modelled with two pipes interconnected
by cross-connections. The
inner pool pipe is connected to the reactor
chimney, which is marked in
brown, by a single junction pipe. The
reactor core is modelled with two
representative assemblies. Each is
composed of 18 core cooling channels.
One assembly is representing 28
grouped average assemblies. The
other assembly considers a hot channel
factor on the 19 fuel plates plus
one extra penalised fuel plate. The
nodalisation of both assemblies is
identically and shown in Figure 3-2.
To check the capability of the
nodalisation to reproduce the thermal
hydraulic plant conditions, steady
state calculations were performed.
| | Fig. 2-5.
MTR core layout (left) and TRIGA core layout (right), generated by software for input deck generation.
| | Fig. 3-1.
Overview of whole Nodalisation of the ETRR-2 (left) and one fuel assembly (right) with 18 core channels generated by the software
for input deck generation.
Power
[MW]
Loop mass
flow
[kg/s]
The initial conditions of the experiment
and the calculated parameters
are compared in Table 3-1. The
experiment was performed at 9.5 MW
thermal power. There is good agreement
between the calculated and
experimental stationary data.
As an exemplary TRIGA research
reactor, the IPR-R1 was identified.
The IPR-R1 is a TRIGA Mark I model,
installed in Belo Horizonte in Brazil
and operated since 1960. Several
analytic and experimental studies
were performed and published. As
reference data, experimental results
in [REI2009] were used. The IPR-R1
corresponds to the leftmost research
reactor design in Figure 2-1. It is
operating at 250 kW and consists of
63 fuel elements of TRIGA type.
Further description is presented in
[REI2009]. The main nodalisation of
the generated IPR-R1 model in
ATHLET is shown in Figure 3-2. On
the left side, the coolant loop is
Core mass
flow
[kg/s]
Core outlet
temperature
[°C]
Core pressure
drop
[bar]
| | Tab. 3-1.
Overview of whole Nodalisation of the ETRR-2 (left) and one fuel assembly (right) with 18 core channels generated by the software
for input deck generation.
presented in bright blue. The reactor
pool is modelled with two pipes interconnected
by cross-connections. The
inner pool pipe is connected to the
core entrance and core outlet. 13 core
channels, interconnected by crossconnections,
with 63 fuel elements
represent the reactor core (see Figure
3-2 right). The core nodalisation
based on the nodalisation presented
in Figure 2-4.
The experiment was performed at
50 kW thermal power. In Table 3-2,
the calculated steady state results are
compared to measured core inlet and
outlet temperatures. At different
positions, measuring devices were
installed (see [REI2009]). There are
small deviations but overall the results
are consistent.
Further, the ATHLET simulation is
compared to published RELAP steady
state calculation in [REI2009], which
reaches steady state conditions after
about 2000 s simulation time.
Reference
pressure
[bar]
Calculation 9.5 309.24 302.86 35.01 0.42 2.2
Reference
[ABD2015]
9.4 309.24 302.87 34.9 0.31* 2.0
*in [ABD2015] core pressure drop of 3.1 bar is mentioned, but in /IAEA2005/ 0.6 bar pressure drop at 100 % core power is referred
467
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
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Heuristic Methods in Modelling Research Reactors for Deterministic Safety Analysis ı Vera Koppers and Marco K. Koch
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References
468
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Fig. 3-1.
Overview of whole Nodalisation of the IPR-R1 (left) and 13 core channels (right) generated by the software for input deck
generation.
Power
[kW]
| | Tab. 3-2.
Thermal hydraulic data IPR-R1.
Core inlet
temperature
(Position 3)
[°C]
Core outlet
temperature
(Position 3)
[°C]
There is good agreement between
the published RELAP calculations in
[REI2009] and the calculated ATHLET
data.
4 Summary
A new method based on a heuristic
approach for modelling selected
research reactor types in thermal
hydraulic analysis codes is presented.
This new approach allows a fast and
reliable generation of the input deck’s
fundamental elements despite limited
technical documentation. Focusing on
one MTR and one TRIGA design, the
main steps of developing process and
the characteristics of the new method
are highlighted. This includes the
Core inlet
temperature
(Position 8)
[°C]
Core outlet
temperature
(Position 3)
[°C]
Calculation 51 20.87 27.97 20.87 23.94
Reference
[REI2009]
50 20.95 26.95 22.95 24.95
abstraction and modularisation of
research reactor plant designs as well
as the conception of type-specific
nodalisation. At the end of this paper,
an exemplary MTR and TRIGA
research reactor is presented, generated
by the developed software.
Focusing on the stationary conditions,
there is a good agreement between
the calculated and experimental data.
This proves the basic functionality of
the developed modelling system by
generating a realistic plant model for
TRIGA and MTR type. In future work,
the nodalisation for both research reactor
designs will be reviewed and
tested against a range of safety transients
and accidents.
ABD2008A
ABD2008B
ABD2015
I.D. Abdelrazek, E.A. Villarino:
ETRR-2 Nuclear Reactor: Facility
Specification; Coordinated
Research Project on Innovative
Methods in Research Reactor
Analysis, organised by IAEA,
October 2008.
I.D. Abdelrazek, E.A. Villarino:
ETRR-2 Nuclear Reactor:
Experimental Results
Coordinated Research Project
on Innovative Methods in
Research Reactor Analysis, organised
by IAEA, October 2008.
I.D. Abdelrazek, et al.: Thermal
hydraulic analysis of ETRR-2
using RELAP5 code, Kerntechnik
80, 2015.
ATH2016 G. Lerchl et.al.: ATHLET 3.1A
User’s Manual, GRS-P-1/Vol.1,
Ref.7, March 2016.
IAEA2005
IAEA2016
IAEA2016B
REI2009
RRDB2018
Authors
IAEA: Research reactor
utilization, safety, decommissioning,
fuel and waste management,
ISBN 92-0-113904-7,
IAEA 2005.
IAEA: Safety of Research
Reactors, IAEA Safety Standards
Series No. SSR-3, Vienna
Austria, 2016, ISSN 1020-525X.
IAEA: History, development and
future of TRIGA research
reactors, Technical Report
Series No. 482, ISBN 978-92-0-
102016-1, IAEA 2016.
P. A. L. Resi, et al.: Assessment of
a RELAP5 model for the IPR-R1
TRIGA research reactor, International
Nuclear Atlantic
Conference – INAC 2009,
ISBN: 978-85-99141-03-8.
IAEA: Research Reactor
Database, Website URL:
https://nucleus.iaea.org/RRDB/
RR/ReactorSearch.aspx?rf=1
(01.02.2018).
Vera Koppers
Prof. Dr.-Ing. Marco K. Koch
Responsible Professor
Ruhr-Universität Bochum (RUB)
Universitätsstraße 150
44801 Bochum, Germany
| | Fig. 3-2.
Core inlet (left) and core outlet (right) temperature.
AMNT 2018 | Young Scientists' Workshop
Heuristic Methods in Modelling Research Reactors for Deterministic Safety Analysis ı Vera Koppers and Marco K. Koch
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Development and Validation of a CFD
Wash-Off Model for Fission Products
on Containment Walls
Katharina Amend and Markus Klein
The research project aims to develop a CFD model to describe the run down behavior of liquids (wall films, transition
of film flow into a discrete number of rivulets, droplets) and the resulting wash-down of fission products on surfaces in
the reactor containment. Numerical experiments allow for a deeper physical understanding, which is the basis for an
improved semi-empirical modeling.
This paper presents a three-dimensional
numerical simulation for water
running down inclined surfaces
coupled with an aerosol wash-off
model and the resulting particle transport
using the software package
OpenFOAM. The wash-off model is
based on the procedure used in AULA
(German: Abwaschmodell für unlösliche
Aerosole, wash-off of insoluble
aerosol particles) in the lumped
parameter code COCOSYS [1]. A
parameter variation was conducted
and the simulation results are compared
to the laboratory experiments
performed by Becker Technologies
[2].
1 Introduction
The desired goal is the prevention of
environmental contamination with
radioactive particles after a core
meltdown in a light water reactor. The
containment in a nuclear reactor
building prevents high pressure radioactive
steam from escaping in the
event of an emergency. During such a
critical accident in a light water
reactor, most of the fission products
enter the containment building in the
form of soluble and insoluble aerosols.
These particles might deposit
on walls and installation surfaces.
Condensing steam that is also released
into the containment can wash down
even insoluble particles into the
containment sump.
In previous studies [3, 4] the understanding
of the run down behavior
of water, the formation of film flow,
rivulets or droplets, was the main
subject of interest. This study investigates
the wash-off of insoluble
particles based on the run down behavior
of water on inclined plates and
the developing flow patterns using
CFD simulations.
2 Laboratory experiments
The laboratory tests are part of the
THAI AW3 test program [5]. They
investigate the aerosol wash-down
behavior of non-soluble silver from
inclined walls by steam condensate.
Trapezoidal plates (plain stainless
steel or decontamination paint coating)
with different inclinations
are loaded with dry silver aerosol. At
the uppermost part water enters
the plate via a tubular distributor
with a given flow rate. The water
flows down the plate, washes off
part of the particles and is finally
collected in cups, which get exchanged
after a specified time period.
The samples are put into a cabinet
dryer and the remaining aerosol mass
is weighed to quantify the wash-off.
Pictures taken during the experiments
show the flow patterns and run
down behavior of the water on the
plates, see Figure 1.
Two kinds of silver aerosol particles
are used: a fine silver powder
and coarse silver powder. The fine
silver powder is specified with a particle
diameter of 0.7-1.2 μm for 99.9 %
of the particles and as averaged
par ticle diameter of the undisturbed
powder d p = 1 μm can be assumed.
It has a bulk density of ρ bulk =
1.1 g/(cm 3 ) and a specific surface of
A sp = 2.5 m 2 /g. For the coarse silver
powder the specification of particle
diameter is 1.5 – 2.5 μm (99.9 %).
Here the averaged particle diameter is
d p =2 μm, the bulk density is also
ρ bulk = 1.1 g/(cm 3 ) and the specific
surface A sp = 1.21 m 2 /g.
3 Simulation of the water
field
In previous studies the simulation
of the flow field with three different
inclinations, namely 2°, 10° and 20°,
and with empirical contact angle field
and filtered randomized initial contact
angle distribution (FRICAF) were
presented [4, 6]. The computational
domain is a trapezoidal geometry
(length = 1.215 m, small base =
0.09 m, large base = 0.475 m, Figure
2), as used in the laboratory
experiments [5].
| | Fig. 1.
Pictures of lab tests [5] with 2° inclination and
a mass flow rate of 11 g/s after 2 min and
15 min.
| | Fig. 2.
Schematic of computational domain of
inclined trapezoidal plate, dimensions in mm.
The simulations are carried out
with the software package Open-
FOAM using the standard two-phase
solver InterFoam. The Navier-Stokes
equations for isothermal and incompressible
multiphase flow are solved
and the phase interface is captured
by the Volume-of-Fluid method. The
time step is adjusted such that the
maximum Courant number is below
0.4 to ensure a sufficient level of accuracy.
The time is discretized via Euler
implicit. The inlet is extending over
the entire upper boundary with a
Young
Scientists'
Workshop
Awarded
Katharina Amend
was awarded with the
2 nd price of the 49 th
Annual Meeting on
Nuclear Technology
(AMNT 2018) Young
Scientists' Workshop.
469
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
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Development and Validation of a CFD Wash-Off Model for Fission Products on Containment Walls ı Katharina Amend and Markus Klein
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AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Fig. 3.
Comparison of simulations with empirical contact angle model and the laboratory experiments by Becker
Technologies [5] (in false color representation) with mass flow rate ṁ = 11 g/s (ṁ =12 g/s for inclination
of 10°), three different inclinations (left 2°, middle 10°, right 20°) and without aerosol loading.
given water velocity parallel to the
surface such that a specified mass flow
rate is achieved. The flat plate is
bounded by vertical sidewalls and
has an inclination angle α. Material
properties of water and air are used.
For snapshots of the resulting flow
fields see Figure 3.
The simulations are conducted
with the empirical contact angle
model and the filtered initial
randomized contact angle field [6].
The contact angle is specified in the
boundary conditions of the water field
and is taken into account to calculate
the curvature of the water-air interface.
The contact angle has a huge
impact on the formation of rivulets
and their stability as shown in previous
studies [7]. The empirical contact
angle model accounts for the
wetted history and therefore enforces
a spatially and temporally stable
rivulet flow.
4 Simplified geometry
This study also considers a simplified
geometry with dimensions of 6 cm
x 5 cm, 60° inclination and different
water loadings. As a first step the
simplified geometry, for which additional
benchmark data from CFD
simulations and experiments are
available, is used for the parameter
variation to save computational effort
and time. Later the findings are
transferred to the larger laboratory
geometry. Also the experimental
data can be used to investigate the
empirical contact angle model [6] in
another scenario than the laboratory
geometry where it was developed. For
the simplified geometry Singh et al.
[8] provide results of CFD simulations,
as do Hoffmann [9] and Iso et.
al [10]. Experiments are conducted by
Ausner [11]. All of the latter use the
identical geometry, but different inlet
conditions (overflow weir and feed
tube) and various simulation tools
(Singh and Iso Fluent, Hoffmann
CFX). In the present study simulations
with constant contact angle and with
empirical contact angle model are
performed. The results for different
Weber numbers are evaluated and
compared to the results of the studies
mentioned above for validation. Five
different Weber numbers (We = 0.02,
We = 0.24, We = 0.47, We = 0.76 and
We = 1.10) are investigated, which
correspond to an increasing water
mass flow rate:
We =
with liquid density ρ l , inclination
angle α, volumetric mass flow rate Q,
surface tension σ, plate width W
and viscosity μ. As the water load
increases, the flow pattern changes
from a thin rivulet to a more pronounced
rivulet to a fully wetting
water film (see Figure 4).
The influence of the side walls
is also clearly visible and was also
observed by Hoffmann [9] and Ausner
[11]. With a constant contact angle of
70° (which is the value frequently
quoted in the literature for the material
combination water on steel) the
percentage of wetted area in the
present CFD calculations and in the
calculations of Hoffmann and Iso
tends to be underestimated, whereas
the similar setup of Singh yields, for
an unknown reason, larger values
of wetted area. In Figure 5 the
measurements are shown as blue
triangles, the results of Hoffmann, Iso
and Singh in purple, yellow and green,
respectively, and the current calculations
with the different contact angles
in red, gray and black. The simulations
with constant contact angle and
empirical contact angle model with
70° for a dry surface and 50° for a wet
one still slightly underestimate the
wetted surface. With the empirical
contact angle model 30°/70° the
results are very well within the variation
of the experiments.
5 Wash-off model
The particle wash-off consists of a
two-stage process. First the sedimented
particles on the plate floor are
| | Fig. 4.
Comparison of simulations with empirical contact angle model with
θ dry = 70° and θ wet = 30° for different Weber numbers We. The water height
is indicated by color.
| | Fig. 5.
Normalized wetted surface A wn for different Weber numbers. Blue triangles
indicate the experimental results; results of CFD simulations are displayed
with differently colored lines.
AMNT 2018 | Young Scientists' Workshop
Development and Validation of a CFD Wash-Off Model for Fission Products on Containment Walls ı Katharina Amend and Markus Klein
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resuspended into the water flow, and
then they are transported by the water
flow down the plate and through the
outlet. The model is based on the
approach suggested and investigated
in [1], which is also implemented in
AULA.
5.1 Shields criterion
In this section wash-off criteria, i.e.
the circumstances that have to be met
to resuspend settled particles, are
presented.
Many forces act upon a particle
lying in a sediment bed. The particle
starts to move, if the hydrodynamic
forces and the buoyancy exceed the
forces of gravity, friction, cohesion
and adhesion. Shields proposes a
criterion, which states, that the
incipient motion occurs, when the
shear velocity acting on the particle
exceeds a critical threshold, the critical
shear velocity. This critical shear
velocity u c can be approximated with
the help of the Shields-Rouse equation
[12]. Using the dimensionless
Rouse Reynolds number R *
(1)
(with the specific gravity of sediment
, the particle density ρ p , the
gravitational acceleration g and the
kinematic viscosity of water ν) the
critical dimensionless shear stress τ c
*
and further the critical shear velocity
u c can be calculated via
.
, (2)
(3)
In this relation the adhesion and
cohesion forces are neglected. Thus
according to this criterion all particles
with the same diameter and density
would erode exactly at the same time
as soon as u > u c holds. This leads to
the so called instantaneous total
wash-off. One way to also take the
adhesion and cohesion forces into
account is to model the wash-off as an
exponential decay of the sedimented
particle concentration c s (t) with a
mass erosion rate r e [13], defined as:
,(4)
(5)
and erosion constant (or wash-off
coefficient) ~ r e [13] which has to be
estimated.
5.2 Particle transport
The second stage is the transport of
the volumetric particle concentration
c with [c] = kg/m 3 . It is based on the
OpenFOAM solver scalarTransport-
Foam, which solves a simple transport
equation for a scalar volume field
.
(6)
The resuspended aerosol concentration
is treated as massless particles
that follow the flow perfectly. The
velocity field v, shared by the water
and air phase, is set to zero in cells
without water. Thus particles are
transported only within water and
not within air. The concentration of
eroded particles in each floor face at
each time step serves as the source
term S in the corresponding cell above
the floor. The result of the simulations
Name
is the time-resolved particle mass that
is transported through the outlet.
6 Results of the parameter
variation
In this section parameters such as
particle density and the wash-off
coefficient are varied. Detailed correlations
or influences of the parameters
on the total washed-off mass are
analyzed. Table 1 summarizes the
constant particle properties, the initial
plate loading and the properties of
the water flow. To investigate the
influence of the particle density ρ p
three different densities are used:
10 000 kg/m 3 which resembles the
density of silver, 5000 kg/m 3 and
2500 kg/m 3 which is the effective
density of the aerosol.
The erosion constant ~ r e is also
varied with values of 0.027 s –1 ,
0.135 s –1 and 0.27 s –1 . Together with
the five different Weber numbers (see
Sec. 4) the parameter variation covers
a total number of 45 simulations that
are evaluated hereinafter.
The parameter variation is conducted
with the simplified geometry.
Three seconds of water field simulations
are calculated. The water field is
then in a pseudo-stationary state
and the water, the velocity and the
pressure fields are kept constant.
Overall the particle wash-off and
Value
α Inclination 60°
ρ l Density of water 1000 kg/m^3
c s Initial loading 27 g/m^2
θ dry Contact angle dry 70°
θ wet Contact angle wet 30°
d p Particle diameter 2 μm
| | Tab. 1.
Parameters for simulations with simplified geometry.
471
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Fig. 6.
Time resolved washed off mass for different Weber numbers. Parameters
according to Table 1, ρ p =2500 kg/m 2 and r ~ e = 0.027 s –1 .
| | Fig. 7.
Time resolved washed off mass for different particle densities. Parameters
according to Table 1 and r ~ e = 0.027 s –1 .
AMNT 2018 | Young Scientists' Workshop
Development and Validation of a CFD Wash-Off Model for Fission Products on Containment Walls ı Katharina Amend and Markus Klein
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AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Fig. 8.
Time resolved washed off mass for different wash-off coefficients.
Parameters according to Table 1 and ρ p = 5000 kg/m 2 .
transport simulations last for 30 s.
First the influence of the Weber
number is investigated, see Figure 6.
Increasing Weber numbers correspond
to larger water velocities and
an increasing percentage of wetted
surface. Consequently for larger
Weber numbers more particle mass is
washed off. Thus two effects manifest
in the results: first the larger velocities
are able to wash-off even particles
with larger density. And secondly the
enlarged percentage of wetted surface
enhances the particle wash-off, since
much more particles can be eroded
by the water.
Figure 7 shows the variation of the
particle density. Particles with larger
density cannot be eroded that easily
and hence the total washed off mass
decreases with increasing particle
density, as expected. In Figure 8 the
influence of the wash-off coefficient is
investigated. The total washed off
mass, which is to a large extent determined
by the area of wetted surface,
does not change with different values
of ~ r e but the temporal behavior does.
For a large value of ~ r e a large fraction
of particles erodes in a short timespan.
Asymptotically for t → ∞ the
total washed off mass converges
always to the same amount.
In order to compare the simulations
with experimental data a parameter
set based on Weber et. al [1]
is chosen. Figure 9 displays the results
of the simulation and the experimental
data of test 4. In the experiments
the particles are collected in intervals
of 10 s for a total duration of 130 s.
Due to this sampling strategy the time
resolved washed off particle mass in
the simulations is presented in the
same manner and for the same
duration. A good agreement for the
temporal course of the wash-off as
well as for the total washed off mass
can be achieved.
7 Conclusions and
discussion
This paper presents a CFD particle
wash-off model and particle transport
by gravity driven flows. A parameter
variation was conducted within the
setting of a simplified geometry and
with the geometry of the laboratory
tests. The particle wash-off model,
which is based on Shields criterion
[12] and Weber et. al [1], shows the
expected behavior for varying particle
properties such as particle density and
wash-off coefficient. One key influencing
parameter for the resulting
washed off mass is the percentage of
area covered by water in each case,
which differs with inclination and
mass flow rate. First simulations
with the laboratory geometry show
satisfactory agreement when compared
to the experiments. Nevertheless,
the prediction of particle
wash-off for a large variety of setups
as in the laboratory experiments
( different inclinations, particle and
surface properties and initial loadings)
remains a great challenge and
further comparisons for different
parameter sets are current work in
progress. This study contributes to
the development of a semi-empirical
model to quantify the aerosol washoff
and the wetted surface area during
an accident in a light water reactor.
Acknowledgment
The project underlying this report
is funded by the German Federal
Ministry of Economic Affairs and
Energy under grant number 1501519
on the basis of a decision by the
German Bundestag. The THAI project
was carried out on behalf of the
Federal Ministry for Economic Affairs
and Energy under grant number
1501455 on the basis of a decision by
the German Bundestag. We are also
grateful for the support from Becker
Technologies and the GRS.
References
| | Fig. 9.
Comparison of test 4 of the laboratory experiments with the simulations of particle wash-off
with inclination α = 20°, mass flow rate m = 11 g/s, initial loading c s = 12.5 g/m 2 ,
particle diameter d p = 2 μm, particle density ρ p = 5000 kg/m 3 and wash-off coefficient ~ r e = 0.025 s –1 .
[1] G. Weber, F. Funke, W. Klein-Hessling,
and S. Gupta. Iodine and silver washdown
modelling in COCOSYS-AIM by
use of THAI results. Proceedings of the
International OECD-NEA/NUGENIA-
SARNET Workshop on the Progress in
Iodine Behaviour for NPP Accident
Analysis and Management, 2015.
[2] S. Gupta, F. Funke, G. Langrock, G.
Weber, B. von Laufenberg, E. Schmidt,
M. Freitag, and G. Poss. THAI Experiments
on Volatility, Distribution and
Transport Behaviour of Iodine and
Fission Products in the Containment.
Proceedings of the International
OECD-NEA/NUGENIA-SARNET Workshop
on the Progress in Iodine
Behaviour for NPP Accident Analysis
and Management, p. 1-4, 2015.
[3] M. Freitag, B. von Laufenberg, M.
Colombet, K. Amend, and M. Klein.
Particulate fission product wash-down
from containment walls and installation
surfaces. Proceedings of the 47 th
Annual Meeting on Nuclear
Technology, Hamburg, 2016.
AMNT 2018 | Young Scientists' Workshop
Development and Validation of a CFD Wash-Off Model for Fission Products on Containment Walls ı Katharina Amend and Markus Klein
atw Vol. 63 (2018) | Issue 8/9 ı August/September
[4] K. Amend and M. Klein. Modeling and
Simulation of Water Flow on Containment
Walls with Inhomogeneous
Contact Angle Distribution. ATW
International Journal for Nuclear
Power, 62(7):477-481, 2017.
[5] B. von Laufenberg, M. Colombet, and
M. Freitag. Wash-down of insoluble
aerosols Results of the Laboratory Test
related to THAI AW3 Test. Technical
report, Becker Technologies, 2014.
[6] K. Amend and M. Klein. Simulation of
Water Flow down inclined Containment
Walls. 14 th Multiphase Flow
Conference, Dresden, 2016.
[7] K. Amend and M. Klein. Influence of the
contact angle model on gravity driven
water films. 13 th Multiphase Flow
Conference, Dresden, 2015.
[8] R. K. Singh, J. E. Galvin, and X. Sun.
Three-dimensional simulation of rivulet
and film flows over an inclined plate:
Effects of solvent properties and contact
angle. Chemical Engineering Science,
142:244–257, 2016.
[9] A. Hoffmann. Untersuchung mehrphasiger
Filmströmungen unter
Verwendung einer Volume-Of-Fluidähnlichen
Methode. PhD thesis,
Technische Universität Berlin, 2010.
[10] Y. Iso, X. Chen. Flow transition behavior
of the wetting flow between the film
flow and rivulet flow on an inclined
wall. Journal of Fluids Engineering
133.9:091101, 2011.
[11] I. Ausner. Experimentelle Untersuchungen
mehrphasiger Filmströmungen.
PhD thesis, Technische
Universität Berlin, 2006.
A Preliminary Conservative Criticality
Assessment of Fukushima Unit 1 Debris
Bed
María Freiría López, Michael Buck and Jörg Starflinger
[12] J. Guo. Hunter Rouse and Shields
diagram. Advances in Hydraulic and
Water Engineering, 2:1096–1098,
2002.
[13] R. Ariathurai. A finite element model of
cohesive sediment transportation. PhD
thesis, University of California, Davis,
California, 1974.
Authors
Katharina Amend
Prof. Dr.-Ing. habil. Markus Klein
Responsible Professor
Institute for Numerical Methods in
Aerospace Engineering Universität
der Bundeswehr München
Werner Heisenberg Weg 39
85577 Neubiberg, Germany
Young
Scientists'
Workshop
Awarded
473
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
1 Introduction On March 11, 2011, a big severe accident occurred at Fukushima Daiichi nuclear power plant
(NPP) in Japan resulting in largely melted cores of Units 1, 2 and 3. After the corium solidification, debris beds
were formed and they are considered to be distributed not only in the reactor pressure vessel but also in the primary
containment. If such debris enter in contact with water, recriticality becomes possible. To prevent recriticality, severe
accident mitigation measures prescribe the injection of borated water into the reactor core. However, some leakage of
cooling water and the inflow of groundwater into the reactor building make it very difficult to maintain the necessary
boron concentration to secure the subcritical condition. Currently, the subcriticality of the debris bed is being monitored
by measurements of short lifetime fission products gas (e.g. Xe 133 or Xe 135 ) and water temperature [1]. As no sign of
criticality has been detected until now, the fuel debris is estimated to be subcritical and no preventive measure against
a possible recriticality event is being taken [2]. Nonetheless, this apparently critical-stable condition can change at any
moment due to changes in debris conditions. During the retrieval operations, changes in the water level and debris
shape are expected to occur that will endanger this stability. Thus, using borated water is then planned to ensure the
subcriticality [3].
María Freiría López
was awarded with the
3 rd price of the 49 th
Annual Meeting on
Nuclear Technology
(AMNT 2018) Young
Scientists' Workshop.
A recriticality scenario would lead to a
power increase, new fission products
release and may have severe consequences
even causing a secondary
criticality accident. Prevention and
controlling core sub-criticality is
there fore one of the main accident
management objectives. A risk evaluation
of recriticality is necessary for
the safe preservation and handling of
fuel debris.
This study is part of a larger project,
which pursues to assess the
recriticality potential of fuel debris
after a severe accident taking
Fukushima as reference. The final
aim is to develop a criticality map that
will be used to evaluate the potential
risk of criticality of a fuel debris
taking the debris conditions as input
parameters. The criticality situation of
Fukushima damaged reactors will be
assessed by placing onto the map the
fuel debris conditions revealed by
observations or sample analyses.
In this study, a conservative
criticality evaluation of the Fukushima
Daiichi Unit 1 debris bed was carried
out. Parameters, such as debris size,
porosity, particle size, fuel burnup
and the coolant conditions including
the water density and the content of
boron were considered. The effect of
these parameters on the neutron
multiplication factor was analysed
and safety parameter ranges, i.e.
zones where the recriticality can be
totally excluded, have been identified.
The objective is to fix some boundaries
for the selected parameters
and define the ranges in which the recriticality
could be an issue. This will
provide the starting point for a future
more detailed criticality evaluation.
The Monte Carlo code MCNP6.1
was used to model the hypothetical
debris bed and to calculate the
neutron multiplication factor (k eff )
[4]. The ENDF/B-VII.1 cross section
libraries were used to perform the
calculations.
2 Criticality of debris bed
after a severe accident
After a severe accident (SA), recriticality
occurs when the whole or part of the
reactor becomes unintentionally critical
after the reactor shutdown. This
study focuses on the analysis of recriticality
in debris beds that are formed
either at the bottom of the reactor
vessel (in-vessel debris bed) or in the
reactor containment (ex-vessel debris
bed) after the cool down of the reactor.
Debris beds are formed during a SA
after the solidification of the melted
AMNT 2018 | Young Scientists' Workshop
A Preliminary Conservative Criticality Assessment of Fukushima Unit 1 Debris Bed ı María Freiría López, Michael Buck and Jörg Starflinger
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AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
corium resulting in a porous rubble
structure that mainly consists of fuel
and control rods. If this porous structure
enters in contact with the right
amount of water acting as moderator,
there is a potential for recriticality.
In order to avoid recriticality and its
adverse consequences, a criticality
evaluation of the debris bed needs to
be carried out.
The conditions of the debris bed
can be very diverse and strongly
depend on the accident scenario. The
criticality safety control of the fuel
debris is a challenge principally due to
the large uncertainty of the fuel debris
conditions (location, geometry, composition,
temperature, etc.). Severe
accident codes are able to simulate the
accident progression and can be used
to estimate the debris bed conditions,
however, an adequate observation,
sample taking and analysis of the real
fuel debris are crucial to perform an
accurate criticality evaluation.
Due to the high uncertainty of fuel
debris properties, it is necessary to
prepare a comprehensive and extensive
database, which embraces criticality
data of any possible debris bed.
The main factors on the criticality
evaluation of the fuel debris after a SA
are listed below:
• Total amount of corium
• Composition of corium
• Fuel debris geometry
• Coolant conditions
3 Calculation model
3.1 Geometrical model
of the debris bed
Figure 1 shows the conceptual
geometric model of the debris bed
for the Monte Carlo criticality calculations.
The innermost region of the
model represents the debris itself, as a
porous structure consisting of fuel
| | Fig. 1.
Geometric model of debris bed.
Parameter Range Boundary value
Particle size 1 to 14 mm 10.7 mm
Porosity 0.32 to 0.8 Optimum Porosity
Water void fraction 0 to 90 % 0
Fuel burnup 0 to 60 GWd/t HM
25.8 GWd/t HM
(accident conditions)
Debris bed size 10 to 200 cm 200 cm
Water boration 0 to 2,000 ppm B 0
| | Tab. 1.
Criticality parameters and ranges.
particles and water. For conservative
results, the shape of the debris was
spherically arranged minimizing the
neutron leakage and the critical mass.
Surrounding the fuel debris there is a
water reflector of effectively infinite
thickness (approx. 30 cm). Such configuration
was already used for a
criticality safety evaluation for the
TMI-2 safe fuel mass limit [5].
Debris beds comprise particles of
different shapes and sizes, which are
chaotically arranged in the space. In
order to reduce the computational
effort for the criticality calculations,
some simplifications have been
applied to model the porous structure
of the debris: the particles were
assumed to be spherical, all the particles
were assumed to have the same
size and the particles were assumed to
be regularly distributed in the space
following a Body Centered Cubic
(BCC) lattice [6].
3.2 Corium composition
In this study, the Unit 1 of Fukushima
Daiichi NPP was used as reference
[7, 8].
Conservatively, it was assumed
that there was nothing present in the
fuel debris but fuel pellets and water.
Thus, the negative reactivity effects
due to the possible presence of cladding,
fixed absorbers and structural
materials are ignored. As boundary
conditions, room temperature and a
fuel density of 10.4 g/cm 3 are considered.
ORIGEN 2.1 [9] was used to calculate
the radionuclide inventory for
different average burnups, from fresh
fuel up to a burnup of 60 GWd/t HM .
The average burnup in the reactor of
Unit 1 at the moment of the accident
was calculated to be 25.8 GWd/t HM
[8] and was used as reference model.
To perform the burnup calculations,
fresh fuel UO 2 with an initial enrichment
of 3.7 % wt 235 U was irradiated
considering a specific power of
20 MW/t HM in the reactor.
3.3 Coolant composition
Light water is used as moderator. The
density of the water (or void fraction)
was varied to analyse the influence on
the neutron multiplication factor.
Additionally, boron was added in
every scenario in order to know the
required concentration that guarantee
a subcritical condition of the
debris. Room temperature was considered
for all the calculations.
4 Criticality calculations
Criticality calculations have been
performed for multiple scenarios
using the calculation model described
before. Six parameters have been considered
for these calculations: particle
size, porosity, water void fraction, fuel
burnup, debris size and water boration.
The parameters and ranges of
variation are resumed in Table 1.
In order to analyse all the possible
dependencies between these parameters,
they all have been combined by
pairs, resulting in 15 possible combinations
or calculations sets. In each
calculation set, the paired parameters
have been varied over their whole
ranges, giving to the rest of parameters
a boundary value. The neutron multiplication
factor k eff was then calculated
for all the possible combinations. All
the boundary values have been chosen
to be conservative, except the burnup,
where the value at the moment of
AMNT 2018 | Young Scientists' Workshop
A Preliminary Conservative Criticality Assessment of Fukushima Unit 1 Debris Bed ı María Freiría López, Michael Buck and Jörg Starflinger
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Calculation
set
Particle size
/ mm
| | Tab. 2.
Criticality calculation matrix.
Porosity
/ -
the accident was selected. This allows
focusing on the current criticality
situation of the debris bed of
Fukushima Daiichi Unit 1.
Table 2 summarizes all the criticality
calculations of this study. The
paired parameters of a set of calculations
appear in grey cells where the
variation ranges are given. The white
cells represent the values of the rest of
parameters, the boundary values,
which are kept constant during this
set of calculations. For example, in the
calculation set 3, the particle size
and the fuel burnup are combined;
particles size ranges from 1 to 14 mm
and burnup from fresh fuel up to
60 GWd/t HM . The neutron multiplication
factor for all the possible combinations
of these two parameters was
calculated, while the rest of parameters
maintained the boundary values:
the porosity is set to the optimum value
that maximizes the k eff , no void
fraction nor boration in water is
considered and a debris bed size of
200 cm is modelled.
MCNP6.1 code [4] and ENDF/B-
VII.1 cross section libraries were used
to perform the criticality calculations
of the reactor corium. The standard
deviations of the estimated the
neutron multiplication factors were
always kept below the 0.1 % for all
the calculations of this study.
5 Results
Some of the most important results
of the previously explained criticality
calculations will be shown and
discussed in this section.
Figure 2 corresponds to the calculation
set 1 and shows the influence of
the geometrical arrangement of fuel
particles (porosity and particle size) on
Water void fraction
/ %
| | Fig. 2.
Porosity – Particle Size Unit 1 Fukushima Daiichi criticality map.
| | Fig. 3.
Water void fraction – Boration Unit 1 Fukushima Daiichi criticality map.
the neutron multiplication factor. The
rest of parameters are set to conservative
values. Two different representations
can be distinguished: a 3D criticality
surface and a contour criticality
plot. It can be clearly seen that the k eff
increases slightly with the particle size.
The influence of the porosity is substantially
larger and the k eff reaches a
maximum value for optimum porosities
between 0.74 and 0.79.
The critical level was conservatively
set to k eff = 0.95 as prescribed by
the Nuclear Safety Standards Commission
(KTA) [10]. Thus, the contour
Fuel burnup Debris bed size
/ GWd/t HM / cm
Water boration
/ ppm B
1 1 to 14 0.32 to 0.8 0 25.8 200 0
2 1 to 14 Opt. 0 to 90 25.8 200 0
3 1 to 14 Opt. 0 0 to 60 200 0
4 1 to 14 Opt. 0 25.8 10 to 200 0
5 1 to 14 Opt. 0 25.8 200 0 to 2000
6 10.7 0.32 to 0.8 0 to 90 25.8 200 0
7 10.7 0.32 to 0.8 0 0 to 60 200 0
8 10.7 0.32 to 0.8 0 25.8 10 to 200 0
9 10.7 0.32 to 0.8 0 25.8 200 0 to 2000
10 10.7 Opt. 0 to 90 0 to 60 200 0
11 10.7 Opt. 0 to 90 25.8 10 to 200 0
12 10.7 Opt. 0 to 90 25.8 200 0 to 2000
13 10.7 Opt. 0 0 to 60 10 to 200 0
14 10.7 Opt. 0 0 to 60 200 0 to 2000
15 10.7 Opt. 0 25.8 10 to 200 0 to 2000
line k eff = 0.95 indicates the limit
values from which the subcriticality is
guaranteed. For porosities lower than
0.4 the recriticality can be totally
excluded. In the case of the particle
size there is no threshold value.
Figure 3 shows a criticality map
with the evolution of the neutron multiplication
factor in dependence of
the water properties (void fraction
and boration). As the void fraction
and boration increase, k eff significantly
decreases. For a water void fraction
higher than 78 %, there is not
enough moderator in the system and
475
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
AMNT 2018 | Young Scientists' Workshop
A Preliminary Conservative Criticality Assessment of Fukushima Unit 1 Debris Bed ı María Freiría López, Michael Buck and Jörg Starflinger
atw Vol. 63 (2018) | Issue 8/9 ı August/September
References
476
1. Tsuchiya A, Kondo T, Maruyama H.
Criticality calculation of fuel debris in
Fukushima Daiichi nuclear power station.
In: PHYSOR 2014. Kyoto, Japan; 2014.
AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP
| | Fig. 4.
Debris size – Burnup Unit 1 Fukushima Daiichi criticality map.
critica lity cannot be reached. A boration
of 1,600 ppm B will ensure the
subcriticality independently of the
debris bed conditions.
Figure 4 provides criticality data
as function of the debris size and
burnup. It can be noticed how the k eff
decreases progressively with the
burnup of the core. If the SA happens
at the very end of a fuel cycle, when
the average burnup of the fuel is larger
than 53 GWd/t HM , recriticality will
not be reached under any conditions.
Additionally, the graph provides
the information about the criticality
condition of a debris bed depending of
its size. With these data, the critical
masses for the different burnups
can be calculated. The burnup of
Fukushima Unit 1 at the moment of
the accident was estimated to be
25.8 GWd/t HM . The minimum critical
size of a debris bed for this case is
about 55 cm. For these conditions, the
optimum porosity was calculated to
be 0.75. This results in critical mass of
226.5 kg, which represents only the
2.4 % of the core.
Conclusions
In this study, a conservative criticality
evaluation of the current debris bed
of Fukushima Daiichi Unit 1 was
performed. The lack of knowledge
regarding the debris bed properties
has compelled the use of very conservative
assumptions in the debris
bed models. Six of the most influencing
parameters on the k eff were considered:
debris size/mass, particle size,
porosity, water density and content of
boron in water. The effect of these parameters
on the criticality condition of
Fukushima Daiichi Unit 1 debris bed
was calculated and discussed. Finally,
it was concluded that recriticality can
be totally excluded if:
1. Porosity of the debris bed is lower
than 0.4 or
2. Void fraction of water is higher
than 78 % or
3. Debris mass is lower than 226.5 kg
or
4. Boration in water is equal or
greater than 1,600 ppm B
Additionally, for a reactor core with
UO 2 fuel and initial enrichment of
3.7 % wt 235 U it was found that if a
SA occurred at the very end of a fuel
cycle when the average burnup is
53 GWd/t HM or higher, recriticality is
not achievable under any conditions.
Taking severe accident scenarios
into account, the void fraction threshold
(2) and the debris mass threshold
(3) will be violated under almost all
circumstances. The molten mass
easily reaches values higher than
226 kg, which represents only 2 % of
the core mass, and the void fraction
does not stay at values higher than
78 % for the range of cool temperatures
considered. However, experiments
like DEFOR [11] or FARO [12]
indicate average porosities of about
38 %, which is slightly underneath
the “criticality safe” threshold (1) for
porosity.
As a next step, it is planned to
include new parameters, for example,
the presence of zirconium, control
rods or other reactor structural materials
in order to evaluate their
influence on the criticality of debris
beds. Additionally, new debris bed
configurations will be also investigated.
The first samples and explorations
of debris beds in Fukushima are
planned for this year 2018. This
will provide more information
about the debris characteristics and
will allow a less conservative
and more accurate criticality evaluation.
Acknowledgments
The presented work was funded by
the German Ministry for Economic
Affairs and Energy (BMWi. Project no.
1501533) on basis of a decision by the
German Bundestag.
2. Kotaro Tonoike, Hiroki Sono, Miki Umeda,
Yuichi Yamane, Teruhiko Kugo, Kenya
Suyama. Options of Principles of Fuel Debris
Criticality Control in Fukushima Daiichi
Reactors. In: Ken Nakajima, editor. Nuclear
Back-end and Transmutation Technology
for Waste Disposal. Springer Open;
2015. p. 251–60.
3. Nuclear Damage Compensation and
Decommissioning Facilitation Corporation.
Technical Strategic Plan 2016 for
Decommissioning of the Fukushima
Daiichi Nuclear Power Station of Tokyo
Electric Power Company Holdings, Inc.
2016 Jul.
4. Goorley, John T., James, Michael R.,
Booth, Thomas E., Brown, Forrest B., Bull,
Jeffrey S., Cox, Lawrence J., et al. Initial
MCNP6 Release Overview – MCNP6 version
1.0. Los Alamos National Laboratory
(LA-UR-13-22934); 2013.
5. GPU NUCLEAR. Three Mile Island
Nuclear Station Unit II Defueling
Completion Report. 1990.
6. Freiría López M, Buck M, Starflinger J.
Neutronic Modelling of Fuel Debris for a
Criticality Evaluation. In: PHYSOR 2018.
Cancun, Mexico; 2018.
7. International Atomic Energy Agency
(IAEA). The Fukushima Daiichi Accident
Technical Volume 1/5 Description and
Context of the Accident Annexes.
Vienna (Austria): International Atomic
Energy Agency (IAEA); 2015.
8. Nishihara K, Iwamoto H, Suyama K.
Estimation of fuel compositions in
Fukushima-Daiichi nuclear power plant.
Japan Atomic Energy Agency; 2012.
9. Croff AG. ORIGEN 2.1. Oak Ridge
National Laboratory; 1991.
10. Nuclear Safety Standards Commission
(Kerntechnischer Ausschuss, KTA).
Storage and Handling of Fuel Assemblies
and Associated Items in Nuclear
Power Plants with Light Water Reactors.
2003 Nov. Report No.: KTA 3602.
11. Kudinov P, Karbojian A, Tran C-T,
Villanueva W. Agglomeration and size
distribution of debris in DEFOR-A
experiments with Bi2O3–WO3 corium
simulant melt. Nucl Eng Des.
2013;263(Supplement C):284–95.
12. Magallon D. Characteristics of corium
debris bed generated in large-scale
fuel-coolant interaction experiments.
Nucl Eng Des. 2006;236(19):1998–
2009.
Authors
María Freiría López
Dr.-Ing. Michael Buck
Prof. Dr.-Ing. Jörg Starflinger
Responsible Professor
Institute of Nuclear Technology
and Energy Systems (IKE)
University of Stuttgart
Pfaffenwaldring 31
70569 Stuttgart, Germany
AMNT 2018 | Young Scientists' Workshop
A Preliminary Conservative Criticality Assessment of Fukushima Unit 1 Debris Bed ı María Freiría López, Michael Buck and Jörg Starflinger
atw Vol. 63 (2018) | Issue 8/9 ı August/September
49 th Annual Meeting on Nuclear Technology (AMNT 2018)
Key Topic | Outstanding Know-How
& Sustainable Innovations
The following report summarises the presentations of the Focus Session International Regulation | Radiation
Protection: The Implementation of the EU Basic Safety Standards Directive 2013/59 and the Release of
Radioactive Material from Regulatory Control presented at the 49 th AMNT, Berlin, 29 to 30 May 2018.
The other Focus, Topical and Technical Sessions will be covered in further issues of atw.
477
AMNT 2018
Key Topic: Outstanding
Know-How & Sustainable
Innovations
Focus Session International
Regulation: Radiation Protection:
The Implementation of the EU
Basic Safety Standards Directive
2013/59 and the Release of
Radioactive Material from
Regulatory Control
Christian Raetzke
The topical session Radiation Protection:
The Implementation of the EU
Basic Safety Standards Directive
2013/59 and the Release of Radioactive
Material from Regulatory
Control was coordinated and chaired
by the author of this report.
As the chairman explained in his
short introductory statement, the implementation
of the Basic Safety
Standards (BSS) Directive 2013/59/
Euratom, which has introduced many
changes in radiation protection, has
posed considerable challenges to EU
Member States. In Germany, it became
the occasion for a major revision of the
legal framework and the creation of a
new Act on Radiation Protection. The
chairman expressed his delight that
two distinguished speakers had consented
to talk about implementation
in Germany and Sweden: Dr. Goli-
Schabnam Akbarian from the Federal
Ministry for Environment, Nature
Conservation and Nuclear Safety (BMU)
and Dr Jack Valentin from Sweden.
An important aspect in the regulation
of radiation protection is the
release of radioactive substances from
regulatory control. This is a topic
particularly discussed in Germany
where huge amounts of debris are
produced, and will continue to be produced,
by the dismantling of the fleet
of nuclear power plants. Two eminent
speakers had agreed to shed light on
this issue under a multinational, comparative
angle: Dr Edward Lazo from
the OECD Nuclear Energy Agency and
Dr. Jörg Feinhals from DMT.
As the first speaker, Dr. Goli-
Schabnam Akbarian (Head of Division
“Radiation Protection Law [ionising
radiation]” at the German Federal
Ministry for the Environment, Nature
Conservation and Nuclear Safety) outlined
The Implementation of the
New Euratom BSS in Germany. First,
Ms. Akbarian explained the genesis of
the new Act on Radiation Protection
(Strahlenschutzgesetz, StrlSchG). It
was triggered by the need to transpose
the BSS Directive 2013/59 into
national German law. However, there
were additional reasons for laying a
new foundation for German radiation
protection law which had hitherto
been regulated “merely” by a Government
ordinance (Strahlenschutzverordnung,
StrlSchV). For example, after
Fukushima a need was perceived to
revise the provisions on emergency
preparedness and response which
were scattered among different legal
texts and guidelines. The main body
of the German Strahlenschutzgesetz of
27 June 2017 was to enter into force
on 31 December 2018. It was to be
supplemented by a set of new
ordinances which, as Ms. Akbarian
explained, were currently under
preparation.
Next, she turned to the new structure
introduced by the Directive
2013/59, namely the three exposure
situations: planned, existing and
emergency exposure situations. The
Directive has a greatly enlarged scope
of application as compared to its
predecessor, the Directive 96/29/
Euratom, especially regarding NORM
(naturally occurring radioactive material)
and existing exposure situations.
However, Ms. Akbarian focused
on the category of planned exposure
situations which regards practices. i.e.
human activities that can increase the
exposure of individuals to radiation
from a radiation source. In this area of
particular importance to the nuclear
industry, she highlighted some areas
where meaningful changes had been
introduced. One example was exemption
values which were – though not
too substantially – adapted, which
may result in some activities to require
a licence which had hitherto been
exempted. New requirements were
also introduced concerning the
handling of high-activity sealed
sources. Further modifications affected
the transport of radioactive substances,
a slight change in the dose
limits for occupational exposure and
the introduction of an inspection
programme. For all of these issues, the
new Act included transitional provisions
to allow smooth adaptation.
Ms. Akbarian concluded by
mentioning a host of other aspects
regulated by the new Act, such as
type approval, clearance, radon in
dwellings and at workplaces, and
many others. It became apparent that
the new Act is of fundamental importance,
laying a new foundation for
an area of nuclear law – the law of
radiation protection – which will
become even more important in the
future.
In the ensuing discussion, Ms.
Akbarian was asked about how the BSS
Directive's concept of radiation protection
expert (RPE) and radiation protection
officer (RPO) had been taken
into account in the German Act. She
replied that the traditional two roles
defined in German radiation protection
law, namely the person responsible
for radiation protection (Strahlenschutzverantwortlicher,
SSV) and the
expert entrusted with operational
radiation protection (Strahlenschutzbeauftragter,
SSB), had been retained
as they fulfil this concept. The SSB
basically performed both the role of
the RPE and the RPO. The new Act
strengthened his position, e.g. by introducing
protection against dismissal by
the employer. Another question from
the audience concerned the new
notion of dose constraints and how
stringent requirements for the operator
were. Ms. Akbarian explained
that dose constraints (Dosisrichtwerte)
were included in the new Act and in
supplementing ordinances but that
they were mainly an instrument of
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optimisation to be used by the regulator.
However, there were some requirements
on persons starting a new
practice to analyse whether dose
constraints were useful for this practice,
and to document this analysis
and, if asked, provide the analysis to
the authority.
Next, Dr Jack Valentin (independent
consultant, Sweden, former
Scientific Secretary of the ICRP,
former senior radiation protection
regulator in Sweden) gave a presentation
on Implementation of the EU
BSS Directive in Sweden. First, Jack
Valentin outlined the genesis of the
new radiation protection requirements,
particularly the role of the
International Commission on Radiation
Protection (ICRP) and its
Recommendation no. 103 which was
the basis for the BSS Directive.
He highlighted four essential new
features of ICRP 103: The focus on the
exposure situation (planned/emergency/existing),
not the process
(practice/intervention); the optimisation
of radiological protection in all
exposure situations; the modulation
of optimisation using dose and risk
constraints, and finally, enhanced
protection of the environment by
maintaining biodiversity and ecosystems.
Interestingly, Jack Valentin
highlighted two issues where the BSS
Directive – and, as a consequence,
Swedish legislation and regulation –
was not fully in line with ICRP 103.
One concerned dose limits for occupational
exposure where the Directive
fixes an annual dose of 20 mSv with
no automatic averaging over five years
as had been the case before. As Jack
Valentin pointed out, averaging over a
5-year period facilitated the operator’s
optimisation of protection; for
example, in the case of rare major
jobs, the lowest collective dose was
achieved if a few specialists took
relatively high individual dose.
Concerning emergency worker dose
levels, whereas ICRP did not introduce
any dose limit for a life-saving
informed volunteer, relying instead on
an individual risk/benefit assessment,
the Directive featured a dose limit of
500 mSv. In Jack Valentin's view,
inexperienced rescue leaders might in
future be likely to omit life-saving for
fear of transgression (although doses
will rarely be higher than 500 mSv).
He next depicted the implementation
of the BSS Directive in Sweden
on three levels: the 2018 Radiation
Protection Act, the 2018 Radiation
Protection Ordinance and the 2018
Radiation Protection Regulations
(which have legal force and usually
also include a separate section giving
advice). Like in Germany, these new
or modified texts brought the law fully
into line with the Directive; in some
instances, they use a wording somewhat
different from that of the Directive
(e.g. Swedish law retained the
denomination “activities with ionising
radiation” for planned exposure
situations). And, like in Germany,
there were other reasons for the
legislative and regulatory overhaul
besides the BSS Directive.
When asked about why dose limits
in Sweden were contained in the regulations
rather than in the Act or the
Ordinance, Jack Valentin replied that
this provided some flexibility since
they could more easily be changed. Dr.
Akbarian noted that this was an interesting
viewpoint; she observed the
German view was rather to enshrine
them in legislation because of their
basic importance. Jack Valentin consented
that either view is perfectly
reasonable from its respective angle.
Responding to another comment, Jack
Valentin highlighted the importance
of participation of the public which
had always been a prominent feature
of Swedish nuclear and radiation
protection law and of more general
environmental law.
Next, Dr Edward (Ted) Lazo (Principal
Administrator, Division of Radiological
Protection and Human Aspects
of Nuclear Safety, OECD Nuclear
Energy Agency, Paris) spoke about
The NEA Report on Recycling and
Reuse of Materials Arising from
Decommissioning of Nuclear Facilities.
As Ted Lazo explained, significant
volumes of materials will be gen erated
from decommissioning of nuclear
facilities throughout the world. In
Europe, more than a third of currently
operating reactors were due to be shut
down by 2025. The importance of the
management of slightly contaminated
material was likely to grow and the
inherent value of these materials and
the need to reduce radioactive waste
to be disposed required attention.
However, the international community
was far from a complete
harmonization of the strategies and
regulations on this issue.
In order to rise to this challenge,
the NEA Cooperative Programme on
Decommissioning (CPD) Task Group
on Recycling and Reuse of Material
was created. The Task Group had produced
its first report in 1996; a new
report, updating and extending the
previous one, was released in 2016.
This recent report noted that in the
past two decades, international guidance
had been issued, notably the
IAEA guide RS-G1.7 and several
recommendations of the expert group
under article 31 of the Euratom Treaty.
Still, there was only a limited degree
of alignment of national regulations.
As the report noted, unconditional
clearance – which is normally preferred
to conditional clearance if
possible – is well-regulated in all
countries the report looked at, however
some differences between countries
remained, e.g. in the disposal of
rubble and concrete blocks from
dismantling. For conditional clearance,
in the absence of international
guidance, regulatory systems varied
greatly. As Ted Lazo pointed out, the
BSS Directive may help to achieve
greater consistency.
Generally, as he noted, since the
first report of 1996 a greater consolidation
and alignment of the requirements
to control dose and
exposure to workers, members of the
public and the environment had been
achieved; there was also an increase
in general public awareness but issues
over public acceptability remained.
Education, information sharing and
awareness-raising through direct
and public communications could be
utilized to alleviate many of the fears
surrounding recycling and reuse of
materials. Besides, a well-established
relationship between the nuclear
industry and the recycling industry
could have a considerably positive
effect to ensuring stakeholder and
public acceptance of materials. Ted-
Lazo concluded by saying that numerous
challenges to recycling and reuse
of materials persisted internationally
and that the Task Group felt that
success stories, such as those included
in its report, needed to be shared
internationally to help build consensus
for the safe recycling and reuse of
valuable materials.
Last not least, Dr. Jörg Feinhals
(Head of Project Group “Radiation
Protection and Disposal” at DMT,
Hamburg; Member of the Directorate
of the German-Swiss Association for
Radiation Protection) took the floor on
the topic Necessary Modifications
on Clearance Regulations in Germany
and Switzerland – Comparative
Analysis. Jörg Feinhals first
remarked that comparison between
the two countries is rendered more
difficult by the fact that sometimes
the same (German) word is used
with different meanings – a difficulty
which remarkably cannot arise with
English where there is a common
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understanding in the international
community. Next, Jörg Feinhals
depicted the Swiss situation. The
Swiss Radiation Protection Ordinance
(Strahlenschutzverordnung, StSV) was
revised with effect from 1 st January
2018 in order to keep up with the state
of the art (ICRP 103 and IAEA BSS)
and to be in compliance with EU BSS
Directive in most cases, however without
changing things being tried and
trusted. Besides, complementary
regulations were still in the making.
Jörg Feinhals analysed the criteria for
exemption and clearance in the Swiss
system, namely surface contamination,
net dose rate and activity. He
compared the new Swiss clearance
criteria to the German ones and concluded
that average parameters were
no longer more restrictive in Switzerland
than in Germany.
With a view to the revision of
German radiation protection law explained
by Goli-Schabnam Akbarian
in the first presentation, Jörg Feinhals
focussed on clearance. Clearance,
until now regulated in section 29 of
the existing Radiation Protection
Ordinance, was the object of section
68 of the new Act on Radiation Protection;
however, this section merely
empowered government to regulate
clearance in a new ordinance, which
was still in the making. Based on analysis
of a draft version of this new ordinance,
Jörg Feinhals concluded that
most values in a table appended to the
new ordinance were unchanged as
compared to the existing values in Appendix
3 Table 1 of the existing Ordinance.
However, there were some
changes in detail, most notably a new
term for specific clearance (Spezifische
Freigabe) and mass limits of 10.000
Mg/a for Cs-137 in concrete debris
and 10 Mg/a for scrap, if only one specific
nuclide is detected. As to the effects
of these differences
in terms of masses and cost, Jörg
Feinhals stated that there was a
tendency towards shifting between
clearance pathways (e.g. Cs-137) in a
direction from clearance to specific
clearance, from there to decay storage
and thence to long-term storage.
Besides, he expected in some cases
an increased time expenditure for
measurement or new equipment (e.g.
in the case of Eu-152/154). This
was somewhat offset by increase
of values for some nuclides (e.g. Pu-
238/39/40/41, Am-241). Whereas
the mass limit for concrete debris
Cs-137 was acceptable, the limit for
clearance of metal scrap in case of
single nuclides seemed to be out of
practice, Jörg Feinhals noted. Overall,
he predicted a (merely) moderate
increase of effort and costs, provided
however that the path of specific
clearance proved to be fully operational.
He concluded his presentation
by pointing out some constellations
where difficulties could arise due to a
lack of transitory or grace periods for
specific cases.
When asked about contaminated
soil after accidents, Jörg Feinhals
stated that from the view of emergency
preparedness and response it
was very necessary to have a plan for
the disposal of large amounts of contaminated
soil and of other materials.
This should not be based on the de
minimis concept but rather on the
basis of an existing exposure situation,
i.e. a 1 mSv/a dose limit for the
public.
The session closed with a panel discussion
with the four speakers and the
audience. The chairman opened the
discussion by sharing his impression
that while the BSS Directive and the
implementing legislation in EU
Member States introduced many new
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REPORT
factors such as the structuring along
exposure situations and the inclusion
of many situations with natural
radiation which had hitherto not been
regulated, it seemed to him that there
were no dramatic changes to the
regulation of the nuclear industry.
Goli-Schabnam Akbarian basically
agreed, nevertheless pointing out
there were some issues (such as the
new dose limit for the lens of the eye)
where a solution would have to be
found to demonstrate compliance in
practice. Jörg Feinhals, when looking
at clearance, took a balanced position:
changes were basically moderate but
there was some increase in risk for
nuclear industry due to the fact that
concerning some substances there
was a shift from unconditional to
specific clearance; the latter was liable
to be more prone to public controversy.
On the other hand, nuclear
industry could be happy that specific
clearance as such had been retained in
legislation at all. Jack Valentin tended
to agree that nuclear industry was not
overly affected. He said that in this
respect there was a clear divide
between the nuclear and non-nuclear
area and that most problems would
arise outside the nuclear industry. He
also mentioned that some changes
were likely to have an influence on
public perception. Ted Lazo agreed
and emphasised the role of stakeholder
participation, which he
expected to grow in importance; it
was essential, he noted, to take this
into account.
The chairman remarked that radiation
protection experts so far, in his
view, had not entirely succeeded in
educating the public, and asked how
participation could be meaningful
given the limited knowledge of the
average member of the public. Ted Lazo
responded that education in radiation
protection indeed was not feasible on a
general basis; how ever, his personal
experience from Fukushima had
shown that those persons actually
affected by a crisis were very knowledgeable
and had a good perception of
what mattered in radiation protection.
Jack Valentin agreed: it was essential to
utilise people's common sense. This
was supported by Jörg Feinhals who
emphasised that communication needed
to be kept easy, simple and truthful.
Statements by NGOs in Germany about
lethal effects of clearance under the
10-Micro sievert-concept showed that
much could go wrong if calculation
was done with inappropriate numbers.
Next, the topic of clearance vs.
exemption levels was brought up. The
BSS Directive (recital 37) follows the
philosophy that the activity concentration
limits for both clearance and
exemption should be the same. The
chairman stated this seemed logical to
him and asked whether this wasn't an
aspect of the new Directive which was
welcome to everyone. Jörg Feinhals explained
that there may be different
conditions and different reasons for
clearance and exemption assumptions
and limits. Historically, the – very
influential – values in the IAEA RS-G1.7
document were meant for exemption
and not for clearance of huge amounts
of materials. There was also an issue
about the efforts for licensing due to
the reduction of exemption values.
Jörg Feinhals explained that in nearly
all cases not the exemption values in
column 3 of the relevant table in
the Strahlenschutzverordnung (specific
activity) but the exemption values in
column 2 (total activity) are relevant
for the licensing procedure. These
exemption values are not changed.
Differences between exemption and
clearance are mainly based on different
scenarios for exemption (do I need
a license for a small amount of mass
with radio activity?) and clearance
(can I dispose of large amounts of
contaminated/activated material?).
Nevertheless, Jörg Feinhals saw a certain
benefit in adopting a plain and
easy approach by taking the same
values. Ted Lazo agreed and proposed
that a new terminology may be needed
to introduce the differentiation which
was necessary in some cases.
Finally, a participant asked about
averaging criteria. He stressed their
importance and asked whether any
international regulations will be published
to this issue. Jörg Feinhals agreed
about the relevance of averaging criteria
and noted that this topic has been
brought to the attention of the IAEA for
establishing guidance for member
states.
At the end of the session, there
was a strong final applause for the
excellent speakers.
Report: GRS Workshop
“Safety of Extended Dry Storage
of Spent Nuclear Fuel”
Klemens Hummelsheim, Florian Rowold and Maik Stuke
Since up to now all NPP-operating countries are lacking a disposal site for high-level waste and thus are confronted
with the necessity of prolonged storage periods, an increase of scientific working effort was notable in the past years.
From the German perspective, irradiated fuel assemblies from nuclear power plants are packed in transport and storage
casks, e.g. of CASTOR® type, following the wet storage in the spent fuel pool of the reactor. The originally planned
storage period of a maximum of 40 years will not be sufficient in all cases. According to the German Atomic Energy Act,
a license “may only be renewed on imperative grounds and after it has been discussed in the German Bundestag”. On the
technical side, the availability of all safety functions of the storage system and thus the compliance with the respective
safety goals of both the aged casks including their components and structures as well as the inventories have to be
demonstrated for the envisaged prolongation. Special and unique features of Germany’s spent fuel situation are the
very high burn-up of the fuel, the use of mixed oxide fuels (MOX) and a large variety in casks, fuel assembly types and
cladding materials. To address these technical aspects that may be important for extended storage, the Gesellschaft für
Anlagen- und Reaktorsicherheit (GRS) gGmbH in Garching initiated in 2017 an annual workshop. This year it took
place from 6 th to 8 th June entitled “Safety of Extended Dry Storage of Spent Nuclear Fuel”. Nearly 60 experts from
Report
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30 institutes of 10 countries as well as representatives of the International Atomic Energy Agency (IAEA) attended the
event. The experts focused on scientific and technical aspects that may be important for extended storage. With 18 oral
contributions the science-focused agenda of the workshop reflected the broad diversity in current research projects.
The subjects ranged from cladding material behavior to the thermo-mechanical simulation of fuel rods and fuel
assemblies. Furthermore, specific aspects were addressed such as non-destructive testing of casks or management
issues, as well as analysis of the still unresolved technical issues that need to be closed by further research programs.
The sessions started with a talk given
by Maik Stuke from GRS, Germany,
entitled “Current Research Activities
at GRS”. The presented activities focus
on the long-term behavior of drystored
fuel assemblies with special
emphasis on high burn-up values of
65 GWd/tHM UO2 and MOX fuel. The
presentation included detailed maps
of temperature fields of loaded casks.
The thermo-mechanical behavior of
the fuel rods was investigated using
the TESPA-ROD code. Furthermore,
research on the influence of hydride
behavior in cladding materials was
presented e.g. an in-depth analysis of
hydrogen terminal solid solubility.
Representing the IAEA, Alena
Zavazanova provided in her talk “IAEA
safety standards for dry storage of
SNF” an overview of the regulatory
considerations concerning nuclear fuel
management. Some of the IAEA Safety
Standards concerning the storage of
spent nuclear fuel were discussed in
greater detail, e.g. the “General Safety
Requirements” part 5 and 6 of the IAEA
Safety Standard “Predisposal Management
of Radio active Waste”, and the
“Specific Safety Guide 15: Storage of
Spent Nuclear Fuel”.
In their joint presentation
“ Response of Irradiated Nuclear Fuel
Rods to Quasi-Static and Dynamic
Loads” Efstathios Vlassopoulos and
Dimitri Papaioannou presented a
collaborative effort of the École
polytechnique fédérale de Lausanne
(EPFL) in Lausanne, Switzerland, the
Swiss National Cooperative for the
Disposal of the Radioactive Waste
(Nagra), the European Commission
Joint Research Center (JRC) in Karlsruhe,
Germany, and CADFEM (Suisse)
AG in, Aadorf, Switzerland. The group
investigates the response of spent
nuclear fuel in various loading conditions.
The focus lies on the determination
and the study of the
mechanical properties and rod failure
processes using experimental and
numerical techniques.
Jesus Ruiz-Hervias from the Technical
University of Madrid, Spain,
presented in his talk “Effect of Zirconium
Hydrides on the Mechanical
Behaviour of Cladding” investigations
on the effect of hydrogen embrittlement
on the mechanical behaviour
of un-irradiated cladding. One of the
objectives of the work was to develop
operative failure criteria to predict the
cladding behaviour during dry storage
and transport operations. He presented
experimental and numerical
results for ring compression tests of homogeneously
hydrogen loaded samples
and the derived failure criteria.
As chairperson of the Extended
Storage Collaboration Program
( ESCP) Steering Committee of the
Electric Power Research Institute
(EPRI), USA, Hatice Akkurt provided
in her talk “Extended Storage Collaboration
Program (ESCP) for Addressing
Long-Term Dry Storage
Issues” the actual ESCP Program. The
collaboration aims at enhancing the
technical bases to ensure a continued
safe long term used fuel storage and
transportability. It involves about
575 members from 19 countries and
is organized in 6 subcommittees:
Fuel Assembly, Thermal Modelling,
CISCC, Non-Destructive Examination,
Canister Mitigation/Repair, and International.
Amongst other topics she
discussed results from the Demo Project
in which a cask that has been
loaded in 2017 is investigated under
defined conditions.
In his capacity as Sub-Coordinator
Stefano Caruso of the Swiss NAGRA
presented the proposal for the Joint
Programme on Radioactive Waste
Management and Disposal in Europe
(RWMD-EJP). He discussed the aims
of this programme and its current
state of definition with focus on the
budgetary and time planning. As it
involves several authorities, it is
subjected to many constraints. The
proposal is currently undergoing the
second review; final submission is
planned for the end of September. The
first implementation phase will extend
over five years (EJP1 2019-2024),
with a maximum budget of 32.5 M€.
In his talk “Sensitivity Tests of
Several Factors Affecting Dynamic
Buckling Strength of Spacer Grids of a
Spent Nuclear Fuel”, Jae-Yong Kim
from the Korea Atomic Research
Institute (KAERI) reported about a
research program on spent nuclear
fuel. He discussed a pendulum impact
tester, installed in 2017 to improve
analytical skills of very limited impact
test results in hot cells. The tests were
established to assure consistency and
qualification of impact test results.
The functional verification tests are
performed to confirm the hammer’s
impact velocity, initial impact energy
and heating conditions of an electric
furnace. Finally, impact tests were
performed with simulated spacer
grids replacing the spent fuel spacer
grids by changing ambient temperature
and cell size.
Michel Herm from the Institute for
Nuclear Waste Disposal of Karlsruhe
Institute of Technology (KIT-INE), Germany,
presented “Research activities
on safety of extended dry storage of
spent nuclear fuel at KIT-INE”.
Using irradiated fuel rod segments
from the PWR Gösgen, Switzerland,
and Obrigheim, Germany, radionuclide
inventories of Zircaloy-4
samples were determined and compared
to theoretical predictions. UO 2
and MOX samples were used and
different methods applied according
to the nuclides. Nuclide inventories
were investigated in the fuel region, as
well as in the plenum. Separation
methods for Chlorine and Iodine are
currently under development.
A second talk from the KIT
was given by Mirko Grosse from
the Institute of Applied Materials
(KIT-IAM). He presented his work
entitled “Investigation of the hydrogen
diffusion and distribution in
Zirconium by means of Neutron
Imaging”. The work conducted in an
international effort described the
hydrogen diffusion and distribution
in zirconium, analysed by using
neutron imaging facilities CoNRad
(Berlin, Germany), ANTARES (Garching,
Germany), and ICON (Villigen,
Switzerland). Neutron imaging
enables generally in-situ measurements
with high accuracy. It was
especially used to study the hydrogen
diffusion and redistribution in case of
stressed samples. Delayed Hydride
Cracking (DHC) is of high interest and
will be further investigated.
Uwe Hampel from the Technical
University of Dresden, Germany presented
results from the Project
DSC-Monitor in his talk titled “Potential
Methods for the Long-term Monitoring
of the State of Fuel Elements in
Dry Storage Casks”. The fundamental
investigations aim on the feasibility
and applicability of potential methods
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for non-intrusive monitoring of the
state of fuel elements in dry storage
casks. In particular radiation-based
methods, thermography and acoustic
methods were discussed. The assessment
of the applicability, sensitivity,
and uncertainty of the proposed
methods are underway using numerical
and experimental techniques.
As a representative of the German
Federal Office for the Safety of Nuclear
Waste Management (BfE), Tobias
Zweiger briefly outlined the current
state of spent fuel storage in Germany,
the new structure of the BfE and the
work areas of the respective divisions.
A summary of ongoing work in the
spent fuel storage division and an outlook
on future research activities and
interests of the BfE was discussed.
Gerold Spykman from TÜV NORD
EnSys GmbH & Co. KG, Hannover,
Germany, provided in his talk “Dry
storage of high level waste in Germany
– Safety assessments for 40 years
and beyond” his view on the licensing
of cask inventories and on the
licensing of the storage facilities in
Germany. The formulation and the
ranking of the influencing factors on
storage, transportability and final
disposal were presented as a gap
analysis based on the experiences
from the licensing and surveillance
procedures in Germany from the TÜV
NORD EnSys point of view.
Francisco Feria from CIEMAT,
Spain, provided an overview entitled
“CIEMAT response to challenges on
fuel safety research during dry
storage”. The research focuses on
developing predictive capabilities on
fuel rod performance during dry
storage including extended storage. To
assess the spent nuclear fuel integrity
along dry storage and to determine its
characteristics prior to transport,
CIEMAT’s strategy consists of the
extension of the FRAPCON code
( FRAPCON-xt) to the dry storage and
thus to enable predictions of in-clad
hydrogen radial distribution and characterization
of the outward cladding
creep. The adoption of best- estimateplus-uncertainty
methodology (BEPU)
allows determining the code’s uncertainty.
The talk “Considerations on spent
fuel behaviour for transport after
extended storage” was given by
Konrad Linnemann from the Safety of
Transport Containers Division of the
German Bundesanstalt für Materialforschung
(BAM). His presentation
focused on the fuel rod failure in the
transport package safety assessment
and the assumptions for criticality
safety analysis, leading to the discussion
of aspects about transport after
extended storage. A stress limit was
determined, beyond which rod failure
is assumed to occur, leading to fissile
material release in the cask cavity.
As a conclusion, further experimental
investigations were described as
desirable.
A further talk entitled “R&D initiatives
at BAM concerning spent nuclear
fuel integrity during long term storage”
was given by Teresa Orellana
Pérez from the Safety of Storage
Containers Division of BAM. The
research project aims at developing
numerical methods that will enable
brittle failure probability assessments
of fuel claddings and the estimation of
boundary conditions to prevent cladding
failure. Experimental data
from ring compression tests will be
analysed in cooperation with the
University of Madrid. In addition, the
perspective to contribute to a comprehensive
fuel cladding characterization
in the frame of the EJP was discussed.
Julia Neles from the Öko-Institut
e.V., Germany, provided a talk entitled
“Organizational and management
aspects in extended storage”. One
focus was on the German Act on
Reorganization of Nuclear Waste
Responsibilities from 2017, which
regulates the transition of responsibilities
for the waste management
from the waste producers to the
public-owned operator BGZ (Gesellschaft
für Zwischenlagerung). Knowledge
management has to be applied at
authorities and the long-term preservation
of expert organisation
knowledge has to be clarified. It was
also pointed out, that the periodic
safety revisions should be strengthened
as an inspection tool for organizational
and management topics.
In his talk “Hydrides and Zr-
Cladding Mechanics”, Weija Gong of
the Swiss Paul Scherrer Institute (PSI)
presented an overview of ongoing research
topics at PSI. Using neutron
imaging, investigations were conducted
on hydrogen diffusion in
Zr-Materials under stress. Combining
experimental results and Finite-
Element-Modelling for the stress field,
a thermodynamic modelling was
achieved defining a stress-dependant
chemical potential. Liner claddings
were also carefully studied at PSI,
especially for hydrogen redistribution
issues during cooling and under stress
conditions. Also, some tests were
performed to determine the impact of
hydride reorientation on the fatigue
of the material.
The workshop was concluded with
the talk “Open questions on the road
to reliable predictions” presented
by Florian Rowold from GRS. He
discussed the large number of
parameters governing the cladding
hoop stress and their strong interdependencies.
Due to the latter it
seems indispensable to establish an
integrated calculation method which
covers the entire lifetime of a fuel
rod. It was shown that this is also
important with respect to conservatism
and predictability for long time
scenarios. An integrated approach
combines reliable end-of-life fuel
data, thermal modelling and fuel
performance code enhancement as
well as improved material behaviour
understanding and simulation.
Prior to the GRS workshop and in
conjunction to it, a two-day meeting
of the International Subcommittee of
the Extended Storage Collaboration
Program of the Electric Power
Research Institute was also hosted at
GRS in Garching. The objective of this
meeting was to further establish an
international network that shares
essential information in the field of
long-term storage of spent fuel.
Both events showed the need of
intensive exchange of knowledge
with a clear focus on scientific and
technical aspects. The implications of
an extended storage of used nuclear
fuel cover a large variety of features,
phenomena and effects. Due to the
existing similarities in the international
context of the spent fuel characteristics,
it seems to be obvious to
involve experts from other countries.
This gives the opportunity for synergetic
effects, especially in the light of
large-scale experiments and limited
national research funding. The large
number of participants fortified the
general opinion that an exchange of
scientific and technical knowledge is
needed to identify and prioritize
the knowledge gaps for the German
situation. All participants valued the
workshop as a great success. The next
workshop “Safety of Extended Dry
Storage of Spent Nuclear Fuel” is
planned again as a three-day event at
GRS in Garching during the first week
of June 2019.
Authors
Klemens Hummelsheim
Florian Rowold
Maik Stuke
Gesellschaft für Anlagen- und
Reaktorsicherheit (GRS) gGmbH
Boltzmannstr. 14
85748 Garching (München),
Germany
Report
Report: GRS Workshop “Safety of Extended Dry Storage of Spent Nuclear Fuel”
ı Klemens Hummelsheim, Florian Rowold and Maik Stuke
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Inside
483
KTG-Sektion NORD
Einladung: Erfolgreicher Nachweis
von kohärenten Neutrinos
im Kernkraftwerk Brokdorf
Neutrinos sind sogenannte „Geisterteilchen“, weil sie viele
Lichtjahre Flugweg Materie durchdringen können ehe sie
mit ihr wechselwirken. Sie entstehen in verschiedenen
Quellen, wie etwa im Herzen der Sonne bei Fusionsprozessen.
Kernreaktoren emittieren ebenfalls einige
Prozent der frei gesetzten Energie in Form von Neutrinos,
weswegen in unmittelbarer Nähe eines Reaktors sehr
interessante Experimente mit Neutrinos möglich sind.
Im Vortrag wird erklärt, wie man diese Neutrinos nachweisen
kann, welche spannenden Fragestellungen sich damit
verbinden und welche Rolle das Kernkraftwerk Brokdorf
dabei spielt.
Der Referent, Prof. Dr. Dr. h.c. Manfred Lindner ist
Direktor am Max-Planck-Institut für Kernphysik in Heidelberg.
Er forscht auf dem Gebiet der Teilchen- und Astroteilchenphysik
mit dem Ziel, die elementare Struktur und
Entstehung der Materie zu erklären. Dazu ist er führend
an internationalen Projekten aus dem Bereich der
Neutrino- Physik und der Suche nach Dunkler Materie
beteiligt. Im Anschluss an den etwa einstündigen Vortrag
wird Gelegenheit zur weiteren Diskussion sein.
Interessierte KTG-Mitglieder sowie Freunde und
Bekannte sind herzlich eingeladen am Mittwoch, den
17. Oktober 2018 um 13:00 Uhr, bei der PreussenElektra
GmbH, Tresckowstraße 5, Hannover, teilzunehmen.
Wir danken der PreussenElektra GmbH für die Initiative
zum und die Unterstützung des Vortrags.
Wir bitten um eine namentliche Anmeldung
der Teilnehmer bis zum 4. Oktober 2018 unter
Telefon 0511 439-2184 oder an
thomas.froehmel@preussenelektra.de
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Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag und wünscht ihnen weiterhin alles Gute!
September 2018
99 Jahre | 1919
27. Dipl.-Ing. Werner H.F. Hünlich,
Baden Baden
90 Jahre | 1928
16. Dr. Walter Schueller, Weingarten
89 Jahre | 1929
15. Dipl.-Ing. Dankward Jentzsch,
Bergisch Gladbach
22. Dipl.-Ing. Herbert Küster, Bochum
23. Dr. Hubert Eschrich, Geel
88 Jahre | 1930
22. Dr. Wilhelm Peppler, Dobel
87 Jahre | 1931
04. Dr. Klaus Schifferstein, Erftstadt
22. Dipl.-Ing. Emile A. Fossoul, Kraainem
22. Dipl.-Ing. Ludwig Seyfferth, Egelsbach
86 Jahre | 1932
12. Dipl.-Ing. Richard Ruf, Eckental
85 Jahre | 1933
17. Dr. Ing. Manfred Mach, Breitenfelde
20. Dr. Willy Marth, Karlsruhe
84 Jahre | 1934
13. Dipl.-Phys. Veit Ringel, Dresden
13. Dr. Richard von Jan,
Fürth-Burgfarrnbach
30. Dr. Klaus Ebel,
Ingersleben OT Morsleben
83 Jahre | 1935
27. Dipl.-Ing. Klaus Kleefeldt,
Karlsdorf-Neuthard
82 Jahre | 1936
7. Dr. Harald Stöber,
Eggenstein-Leopoldshafen
13. Dipl.-Ing. Jakob Geissinger, Ettlingen
13. Dipl.-Ing. Harald Gruhl, Hemhofen
17. Dipl.-Ing. Hermann Buchholz,
Neunkirchen-Seelscheid
19. Dr. Ludwig Lindner, Marl
81 Jahre | 1937
2. Dipl.-Ing. Dieter Ewers,
Mühlheim/Main
15. Dr. Jochem Eidens, Aachen
17. Dr. Thomas Roser,
Bonn – Bad Godesberg
22. Dr. Uwe Schmidt, Obertshausen
80 Jahre | 1938
17. Prof. Dr. Heiko Barnert, Baden bei Wien
79 Jahre | 1939
17. Dr. Klaus Böhnel, Karlsruhe
21. Dr. Helmut Wilhelm, Rösrath
77 Jahre | 1941
5. Prof. Dr. Manfred Popp, Karlsruhe
14. Dr. José Lopez-Jimenez,
Majadahonda/ESP
14. Dr. Werner Rosenhaue, Rösrath
19. Dipl.-Ing. Horst Heckermann,
Heiligenhaus
21. Dr. Wolfgang Köhler, Kalchreuth
75 Jahre | 1943
13. Günter Reiche, Berlin
70 Jahre | 1948
6. Dr. Heinz-Peter Berg, Braunschweig
8. Bärbel Leibrecht, Krefeld
10. Dr. Eberhard Hoffmann, Bochum
17. Robert Holzer, Bad Homburg
65 Jahre | 1953
8. Bernhard Lehmann, Hochdorf
22. Gerhard Koehler, Sandhausen
23. Prof. Dr. Thomas Schulenberg,
Walzbachtal
60 Jahre | 1958
10. Stefan Busch, Bad Bentheim
50 Jahre | 1968
10. Dr. Martin Filss, München
14. Karsten Beier
KTG Inside
atw Vol. 63 (2018) | Issue 8/9 ı August/September
484
NEWS
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Verantwortlich
für den Inhalt:
Die Autoren.
Lektorat:
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Kerntechnische
Gesellschaft e. V.
(KTG)
Robert-Koch-Platz 4
10115 Berlin
T: +49 30 498555-50
F: +49 30 498555-51
E-Mail:
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Oktober 2018
91 Jahre | 1927
23. Dr. Helmut Krause, Bad Herrenalb
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89 Jahre | 1929
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Großhesselohe
87 Jahre | 1931
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Bergisch Gladbach
84 Jahre | 1934
31. Prof. Dr. Rudolf Taurit, Lübeck
83 Jahre | 1935
15. Dr. Dietrich Budnick, Erlangen
82 Jahre | 1936
1. Dr. Hans-Jürgen Dibbert,
Heiligenhaus
10. Hans-Jürgen Rokita, Schnakenbek
31. Prof. Dr. Hans-Dieter Schilling,
Hattingen
81 Jahre | 1937
21. Dipl.-Ing. Gerhard Hendl, Freigericht
80 Jahre | 1938
3. Dr. Hans-Jörg Wingender, Mömbris
4. Dr. Helmut Albrecht,
Eggenstein-Leopoldshafen
26. Dr. Knut Scheffler, Beckedorf
79 Jahre | 1939
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Karlsruhe
10. Dipl.-Ing. Siegfried Jackem Bonn
13. Helmut Goebel, Jülich
21. Dipl.-Ing. Michael Will, Morsbach
78 Jahre | 1940
19. Dr. Gustav Katzenmeier, Karlsruhe
24. Dr. Peter Wirtz,
Eggenstein-Leopoldshafen
30. Dr. Fritz Ruess, Forchheim
77 Jahre | 1941
21. Ing. Peter Schween,
Stutensee-Blankenloch
31. Dr. Eike Roth, Klagenfurt
76 Jahre | 1942
7. Dr. Klaus W. Stork, Bad Dürkheim
20. Dipl.-Ing. Norbert König, Baiersdorf
21. Dr. Enrique Horacio Toscano,
Stutensee
22. Dr. Alexander Alexas, Stutensee
75 Jahre | 1943
4. Klaus Günther, Bergisch Gladbach
9. Alfred Kapun, Obertshausen
70 Jahre | 1948
9. Bernd Müller-Kiemes, Bingen
14. Claus Fenzlein, Erlangen
65 Jahre | 1953
17. Edgar Albrecht, Beckedorf
20. Dieter Gaeckler, Lingen
Top
First Westinghouse AP1000
nuclear plant Sanmen 1
completes commissioning
(westinghouse) On 6 June 2018,
Westinghouse Electric Company,
China State Nuclear Power Technology
Corporation (SNPTC) announced
that the world’s first AP1000 nuclear
power plant located in Sanmen,
Zhejiang Province, China has successfully
completed initial criticality.
“Today we completed the final
major milestone before commercial
operation for Westinghouse’s AP1000
nuclear power plant technology,” said
José Emeterio Gutiérrez, Westinghouse
president and chief executive
officer. “We are one step closer to
delivering the world’s first AP1000
plant to our customer and the world –
with our customers, we will provide
our customers in China with safe,
reliable and clean energy from
Sanmen 1.”
| | First Westinghouse AP1000 nuclear plant Sanmen 1 completes
commissioning (Photo: Westinghouse)
Following initial criticality will
be connection to the electrical grid.
Once plant operations begin at
Sanmen 1, it will be the first AP1000
nuclear power plant in operation,
offering innovative passive safety
system technology, multiple layers of
defense and advanced controls for
unequaled reliability and safety.
Commenting on Westinghouse’s
strong partnership with the China
customer, Gavin Liu, president –
Asia Region stated, “Westinghouse’s
success in China is the joint effort
between Westinghouse and our China
customers.” He added, “This partnership
and cooperation model can help
to deploy a fleet of AP1000 units in the
world for many years to come.”
On 30 June 2018 the Sanmen
nuclear power plant has begun initial
connection to the electrical grid.
Sanmen 1’s turbine generator is now
initially connected to the electrical
grid and has begun generating
electricity.
Sanmen 1 is capable of generating
1,117 megawatts of electricity when at
full power. It’s also the first of a fleet of
four new AP1000 plants in eastern
China and will provide safe, reliable
and environmentally-friendly energy
for the next 60+ years.
Commenting on Westinghouse’s
recent successes in China, David
Durham, Westinghouse senior vice
president, New Projects Business
stated, “It’s such an exciting time for
Westinghouse, our China customer
and the nuclear industry, as we
proudly move closer and closer to
100 percent power and commercial
operation at Sanmen 1.”.
Westinghouse currently has six
AP1000 nuclear power plants progressing
through construction, testing
and start-up. These projects include
two units in Sanmen, Zhejiang
Province, China, two units in Haiyang,
Shandong Province, China, as well as
two units under construction at the
Alvin W. Vogtle Electric Generating
Plant near Waynesboro, Georgia, USA.
| | www.westinghousenuclear.com
World
Belarusian nuclear station
meets ‘Stress Test’ standards,
EU Peer Review concludes
(nucnet) EU regulators have concluded
that the Belarusian nuclear
power station under construction near
the town of Ostrovets complies with
the bloc’s risk and safety assessments
– so-called “stress tests” – but made a
number of recommendations to the
national regulator.
A European Nuclear Safety Regulators
Group (Ensreg) peer review gave
News
atw Vol. 63 (2018) | Issue 8/9 ı August/September
the Ostrovets nuclear power plant,
which is close to the Lithuanian
border, an “overall positive” review,
following a site investigation that took
place in March.
The stress tests are meant to ensure
nuclear power plants comply with
strict criteria established by the International
Atomic Energy Agency and
were established by the European
Commission and Ensreg as a direct
reaction to the earthquake and
tsunami that caused the shutdown of
the Fukushima-Daiichi nuclear station
in Japan in March 2011.
The peer review team, which
reviewed an earlier stress test report
prepared by Belarus, comprised of 17
members, two representatives from
the EC and three observers: one from
the IAEA, one from Russia and one
from Iran.
The team praised the Belarusian
authorities for complying with the
review, even though Belarus had no
obligation to do so because it is not an
EU member state.
Following the Fukushima-Daiichi
accident, the EU carried out stress
tests of all its nuclear power plants
and also invited interested non-EU
countries to take part in the exercise.
In a detailed report, Ensreg
addressed three main areas: the site’s
resilience to extreme natural events
like earthquakes and flooding; the
capacity of the plant to respond to
electric power outages and loss of
heat sink; and severe accident
management.
According to the findings, the site
is resistant to earthquakes, flooding
and extreme weather, although the
investigators warned that seismic data
was not fully available and called on
the regulator to make sure run-off water
cannot enter safety-related buildings.
There are two 1,109-MW Russian
VVER-1200 reactor units under construction
at the Belarusian nuclear
station. Construction of Unit 1 began
in November 2013 and of Unit 2 in
April 2014.
The final peer review report is
online: https://bit.ly/2NnOixf
| | europa.eu, www.ensreg.eu,
www.dsae.by
Japan: Approval of energy
plan paves way for reactor
restarts
(nucnet) Nuclear reactor restarts in
Japan have become more likely after
the government approved an energy
plan today confirming that nuclear
power will remain a key component of
Japan’s energy strategy.
The plan, known as the Basic
Energy Plan, calls for a nuclear
share of around 20-22% by 2030. The
nuclear industry group, the Japan
Atomc Industrial Forum (Jaif) has
said about 30 reactors must be
brought back online to meet the
target.
Japan shut down all 42 com mercial
nuclear reactors after the Fukushima-
Daiichi accident. According to the
International Atomic Energy Agency,
the country’s nuclear share in 2017
was about 3.6%. Before Fukushima,
Japan generated about 30% of its
electricity from nuclear and planned
to increase that to 40%
Nine units have been restarted in
Japan since the Fukushima accident.
They are: Ohi-3, Ohi-4, Genkai-3,
Genkai-4, Sendai-1, Sendai-2, Ikata-3,
Takahama-3 and Takahama-4.
The energy plan also strengthens
the government’s commitment to
giving renewables such as solar and
wind power a major role in energy
generation.
The plan, which charts the nation’s
mid- and long-term energy policy,
marks the fifth in a series that is
required by law to be reviewed about
every three years.
The plan also maintains a reliance
on coal-fired thermal power as a
485
NEWS
| | Editorial Advisory Board
Frank Apel
Erik Baumann
Dr. Maarten Becker
Dr. Erwin Fischer
Carsten George
Eckehard Göring
Florian Gremme
Dr. Ralf Güldner
Carsten Haferkamp
Dr. Petra-Britt Hoffmann
Christian Jurianz
Dr. Guido Knott
Prof. Dr. Marco K. Koch
Dr. Willibald Kohlpaintner
Ulf Kutscher
Herbert Lenz
Jan-Christian Lewitz
Andreas Loeb
Dr. Thomas Mull
Dr. Ingo Neuhaus
Dr. Joachim Ohnemus
Prof. Dr. Winfried Petry
Dr. Tatiana Salnikova
Dr. Andreas Schaffrath
Dr. Jens Schröder
Norbert Schröder
Prof. Dr. Jörg Starflinger
Prof. Dr. Bruno Thomauske
Dr. Brigitte Trolldenier
Dr. Walter Tromm
Dr. Hans-Georg Willschütz
Dr. Hannes Wimmer
Ernst Michael Züfle
Imprint
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Im Tal 121, 45529 Hattingen, Germany
Phone: +49 2324 4397723
Fax: +49 2324 4397724
E-mail: editorial@nucmag.com
| | Official Journal of Kerntechnische Gesellschaft e. V. (KTG)
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News
atw Vol. 63 (2018) | Issue 8/9 ı August/September
486
NEWS
baseload energy source despite high
emissions of carbon dioxide.
The administration of prime
minister Shinzo Abe decided to promote
nuclear energy when it revised
the plan in 2014, reversing the policy
of the previous government led by
the then-Democratic Party of Japan,
which pledged to phase out nuclear
power by 2039 in the face of public
concern over safety.
Under the latest plan, the ratio of
nuclear energy, renewables and coal
thermal power in the nation’s overall
energy as of fiscal 2030 will remain at
20-22%, 22-24% and 26%, respectively,
in line with the government’s
target set three years ago.
The plan doe not make any
mention of the need for building new
nuclear plants.
However, it re-endorses using the
nuclear fuel cycle, in which plutonium
extracted from spent nuclear fuel at
nuclear plants is used to generate
power.
But the plan, noting calls from the
US, says that Japan will make efforts
to cut its stockpile of plutonium,
which can be used in making nuclear
weapons.
Japan holds about 47 tonnes of
plutonium, a source of criticism from
the US and other countries. Spent
nuclear fuel containing plutonium
from nuclear power plants in Japan
is sent to the UK and France for
reprocessing and eventual fabrication
into uranium-plutonium mixed oxide
(MOX) fuel before being returned to
Japan.
| | www.japan.go.jp
Reactors
Kola-1 becomes first Russian
nuclear plant to get operating
extension
(rosatom, nucnet) Russia’s state
nuclear operator Rosenergoatom has
been granted a licence by regulator
Rostekhnadzor to operate the Kola-1
nuclear power unit in the north of the
country for an additional 15 years
until 2033.
In a statement on its website, state
nuclear corporation Rosatom said this
is the first time a nuclear power plant
in Russia has been given such an
extension.
In April 2018 Rosenergoatom said
it had begun an extensive refurbishment
and modernisation programme
at Kola-1, a 411-MW VVER which
Operating Results March 2018
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 910 880 743 670 174 1 971 306 256 625 492 100.00 100.00 98.15 99.28 99.12 100.34
OL2 Olkiluoto BWR FI 910 880 743 687 017 1 996 275 246 295 456 100.00 100.00 99.88 99.88 100.51 100.50
KCB Borssele PWR NL 512 484 743 381 342 1 107 719 159 314 638 99.81 99.83 99.81 99.83 100.59 100.54
KKB 1 Beznau 1,2,7) PWR CH 380 365 296 108 560 108 560 124 854 647 39.84 13.71 38.07 13.10 37.85 13.03
KKB 2 Beznau 7) PWR CH 380 365 743 285 428 829 214 131 994 087 100.00 100.00 100.00 100.00 101.14 101.09
KKG Gösgen 7) PWR CH 1060 1010 743 793 650 2 308 759 307 503 346 100.00 100.00 99.98 99.98 100.77 100.88
KKM Mühleberg BWR CH 390 373 724 276 060 823 500 125 161 645 97.44 99.12 96.02 98.58 95.27 97.80
CNT-I Trillo PWR ES 1066 1003 743 775 921 2 278 694 241 303 118 100.00 100.00 99.94 99.98 97.56 98.60
Dukovany B1 PWR CZ 500 473 743 371 963 1 083 856 109 714 339 100.00 100.00 99.97 99.95 100.13 100.40
Dukovany B2 1,2) PWR CZ 500 473 209 102 246 747 959 105 370 496 28.13 69.38 27.75 68.97 27.52 69.29
Dukovany B3 PWR CZ 500 473 743 369 915 1 075 918 103 698 345 100.00 100.00 100.00 100.00 99.57 99.67
Dukovany B4 PWR CZ 500 473 743 372 191 1 080 365 104 352 106 100.00 100.00 100.00 100.00 100.19 100.08
Temelin B1 PWR CZ 1080 1030 721 772 094 772 094 107 253 388 97.04 33.40 95.25 32.78 96.22 33.11
Temelin B2 PWR CZ 1080 1030 743 813 415 2 356 815 103 846 761 100.00 100.00 100.00 100.00 101.37 101.08
Doel 1 PWR BE 454 433 743 338 988 984 072 135 198 820 100.00 100.00 99.98 99.99 100.59 100.39
Doel 2 PWR BE 454 433 743 337 020 984 599 133 236 867 100.00 100.00 99.15 99.61 99.79 100.28
Doel 3 3) PWR BE 1056 1006 0 0 0 251 169 221 0 0 0 0 0 0
Doel 4 PWR BE 1084 1033 743 817 209 2 371 746 256 917 588 100.00 100.00 100.00 100.00 100.48 100.33
Tihange 1 PWR BE 1009 962 726 739 606 2 191 381 293 030 257 97.66 99.20 96.92 98.94 99.17 101.03
Tihange 2 PWR BE 1055 1008 743 794 003 2 296 556 251 246 094 100.00 100.00 100.00 99.68 101.96 101.44
Tihange 3 PWR BE 1089 1038 722 784 314 2 332 443 271 227 273 97.15 99.02 96.80 98.90 96.84 99.13
Operating Results May 2018
Plant name
Type
Nominal
capacity
gross
[MW]
net
[MW]
Operating
time
generator
[h]
Energy generated, gross
[MWh]
Time availability
[%]
Energy availability Energy utilisation
[%] *) [%] *)
Month Year Since Month Year Month Year Month Year
commissioning
KBR Brokdorf 1,2) DWR 1480 1410 641 848 394 3 707 680 343 899 739 86.10 77.42 79.28 72.34 76.74 68.88
KKE Emsland 2,4) DWR 1406 1335 596 777 929 4 790 392 340 113 675 80.09 95.91 79.82 95.86 74.19 94.04
KWG Grohnde DWR 1430 1360 744 1 001 747 4 022 931 370 650 510 100.00 82.75 99.53 80.46 93.53 77.17
KRB C Gundremmingen 1) SWR 1344 1288 136 153 154 3 540 364 324 120 256 18.31 76.80 15.43 75.98 15.18 72.25
KKI-2 Isar DWR 1485 1410 744 1 067 384 5 291 198 346 889 521 100.00 100.00 100.00 99.99 96.24 98.06
KKP-2 Philippsburg 1,2,4) DWR 1468 1402 256 300 335 4 349 845 359 517 361 34.41 86.53 33.94 86.36 26.80 80.44
GKN-II Neckarwestheim DWR 1400 1310 744 1 011 600 4 974 300 325 097 434 100.00 100.00 100.00 99.87 97.29 98.32
News
atw Vol. 63 (2018) | Issue 8/9 ı August/September
began commercial operation in
December 1973.
The work was scheduled to take
about six months, Rosatom said at the
time.
The Kola station, 200 km south of
the city of Murmansk on the shore of
Imandra Lake, generates about 60%
of electricity in the Murmansk region,
Rosatom said.
All four units at Kola are Sovietdesigned
pressurised water reactors.
Units 1 and 2, of the older V-230
model, began commercial operation
in the mid-1970s and Units 3 and 4, of
the newer V-213 model, in the
mid-1980s.
| | www.rosatom.ru
Company News
USA: Framatome completes
major refurbishment of 31
reactor coolant pump motors
(framatome) Framatome recently
completed the refurbishment of 31
reactor coolant pump motors for
three southeastern nuclear energy
facilities. From 2002 to May 2018, the
company modified and upgraded
these components, which resulted
in a 100 percent reliability and
zero- failure performance record since
being re-installed.
The motors in reactor coolant
pumps help move coolant around the
primary circuit of a nuclear reactor
core. This keeps the reactor from overheating
while ensuring the safe heat
transfer from a reactor core to steam
generators.
“The success of this refurbishment
campaign is a tribute to Framatome’s
dedicated and experienced employees,”
said Craig Ranson, senior vice president
of the Installed Base Business Unit at
Framatome in North America. “Their
unmatched expertise, bolstered by
access to world-class facilities, allows
us to provide our customers with solutions
that, in many cases, are
more innovative and cost effective
than their plant’s original equipment
manufacturer.”
Members of Framatome’s Installed
Base services team worked with the
plants’ personnel to remove each
motor. They then brought the motors
to the company’s 70,000-square-foot
Pump and Motor Service Center in
Lynchburg, Virginia. While at the
center, experts inspected the components,
completed necessary repairs
and replacements, and tested each
motor. Such refurbishments allow
these components, and thus their
nuclear facilities, to operate safely and
reliably for longer durations.
Following successful testing, pump
and motor specialists re-installed the
motors and assessed their performance
on-site.
| | www.framatome.com
URENCO to supply EDF with
new uranium enrichment
services
(urenco) URENCO and EDF have
signed a new enrichment contract to
serve EDF’s French reactor fleet.
The high value and long-term
contract supports the recycling of
nuclear fuel by enriching uranium
recovered from fuel which has been
previously used and reprocessed.
The technical complexities of
enriching this material will involve
expertise from across URENCO and
upgrading our facilities.
Dominic Kieran, URENCO’s Chief
Commercial Officer, said: “URENCO is
proud to be part of EDF’s endeavour to
recycle spent nuclear fuel. It is a
significant step in further proving the
sustainability of nuclear energy and a
testimony to URENCO’s technical
capabilities.”
| | www.urenco.com
Full core of Westinghouse fuel
achieved at South-Ukraine
nuclear power plant unit 3
(westinghouse) Westinghouse Electric
Company announced that Ukraine’s
State Enterprise National Nuclear
Energy Generation Company (SE
*)
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
487
NEWS
Operating Results April 2018
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 910 880 720 651 030 2 622 336 257 276 522 100.00 100.00 98.66 99.13 99.36 100.09
OL2 Olkiluoto BWR FI 910 880 522 480 632 2 476 907 246 776 088 72.50 93.12 72.18 92.95 72.56 93.51
KCB Borssele PWR NL 512 484 720 361 216 1 468 935 159 675 854 97.83 99.33 97.81 99.33 98.13 99.94
KKB 1 Beznau 1,2,7) PWR CH 380 365 720 276 656 385 216 125 131 303 100.00 35.29 100.00 34.83 101.19 35.07
KKB 2 Beznau 7) PWR CH 380 365 720 275 430 1 104 644 132 269 517 100.00 100.00 100.00 100.00 100.72 101.00
KKG Gösgen 7) PWR CH 1060 1010 720 759 700 3 068 459 308 263 046 100.00 100.00 99.91 99.96 99.54 100.55
KKM Mühleberg BWR CH 390 373 720 277 490 1 100 990 125 439 135 100.00 99.34 99.89 98.91 98.82 98.06
CNT-I Trillo PWR ES 1066 1003 720 762 241 3 040 935 242 065 359 100.00 100.00 100.00 99.98 98.85 98.66
Dukovany B1 PWR CZ 500 473 720 357 871 1 441 727 110 072 210 100.00 100.00 99.43 99.82 99.41 100.15
Dukovany B2 1,2) PWR CZ 500 473 0 0 747 959 105 370 496 0 52.03 0 51.72 0 51.96
Dukovany B3 PWR CZ 500 473 529 258 528 1 334 447 103 956 874 73.47 93.37 72.49 93.12 71.81 92.70
Dukovany B4 PWR CZ 500 473 496 240 156 1 320 521 104 592 262 68.89 92.22 66.94 91.73 66.71 91.73
Temelin B1 PWR CZ 1080 1030 720 777 874 1 549 968 108 031 262 100.00 50.05 99.96 49.60 99.85 49.83
Temelin B2 PWR CZ 1080 1030 720 783 901 3 140 716 104 630 662 100.00 100.00 100.00 100.00 100.81 101.01
Doel 1 PWR BE 454 433 541 245 643 1 229 715 135 444 462 75.19 93.80 75.00 93.74 75.06 94.05
Doel 2 PWR BE 454 433 720 328 895 1 313 494 133 565 762 100.00 100.00 99.98 99.70 100.44 100.32
Doel 3 3) PWR BE 1056 1006 0 0 0 251 169 221 0 0 0 0 0 0
Doel 4 PWR BE 1084 1033 720 787 926 3 159 672 257 705 513 100.00 100.00 99.86 99.97 99.90 100.23
Tihange 1 PWR BE 1009 962 720 732 505 2 923 886 293 762 762 100.00 99.40 99.94 99.19 101.25 101.09
Tihange 2 PWR BE 1055 1008 720 753 721 3 050 277 251 999 815 100.00 100.00 98.60 99.41 99.80 101.03
Tihange 3 PWR BE 1089 1038 0 0 2 332 443 271 227 273 0 74.25 0 74.16 0 74.34
News
atw Vol. 63 (2018) | Issue 8/9 ı August/September
488
NEWS
NNEGC) Energoatom’s South-Ukraine
NPP Unit 3 near Yuzhnoukrainsk
in Mykolaiv province was loaded with
a full core of Westinghouse VVER-1000
fuel. This is the first unit in Ukraine to
operate with Westinghouse VVER-
1000 fuel assemblies as the sole fuel
source.
“Westinghouse began supplying
fuel to Ukraine in 2005, when the first
lead test assemblies were delivered to
South-Ukraine Unit 3,” said Aziz Dag,
vice president and managing director,
Northern Europe. “We are proud
to continue supporting Ukraine
with their energy diversification by
supplying a full core of Westinghouse
VVER-1000 fuel to our customer,
Energoatom.”
Westinghouse currently supplies
fuel to six of Ukraine’s 15 nuclear
power reactors. Beginning in 2021,
the number of reactors with Westinghouse
fuel will increase to seven.
“Westinghouse has made significant
investments over the last several
years in order to further enhance our
fuel delivery support to Energoatom,”
said Michele DeWitt, senior vice
president, Nuclear Fuel. “We have
dedicated production lines for
VVER-1000 fuel and stand ready to
supply fuel for further contract
expansions.”
The nuclear fuel delivered by
Westinghouse is manufactured in its
fuel fabrication facility in Västerås,
Sweden. Nuclear power continues to
be an important energy source for the
country of Ukraine, accounting for
approximately 50% of its electricity
production.
| | www.westinghousenuclear.com
Forum
GRS-IRSN Workshop zu
Sicherheitskriterien von
Brennelementen
(grs) In den vergangenen Jahren
wurde in Frankreich aufgrund neuerer
experimenteller Erkenntnisse das
kerntechnische Regelwerk hinsichtlich
der Sicherheitskriterien für
Brennelemente und deren Verhalten
bei Betrieb und in Störfällen überarbeitet
und aktualisiert. Da ähnliche
Fragestellungen in der Vergangenheit
auch Thema in Deutschland waren
und zu Regelwerksänderungen geführt
hatten, veranstalteten die
Gesellschaft für Anlagen- und Reaktorsicherheit
(GRS) gGmbH und das
Institut de Radioprotection et de
Súreté Nucléaire (IRSN) einen
gemein samen Workshop zum Thema
„Fuel Safety Criteria“, welcher am
20./21. Juni 2018 in Paris in den
Räumen des IRSN stattfand. Neben
Experten der GRS und des IRSN nahmen
Vertreter aus Belgien, Tschechien
und Litauen, sowie der deutschen Reaktorsicherheitskommission
und des
Betreibers PreussenElektra an der Veranstaltung
teil.
In fünf Sitzungen wurden Informationen
und Erfahrungen zu Sicherheitskriterien
und zugehörigen
Nachweisverfahren hinsichtlich betrieblicher
und störfallbedingter
Phänomene wie Hüllrohrkorrosion,
-oxidation, Wasserstoffversprödung,
Reaktivitäts- und Kühlmittelverluststörfälle,
mechanische Pellet-Hüllrohr-
Wechselwirkungen (Pellet Cladding
Mechanical Interaction, PCMI),
Brennstoff-Verlagerung und -Auswurf
bei Hochabbrand sowie Brennelementverbiegungen
ausgetauscht.
Es wurde deutlich, dass beide Länder
trotz mitunter unterschiedlicher
Sicherheitsphilosophien, regulatorischer
Anforderungen und Brennelement-Ausführungen
mit weitgehend
übereinstimmenden Problemstellungen
konfrontiert waren und
entsprechende Änderungen in den
ihren einschlägigen Regelwerken
umgesetzt haben. Der Workshop ist
daher als Startpunkt für ein gemeinsames
Verständnis brennstoffbezogener
Sicherheitskriterien und
zugehöriger Nachweisverfahren zu
verstehen. Weitere Detail-Diskussionen
zu ausgewählten Teilaspekten
sind geplant. Das nächste Expertentreffen
in diesem Themenfeld wird
voraussichtlich in Berlin stattfinden.
| | www.grs.de
People
Dipl.-Ing. Christoph Michel
wird zum 1. Januar 2019
Nachfolger von Dr.-Ing. Hans
Fechner als Sprecher der
Geschäftsführung der
Siempelkamp Gruppe
(siempelkamp) Mit Wirkung zum
1. August 2018 ist Dipl.-Ing. Christoph
Michel zum weiteren Mitglied der
Geschäftsführung der G. Siempelkamp
GmbH & Co. KG bestellt worden.
Er wird ab dem 1. Januar 2019 als
Sprecher der Geschäftsführung der
Siempelkamp Gruppe die Nachfolge
von Dr.-Ing. Hans Fechner übernehmen,
der nach vielen Jahren
erfolgreicher Tätigkeit in den Ruhestand
geht.
Christoph Michel hat Luft- und
Raumfahrttechnik an der Universität
Stuttgart studiert und später berufsbegleitend
einen MBA an der Duke
University, USA abgeschlossen. Er
blickt auf eine 18-jährige erfolgreiche
Karriere im Maschinen- und Großanlagenbau
zurück.
| | www.siempelkamp.com
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)].
2014
• Uranium: 28.10–42.00
• Conversion: 7.25–11.00
• Separative work: 86.00–98.00
2015
• Uranium: 35.00–39.75
• Conversion: 6.25–9.50
• Separative work: 58.00–92.00
2016
• Uranium: 18.75–35.25
• Conversion: 5.50–6.75
• Separative work: 47.00–62.00
2017
• Uranium: 19.25–26.50
• Conversion: 4.50–6.75
• Separative work: 39.00–50.00
2018
January 2018
• Uranium: 21.75–24.00
• Conversion: 6.00–7.00
• Separative work: 38.00–42.00
February 2018
• Uranium: 21.25–22.50
• Conversion: 6.25–7.25
• Separative work: 37.00–40.00
March 2018
• Uranium: 20.50–22.25
• Conversion: 6.50–7.50
• Separative work: 36.00–39.00
April 2018
• Uranium: 20.00–21.75
• Conversion: 7.50–8.50
• Separative work: 36.00–39.00
May 2018
• Uranium: 21.75–22.80
• Conversion: 8.00–8.75
• Separative work: 36.00–39.00
June 2018
• Uranium: 22.50–23.75
News
atw Vol. 63 (2018) | Issue 8/9 ı August/September
• Conversion: 8.50–9.50
• Separative work: 35.00–38.00
| | 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
2017: 91.28, 25,739,010
| | Uranium spot market prices from 1980 to 2018 and from 2008 to 2018. The price range is shown.
In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.
489
NEWS
2018
I. quarter: 89.88; 5.838.003
| | Source: BAFA, some data provisional
www.bafa.de
EEX Trading Results
June 2018
(eex) In June 2018, the European
Energy Exchange (EEX) increased
volumes on its power derivatives
markets by by 28% to 231.1 TWh
(June 2017: 181.2 TWh). On the
Dutch power market, volumes increased
by 141% to 3.2 TWh (June
2017: 1.3 TWh). EEX achieved strong
double-digit growth in the markets for
France (22.0 TWh, +22%), Italy
(44.5 TWh, +46%) as well as in
power options (9.3 TWh, +45%).
Volumes in Phelix-DE Futures increased
to 132.7 TWh.
On the EEX markets for emission
allowances, the total trading volume
almost tripled to 297.4 million tonnes
of CO 2 in June (June 2017:
105.1 million tonnes of CO 2 ). On the
EUA secondary market (including
options), volumes increased sixfold to
217.8 million tonnes of CO 2 (June
2017: 30.6 million tonnes of CO 2 ).
Primary market auctions contributed
79.6 million tonnes of CO 2 to the total
volume.
The Settlement Price for base load
contract (Phelix Futures) with
delivery in 2019 amounted to
43.14 €/MWh. The Settlement
Price for peak load contract (Phelix
Futures) with delivery in 2019
amounted to 53.55 €/MWh.
The EUA price with delivery in
December 2018 amounted to
14.24/16.14 €/ EUA (min./max.).
July 2018
(eex) In July 2018, the European
Energy Exchange (EEX) increased
volumes on its power derivatives
| | Separative work and conversion market price ranges from 2008 to 2018. 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.
markets by 46% to 213.8 TWh (July
2017: 146,2 TWh). On the Spanish
power market, volumes exceeded
the mark of 10 TWh for the first time,
doubling last year’s volume
(10.6 TWh, July 2017: 4.3 TWh).
Furthermore, the markets for France
(18.2 TWh, +18%) and Italy
(37.3 TWh, +70%), in particular,
developed positively. In Phelix-DE
Futures, trading volumes amounted
to 128.7 TWh which is clearly above
the total July volume in 2017 in the
products for the German market
( Phelix-DE and Phelix-DE/AT in July
2017: 98.1 TWh).
The Settlement Price for base load
contract (Phelix Futures) with
delivery in 2019 amounted to
43.79 €/MWh. The Settlement Price
for peak load contract (Phelix
Futures) with delivery in 2019
amounted to 53.93 €/MWh.
The EUA price with delivery
in December 2018 amounted to
15.08/17.40 €/ EUA (min./max.).
| | www.eex.com
MWV Crude Oil/Product Prices
May 2018
(mwv) According to information and
calculations by the Association of the
German Petroleum Industry MWV e.V.
in May 2018 the prices for super fuel,
fuel oil and heating oil noted higher
compared with the pre vious month
April 2018. The average gas station
prices for Euro super consisted of
145.62 €Cent ( April 2018:
138.96 €Cent, approx. +6.6 % in
brackets: each information for previous
month or rather previous month
comparison), for diesel fuel of
126.22 €Cent (121.09; +5.13 %) and
for heating oil (HEL) of 67.93 €Cent
(63.12 €Cent, +4.81 %).
Worldwide crude oil prices
(monthly average price OPEC/Brent/
WTI, Source: U.S. EIA) were higher,
approx. +4.21 % (+7.39 %) in May
2018 compared to April 2018.
The market showed a stable
development with slightly higher
prices; each in US-$/bbl: OPEC
basket: 73.22 (68.43); UK-Brent:
74.40 (72.11); West Texas Intermediate
(WTI): 67.87 (66.25).
June 2018
In June 2018 the prices for super
fuel, fuel oil and heating oil noted
inconsistent compared with the
pre vious month May 2018. The
average gas station prices for Euro
super consisted of 147.60 €Cent (May
2018: 145.62 €Cent, approx. +1.98 %
in brackets: each information for previous
month or rather previous month
comparison), for diesel fuel of
129.41 €Cent (126.2; +3.19 %) and
for heating oil (HEL) of 67.67 €Cent
(67.937 €Cent, -0.38 %).
| | www.mwv.de
News
atw Vol. 63 (2018) | Issue 8/9 ı August/September
Why do We Allow Nuclear to Take the ‘Silly Season’ Media Heat?
490
NUCLEAR TODAY
John Shepherd is a
UK-based energy
writer and editor- inchief
of Energy
Storage Publishing.
Links to reference
sources:
ICL briefing paper:
https://bit.ly/
2vkIM6Y
Rick Perry’s remarks:
https://bit.ly/
2Kz3WTB
GMB union statement
on UK nuclear:
https://bit.ly/2OPk3jd
John Shepherd
As I started to write this article, we were approaching the end of what is often referred to in the UK as the ‘silly season’
– the main summer holiday period when hard news is hard to come by.
The time of year always means all manner of weird and
wonderful stories finding their way into newspapers and
broadcast news, stories that would most probably never
see the light of day outside the silly season. This year
has been slightly different, because the lengthy spell of
hot weather that many of us across Europe experienced
generated much of the journalistic ‘heat’.
But for the nuclear industry, the hot spell fanned the
media flames of an old anti-nuclear favourite, as it became
clear operations at some nuclear power plants were being
halted temporarily to comply with restrictions that prevent
cooling water further heating local rivers and waterways.
Some media outlets preferred the alarmist over the
factual. I was dismayed to hear one BBC report claim “one
ageing (European) nuclear power plant” had been “forced”
by the heat wave to cut back on production “to keep vital
equipment cool”. That statement was misleading – albeit
probably more out of ignorance than malice.
I don’t recall hearing from any of our industry representatives
early on in the summer, communicating the
facts on the cooling issue to the public and journalists. If
there was no industry-wide effort on this there should
have been. It’s not a new situation for our industry and
every opportunity should be taken to head off misinformation
that experience tells us is just around the
corner. PR directors should be making a note in their
diaries for next year just in case – because forewarned is
forearmed.
But there was more refreshing news out of the UK over
the summer in the form of a briefing paper by researchers
at Imperial College London (ICL). According to the paper by
ICL’s Grantham Institute – Climate Change and the
Environment, nuclear power “will be essential for meeting
the UK’s greenhouse gas emissions reduction target, unless
we can adapt to depend largely on variable wind and solar,
or there is a breakthrough in the commercialisation of
carbon capture and storage”.
The paper acknowledged the difficulties involved in
attracting private investment to build new nuclear projects,
but said the UK government’s decision to procure the
3.2 gigawatt Hinkley Point C nuclear plant “represents a
crucial opportunity for the conventional nuclear industry,
which is under significant financial stress, to rebuild itself”.
There certainly does appear to be a new realism in the
UK about the urgent need to turn talk about investments in
nuclear into real action. One of the country’s major trade
unions, the GMB, put new nuclear firmly on the agenda.
The union was quick to respond to reports that the UK’s
planned Moorside nuclear plant in Cumbria, northwest
England, could be scrapped unless a buyer is found.
Moorside is being developed by NuGen, which is owned
by Toshiba. NuGen has been put up for sale as Toshiba
restructures its operations in the aftermath of financial
issues triggered by losses in its US nuclear business,
Westinghouse. The three AP1000 reactor units proposed for
Moorside were to have come from Westinghouse.
Now the GMB has reiterated its call for the UK government
to take a stake in the financing of the Moorside
project, “rather than leaving this vital project at the mercy
of foreign companies”.
GMB national secretary Justin Bowden said: “As well
as eradicating the uncertainty, by the government taking a
stake and taking control at Moorside, the price to consumers
will be greatly reduced making good all round
sense, not just the obvious benefits to bill payers but
because the government is ‘the lender of last resort’ when
it comes to guaranteeing the country’s energy supply and
so direct public funding of the construction does away
with the nonsensical pretence that this is some other
country or company’s responsibility.”
And the union cautioned the UK against an over reliance
on renewables in energy policy. According to the GMB, “for
the 12 months from 7 March 2017, every one in 5.6 days
was a low wind day (65 days in total) when the output of
the installed and connected wind turbines in the UK
produced less than 10 % of their installed and connected
capacity for more than half of the day”.
“For 341 days in the year, solar output was below 10 %
of installed capacity for more than half of the day,” the
union said.
Such championing of public investment in nuclear from
the union is welcome as the UK struggles to advance its
civil nuclear ambitions.
However, it’s a different story for one of the world’s
nuclear newcomer nations – the United Arab Emirates –
where nuclear development continues apace. In August,
the Emirates Nuclear Energy Corporation (ENEC) announced
the successful completion of hot functional testing
at unit 2 of the Barakah nuclear plant, which is under
construction around 240 kilometres west of Abu Dhabi.
ENEC said that as of June 2018, the construction progress
rate of unit 2 was 93 % and overall construction progress
rate for the four Barakah units is now more than 89 %.
Meanwhile, in the US, energy secretary Rick Perry made
his first visit to a nuclear power plant since his appointment
16 months earlier. Speaking at the James A FitzPatrick
plant, Perry gave a ringing endorsement of nuclear on
behalf of the Trump administration.
Perry said: “Nuclear provides approximately 20 % of
the electricity generated in the United States. It is one of
our most reliable sources of baseload power, and it is also
one of our cleanest sources of power, providing about 60 %
of our carbon-free energy output.”
And a day after Perry’s visit, the Department of Energy
announced $ 36.4 million (€ 31.5 m) in funding for 37
research awards at universities, national laboratories, and
private industry on a range of topics in fusion energy
sciences. The Department said the research “is designed to
help lay the groundwork for the development of nuclear
fusion as a future practical energy source”.
Investment in nuclear construction and research should
be welcomed wherever it comes and our industry should
not be afraid to campaign for public investment. The
renewables lobby has been doing this successfully for some
time. Nuclear should not shy away from speaking up too.
Author
John Shepherd
Shepherd Communications
3 Brooklands
West Sussex
BN43 5FE
Nuclear Today
Why do We Allow Nuclear to Take the ‘Silly Season’ Media Heat? ı John Shepherd
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Seminarinhalte
Genehmigungen für die Stilllegung und den Rückbau
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Gestaltung der Übergänge zwischen der Betriebs- und der Stilllegungsgenehmigung
Gestaltung der Übergänge zwischen Genehmigungs- und Aufsichtsphase
Rechtliche Gestaltung des Genehmigungsverfahrens
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ı
Vorstellung von Reststoffkonzepten
Bewertung der Reststoffkonzepte
Neue Regelungen zum Übergang der Entsorgungsverantwortung
Entsorgungsfragen rund um Abfälle aus dem Rückbau
Zielgruppe
Die 2-tägige Schulung wendet sich an Fach- und Führungskräfte, Mitarbeiterinnen und Mitarbeiter
von Betreibern, Industrie und Dienstleistern, die sich mit der Thematik aktuell bereits beschäftigen
oder sich künftig damit auseinander setzen werden.
Maximale Teilnehmerzahl: 12 Personen
Referenten
Dr. Matthias Bauerfeind
Dr. Christian Raetzke
Wir freuen uns auf Ihre Teilnahme!
ı Abteilung Stilllegung, Entsorgung, Reaktorphysik, TÜV SÜD Energietechnik
GmbH Baden-Württemberg
ı Rechtsanwalt, Leipzig
Bei Fragen zur Anmeldung rufen Sie uns bitte an oder senden uns eine E-Mail.
Termin
2 Tage
24. bis 25. September 2018
Tag 1: 10:30 bis 17:45 Uhr
Tag 2: 09:00 bis 16:45 Uhr
Veranstaltungsort
Geschäftsstelle der INFORUM
Robert-Koch-Platz 4
10115 Berlin
Teilnahmegebühr
1.598,– € ı zzgl. 19 % USt.
Im Preis inbegriffen sind:
ı Seminarunterlagen
ı Teilnahmebescheinigung
ı Pausenverpflegung
inkl. Mittagessen
Kontakt
INFORUM
Verlags- und Verwaltungsgesellschaft
mbH
Robert-Koch-Platz 4
10115 Berlin
Petra Dinter-Tumtzak
Fon +49 30 498555-30
Fax +49 30 498555-18
seminare@kernenergie.de
Media Partner
www.nucleartech-meeting.com
Save the Date
7 – 8 May 2019
Estrel Convention Center Berlin, Germany
Key Topics
Outstanding Know-How & Sustainable Innovations
Enhanced Safety & Operation Excellence
Decommissioning Experience & Waste Management Solutions
The International Expert Conference on Nuclear Technology