atw 2018-09v3

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

SCHWER WEGZUDENKEN 

als Ihr Partner 

für Schwergutlogistik 

und KTA-Hubgerüste 

August Alborn GmbH & Co. KG 

Planetenfeldstraße 102 

44379 Dortmund-Marten 

Telefon +49(0)231 961080-0 

Telefax +49(0)231 961080-9 

alborn@alborn.de 

www.alborn.de 

WWW.ALBORN.DE

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


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 


Maria Poetsch 


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 

INFORUM Verlags- und Verwaltungs gesellschaft mbH ı Robert-Koch-Platz 4 ı 10115 Berlin 

Petra Dinter-Tumtzak ı Fon +49 30 498555-30 ı Fax +49 30 498555-18 ı seminare@kernenergie.de 

Die INFORUM-Seminare können je nach 

Inhalt ggf. als Beitrag zur Aktualisierung 

der Fachkunde geeignet sein.


430 

Issue 8/9 

August/September 

CONTENTS 

437 


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 


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


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 


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


432 

ABSTRACTS | ENGLISH 


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. 


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. 


Neutron Flux Fluctuations in German PWR 


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. 


(AMNT) Key Topic | Outstanding Know-How 


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. 


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


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. 


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 


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


434 

INSIDE NUCLEAR WITH NUCNET 


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. 


A Stark Warning to Trump on China, Russia and the ‘Crisis’ Facing US Nuclear Industry ı NucNet, David Dalton


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 

Advertisement 

The International Expert Conference on Nuclear Technology 

Estrel Convention 

Center Berlin 

Germany 

Proceedings 

Key Topics 

Outstanding Know-How & 

Sustainable Innovations 

Enhanced Safety & 

Operation Excellence 

Decommissioning Experience & 

Waste Management Solutions 

All-in-one: Proceedings of AMNT 2018 

3 Get presentations of all sessions on disc 

Estrel Convention 

Center Berlin 

Germany 

www.nucleartech-meeting.com 

29 – 30 May 


Proceedings 

3 Order online 

89.– € * 

www.kernenergie.de/shop 

* Price excluding postage. Price including 19% VAT for Germany and all EU member states without VAT number. 

For EU member states with VAT number and all other countries VAT will not be added. For participants of the AMNT 2018 free of charge. 

AMNT2018 Proceedings 179x80.indd 1 


07.08.18 14:13 

A Stark Warning to Trump on China, Russia and the ‘Crisis’ Facing US Nuclear Industry ı NucNet, David Dalton


436 

CALENDAR 

Calendar 


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 


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 


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 


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



Preparations for Premiere in Full Swing 


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 


Akademik Lomonosov: Preparations for Premiere in Full Swing ı Roman Martinek


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. 




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 


Expert for Communication 

Czech Republic 

ENERGY POLICY, ECONOMY AND LAW 439 




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)). 


Nuclear Phase-out Last Act? Are the New Compensation Regulations for Frustrated Expenses in Accordance with the Constitution? ı Tobias Leidinger


| | 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 


Nuclear Phase-out Last Act? Are the New Compensation Regulations for Frustrated Expenses in Accordance with the Constitution? ı Tobias Leidinger


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


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 



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 



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 



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 


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 


Analyses of Possible Explanations for the Neutron Flux Fluctuations in German PWR ı Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov


| | 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. 





| | 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 




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. 






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. 

1. M. Seidl et al., Review of the historic 

neutron noise behavior in German 

GWU built PWRs, Progress in Nuclear 

Energy 85, pp 668-675, 2015. 

2. Reaktor-Sicherheitskommission, 

Stellungnahme DWR-Neutronenflussschwankungen, 

457. Sitzung vom 

11.04.2013. 

3. Bundesamt für Strahlenschutz, 

Kurzbeschreibung und Bewertung der 

meldepflichtigen Ereignisse in Kernkraftwerken 

und Forschungsreaktoren 

der Bundesrepublik Deutschland im 

Zeitraum Januar 2011, Stand 

14.12.2012. 




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 

Energy for New Europe, Portoroz, 

Slovenia, 2014. 

12. G. Verdú et al., Study of the Noise 

Propagation in PWR with Coupled 

Codes, International Conference on 

Mathematics and Computational 

Methods Applied to Nuclear Science 

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, 

1975. 

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 

instrumentation, Progress in Nuclear 

Energy 9, pp. 437–452, 1982. 

17. Hashemian et al., Sensor Response 

Time Monitoring Using Noise Analysis, 

Progress in Nuclear Energy 21, 

pp. 583-592, 1988. 

18. R. Sunder, Sammlung von Signalmustern 

zur DWR-Schwingungs überwachung 

– Informationsgehalt der 

Neutronenflussrauschsignale, 

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 

position on the ex-core neutron flux 

detection in a PWR, Journal of Nuclear 

Science and Technology 54(2), 

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 


Advertisement 

WEITERBILDUNGSKURS NUKLEARFORUM SCHWEIZ 

IT-SICHERHEIT IM ALLTAG – PRAXISWISSEN 

FÜR MITARBEITER IN DER NUKLEARTECHNIK 

Donnerstag, 22. November 2018, Trafo, Baden 

Der eintägige Weiterbildungskurs gibt einen Einblick in die Gesamt problematik 

der Cybersecurity und die vielfältigen Formen der Be drohung. Das an der 

betrieblichen Praxis orientierte Programm hilft, potenzielle Gefahrenstellen 

im Alltag zu erkennen und den Hinter grund betrieblicher Massnahmen zur 

IT-Sicherheit zu verstehen. 

Der Kurs richtet sich an Mitarbeitende aller Funktionsstufen im Bereich der 

Nuklear technik, die in ihrem Büroalltag computerbasierte Informations- und 

Kommunikations systeme benutzen. Besondere IT-Kenntnisse sind 

nicht erforderlich. 

Programm und Anmeldung unter www.nuklearforum.ch 

SwissNuklear_Weiterbildung_Inserat_166x115mm_20180730.indd 1 30.07.18 14:30 





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]. 


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


| | 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 


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. 





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) 




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 


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 ). 




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


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 


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


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 




| | 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 


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 






| | 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 




| | 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 


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 






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 




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


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]. 


Heuristic Methods in Modelling Research Reactors for Deterministic Safety Analysis ı Vera Koppers and Marco K. Koch


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 


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). 




466 


| | 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 




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 





References 

468 


| | 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. 




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 





469 



Development and Validation of a CFD Wash-Off Model for Fission Products on Containment Walls ı Katharina Amend and Markus Klein


470 


| | 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. 




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 


| | 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 . 




472 


| | 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. 




[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 


Institute for Numerical Methods in 

Aerospace Engineering Universität 

der Bundeswehr München 

Werner Heisenberg Weg 39 

85577 Neubiberg, Germany 

Young 

Scientists' 

Workshop 

Awarded 

473 


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 





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 


A Preliminary Conservative Criticality Assessment of Fukushima Unit 1 Debris Bed ı María Freiría López, Michael Buck and Jörg Starflinger


474 


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 




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 





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. 


| | 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 


Institute of Nuclear Technology 

and Energy Systems (IKE) 

University of Stuttgart 

Pfaffenwaldring 31 

70569 Stuttgart, Germany 




49 th Annual Meeting on Nuclear Technology (AMNT 2018) 

Key Topic | Outstanding Know-How 


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 


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 


Key Topic | Outstanding Know-How & Sustainable Innovations ı Christian Raetzke


478 


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 




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 

479 


Advertisement 




480 

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 

Report: GRS Workshop “Safety of Extended Dry Storage of Spent Nuclear Fuel” 

ı Klemens Hummelsheim, Florian Rowold and Maik Stuke


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 

481 

REPORT 


Report 



482 

REPORT 

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 




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 

Dr.-Ing. Hans-Georg Willschütz 

Sprecher KTG-Sektion NORD 

Thomas Fröhmel 

Stellv. Sprecher der KTG-Sektion NORD 

KTG INSIDE 

Herzlichen Glückwunsch! 

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


484 

NEWS 

Wenn Sie keine 

Erwähnung Ihres 

Geburtstages in 

der atw wünschen, 

teilen Sie dies bitte 

rechtzeitig der KTG- 

Geschäftsstelle mit. 

KTG Inside 

Verantwortlich 

für den Inhalt: 

Die Autoren. 

Lektorat: 

Natalija Cobanov, 

Kerntechnische 

Gesellschaft e. V. 

(KTG) 


10115 Berlin 

T: +49 30 498555-50 

F: +49 30 498555-51 

E-Mail: 

natalija.cobanov@ 

ktg.org 

www.ktg.org 

Oktober 2018 

91 Jahre | 1927 

23. Dr. Helmut Krause, Bad Herrenalb 

90 Jahre | 1928 

8. Dipl.-Ing. Rainer Rothe, Möhrendorf 

89 Jahre | 1929 

23. Prof. Dr. Helmut Karwat, 

Großhesselohe 

87 Jahre | 1931 

6. Dr. Edmund Ruppert, 

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, 


26. Dr. Knut Scheffler, Beckedorf 

79 Jahre | 1939 

5. Dipl.-Ing. Günter Langetepe, 

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, 


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


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 

| | Editorial 

Christopher Weßelmann (Editor in Chief) 

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) 

| | Publisher 

INFORUM Verlags- und Verwaltungsgesellschaft mbH 

Robert-Koch-Platz 4, 10115 Berlin, Germany 

Phone: +49 30 498555-30, Fax: +49 30 498555-18 

www.nucmag.com 

| | General Manager 

Christian Wößner, Berlin, Germany 

| | Advertising and Subscription 

Petra Dinter-Tumtzak 

Robert-Koch-Platz 4, 10115 Berlin, Germany 

Phone: +49 30 498555-30, Fax: +49 30 498555-18 

E-mail: petra.dinter@kernenergie.de 

| | Prize List for Advertisement 

Valid as of 1 January 2018 

Published monthly, 9 issues per year 

Germany: 

Per issue/copy (incl. VAT, excl. postage) 24.- € 

Annual subscription (incl. VAT and postage) 176.- € 

All EU member states without VAT number: 

Per issue/copy (incl. VAT, excl. postage) 24.- € 

Annual subscription (incl. VAT, excl. postage) 176.- € 

EU member states with VAT number 

and all other countries: 

Per issues/copy (no VAT, excl. postage) 22.43 € 

Annual subscription (no VAT, excl. postage) 164.49 € 

| | Copyright 

The journal and all papers and photos contained in it 

are protected by copyright. Any use made thereof outside 

the Copyright Act without the consent of the publisher, 

INFORUM Verlags- und Verwaltungsgesellschaft mbH, 

is prohibited. This applies to reproductions, translations, 

microfilming and the input and incorporation into 

electronic systems. The individual author is held 

responsible for the contents of the respective paper. 

Please address letters and manuscripts only to the 

Editorial Staff and not to individual persons of the 

association´s staff. We do not assume any responsibility 

for unrequested contributions. 

Signed articles do not necessarily represent the views 

of the editorial. 

| | Layout 

zi.zero Kommunikation 

Berlin, Germany 

Antje Zimmermann 

| | Printing 

inpuncto:asmuth 

druck + medien gmbh 

Baunscheidtstraße 11 

53113 Bonn 

|| 

| | 

ISSN 1431-5254 

News


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 


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 



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] 


[%] 

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


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 


[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Month Year Month Year 



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 B2 1,2) PWR CZ 500 473 0 0 747 959 105 370 496 0 52.03 0 51.72 0 51.96 





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


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 



2016 

• Uranium: 18.75–35.25 



2017 

• Uranium: 19.25–26.50 




January 2018 

• Uranium: 21.75–24.00 



February 2018 

• Uranium: 21.25–22.50 



March 2018 

• Uranium: 20.50–22.25 



April 2018 

• Uranium: 20.00–21.75 



May 2018 

• Uranium: 21.75–22.80 



June 2018 

• Uranium: 22.50–23.75 

News




| | 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 


I. quarter: 89.88; 5.838.003 

| | Source: BAFA, some data provisional 

www.bafa.de 

EEX Trading Results 


(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). 


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


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: 


2vkIM6Y 

Rick Perry’s remarks: 


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

Kommunikation und 

Training für Kerntechnik 

Wen ig e P l ätze f rei . 

Jetzt anmelden! 

Rückbau kerntechnischer Anlagen 

In Kooperation mit 

TÜV SÜD Energietechnik GmbH 

Baden-Württemberg 

Seminar: 

Stilllegung und Rückbau in Recht und Praxis 

Seminarinhalte 

Genehmigungen für die Stilllegung und den Rückbau 

ı 

ı 

Gestaltung der Übergänge zwischen der Betriebs- und der Stilllegungsgenehmigung 

Gestaltung der Übergänge zwischen Genehmigungs- und Aufsichtsphase 

Rechtliche Gestaltung des Genehmigungsverfahrens 

ı 

ı 

ı 

ı 

Elemente des Umweltbereiches 

UVP und Erörterungstermin 

Anfechtung von Genehmigungen 

Stellungnahmen der EU-Kommission 

Reststoffe und Abfälle 

ı 

ı 

ı 

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 


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 


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

atw 2018-09v3

Create successful ePaper yourself

Delete template?

Save as template?