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

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

Nuclear Energy: The Dead Live Longer or

the Summer of 2018

Dear Reader, Although nuclear energy offers both comprehensive technical potential with further development

prospects for use in power generation and attractive economic conditions, both for existing plants and for new plants –

assuming a reliable regulatory and political environment – there was no visible impetus for this for a long time.

Nuclear energy has also been or is facing serious market

challenges. There are two reasons why it cannot exploit its

economic advantages: On the one hand, there are hardly

any free electricity markets left; regulated markets with

subsidy systems, some of which are excessive and barely

manageable, prevent any market development towards

­efficient systems as a whole. On the other hand, plants

with long depreciation periods, as is the case with nuclear

energy at around 20 years, are not very attractive.

Remarkable developments in spring/summer 2018 set

clear signals for future impulses, especially with their

technical accents:

1. At the end of April 2018, the Akademik Lomonosov was

launched in St. Petersburg, Russia. The lighter is

equipped with two KLT-40S type nuclear reactors,

which have been successfully used in icebreakers for

many decades. Each reactor can supply up to 35 MW of

electricity and 200 GJ/h of district heating, sufficient to

supply around 100,000 people in polar regions. After

the launch, the lighter was towed through the Baltic

and North Sea to Murmansk, where it is loaded with

nuclear fuel. Next year, the Akademik Lomonosov will

be towed to the Chukchi region in eastern Russia to its

final location.

2. On 6 June 2018, the Taishan 1 nuclear power plant unit

in the province of Guangdong in southern China

achieved first criticality. This is the first active EPR type

plant in the world and thus the second Generation III+

reactor type to go into operation after the Russian

VVER-1200 in Novovoronezh, which went into operation

in 2016. With a gross nominal output of 1750 MW, it is

the world's most powerful type of nuclear power plant.

Construction of the plant began in 2009. 2 blocks of the

same type have been under construction in Europe

since 2005 (Olkiluoto 3, Finland) and 2007 ( Flamanville

3, France). Originally, EPR reactors were developed

for a Western European expansion program and are

supplied by Framatome. A second unit is currently being

commissioned at the Taishan site in China. French

President Emmanuel Macron and Indian President

Narenda Modi signed a contract in March 2018 to build

six EPRs in India.

3. On 21 June 2018, the Sanmen 1 nuclear power plant

unit in the Chinese province of Zhejiang achieved first

criticality. This is the first AP1000 plant worldwide

and thus the third Generation III+ reactor type in

operation. Construction of the plant began in 2009 and

on 8 August 2018 the identical Haiyang 1 block in the

Chinese province of Shandong also achieved first

criticality. A further block is under construction at each

of the two sites. The AP1000 with a gross output of

around 1250 MW is a development of Westinghouse. In

the USA, four units are under construction at the Vogtle

and Summer sites; construction of the two Summer

units was suspended in August 2017, partly because the

Westinghouse Electric Company, as the manufacturer,

had to initiate Chapter 11 insolvency procedure.

Meanwhile, the Canadian Brookfield Business Partners

has taken over the nuclear technology company. Among

others, the Indian government is confident of signing a

contract for the construction of 6 AP1000s in India in

the near future.

These start-ups not only mark the fact that, despite all the

challenges and the associated delays, new technical

ground can be successfully broken in nuclear technology.

EPR, AP1000 or VVER-1200 can now provide impetus for

the marketing of nuclear energy in the new markets

available - even if these markets are not necessarily located

in Europe at present.

Oh yes, Europe ... two sentences about the Old World:

1. Nuclear energy, and thus the reactors at the Belgian

sites of Tihange and Doel, which are almost prayer- milllike

in some media, have so far this year covered around

60 % of the country's electricity requirements. In April

2018, the current Belgian government had confirmed

an “energy pact” for the country's nuclear power plants,

which intends for the plants to be decommissioned

between 2022 and 2025. This is about the seventh exit

announcement by a Belgian government.

2. The UK government is promoting the development

and construction of small modular reactors (SMR). A

£ 200 million investment programme as part of the

country's long-term industrial strategy is to accelerate

the construction of a pilot plant at Trawsfynydd in

northern Wales.

So it is not only exciting with regard to the future of nuclear

energy worldwide, there are now also future prospects for

expansion worldwide with currently 454 commercial units

in operation, as many as never before.

Christopher Weßelmann

– Editor in Chief –

427

EDITORIAL

Editorial

Nuclear Energy: The Dead Live Longer or the Summer of 2018

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

EDITORIAL 428

Kernenergie: Totgesagte leben länger

oder der Sommer 2018

Liebe Leserin, lieber Leser, obgleich die Kernenergie ein sowohl umfassendes technisches Potenzial mit

weiteren Entwicklungsperspektiven für den Einsatz in der Energieerzeugung als auch attraktive betriebswirtschaftliche

Rahmenbedingungen, sowohl für bestehende Anlagen als auch für Neuanlagen – ein verlässliches regulatorisches und

politisches Umfeld vorausgesetzt – bietet, fehlten hierzu lange sichtbare Impulse.

Die Kernenergie wurde bzw. wird zudem mit ernsten

Herausforderungen der Märkte konfrontiert. So kann sie

ihre wirtschaftlichen Vorteile aus zwei Gründen nicht

ausspielen: Zum einen existieren kaum noch freie Strommärkte;

regulierte Märkte mit teils überbordenden und

kaum noch überschaubaren Subventionssystemen verhindern

jegliche Marktentwicklung in Richtung effizienter

Systeme überhaupt. Zum anderen sind Anlagen mit langen

Abschreibungszeiten, wie es bei der Kernenergie mit rund

20 Jahren der Fall ist, wenig attraktiv – langer Atem ist für

Kernkraftwerksbetreiber erforderlich.

Bemerkenswerte Entwicklungen im Frühjahr/Sommer

2018 setzen insbesondere mit ihren technischen Akzenten

deutliche Zeichen für Zukunftsimpulse:

1. Ende April 2018 lief in St. Petersburg, Russland, die

Akademik Lomonosov vom Stapel. Der Leichter ist ausgerüstet

mit zwei Kernreaktoren vom Typ KLT-40S, wie sie

erfolgreich seit vielen Jahrzehnten in Eisbrechern zum

Einsatz kommen. Jeder Reaktor kann bis zu 35 MW Strom

liefern sowie 200 GJ/h Fernwärme, ausreichend für die

Versorgung von rund 100.000 Menschen in polaren

Regionen. Der Leichter wurde nach dem Stapellauf durch

Ost- und Nordsee nach Murmansk geschleppt, wo die

Kernbrennstoffbeladung erfolgt. Im kommenden Jahr

wird die Akademik Lomonosov in die Tschuktschen-Region

im Osten Russlands zu ihrem endgültigen Einsatzort

geschleppt.

2. Am 6. Juni 2018 erreichte der Kernkraftwerksblock

Taishan 1 in der im Süden Chinas gelegenen Provinz

Guangdong Erstkritikalität. Es ist dies die erste Anlage

weltweit vom Typ EPR und damit nach dem 2016 in Betrieb

gegangenen russischen WWER-1200 in Nowoworonesch

der zweite Reaktortyp der Generation III+ in Betrieb. Mit

einer Nennleistung von 1750 MW brutto ist es der weltweit

leistungsstärkste Kernkraftwerkstyp. Der Bau der Anlage

begann im Jahr 2009. In Europa sind 2 typgleiche Blöcke

seit 2005 (Olkiluoto 3, Finnland) bzw. 2007 (Flamanville 3,

Frankreich) in Bau. Ursprünglich waren Reaktoren des

Typs EPR für ein westeuropäisches Zubauprogramm entwickelt

worden und werden von Framatome geliefert. Am

chinesischen Standort Taishan befindet sich ein zweiter

Block in der Inbetriebnahme. Der französische Staatspräsident

Emmanuel Macron und der indische Präsident

Narenda Modi unterzeichneten im März 2018 einen

Vertrag, der zum Bau von sechs EPR in Indien führen soll.

3. Am 21. Juni 2018 erreichte der Kernkraftwerksblock

Sanmen 1 in der chinesischen Provinz Zhejiang Erstkritikalität.

Es ist dies die erste Anlage weltweit vom Typ

AP1000 und damit der dritte Reaktortyp der Generation

III+ in Betrieb. Der Bau der Anlage begann im Jahr 2009.

Am 8. August 2018 erreichte der baugleiche Block Haiyang

1 in der chinesischen Provinz Shandong ebenfalls Erstkritikalität.

An beiden Standorten ist jeweils ein weiterer

Block in Bau. Der AP1000 mit einer Bruttoleistung von rd.

1250 MW ist eine Entwicklung von Westinghouse. Der Bau

begann im Jahr 2009. In den USA sind an den Standorten

Vogtle und Summer vier Blöcke in Bau; für die beiden

Blöcke Summer wurde im August 2017 ein Baustopp

beschlossen, u.a. da die Westinghouse Electric Company

als Hersteller ein sog. „Chapter 11-Insolvenzverfahren“

ein leiten musste. Inzwischen hat die kanadische Brookfield

Business Partners das Kerntechnikunternehmen übernommen.

Unter anderem die indische Regierung ist

zuversichtlich, einen Vertrag über den Bau von 6 AP1000

in Indien in der nächsten Zukunft unterzeichnen zu

können.

Diese Inbetriebnahmen kennzeichnen nicht nur, dass bei

allen Herausforderungen und auch damit verbundenen

Verzögerungen, technisches Neuland in der Kerntechnik

erfolgreich beschritten werden kann. EPR, AP1000 oder

auch WWER-1200 können jetzt Impulse mit sich bringen,

die der Vermarktung auf den bereit stehenden neuen

Märkten für die Kernenergie Schwung liefern – auch wenn

diese Märkte derzeit nicht unbedingt in Europa liegen.

Ach ja Europa ... zwei Sätze zur Alten Welt:

1. Die Kernenergie und damit die in manchen Medien fast

gebetsmühlenartig gescholtenen Reaktoren an den

belgischen Standorten Tihange und Doel haben im

bisherigen Jahresverlauf rund 60 % des Strombedarfs des

Landes gedeckt. Die derzeitige belgische Regierung

hatte im April 2018 für alle Kernkraftwerke des Landes

einen „Energiepakt“ bestätigt, der eine Stilllegung der

Anlagen in den Jahren 2022 bis 2025 vorsieht. Es ist

dies ungefähr die siebte Ausstiegsankündigung einer

belgischen Regierung.

2. Die Regierung Großbritanniens fördert die Entwicklung

und den Bau von modularen Kernreaktoren kleiner

Leistung (SMR: small modular reactor). Ein 200-Mio.-

Pfund Investitionsprogramm im Rahmen der langfristigen

Industriestrategie des Landes soll den Bau einer Pilotanlage

am Standort Trawsfynydd im Norden Wales

forcieren.

Es bleibt also nicht nur spannend, was die Zukunft der

Kernenergie weltweit betrifft, es gibt jetzt auch Zukunftsperspektiven

sogar für einen Ausbau weltweit– mit derzeit

454 Kernkraftwerken weltweit in Betrieb...so viele wie

noch nie zuvor.

Christopher Weßelmann

– Chefredakteur –

Editorial

Nuclear Energy: the Dead Live Longer or the Summer of 2018

---

Kommunikation und

Training für Kerntechnik

Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?

Wählen Sie aus folgenden Themen: Dozent/in Termin/e Ort

3 Atom-, Vertrags- und Exportrecht

Ihr Weg durch Genehmigungs- und Aufsichtsverfahren RA Dr. Christian Raetzke 18.09.2018

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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

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04.06.2019 Berlin

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Schlüsselfaktor Interkulturelle Kompetenz –

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Public Hearing Workshop –

Öffentliche Anhörungen erfolgreich meistern

Kerntechnik und Energiepolitik im gesellschaftlichen Diskurs

– Themen und Formate

Angela Lloyd 26.09.2018 Berlin

Dr. Nikolai A. Behr 16.10. - 17.10.2018

05.11. - 06.11.2019

Berlin

N.N. 12.11. - 13.11.2018 Gronau/

Lingen

3 Rückbau und Strahlenschutz

In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:

Stilllegung und Rückbau in Recht und Praxis

Das neue Strahlenschutzgesetz –

Folgen für Recht und Praxis

Dr. Matthias Bauerfeind

RA Dr. Christian Raetzke

Maria Poetsch

RA Dr. Christian Raetzke

24.09. - 25.09.2018 Berlin

05.11. - 06.11.2018

12.02. - 13.02.2019

25.06. - 26.06.2019

Berlin

3 Nuclear English

Enhancing Your Nuclear English Devika Kataja 22.05. - 23.05.2019 Berlin

Advancing Your Nuclear English (Aufbaukurs) Devika Kataja 10.10. - 11.10.2018

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3 Wissenstransfer und Veränderungsmanagement

Berlin

Veränderungsprozesse gestalten – Heraus forderungen

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Erfolgreicher Wissenstransfer in der Kern technik –

Methoden und praktische Anwendung

Dr. Tanja-Vera Herking

Dr. Christien Zedler

Dr. Tanja-Vera Herking

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28.11. - 29.11.2018

26.11. - 27.11.2019

Berlin

26.03. - 27.03.2019 Berlin

Haben wir Ihr Interesse geweckt? 3 Rufen Sie uns an: +49 30 498555-30

Kontakt

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Die INFORUM-Seminare können je nach

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

430

Issue 8/9

August/September

CONTENTS

437

Akademik Lomonosov:

Preparations for Premiere

in Full Swing

| | The world’s only floating power unit ‘Akademik Lomonosov’ takes the sea. On 28 April 2018, the floating nuclear power unit (FPU)

‘Akademik Lomonosov’ has left the territory of Baltiyskiy Zavod in St. Petersburg, Russia, where its construction had been carried out

since 2009, and headed to its base in Chukotka.

Editorial

Nuclear Energy: the Dead Live Longer

or the Summer of 2018 427

Kernenergie: Totgesagte leben länger

oder der Sommer 2018 428

Abstracts | English 432

Abstracts | German 433

Inside Nuclear with NucNet

A Stark Warning to Trump on China, Russia

and the ‘Crisis’ Facing US Nuclear Industry 434

NucNet, David Dalton

Calendar 436

442

| | Neutron radiographs of U3Si2 pins from ATR.

Energy Policy, Economy and Law

Akademik Lomonosov:

Preparations for Premiere in Full Swing 437

Roman Martinek

440

Spotlight on Nuclear Law

Nuclear Phase-out Last Act?

Are the New Compensation Regulations for

Frustrated Expenses in Accordance

with the Constitution? 440

Atomausstieg letzter Akt?

Sind die neuen Entschädigungs regelungen

für frustrierte Aufwendungen und nicht mehr

verstrombare Elektrizitätsmengen im Atomgesetz

verfassungsgemäß? 440

| | Upper part of a pressurized reactor vessel during maintenance.

Tobias Leidinger

Contents

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

431

Fuel

Innovations for the Future

Westinghouse EnCore® Accident Tolerant Fuel 442

Gilda Bocock, Robert Oelrich, and Sumit Ray

Operation and New Build

Analyses of Possible Explanations for the

Neutron Flux Fluctuations in German PWR 446

457

CONTENTS

Joachim Herb, Christoph Bläsius, Yann Perin,

Jürgen Sievers and Kiril Velkov

| | Coated Ceramic Honeycomb type Passive Autocatalytic Recombiner.

Detailed Measurements and Analyses of the

Neutron Flux Oscillation Phenomenology

at Kernkraftwerk Gösgen 452

A Preliminary Conservative Criticality Assessment

of Fukushima Unit 1 Debris Bed 473

G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff

María Freiría López, Michael Buck and Jörg Starflinger

452

AMNT 2018

Key Topic | Outstanding Know-How

& Sustainable Innovations

Focus Session International Regulation:

Radiation Protection: The Implementation

of the EU Basic Safety Standards Directive 2013/59

and the Release of Radioactive Material

from Regulatory Control 477

Christian Raetzke

| | Schematic representation of the 3002 MW 3-Loop KKG core.

DAtF Notes 456

Research and Innovation

Effects of Airborne Volatile Organic Compounds on

the Performance of Pi/TiO 2 Coated Ceramic Honeycomb

Type Passive Autocatalytic Recombiner 457

Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

AMNT 2018

Young Scientists' Workshop 463

Report

Report: GRS Workshop “Safety of Extended

Dry Storage of Spent Nuclear Fuel” 480

Klemens Hummelsheim, Florian Rowold and Maik Stuke

KTG Inside 483

News 484

Nuclear Today

Why do We Allow Nuclear to Take

the ‘Silly Season’ Media Heat? 490

Jörg Starflinger

John Shepherd

Heuristic Methods in Modelling Research

Reactors for Deterministic Safety Analysis 464

Imprint 485

Vera Koppers and Marco K. Koch

Development and Validation of a CFD

Wash-Off Model for Fission Products

on Containment Walls 469

Katharina Amend and Markus Klein

Aachen Institute for Nuclear Training

AMNT 2019: Call for Papers

Inforum: Seminar Programme 2018/2019

Insert

Insert

Insert

Contents

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

432

ABSTRACTS | ENGLISH

A Stark Warning to Trump on China, Russia

and the ‘Crisis’ Facing US Nuclear Industry

NucNet, David Dalton | Page 434

The US has the largest number of nuclear plants in

the world – 99 in commercial operation at the time

of writing – but its global leadership position is said

to be declining as efforts to build a new generation

of reactors have been plagued by problems, and

aging plants have been retired or closed in the face

of economic, market, and financial pressures. A

recent report by the Atlantic Council issued a

stark warning, arguing that the US nuclear energy

industry is facing a crisis that the Trump administration

must immediately address as a core part of

its “all of the above” energy strategy.

Akademik Lomonosov:

Preparations for Premiere in Full Swing

Roman Martinek | Page 437

At the end of July 2018, the loading of the floating

power unit Akademik Lomonosov with nuclear fuel

started in Murmansk. This is one of the key stages of

the project, which as of today has no analogues in the

world. In 2019, the power unit will begin to supply

local population and industrial facilities in North-

Eastern Siberia with heat and electricity. The project is

expected to open up opportunities for the mass production

of floating nuclear power plants – a number

of countries have already voiced their interest.

The Akademik Lomonosov is intended for providing

energy to remote industrial facilities, port cities, as

well as gas and oil platforms located on the high seas.

Nuclear Phase-out Last Act?

Are the New Compensation Regulations for

Frustrated Expenses in Accordance with the

Constitution?

Tobias Leidinger | Page 440

Shortly before it was passed, the legislature reacted

to the constitutional deficiencies which the Federal

Constitutional Court (BVerfG) objected to in its

judgment of 6 December 2016 on the nuclear

phase-out (BVerfGE 143, 246) and for which a

constitutional situation had to be established by

30 June 2018. However, the newly created compensation

regulations in the 16 th amendment to the

Atomic Energy Act raise new legal questions,

especially those relating to their constitutionality.

Westinghouse EnCore ® Accident

Tolerant Fuel

Gilda Bocock, Robert Oelrich

and Sumit Ray | Page 442

The development and implementation of accident

tolerant fuel (ATF) products, such as Westinghouse’s

EnCore® Fuel, can support the long-term

viability of nuclear energy by enhancing operational

safety and decreasing energy costs. The first introduction

of Westinghouse EnCore Fuel into a commercial

reactor is planned for 2019 as segmented

lead test rods (LTRs) utilizing chromium-coated

zirconium cladding with uranium silicide (U 3 Si 2 )

pellets. The EnCore Fuel lead test assembly (LTA)

program, with LTAs planned for 2022 insertion, will

introduce silicon carbide/silicon carbide composite

cladding with U 3 Si 2 pellets.

Analyses of Possible Explanations for the

Neutron Flux Fluctuations in German PWR

Joachim Herb, Christoph Bläsius, Yann Perin,

Jürgen Sievers and Kiril Velkov | Page 446

During the last 15 years the neutron flux fluctuation

levels in some of the German PWR changed

­significantly. During a period of about ten years, the

fluctuation levels increased, followed by about five

years with decreasing levels after taking actions like

changing the design of the fuel elements. The

­increase in the neutron flux fluctuations resulted in

an increased number of triggering the reactor

limitation system and in one case in a SCRAM.

Several models based on single physical effects are

used to simulate the neutron flux. Each of these

simple models can reproduce some of the characteristics

of the observed neutron flux fluctuations.

Detailed Measurements and Analyses of the

Neutron Flux Oscillation Phenomenology at

Kernkraftwerk Gösgen

G. Girardin, R. Meier, L. Meyer,

A. Ålander and F. Jatuff | Page 452

Recent investigations on measured neutron flux

noise at the Kernkraftwerk Gösgen-Däniken are

summarised. The NPP in operation since 1979 is a

German KWU pre-KONVOI, 3-Loop PWR with a

thermal power of 3,002 MWth (1,060 MWe). In a

period of approx. 7 cycles from 2010 to 2016, an

increase of the measured neutron noise amplitudes

in the in- and out-core neutron detectors has been

observed, although no significant variations have

being detected in global core, thermohydraulic

­circuits or instrumentation parameters. Verifications

of the instrumentation were performed and it was

confirmed that the neutron flux instabilities

increased from cycle to cycle in this period. In the last

two years, the level of neutron flux noise remains

high but seems to have achieved a saturation state.

Effects of Airborne Volatile Organic

Compounds on the Performance of Pi/TiO 2

Coated Ceramic Honeycomb Type Passive

Autocatalytic Recombiner

Chang Hyun Kim, Je Joong Sung,

Sang Jun Ha and Phil Won Seo | Page 457

Ensuring the containment integrity during a severe

accident in nuclear power reactor by maintaining the

hydrogen concentration below an acceptable level

has been recognized to be of critical importance after

Fukushima Daiichi accidents. Although there exist

various hydrogen mitigation measures, a passive

autocatalytic recombiner (PAR) has been considered

as a viable option for the mitigation of hydrogen risk

under the extended station blackout conditions

because of its passive operation char acteristics for

the hydrogen removal. As a post- Fukushima action

item, all Korean nuclear power plants were equipped

with PARs of various suppliers. The capacity and

locations of PAR as a hydrogen mitigation system

were determined through an extensive analysis for

various severe accident scenarios.

49 th Annual Meeting on Nuclear Technology

(AMNT 2018): Young Scientists Workshop

Jörg Starflinger | Page 463

During the Young Scientists Workshop of the 49 th

Annual Meeting on Nuclear Technology (AMNT

2018), 29 to 30 May 2018, Berlin, 13 young

­scientists presented results of their scientific

research as part of their Master or Doctorate theses

covering a broad spectrum of technical areas. Vera

Koppers, Katharina Amend and Maria Freiria were

awarded for their presentations by the jury.

Heuristic Methods in Modelling Research

Reactors for Deterministic Safety Analysis

Vera Koppers and Marco K. Koch | Page 464

A new method for rapid and reliable modelling of

research reactors for deterministic safety analysis is

presented. A rule-based software system is being

developed to support the modelling process in

ATHLET for selected research reactor types in the

light of limited available data. The fundamental

elements of the input deck are generated automatically

by few input data necessary.

Development and Validation of a

CFD Wash-Off Model for Fission Products

on Containment Walls

Katharina Amend and Markus Klein | Page 469

The research project aims to develop a CFD model

to describe the run down behavior of liquids and the

resulting wash-down of fission products on surfaces

in the reactor containment. The paper presents a

three-dimensional numerical simulation for water

running down inclined surfaces coupled with an

aerosol wash-off model and particle transport using

OpenFOAM. The wash-off model is based on Shields

criterion. A parameter variation is conducted and

the simulation results are compared to experiments.

A Preliminary Conservative Criticality

Assessment of Fukushima Unit 1 Debris Bed

María Freiría López, Michael Buck and

Jörg Starflinger | Page 473

A conservative criticality evaluation of Fukushima

Unit 1 debris bed has been carried out. In order to

obtain a multi-dimensional criticality map, parameters,

such as debris size, porosity, particle size, fuel

burnup, water density and boration were varied. As

a result, safety parameter ranges where recriticality

can be excluded have been identified. It was found

that most of the possible debris would be inherently

subcritical because of its porosity and 1600 ppm B

would ensure subcriticality under any conditions.

49 th Annual Meeting on Nuclear Technology

(AMNT) Key Topic | Outstanding Know-How

& Sustainable Innovations

Christian Raetzke | Page 477

The report summarises the presentations of the

Focus Session International Regulation | Radiation

Protection: The Implementation of the EU Basic

Safety Standards Directive 2013/59 and the Release

of Radioactive Material from Regulatory Control

presented at the 49 th AMNT 2018, Berlin, 29 to 30

May 2018.

Report: GRS Workshop “Safety of Extended

Dry Storage of Spent Nuclear Fuel”

Klemens Hummelsheim, Florian Rowold

and Maik Stuke | Page 480

Conference report on the GRS Workshop “Safety of

Extended Dry Storage of Spent Nuclear Fuel”, 6 to 8

June 2018.

Why do We Allow Nuclear to Take the

‘Silly Season’ Media Heat?

John Shepherd | Page 490

The time of year always means all manner of weird

and wonderful stories finding their way into the

news. For the nuclear industry, the hot spell fanned

the media flames of an old anti-nuclear favourite, as

it became clear operations at some nuclear power

plants were being halted temporarily to comply

with restrictions that prevent cooling water further

heating local rivers and waterways. It’s a question

why the nuclear community does not use the time

of year to communicate their important and

interesting topics.

Abstracts | English

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

Eine deutliche Warnung für die

US-Nuklearindustrie – auch vor der

Konkurrenz aus China und Russland

NucNet, David Dalton | Seite 434

In den USA ist die weltweit größte Anzahl von

Kernkraftwerken in kommerziellem Betrieb – 99

Anlagen; aber die globale Führungsposition der

USA schwindet, da die Bemühungen zum Bau einer

neuen Generation von Reaktoren mit Problemen

behaftet ist und ältere Anlagen angesichts wirtschaftlichen

Drucks stillgelegt werden. Ein kürzlich

veröffentlichter Bericht des Atlantic Council warnt

die US-Nuklearindustrie vor einer Krise, der die

Trump-Regierung als Kernstück ihrer „All of the

above“-Energiestrategie begegnen muss.

Akademik Lomonosov: Vorbereitungen

für die Inbetriebnahme in vollem Gange

Roman Martinek | Seite 437

Ende Juli 2018 begann in Murmansk die Kernbrennstoffbeladung

des schwimmenden Kraftwerks

Akademik Lomonosov. Dies ist eine der bedeutenden

Phasen des Projekts, das bis heute weltweit

einzigartig ist. Das Kraftwerk wird ab 2019 eine

ganze Region in Nordostsibirien mit Wärme und

Strom versorgen. Das Projekt soll Möglichkeiten für

die Serienproduktion von schwimmenden Kernkraftwerken

eröffnen – einige Länder haben dafür

bereits ihr Interesse bekundet. Die Akademik

Lomonosov ist für die Energieversorgung abgelegener

Industrieanlagen, Hafenstädte sowie von Gasund

Ölplattformen auf hoher See konzipiert.

Atomausstieg letzter Akt? Sind die neuen

Entschädigungsregelungen für frustrierte

Aufwendungen und nicht mehr verstrombare

Elektrizitätsmengen im Atomgesetz

verfassungsgemäß?

Tobias Leidinger | Seite 440

Kurz vor knapp hat der Gesetzgeber auf die

verfassungs rechtlichen Mängel reagiert, die das

Bundesverfassungsgericht (BVerfG) in seinem Urteil

vom 6. Dezember 2016 zum Atomausstieg (BVerfGE

143, 246) höchstrichterlich beanstandet hat und für

die bis zum 30. Juni 2018 ein verfassungsgemäßer

Zustand herzustellen war. Doch die neu geschaffenen

Entschädigungsregelungen in der 16.

AtG-Novelle werfen neue Rechtsfragen auf, insbesondere

die nach ihrer Verfassungsgemäßheit.

Westinghouse EnCore® Accident

Tolerant Fuel

Gilda Bocock, Robert Oelrich und

Sumit Ray | Seite 442

Entwicklung und Einsatz von „störfalltolerantem

Kernbrennstoff“ wie z.B. EnCore® von Westinghouse,

kann der Kernenergie weitere Zukunftsperspektiven

durch Erhöhung der Betriebssicherheit und

Senkung der Kosten eröffnen. Der erste Einsatz von

Westinghouse EnCore Fuel in einem kommerziellen

Reaktor ist für 2019 geplant. Testbrennstäbe mit

verchromtem Zirkoniummantel und Uransilicid

(U3Si2)-Pellets sind dafür vorgesehen. Das für

2022 geplante EnCore Fuel Lead Test Assembly

(LTA)-Programm sieht ein Siliziumkarbid/Siliziumkarbid-Verbundhüllrohr

mit U3Si2-Pellets vor.

verändert. Während eines Zeitraums von etwa zehn

Jahren nahmen die Schwankungsbreiten zu, gefolgt

von etwa fünf Jahren mit abnehmender Tendenz

nach z.B. einer Änderung der Auslegung der Brennelemente.

Die Zunahme der Neutronenflussschwankungen

führte zu einer erhöhten Anzahl von

Auslösungen des Reaktorbegrenzungssystems und in

einem Fall zu einem SCRAM. Zur Simulation des Neutronenflusses

werden mehrere Modelle verwendet,

die auf einzelnen physikalischen Effekten basieren.

Detaillierte Messungen und Analysen

der Neutronenflussschwingungen

im Kernkraftwerk Gösgen

G. Girardin, R. Meier, L. Meyer, A. Ålander

und F. Jatuff | Seite 452

Aktuelle Untersuchungen zum Neutronenflussrauschen

im Kernkraftwerk Gösgen-Däniken

werden zusammengefasst. Das seit 1979 in Betrieb

befindliche Kernkraftwerk In einem Zeitraum von

ca. 7 Zyklen von 2010 bis 2016 wurde ein Anstieg

der gemessenen Neutronenrauschamplituden beobachtet,

obwohl keine signifikanten Schwankungen

der globalen physikalischen und thermohydraulischen

sowie Instrumentierungsparametern

festgestellt wurden. Überprüfungen der Instrumentierung

wurden durchgeführt und es wurde bestätigt,

dass die Neutronenflussinstabilitäten in diesem

Zeitraum von Zyklus zu Zyklus zunahmen. In den

letzten zwei Jahren blieb das Neutronenflussrauschen

hoch, scheint aber einen Sättigungszustand

erreicht zu haben.

Einfluss von flüchtigen organischen

Verbindungen auf Pi/TiO 2 -beschichtete

keramische Wabenkörpern von passiven

autokatalytischen Rekombinatoren

Chang Hyun Kim, Je Joong Sung, Sang Jun Ha

und Phil Won Seo | Seite 457

Nach den Unfällen von Fukushima Daiichi wurde

festgestellt, dass der Integrität des Sicherheitsbehälters

bei einem schweren Unfall in einem

Kernkraftwerk höchste Priorität gilt, indem die

Wasserstoffkonzentration unterhalb akzeptabler

Werte gehalten wird. Obwohl es verschiedene

Maßnahmen zur Wasserstoffminderung gibt, wird

ein passiver autokatalytischer Rekombinator (PAR)

wegen seiner Betriebseigenschaften als praktikable

Option angesehen. Als Post-Fukushima-Maßnahme

wurden alle koreanischen Kernkraftwerke mit PARs

verschiedener Anbieter ausgestattet. Die Kapazitäten

und optimalen Einbauorte von PARs als

Wasserstoffminderungssystem wurden durch eine

umfangreiche Analyse für verschiedene schwere

Unfallszenarien ermittelt.

49. Jahrestagung Kerntechnik (AMNT 2018):

Young Scientists Workshop

Jörg Starflinger | Seite 463

Im Rahmen des Young Scientists Workshop der

49. Jahrestagung Kerntechnik (AMNT 2018) vom

29. bis 30. Mai 2018 in Berlin stellten 13 Nachwuchswissenschaftlerinnen

und -wissenschaftler im

Rahmen ihrer Master- oder Doktorarbeiten ein breites

Spektrum von Fachthemen vor. Vera Koppers,

Katharina Amend und Maria Freiria wurden für ihre

Präsentationen von der Jury ausgezeichnet.

für die Durchführung von deterministischen

Sicherheitsanalysen vorgestellt. Für ausgewählte

Forschungsreaktor-Typen wird ein regelbasiertes

Softwaresystem konzipiert, das den Modellierungsprozess

für ATHLET unterstützt. Die Entwicklung

wird unter dem Aspekt limitierter verfügbarer

Daten vorgenommen. Die fundamentalen Elemente

des Datensatz werden unter Verwendung weniger

Eingabedaten automatisch generiert.

Entwicklung und Validierung eines

CFD-Modells für das Auswaschen von Spaltprodukten

auf Containment-Oberflächen

Katharina Amend und Markus Klein | Seite 469

Ziel des Forschungsvorhabens ist ein CFD-Modell

für das Ablaufverhalten von Wasser und den

resultierenden Abwasch von Spaltprodukten auf

Oberflächen im Reaktorsicherheitsbehälter. Das

Paper präsentiert eine dreidimensionale numerische

OpenFOAM Simulation von Wasser auf geneigten

Oberflächen gekoppelt mit einem Aerosol-­

Abwaschmodell und dem Partikeltransport. Das

Abwaschmodell basiert auf dem Shields Kriterium.

Es wird eine Parametervariation durchgeführt und

die Simulationsergebnisse mit Experimenten verglichen.

Eine vorläufige konservative

Kritikalitätsbeurteilung des Schüttbetts

des Reaktors Fukushima-1

María Freiría López, Michael Buck und

Jörg Starflinger | Seite 473

Eine konservative Kritikalitätsanalyse des Fukushima

Unit 1 Schüttbetts wurde durchgeführt. Um eine

mehrdimensionale Kritikalitätskarte zu erstellen,

wurden Parameter wie Schüttbettgröße, Porosität,

Partikelgröße, Brennstoffabbrand, Wasserdichte und

Boranteil variiert. Als Resultat, wurden Bereiche

identifiziert, in denen Rekritikalität ausgeschlossen

werden kann. Es stellt sich heraus, dass die meisten

entstehenden Schüttbetten aufgrund seiner Porosität

inhärent unterkritisch sind, und dass auch

1600 ppm B Unterkritikalität sicherstellen.

49. Jahrestagung Kerntechnik (AMNT 2018)

Key Topic | Outstanding Know-How &

Sustainable Innovations

Christian Raetzke | Seite 477

Der Bericht fasst die Vorträge der Focus Session

International Regulation | Radiation Protection:

The Implementation of the EU Basic Safety

Standards Directive 2013/59 and the Release of

Radioactive Material from Regulatory Control

zusammen, die auf der 49. Jahrestagung Kerntechnik

(AMNT 2018) präsentiert wurden.

Report: GRS Workshop “Safety of Extended

Dry Storage of Spent Nuclear Fuel”

Klemens Hummelsheim, Florian Rowold und

Maik Stuke | Seite 480

Tagungsbericht zum Workshop “Sicherheit einer

zeitlich längeren trockenen Lagerung abgebrannter

Brennelemente”, 6 bis 8 Juni 2018.

Warum lassen wir zu, dass die Kernenergie

in der “Saure Gurken Zeit“ Thema wird

433

ABSTRACTS | GERMAN

Analysen zu Neutronenflussschwankungen

in deutschen DWR

Joachim Herb, Christoph Bläsius, Yann Perin,

Jürgen Sievers und Kiril Velkov | Seite 446

In den letzten 15 Jahren haben sich die Neutronenflussschwankungen

in einigen der deutschen DWR

Heuristische Methoden in der Modellierung

deterministischen Sicherheitsanalysen von

Forschungsreaktoren

Vera Koppers and Marco K. Koch | Seite 464

Es wird eine neue Methode zur schnellen und zuverlässigen

Modellierung von Forschungsreaktoren

John Shepherd | Seite 490

In der „Saure Gurken Zeit“ des Jahres werden von

der Presse teils seltsame und teils wunderbare

Geschichten aufgenommen. Immer wieder trifft

dies auch die Kernenergie – warum lassen wir dies

zu, mit den wichtigen positiven Botschaften, die wir

mit der Kernenergie haben?

Abstracts | German

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

434

INSIDE NUCLEAR WITH NUCNET

A Stark Warning to Trump on China, Russia

and the ‘Crisis’ Facing US Nuclear Industry

NucNet, David Dalton

The US has the largest number of nuclear plants in the world – 99 in commercial operation at the time of

writing – but its global leadership position is said to be declining as efforts to build a new generation of

reactors have been plagued by problems, and aging plants have been retired or closed in the face of economic,

market, and financial pressures.

A recent report by the Washington-based think-tank the

Atlantic Council issued a stark warning, arguing that the

US nuclear energy industry is facing a crisis that the Trump

administration must immediately address as a core part of

its “all of the above” energy strategy that is intended to

herald an era of American energy dominance, with tens of

billions of dollars to be spent on drilling and construction

of pipelines, processing plants and liquefied natural gas

export terminals. The administration might be bullish on

energy policy, but the nuclear industry is worried.

Six US nuclear plants have been shut down permanently

since 2013 and 12 more are slated to retire over the

next seven years. The Washington-based Nuclear Energy

Institute, which represents the nuclear industry in the US,

says the US electricity grid is enduring “unprecedented

tumult and challenge” because of the loss of thousands

and thousands of megawatts of carbon-free, fuel-secure

generation that nuclear plants represent. Closing nuclear

plants makes electricity prices go up and is putting

emissions reduction targets hopelessly out of reach, NEI

president and chief executive officer Maria Korsnick said.

The Atlantic Council says the decline of the nuclear

power industry in the US is “an important policy problem”

that is not receiving the attention it deserves. The report

was made public in the same week that Ohio-based utility

FirstEnergy announced plans to permanently shut down

its three nuclear power stations – Davis-Besse, Perry and

Beaver Valley – within the next three years without some

kind of state or federal relief.

The nuclear industry has long argued that electricity

markets should be reformed to recognise the ability of

traditional baseload generation with onsite fuel supplies –

including nuclear power plants – to provide grid resiliency

during extreme events like hurricanes or extreme winter

weather.

To save financially-ailing nuclear plants, state ­legislatures

in Illinois and New York last year approved subsidies to keep

nuclear plants operating after utilities made appeals about

protecting consumers and jobs. But other proposed bailouts

of nuclear plants have stalled in New Jersey, Connecticut,

Massachusetts, Ohio and Pennsylvania. In Minnesota, the

state legislature is considering a bill that would help Xcel

Energy, owner and operator of the Monticello and Prairie

Island nuclear stations, plan for the high costs of maintaining

old nuclear power plants. The proposed legislation would

give utilities earlier notice about how much money they could

recover for costly work, Minnesota Public Radio reported.

The Atlantic Council report says nuclear power should be

elevated in the Trump administration’s national security

strategy because nuclear is an important strategic sector, and

US global leadership and engagement in nuclear power are

“vital to US national security and foreign-policy interests”.

It also argues that nuclear power is an important

­component of a diversified US energy mix, but notes that in

sharp contrast to developments in the US, China and Russia

are pushing to expand their nuclear industries, develop

complete fuel cycles, and build and commercialise new

reactors for both domestic and international markets. The

results of these efforts are striking – nearly two-thirds of the

new reactors under construction worldwide are estimated to

be using designs from China and Russia, countries that have

the advantage of using “state- monopoly and authoritarian

systems” to advance nuclear energy for geopolitical means.

China has the largest nuclear construction programme

in the world by far, with 20 of the 53 total reactors under

construction worldwide. The 13 th Five-Year Plan (2016 to

2020) calls for 58 GW of nuclear capacity online by

2020 to 2021, and an additional 30 GW under construction

at that time.

But what is really worrying the US nuclear industry is

the success of China’s nuclear strategy to establish joint

ventures with Western companies (Toshiba-Westinghouse,

Framatome-Areva, SNC-Lavalin, Energoatom) to build and

evaluate different technologies (AP-1000, EPR, Candu,

VVER-1000), and to incorporate this experience into its

own indigenous designs. Although cost estimates are

­difficult to obtain, China has seemingly been able to build

reactors quicker, and at lower cost, than the US, Europe,

and even South Korea, the report says.

China brings a complete package of design, construction,

labour, technology, and financing, which improves

the economics compared to industries in the West.

Both China and Russia offer attractive financing

packages to fund these projects. China goes into markets

abroad with financing options from its Export-Import

Bank, while Russia uses resources from both the Russian

state budget and the Russia Wealth Fund.

In contrast, says the NEI, the US Export-Import Bank’s

board of directors remains without a quorum and as a

result cannot consider medium- and long-term transactions

exceeding $ 10 m. Typically, commercial nuclear

deals are measured in billions of dollars, not millions:

Turkish President Tayyip Erdoğan said that the investment

in the country’s first nuclear power plant, being built by

Russia’s Rosatom, will exceed $ 20 bn.

While China’s relationship with nuclear power is

­relatively new – with its first nuclear plant completed in

1991 – Russia’s long history with nuclear power dates to

1954, when the first reactor was commissioned in Obninsk.

The industry has since grown to 37 reactors in commercial

operation and five under construction. Nuclear generation

reached a record of 196.3 TWh in 2016, accounting

for 17 % of domestic electricity generation, and further

increased to 202.868 TWh and 19.9 % in 2017.

Inside Nuclear with NucNet

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

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

The Chinese and Russian use of nuclear-power ­financing

and technology as a means of expanding their overseas

physical presence, and their foreign-policy influence in

key countries, has important implications for the US the

Atlantic Council report says.

On one hand, US companies are collaborating with

China on building, developing, and demonstrating new

reactors; GE has won tenders for the supply of turbine

generators for new Russian-supplied units in Hungary and

Turkey. On the other hand, Russia and China are vying for

expanded influence in countries critical to US diplomacy,

namely Iran, Saudi Arabia, Turkey, Jordan, Egypt, and

Pakistan.

The Middle East is emerging as an arena of intense

­nuclear competition and positioning, with the first South

Korean nuclear unit recently completed at Barakah in the

United Arab Emirates, Jordan continuing to negotiate

on financing for two Russian nuclear reactors, Egypt

beginning construction of a nuclear station at Akkuyu

with Russia, and Saudi Arabia announcing its intention

to proceed with two reactors after years of delay. The

Chinese, French, Russians, and South Koreans have

submitted initial bids in Saudi Arabia, and a US has also

submitted a bid on this first phase of the process of

short listing companies. The bid was approved by the US

Department of Energy (DOE), even though the US has not

yet concluded a 123 nuclear framework agreement with

the Saudis, which would be necessary before a US export

deal could be finalised.

Drew Bond, a senior fellow and director of energy

innovation programmes at the American Council for

Capital Formation Centre for Policy Research, agrees

that this is a critical time for the Trump administration,

energy secretary Rick Perry and US domestic nuclear

infrastructure. He says the country’s 30-year hiatus in

building new reactors coupled with the rise of state-owned

competitors abroad has taken “a significant toll on the US

nuclear industry and has seriously undermined America’s

global influence over nonproliferation and other matters”.

The US used to be the overwhelming leader in

designing, building, and fuelling nuclear reactors around

the world, but no longer, said Mr Bond. “Unfortunately, in

recent years we have ceded this role – along with our

­influence – to other nations, particularly Russia, China,

and South Korea. More than a dozen countries have

planned or proposed to build new reactors in the coming

years. Whether those reactors are designed and built up to

US or Russia safety standards is critical, not to mention the

geopolitical implications for the world.”

President Trump and his administration have been

calling for an “all of the above” energy strategy that

achieves US energy dominance. But advanced fossil fuels

and renewables can’t do it alone. According to Mr Bond,

nuclear energy and the supply chain that comes with it

must be a part of the picture.

The Atlantic Council report, said the NEI, shows the

need for the administration and Congress to support

American commercial nuclear exports through concrete

action.

“It’s critical for our industry that, given aggressive

overseas, state-owned competitors, we work with the

White House and Congress to give American companies

the tools they need to compete and win abroad,” NEI

vice-president Dan Lipman said.

“That means reestablishing a quorum at Ex-Im Bank,

ensuring US expot controls for nuclear technology are

more efficient, ensuring Section 123 bilateral nuclear

cooperation agreements are concluded, and fully funding

commercial nuclear energy research and development in

the federal budget.

“It’s not only American jobs that are at stake, but our

influence on safety, security and nonproliferation norms

across the world.”

Author

NucNet

The Independent Global Nuclear News Agency

David Dalton

Editor in Chief, NucNet

Avenue des Arts 56

1000 Brussels, Belgium

www.nucnet.org

The Atlantic Council

report is online:

https://bit.ly/

2GmNx3k

INSIDE NUCLEAR WITH NUCNET 435

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

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

436

CALENDAR

Calendar

2018

02.09.-06.09.2018

19 th International Nuclear Graphite Specialists

Meeting (INGSM-19). Shanghai Institute of Applied

Physics, Shanghai, China, ingsm.csp.escience.cn

03.09.-06.09.2018

Jahrestagung des Fachverbandes Strahlenschutz.

Dresden, Germany, Fachverband für

Strahlenschutz e.V., www.fs-ev.org

04.09.-05.09.2018.

8. Symposium Lagerung und Transport

radioaktiver Stoffe. Hannover, Germany,

TÜV NORD Akademie, www.tuev-nord.de

05.09.-07.09.2018

World Nuclear Association Symposium 2018.

London, United Kingdom, World Nuclear Association

(WNA), www.world-nuclear.org

09.09.-14.09.2018

21 st International Conference on Water

Chemistry in Nuclear Reactor Systems.

San Francisco, CA, USA, EPRI – Electric Power

Research Institute, www.epri.com

12.09.-14.09.2018

SaltMech IX – 9 th Conference on the Mechanical

Behavior of Salt. Hannover, Germany, Federal

Institute for Geosciences and Natural Resources

(BGR) Hannover, the Institute of Geomechanics (IfG)

Leipzig and the Technical University of Clausthal

(TUC), www.saltmech.com

16.09.-20.09.2018

55 th Annual Meeting on Hot Laboratories and

Remote Handling – HOTLAB 2018. Helsinki,

Finland, VTT and International Atomic Energy

Agency (IAEA), www.vtt.fi/sites/hotlab2018/

17.09.-21.09.2018

62 nd IAEA General Conference. Vienna, Austria.

International Atomic Energy Agency (IAEA),

www.iaea.org

17.09.-20.09.2018

FONTEVRAUD 9. Avignon, France,

Société Française d’Energie Nucléaire (SFEN),

www.sfen-fontevraud9.org

17.09.-19.09.2018

4 th International Conference on Physics and

Technology of Reactors and Applications –

PHYTRA4. Marrakech, Morocco, Moroccan

Association for Nuclear Engineering and Reactor

Technology (GMTR), National Center for Energy,

Sciences and Nuclear Techniques (CNESTEN) and

Moroccan Agency for Nuclear and Radiological

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

19.09.-21.09.2018

Workshop Sicherheitskonzepte Endlagerung.

Grimsel, Switzerland. Fachverband für Strahlenschutz

e.V., www.fs-ev.org

26.09.-28.09.2018

44 th Annual Meeting of the Spanish Nuclear

Society. Avila, Spain, Sociedad Nuclear Española,

www.sne.es

30.09.-05.10.2018

14 th Pacific Basin Nuclear Conference (PBNC).

San Francisco, CA, USA, pbnc.ans.org

30.09.-03.10.2018

Fifteenth NEA Information Exchange Meeting on

ctinide and Fission Product Partitioning and

Transmutation. Manchester Hall, Manchester, UK,

OECD Nuclear Energy Agency (NEA), National

Nuclear Laboratory (NNL) in co‐operation with the

International Atomic Energy Agency (IAEA),

www.oecd-nea.org

30.09.-04.10.2018

TopFuel 2018. Prague, Czech Republic, European

Nuclear Society (ENS), American Nuclear Society

(ANS). Atomic Energy Society of Japan, Chinese

Nuclear Society and Korean Nuclear Society,

www.euronuclear.org

01.10.-05.10.2018

3 rd European Radiological Protection Research

Week ERPW. Rovinj, Croatia, ALLIANCE, EURADOS,

EURAMED, MELODI and NERIS, www.erpw2018.com

02.10.-04.10.2018

7 th EU Nuclear Power Plant Simulation ENPPS

Forum. Birmingham, United Kingdom, Nuclear

Training & Simulation Group, www.enpps.tech

08.10.-11.10.2018

World Energy Week. World Energy Council Council’s

Italian Member Committee, www.worldenergy.org

09.10.-11.10.2018

8 th International Conference on Simulation

Methods in Nuclear Science and Engineering.

Ottawa, Ontario, Canada, Canadian Nuclear Society

(CNS), www.cns-snc.ca

10.10.-11.10.2018

IGSC Symposium 2018 – Integrated Group for the

Safety Case; Current Understanding and Future

Direction for the Geological Disposal of Radioactive

Waste. Rotterdam, The Netherlands, OECD

Nuclear Energy Agency (NEA), www.oecd-nea.org

14.10.-18.10.2018

12 th International Topical Meeting on Nuclear

Reactor Thermal-Hydraulics, Operation and

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

www.nuthos-12.org

14.10.-18.10.2018

NuMat 2018. Seattle, United States,

www.elsevier.com

15.10.-18.10.2018

International Conference on Challenges Faced by

Technical and Scientific Support Organizations

(TSOs) in Enhancing Nuclear Safety and Security:

Ensuring Effective and Sustainable Expertise.

Brussels, Belgium, International Atomic Energy

Agency (IAEA), www.iaea.org

16.10.2018

The next steps for nuclear energy projects in the

UK. London, United Kingdom, Westminster Energy,

Environment & Transport Forum,

www.westminsterforumprojects.co.uk

16.10.-17.10.2018

4 th GIF Symposium at the 8th edition of Atoms

for the Future. Paris, France, www.gen-4.org

22.10.-24.10.2018

DEM 2018 Dismantling Challenges: Industrial

Reality, Prospects and Feedback Experience. Paris

Saclay, France, Société Française d’Energie Nucléaire,

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

24.10.-26.10.2018

NUWCEM 2018 Cement-based Materials for

Nuclear Waste. Avignon, France, French

Commission for Atomic and Alternative Energies

and Société Française d’Energie Nucléaire,

www.sfen-nuwcem2018.org

24.10.-25.10.2018

Chemistry in Power Plants. Magdeburg, Germany,

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

05.11.-08.11.2018

International Conference on Nuclear

Decom missioning – ICOND 2018. Aachen,

Eurogress, Germany, Aachen Institute for Nuclear

Training GmbH, www.icond.de

06.11-08.11.2018

G4SR-1 1 st International Conference on

Generation IV and Small Reactors. Ottawa,

Ontario, Canada. Canadian Nuclear Society (CNS),

and Canadian Nuclear Laboratories (CNL),

www.g4sr.org

12.11.-13.11.2018

15. Deutsche Atomrechtssymposium. Berlin,

Germany, Bundesministerium für Umwelt,

Naturschutz und nukleare Sicherheit, Wiss. Ltg. Prof.

Dr. Martin Burgi, www.grs.de/ars_anmeldung

13.11.-15.11.2018

24 th QUENCH Workshop 2018. Karlsruhe, Germany,

Karlsruhe Institute of Technology in cooperation with

the International Atomic Energy Agency (IAEA),

quench.forschung.kit.edu

22.11.2018

Weiterbildungskurs 2018 – IT-Sicherheit im Alltag

– Praxiswissen für Mitarbeiter in der Nukleartechnik.

Baden, Switzerland, Nuklearforum Schweiz,

www.nuklearforum.ch

03.12.-14.12.2018

United Nations, Conference of the Parties –

COP24. Katowice, Poland, United Nations

Framework Convention on Climate Change –

UNFCCC, www.cop24.katowice.eu

06.12.2018

Nuclear 2018. London, United Kingdom, Nuclear

Industry Association (NIA), www.niauk.org

2019

25.02.-26.02.2019

Symposium Anlagensicherung. Hamburg,

Germany, TÜV NORD Akademie, www.tuev-nord.de

10.03.-15.03.2019

83. Annual Meeting of DPG and DPG Spring

Meeting of the Atomic, Molecular, Plasma Physics

and Quantum Optics Section (SAMOP), incl.

Working Group on Energy. Rostock, Germany,

Deutsche Physikalische Gesellschaft e.V.,

www.dpg-physik.de

10.03.-14.03.2019

The 9 th International Symposium On

Supercritical- Water-Cooled Reactors (ISSCWR-9).

Vancouver Marriott Hotel, Vancouver, British

Columbia, Canada, Canadian Nuclear Society (CNS),

www.cns-snc.ca

09.04.-11.04.2019

World Nuclear Fuel Cycle 2019. Shanghai, China,

World Nuclear Association (WNA),

www.world-nuclear.org

07.05.-08.05.2019

50 th Annual Meeting on Nuclear Technology

AMNT 2019 | 50. Jahrestagung Kerntechnik.

Berlin, Germany, DAtF and KTG,

www.nucleartech-meeting.com – Save the Date!

27.10.-30.10.2019

FSEP CNS International Meeting on Fire Safety

and Emergency Preparedness for the Nuclear

Industry. Ottawa, Canada, Canadian Nuclear Society

(CNS), www.cns-snc.ca

Calendar

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

Akademik Lomonosov:

Preparations for Premiere in Full Swing

Roman Martinek

At the end of July, the loading of the floating power unit Akademik Lomonosov with nuclear fuel started in Murmansk.

This is one of the key stages of the project, which as of today has no analogues in the world. In 2019, the power unit will

begin to supply local population and industrial facilities in North-Eastern Siberia with heat and electricity. The project

is expected to open up opportunities for the mass production of floating nuclear power plants – a number of countries

have already voiced their interest.

On July 25, the Russian city of Murmansk, the largest

non-freezing seaport in the world and the largest city

above the Arctic Circle, saw the start of the loading of

­nuclear fuel into the reactors of the world’s only floating

nuclear power unit (FPU) Akademik Lomonosov. The

project, named after the outstanding Russian scientist and

laid down back in 2006, is the first in a series of mobile

transportable small-capacity power units. It is designed to

operate as part of a floating nuclear thermal power plant

(FNPP) and represents a new class of energy sources based

on Russian technologies of nuclear shipbuilding.

The Akademik Lomonosov is intended for the regions in

the High North and the Far East. Its main goal is to provide

energy to remote industrial facilities, port cities, as well as

gas and oil platforms located on the high seas. The permanent

mooring site of the floating NPP will be the Siberian

city of Pevek on the Chukchi Peninsula in the northeastern

extremity of Eurasia. The new plant will replace there two

technologically obsolete generation facilities: Bilibino NPP

and Chaunskaya CHPP. After being brought into operation,

the Akademik Lomonosov will become the northernmost

nuclear power plant in the world.

In the spring of this year, the floating power unit was

towed from the territory of the Baltic Shipyard, where its

construction was carried out from 2009, to the base of

Atomflot in Murmansk. During its transportation, the ship

144 meters long and 30 meters wide travelled the 4,000 km

route through the waters of four seas – the Baltic Sea, the

North Sea, the Norwegian Sea and the Barents Sea –

around the Scandinavian Peninsula and along the coasts

of Estonia, Sweden, Denmark and Norway. On May 19,

the Akademik Lomonosov was successfully moored in

Murmansk, where it was presented to the public in a

ceremonial atmosphere.

Vitaliy Trutnev, Head of Rosenergoatom’s Directorate for

the Construction and Operation of FNPPs, commented on

the current status of the project development: “Here in

Murmansk, we finalize the remaining technological

operations. Specialists have begun to implement one of the

most important tasks – the stage-by-stage loading of

nuclear fuel into the reactor plants. The next key stages

that are planned to be implemented before the end of

this year will be the physical launch of the reactors and the

beginning of complex mooring tests – after obtaining the

appropriate Rostekhnadzor permits (Federal Service for

Environmental, Technological and Nuclear Supervision –

author's note).

The FNPP project is based on the technology of

small modular reactors (SMRs) – this category, according

to IAEA classification, typically includes reactors with

electrical power up to 300 MW. A characteristic feature

of the majority of such designs is the integrated layout

of the reactor plant, in which the active zone, the steam

generator, the pressure compensator and a number of

other types of equipment are assembled in a single unit – a

factory-finished monoblock delivered ready-made to the

437

ENERGY POLICY, ECONOMY AND LAW

Energy Policy, Economy and Law

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

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ENERGY POLICY, ECONOMY AND LAW 438

site. This technology has been known since the 1960s: for

instance, the U.S. floating nuclear power plant Sturgis was

used for ten years to provide energy to the Panama Canal

in case of a threat of an intentional failure of the groundbased

power supply system, but it was decommissioned in

1976. To date, despite the existence of many similar developments

in the world, the Akademik Lomonosov is the only

floating power unit in the world, which gives uniqueness

to the Russian project.

The FPU is equipped with two KLT-40S icebreaker-type

reactors with a capacity of 35 MW each – together they are

able to produce up to 70 MW of electricity and 50 Gcal/h

of heat energy in the nominal operating mode, which is

enough to support the life of a city with a population of

about 100 thousand people. In addition to the floating

power unit itself, the structure of the FNPP project 20870

includes hydrotechnical facilities that provide installation

and detachment of the FPU and transfer of generated

electricity and heat to the shore, as well as onshore

facilities for transmitting this energy to external networks

for distribution to consumers. Currently, specialists are

working on the creation of this infrastructure in Pevek.

One of the main features of the project being implemented

is the placement of two reactor units in a small

hull of the vessel while preserving all the functional

characteristics of the ground-based nuclear power plant

with fewer maintenance personnel. At the same time, the

highest reliability and safety of operation is provided with

no environmental impact.

The floating power unit is supposed to have a lifespan

of from 35 to 40 years. For its operation, low-enriched

uranium will be used, and spent fuel will be accumulated

on the platform itself. Once every three years, fuel will be

reloaded, with the average annual duration of the reactor

refuelling not exceeding 60 days. In addition, on an annual

basis, scheduled shutdowns will be carried out at the plant

for routine maintenance, the average annual duration of

which will be no more than 20 days.

In designing the Akademik Lomonosov, priority was

given to such aspect as the safety of its operation. The

technological solution for the design components of the

FNPP is based on the tried and tested reference technology

used on nuclear icebreakers since 1988. The icebreakers

Taimyr and Vaigach were used as prototypes – their reactor

units have operated without fail for several decades in

the most difficult conditions of the Arctic. At the same

time, it should be noted that the technologies of the reactor

facilities for the icebreaking fleet are constantly being

improved and have made a qualitative step forward since.

This development is taking into account the fact that

increasingly high demands are being placed on nuclear

safety in the world.

Energy Policy, Economy and Law

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

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Thanks to the use of this experience, the Akademik

Lomonosov is today equipped with advanced icebreaker

reactors, and the FPU vessel is designed to withstand a

collision with an iceberg, the pressure exerted by a tsunami

wave as well as hurricanes – this safety margin makes the

ship virtually unsinkable and invulnerable to natural

disasters. From the outside environment, the FPU premises

are insulated with a double hull of the vessel, and reactor

facilities are equipped with special biological barriers that

do not allow radiation to spread beyond the compartments

where these facilities are located.

The FPU vessel design has also taken into account the

climatic conditions in which the FNPP will be operated.

The main body and load-bearing structures are made of

steel, resistant to brittle fracture under low temperature

conditions. In addition, the FPU is equipped with ice

strengthening – additional structural elements that ensure

the vessel’s strength during navigation in ice-covered waters,

as well as all the means necessary for towing with the

help of an icebreaker.

The primary importance of safety in the operation of

small modular reactors is emphasized by Professor Marco

K. Koch, head of the working group Plant Simulation and

Safety at the Ruhr University Bochum, who is also a board

member of the German Nuclear Society (KTG): “Compliance

with all safety standards, including safe nuclear fuel

management, is absolutely imperative”. The expert also

highlighted the advantages of SMRs in this aspect:

“ Depending on the design chosen, it is possible to increase

the safety of small modular reactors by combining active

and passive safety systems. Due to the smaller size

and thus the lower capacity compared to today's power

reactors, in the event of a hypothetical accident, SMRs

have greater capabilities in terms of external cooling, as

well as a higher dynamics of reactor start-up and shutdown.

In addition, due to the lower inventory, absolutely less

­fission products are produced”.

Another important feature of the FPU, which determines

the critical importance of technology for energy

supply to hard-to-reach areas, is its environmental

friendliness. Every day of the FPU operation, either directly

or indirectly due to gas savings, reduces annual consumption

to 200,000 tons of coal and 120,000 tons of fuel oil.

This seems particularly relevant in the light of the global

goals of the Paris Climate Agreement. As part of the fight

against climate change, the Russian side plans to reduce

greenhouse gas emissions by 2030 to 70 percent of the

1990 baseline. At the same time, the only way to achieve

these goals, in terms of the energy sector, is to implement a

program for the development of carbon-free energy.

“ Provided safety aspects are taken into account, small

modular reactor technologies are an environmentally

friendly alternative to energy supply due to the use of

smaller areas and the absence of CO 2 emissions”, agrees

Prof. Marco K. Koch.

The floating power unit Akademik Lomonosov is the first

representative in a series of plants, whose production is

planned to be established in the future, not least for

exports to other countries. “SMR concepts can really be of

interest for countries with decentralized energy supply”,

says Prof. Thomas Schulenberg, director of the Institute

of Nuclear and Energy Technologies at the Karlsruhe Institute

of Technology. “Decentralized energy supply should be

understood as an energy grid that is not interconnected, as

in Europe, but limited to small areas – for example, in

island regions such as Indonesia, or in sparsely populated

regions on land”, the professor explained.

The expert's words have been confirmed by real

experience: Director General of Rosatom Alexei Likhachev

noted interest in the new Russian development coming

from island states, including in South-East Asia. “In the

near future, we plan to move to negotiations on specific

­deliveries, and if the result is achieved, sufficiently large

capacities of Russian shipbuilding will be loaded with

orders”, he added.

Prof. Marco K. Koch notes that small modular reactors

can be used both in countries that already have nuclear

infrastructure on their territory and in the countries that

are new to the industry. Another significant argument

in favor of the development of these technologies is

­significantly lower financial costs compared to large ­energy

facilities. In Prof. Schulenberg’s view, a developing country

is very difficult to find an amount of 10 billion euros for the

construction of a large nuclear power plant – it is much

easier to get a loan for the amount of an order of magnitude

less. These circumstances lead to the conclusion that the

use of small modular reactors in floating power plants is

able to open a wide potential not only for energy supply to

remote regions, but also for expanding the club of states

using atomic energy for peaceful purposes.

Author

Roman Martinek

Expert for Communication

Czech Republic

ENERGY POLICY, ECONOMY AND LAW 439

Energy Policy, Economy and Law

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

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440

SPOTLIGHT ON NUCLEAR LAW

Atomausstieg letzter Akt?

Sind die neuen Entschädigungs regelungen für frustrierte

Aufwendungen und nicht mehr verstrombare Elektrizitätsmengen

im Atomgesetz verfassungsgemäß?

Tobias Leidinger

Kurz vor knapp hat der Gesetzgeber auf die verfassungsrechtlichen Mängel reagiert, die das Bundesverfassungsgericht

(BVerfG) in seinem Urteil vom 6. Dezember 2016 zum Atomausstieg (BVerfGE 143, 246) höchstrichterlich

beanstandet hat. Doch die neu geschaffenen Entschädigungsregelungen in der 16. AtG-Novelle werfen neue

Rechtsfragen auf, insbesondere die nach ihrer Verfassungsgemäßheit.

I. Die Vorgaben des Bundesverfassungsgerichts

Nach dem BVerfG-Urteil vom 6. Dezember 2016 musste

der Gesetzgeber bis zum 30. Juni 2018 in Bezug auf

den Atomausstieg einen verfassungsmäßigen Zustand

herstellen (vgl. dazu Leidinger, atw 2017, S. 26 ff.). Dies

erfolgt jetzt durch Entschädigungsregelungen, die durch

das Sechzehnte Gesetz zur Änderung des Atomgesetzes

(16. AtGÄndG), in das Atomgesetz eingefügt werden (vgl.

BT-Drs. 19/2508). Da das Änderungsgesetz im Hinblick

auf seine beihilferechtlichen Auswirkungen noch der

Überprüfung durch die EU-Kommission bedarf, kann das

Gesetz, das vom Bundestag am 28. Juni 2018 beschlossen

wurde, nicht sofort in Kraft treten.

Das Bundesverfassungsgericht hatte eine Kompensation

in zweifacher Hinsicht gefordert: Zum einen bedarf es

eines angemessenen Ausgleichs für frustrierte Aufwendungen,

die die Betreiber im Vertrauen auf den

Bestand der Ende 2010 zusätzlich gewährten Elektrizitätsmengen

getroffen hatten. Zum anderen ist eine Kompensationsregelung

für die Strommengen erforderlich, die

den Betreibern 2002 im Rahmen des „Energiekonsens“

(Atomausstieg I) zugestanden worden waren, die aber

nunmehr – infolge des endgültigen Atomausstiegs II bis

Ende 2022 – nicht mehr konzernintern verstromt werden

können. Letzteres betrifft allein die Betreiber Vattenfall

und RWE. E.ON verfügt noch über freie Kapazitäten, auch

wenn sämtliche eigenen Mengen verstromt sind. EnBW ist

nach eigenen Angaben nicht betroffen.

Neben dem Deutschen Bundestag hat sich auch der

Bundesrat mit den Regelungen befasst (BR-Drs. 205/18).

Auch eine Sachverständigenanhörung hat es dazu am

13. Juni 2018 im Umweltausschuss des Bundestages

gegeben. Die vom Bundesrat erhobene Forderung, im

Rahmen der gesetzlichen Neuregelung sicherzustellen,

dass Rest strommengen nicht auf norddeutsche Kernkraftwerke

(z.B. Emsland, Brokdorf) im Netzausbaugebiet

übertragen werden dürfen – weil dann die Einspeisung

regenerativer Energien eingeschränkt werde –, hat die

Bundesregierung – zu Recht – zurückgewiesen (BT- Drs.

19/2705). Eine solche Einschränkung von Übertragungsmöglichkeiten

müsste zu weiteren, nicht mehr

erzeugbaren Elektrizitätsmengen führen. Das wirft

erneut verfassungsrechtliche Fragen auf, insbesondere

nach ­einem finanziellen Ausgleich. Im Ergebnis käme es

zu einer noch größeren Belastung für den öffentlichen

Haushalt.

II. „Angemessenheit“ der Kompensation

von zentraler Bedeutung

Von entscheidender Bedeutung ist, ob durch die

neuen Entschädigungsregelungen die verfassungsrechtlich

ge botene Angemessenheit in Bezug auf frustrierte

Auf wendungen und nicht mehr verstrombare Strommengen

hergestellt wird. Denn die „Angemessenheit“

des Ausgleichs ist vom Bundesverfassungsgericht als

zentrales Kriterium einer verfassungskonformen Regelung

bestimmt worden. Fehlt es daran, wären die vom BVerfG

aufgestellten Maßgaben verletzt. Fraglich ist also, ob der

Gesetzgeber das ihm insoweit zukommende Gestaltungsermessen

verfassungskonform ausgeübt hat.

Für den Ausgleich nicht verstrombarer Strommengen

hatte das Gericht drei verschiedene Optionen eröffnet:

Zunächst wäre eine zeitlich auskömmliche Laufzeitverlängerung

bis zu dem Zeitpunkt denkbar, in dem die

ausgleichspflichtigen Strommengen tatsächlich konzernintern

verstromt sind. Das wäre – aus Sicht des Steuerzahlers

– der mit Abstand kostengünstigste Weg. Er wurde

indes nicht beschritten. Es bleibt vielmehr dabei, dass

die Nutzung der Kernenergie „zum frühestmöglichen

Zeitpunkt beendet werden soll“, d.h. es wird am Enddatum

31. Dezember 2022 unverändert festgehalten. Dieses

Datum beruht indes auf einer rein politischen Festlegung,

die bereits in der 13. AtG-Novelle im Jahr 2011 („Atomausstiegsgesetz“)

vorgenommen wurde. Sodann besteht die

Option, eine Weitergabemöglichkeit von Reststrommengen

zu ökonomisch zumutbaren Bedingungen gesetzlich

sicherzustellen oder – als dritte Möglichkeit – einen

angemessenen finanziellen Ausgleich für konzernintern

nicht verstrombare Reststrommengen zu gewähren.

III. Ausgleich für nicht mehr verstrombare

Elektrizitätsmengen

Das neue Gesetz bestimmt mit § 7f AtG (neu) einen

lediglich „konditionierten“ Geldausgleich für nicht mehr

verstrombare Elektrizitätsmengen. Danach müssen sich

die Kraftwerksbetreiber mit nicht verstrombaren Elektrizitätsmengen

zunächst, d.h. primär „ernsthaft darum

bemühen“, diese Mengen an andere Kraftwerksbetreiber

„zu angemessenen Bedingungen zu übertragen“, die zwar

noch über Kernkraftwerke, aber nicht mehr über Elektrizitätskontingente

zur Verstromung verfügen. Nur wenn und

soweit Strommengen zu diesen Bedingungen nicht mehr

übertragen werden konnten, greift dann – sozusagen

­subsidiär – eine finanzielle Kompensation.

Es ist mehr als fraglich, ob das Gesetz mit dieser

Regelung den höchstrichterlichen Vorgaben gerecht wird:

Der vom Bundesverfassungsgericht festgestellte Verstoß

gegen Art. 14 Abs. 1 (Eigentum) und das Gleichheitsgebot

aus Art. 3 Abs. 1 GG resultiert doch gerade daraus, dass es

aufgrund des Ausstiegsgesetzes (13. AtG-Novelle) zu

einem Nachfragemonopol hinsichtlich der nicht mehr

verstrombaren Mengen kommt, also einer Situation, die

per se keine „angemessenen Bedingungen“ für eine

konzernübergreifende Veräußerung der Strommengen

zulässt (vgl. BVerfGE 143, 246 (361)).

Spotlight on Nuclear Law

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

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| | Blick auf den oberen Teil des Reaktordruckbehälters eines Kernkraftwerks in Deutschland während der Revision mit Brennelementwechsel.

SPOTLIGHT ON NUCLEAR LAW 441

Der Gesetzgeber hat sich damit für ein Regelungsmodell

entschieden, das die verfassungsgerichtliche Kritik

am Atomausstieg im Kern ignoriert: Die in ihren Grundrechten

verletzten Konzerne werden nicht etwa entschädigt,

sondern sollen ihre Reststrommengen zu

Bedingungen verkaufen, die das BVerfG als unzumutbar

und gleichheitswidrig qualifiziert hat.

Hinzu kommt, dass das Gesetz keine Regelungen trifft,

die Angemessenheit des Ausgleichs auf der Ebene der

Anteilseigner zu schaffen, sondern es stellt insofern allein

auf die Genehmigungsinhaber ab. Die Feststellungen des

BVerfG bezogen sich indes auf die beschwerdeführenden

Konzerngesellschaften RWE und Vattenfall, die an vorzeitig

abgeschalteten Anlagen wie Krümmel oder in ihren

Laufzeiten verkürzten Anlagen wie Gundremmingen

beteiligt sind. Diese Regelung führt dazu, dass Ansprüche

der Genehmigungsinhaber auf Ausgleich bei den Gemeinschaftsunternehmen,

an denen Vattenfall beteiligt ist, in

Höhe dieser Beteiligungsquote gekürzt werden. Es ist

fraglich, ob die so konzipierte Regelung den Vorgaben des

Urteils entspricht. Das BVerfG hatte es dem Gesetzgeber an

sich leicht gemacht, indem es die verfassungswidrige

Benachteiligung von RWE und Vattenfall in Bezug auf die

Reststrommengen konkret beziffert hatte: Für RWE waren

40 TWh und für Vattenfall 46 TWh bestimmt worden. Die

Gesetzesregelung bleibt hinter diesen höchstrichterlichen

Vorgaben zurück.

Schließlich führt die Entschädigungsregelung in § 7f

dazu, dass die genaue und endgültige Festsetzung des

Ausgleichs erst nach der Abschaltung des letzten deutschen

Kernkraftwerks mit Ablauf des 31. Dezember 2022

erfolgen kann. Das bedeutet weitere Rechtsunsicherheit

für die Ausgleichsberechtigten, denn die behördliche

Entscheidung darüber, ob die Übertragungsangebote

„ angemessen“ sind bzw. waren, ergeht erst nach dem

31. Dezember 2022 – im Zusammenhang mit der Entscheidung

darüber, ob und in welcher Höhe ein Ausgleich

gewährt wird. Wenn sich dann herausstellt, dass ein

Ausgleichsberechtigter die Übertragung zu für den Übernehmenden

günstigeren Konditionen hätte anbieten

müssen, ist sein Ausgleichsanspruch insoweit ausgeschlossen.

IV. Ausgleich für frustrierte Aufwendungen

§ 7e AtG (neu) sieht einen angemessenen Ausgleich für

Investitionen vor, die Kraftwerksbetreiber im Vertrauen

auf die Ende 2010 zusätzlich gewährten Elektrizitätsmengen

getroffen haben. Das Bundesverfassungsgericht

hat das für eine Kompensation relevante „berechtigte

Vertrauen“ auf die Zeit vom 28. Oktober 2010 bis zum

16. März 2011 beschränkt. Dabei kommt es nicht auf den

Zeitpunkt der Leistungserbringung, sondern den der

Vermögensdisposition an, z.B. die Eingehung einer vertraglichen

Verpflichtung. Das im Gesetz formulierte

Kausalitätserfordernis zwischen dem Entzug der 2010

gewährten Zusatzmengen und der Frustration von

Investitionen ist dem Wortlaut nach zu eng gefasst.

Investitionen sind zu berücksichtigen, wenn die Zusatzmengen

dafür ein tragender, nicht aber der alleinige

Grund waren. Auch der jetzt normierte Verweis im Atomgesetz

auf den Rechtsgedanken des § 254 BGB (Mitverschulden)

wirft für die Rechtsanwendung praktisch

schwierige Abgrenzungs-, Bewertungs- und Beweisfragen

auf. Erschwert wird die Problematik dadurch, dass für den

auf die Kompensation gerichteten Ausgleichsantrag eine

Ausschlussfrist von nur einem Jahr ab Inkrafttreten der

neuen Regelung gilt.

V. Rechtsunsicherheit verbleibt

Mit der Neuregelung der §§ 7e-g AtG verbleiben mithin

erhebliche Unsicherheiten: Sie resultieren nicht nur aus

einer Reihe neuer Begriffe, sondern vor allem aus dem

vom Gesetzgeber für die Entschädigung der nicht mehr

verstrombaren Reststrommengen gewählten „konditionierten

Entschädigungsmodell“; das so keiner der vom

Bundesverfassungsgericht eröffneten Regelungsoptionen

entspricht. Damit ist weiterer Streit über den Atomausstieg

vorprogrammiert.

Author

Prof. Dr. Tobias Leidinger

Rechtsanwalt und Fachanwalt für Verwaltungsrecht

Luther Rechtsanwaltsgesellschaft

Graf-Adolf-Platz 15

40213 Düsseldorf

Spotlight on Nuclear Law

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

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

442

FUEL

Innovations for the Future

Westinghouse EnCore® Accident

Tolerant Fuel

Gilda Bocock, Robert Oelrich, and Sumit Ray

EnCore® and

ADOPTTM are trademarks

and registered

trademarks of Westinghouse

Electric

Company LLC, its

affiliates and/or its

subsidiaries in the

United States of

America and may be

registered in other

countries throughout

the world. All rights

reserved. Unauthorized

use is strictly prohibited.

Other names

may be trademarks of

their respective owners

The development and implementation of accident tolerant fuel (ATF) products, such as Westinghouse’s EnCore® Fuel,

can support the long-term viability of nuclear energy by enhancing operational safety and decreasing energy costs. The

first introduction of Westinghouse EnCore Fuel into a commercial reactor is planned for 2019 as segmented lead test

rods (LTRs) utilizing chromium-coated zirconium cladding with uranium silicide (U 3 Si 2 ) pellets. The EnCore Fuel lead

test assembly (LTA) program, with LTAs planned for 2022 insertion, will introduce silicon carbide/silicon carbide

composite cladding with U 3 Si 2 pellets.

Over the past several years, the

Westinghouse EnCore Fuel features

have been tested in autoclaves, in

research reactors, at national laboratories

and in the Westinghouse Ultrahigh

Temperature Test Facility to

­confirm and fully understand the

science behind ATF materials. Based

on the positive results to date, fuel rod

and assembly design in preparation

for the LTR and LTA programs is

underway, as well as licensing efforts

with the U.S. Nuclear Regulatory

Commission (NRC). Accident analyses,

coupled with economic evaluations,

have been continuing to define the

value of ATF to utilities.

These new designs will offer

design- basis-altering safety, greater

uranium efficiency and significant

economic benefits. Adoption of the

Westinghouse ATF, in conjunction

with a transition to 24-month cycle

operation, is the recommended path

forward for implementation of the

Westinghouse EnCore Fuel.

1 Introduction

Nuclear energy remains a fundamental

component of many industrialized

nations’ energy supply mixes due to its

demonstrated reliability in baseload

electrical supply, as well as inherent

carbon-free energy production. Two

factors are critical to maintaining this

capability: (a) enhancing safety to

help safeguard the plant and public

from highly impacting events such as

that which occurred at the Fukushima

Daiichi Nuclear Power Plant and (b)

decreasing operating costs to compete

with other sources of energy. The

development and implementation

of Accident Tolerant Fuel (ATF) products,

such as Westinghouse’s EnCore®

Fuel features, can support both of

these critical factors for long-term

operation.

Development of nuclear fuels with

enhanced accident tolerance is being

accelerated to support implementation

into commercial reactors as soon

as possible. The major objectives for

ATF designs include: 1) improved

cladding reaction to high-temperature

steam; 2) reduced hydrogen generation;

and 3) reduced beyond design

basis accident source term. In addition

to improving safety margins

for light water reactors (LWRs), fuel

designs using advanced, ATF materials

can improve fuel efficiency, ­enhance

debris resistance and extend fuel

management capability. Encore Fuel,

being developed by Westinghouse

Electric Company LLC (Westinghouse),

includes two unique accident tolerant

or fault tolerant fuel designs: chromium

(Cr)-coated zirconium (Zr)

alloy cladding with uranium silicide

(U 3 Si 2 ) fuel pellets, and silicon

carbide (SiC) cladding with U 3 Si 2 fuel

pellets.

The first introduction of Westinghouse

EnCore Fuel into a commercial

reactor is planned for 2019 as segmented

lead test rods (LTRs). The

LTRs will utilize chromium-coated

zirconium cladding with U 3 Si 2 highdensity,

high-thermal conductivity

pellets. The EnCore Fuel lead test

assembly (LTA) program, planned

for 2022 insertion, will introduce

SiC/SiC composite cladding along

with chromium- coated zirconium

cladding and the high-density, /highthermal

conductivity U 3 Si 2 pellets

modified to achieve higher oxidation

resistance.

Over the past several years,

Westinghouse’s ATF test program

has tested the chromium-coated

zirconium and SiC claddings in

autoclaves and in the Massachusetts

Institute of Technology’s (MIT) reactor

and U 3 Si 2 pellets in Idaho National

Laboratory’s (INL) Advanced Test

Reactor (ATR). Tests in the Ultrahigh

Temperature Test Facility at

Westinghouse’s U.S. Materials Center of

Excellence Hot Cell Facility in Churchill,

Pennsylvania, have been carried out to

confirm the time and temperature

limits for the SiC and chromiumcoated

zirconium claddings. Additionally,

an extensive research program to

fully understand the science behind

ATF materials continues with the

Westinghouse-led International Collaboration

for Advanced Research on

Accident Tolerant Fuel (CARAT) group

and at United States (US) and United

Kingdom (UK) national laboratories.

Based on the positive results to date,

fuel rod and assembly design in preparation

for the LTR and LTA programs is

underway, as well as licensing efforts

with the U.S. Nuclear Regulatory Commission

(NRC), and accident analyses

coupled with economic evaluations

for both operating savings and fuel

savings have been continuing to define

the value of ATF to utilities.

2 Lead test rod program

LTR programs are an essential step in

the introduction of new nuclear fuel

technologies into commercial energyproducing

reactors. In the EnCore LTR

program, two Westinghouse 17x17

optimized fuel assemblies (OFA) will

contain up to 20 ATF rods with

Cr-coated Zirconium alloy cladding,

and U 3 Si 2 and enhanced ADOPT fuel

pellets in Exelon’s Byron Unit 2 in

Cycle 22. Coated tubes and U 3 Si 2 and

ADOPT pellets will be delivered to the

Westinghouse Columbia Fuel Fabrication

Facility for manufacturing of

the assemblies. The shipping date for

the assemblies containing the LTRs is

February, 2019.

Westinghouse is continuing development

work with the University of

Wisconsin-Madison to continue the

optimization of coating performance,

and also working with commercial

vendors and the U.S. Army Research

Lab (ARL) to scale-up production

to full-length tubes. The U 3 Si 2 fuel

Fuel

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Material Process Vendor Maximum

Days

Titanium Nitride/ Titanium

Aluminum Nitride

Average

Corrosion Rate

(mg/dm 2 /day)

Std. Dev.

Corrosion Rate

(mg/dm 2 /day)

Average

Zr Corrosion

(mg/dm/day)

Corrosion

Rate

(microns/year)

PVD* PSU** 169 1.07 0.80 2.22 7.67

Chromium Cold spray UW*** 20 0.03 0.06 3.27 0.14

*Physical Vapor Deposition **Pennsylvania State University ***University of Wisconsin

| | Tab. 1.

Autoclave Corrosion Performance for the Top Zirconium Alloy Coatings.

FUEL 443

pellets are being fabricated at INL. The

fuel rod and fuel assembly designs are

progressing and the manufacturing

plan is being refined.

3 Recent testing

3.1 Autoclave testing of ATF

claddings

A primary benefit of ATF coating is to

enhance survivability in high-temperature

steam or water conditions, as

may occur in postulated accident

scenarios. To demonstrate this improved

survivability, Westinghouse

has performed corrosion testing using

the autoclave facility at the Churchill,

Pennsylvania site to screen various

coatings and SiC preparation methods

for corrosion resistance. As part of a

multi-year program, more than 12

types of coatings on zirconium alloys

and approximately 10 versions of SiC

have been tested in autoclaves. As a

result of this testing, two coatings,

titanium-nitride/titanium-aluminumnitride

and chromium (Table 1), were

identified for testing in the MIT

reactor.

Testing in the MIT reactor further

narrowed the options to the chromium

coating (Figure 1). The chromiumcoated

zirconium showed no signs of

peeling and had minimal weight gain

after taking into account the uncoated

inner surface of the tube. The very

positive results from these tests helped

validate the viability of the Cr coating

for use in LTRs being inserted in a

commercial pressurized water reactor

(PWR).

Initial autoclave and reactor testing

resulted in relatively high levels of

SiC corrosion. Autoclave testing with

hydrogen peroxide was used to simulate

the more aggressive oxidation

conditions of the reactor and to

explore coolant conditions that would

minimize SiC corrosion rates. This

testing has been used to refine the

manufacturing parameters of the SiC

composites such that, along with

hydrogen addition to the primary

coolant, above 40 cc/kg [2], the

current corrosion rates for SiC meet or

exceed the target 7of microns/year

recession rate. For a full core of

SiC cladding, this would result in a

maximum of 150 kg of silicon dioxide

(SiO 2 ) or about 350 ppm over an

18-month cycle. This is well below the

solubility limit of ~700 ppm SiO 2 at

the coldest steam generator conditions.

Note also that commercially

available resins to remove SiO 2 could

be added to the current resins used

to maintain water chemistry on a

continuous basis.

In addition to corrosion resistance,

crud buildup on the outside surface

of fuel rod claddings has long been

identified as a potential factor in fuel

rod operation, especially at higher

operating temperatures. Westinghouse

continues to assess the potential for

crud buildup on advanced ATF claddings.

Limits on crud buildup on SiC

claddings are likely to be different than

for coated claddings because the SiC

surface may be corroding underneath

any potential crud buildup. Therefore,

testing in the high heat transfer rate

and crud deposition test loop (WALT

loop) at the Westinghouse facility in

Churchill, Pennsylvania, has been

carried out from mid-2017 and will

continue until 2019 to study heat

transfer rates and crud buildup on the

SiC and chromium-coated cladding

surfaces. Preliminary results indicate a

somewhat higher crud deposition rate

for chromium-coated cladding than for

uncoated zirconium cladding. Surface

treatments are being explored to

reduce the crud deposition rate.

3.2 High-temperature testing

of ATF claddings

One goal of the ATF program is to

develop fuels that can withstand

post-accident temperatures greater

than 1,200 °C without the cladding

igniting in steam or air. Therefore, a

crucial part of the testing carried out

by Westinghouse during the previous

year was aimed at quantifying the

maximum temperature at which the

ATF claddings could operate without

excessive corrosion. The test apparatus

(Figure 2) currently uses a

graphite rod which is inserted into insulation

and then into the test piece.

| | Fig. 1.

Chromium-coated zirconium alloy tubes before and after testing in the MIT

reactor [1].

| | Fig. 2.

SiC rodlet undergoing testing in the ultra-high

temperature apparatus in steam at 1600°C at

Churchill. The sample is mounted inside the

shield tube that is glowing white in the photograph.

The SiC tube is inside the shield tube

with steam injected both above and below

the sample. The steam exits through the hole

that is visible in the shield tube.

This results in a very stable heating of

the test pieces.

Chromium-coated zirconium has

now been tested at up to 1,500 °C. This

is above the chromium- zirconium low

melting eutectic point of 1,333 °C. At

1,400 °C, there was noticeable reaction

between the Cr and the Zr. However,

there was not the rapid oxidation that

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FUEL 444

uncoated zirconium experiences at

1,200 °C. At temperatures of 1,300 °C,

however, the Chromium- coated zirconium

alloy was stable for reasonable

lengths of time. Combined with the

lowering of zirconium oxidation at

normal operating temperatures, which

vastly reduced the formation of zirconium

hydrides, and therefore embrittlement,

the chromium- coated zirconium

provides significant performance

improvements during normal operation,

transients, design basis accidents

and beyond design basis accidents, as

compared to uncoated zirconium.

Similar tests were run with SiC at

temperatures from 1,600 °C up to

1,700 °C. These tests were terminated

only because of excessive corrosion of

the heater element. At 1,600 °C, the

SiC cladding was visually untouched.

At 1,700 °C, there were indications of

small beads on the surface, presumably

SiO 2 from the reaction of SiC

with steam, but on the whole, no

­significant deterioration of the SiC.

Changes are being made to the heating

rod to increase the flow of Helium

cover gas and to allow accurate weight

changes to be made on the SiC rodlets

so that kinetic data can be obtained.

3.3 Testing of Westinghouse

U 3 Si 2 ATF high-density fuel

U 3 Si 2 is a revolutionary material for

LWR fuel service because its inherent

thermal conductivity is much greater

than existing UO 2 -based fuel, resulting

in significantly lower pellet temperatures.

U 3 Si 2 -based fuel can also

have up to 17 percent greater uranium

density than UO 2 -based fuel, so considerably

more energy can be economically

realized from each individual

fuel assembly. However, due to

these differences, considerable data

is required on the behavior of U 3 Si 2

at LWR operating temperatures

(estimated to be from 600 °C and up to

1,200 °C during transients).

To obtain the necessary data,

U3Si2 fuel pellets were manufactured

at INL and put into rodlets in the ATR

in 2015. The first rodlets came out of

the ATR at the end of 2016 (Figure 3)

and post-irradiation examination

(PIE) was performed in the summer

of 2017 at INL [3]. The PIE results

indicate some small amount of

cracking that may have been due to

impurities within the U 3 Si 2 . Fission

gas release and swelling were both

essentially zero with an exit burnup

of 20 MWd/kgU. Considering the

ATR high heat generation rates (12 to

15 kW/ft), which are significantly

above the average of 5 kW/ft and peak

of 9 kW/ft normally found in LWRs,

this was exceptionally good behavior.

The next set of U 3 Si 2 pins is due out in

2018 and will have achieved a burnup

of 40 MWd/kgU.

U 3 Si 2 was tested for air and steam

oxidation and compared to UO 2 using

digital scanning calorimeters at both

the Westinghouse Fuel Fabrication

Facility in Columbia, South Carolina

(USA) [4] and at Los Alamos National

Laboratory (LANL) [5]. The Westinghouse

test results indicate that the

ignition temperatures for UO 2 and

U 3 Si 2 are between 400 °C and 450 °C.

The LANL results indicate an ignition

temperature of about 400 °C. The

reasons for this difference are being

studied. The heat and mass generated

by the oxidation of the U 3 Si 2 is considerably

higher than for UO 2 . The

effect of this difference in heat release

and mass on the stability of the rods

was investigated in rodlet tests in the

autoclaves in the Churchill facility

during the summer of 2017. Unacceptable

tube bulging was found and programs

are now underway to increase

the oxidation resistance of the U 3 Si 2 .

| | Fig. 3.

Neutron radiographs of 20 MWd/kgU U3Si2 pins from ATR. Note the lack of pellet cracking and

distortion.(Ref. 4).

It is noted, however, that ATF cladding

surfaces are much harder than zirconium

alloy cladding and grids, so it is

expected that the likelihood of grid to

rod fretting leakages will be greatly

reduced from the current ppm levels.

4 Accident scenario

evaluations

To assess and demonstrate the performance

of ATF materials in postulated

accident scenarios, Modular Accident

Analysis Program, Version 5 (MAAP5),

calculations were performed for chromium-coated

zirconium and SiC claddings

along with high-density fuels

for the station blackout scenario and

the Three Mile Island Unit 2 (TMI2)

small-break loss-of-coolant (LOCA)

scenario with replenishment of the

primary coolant [6].

The chromium-coated zirconium

option offers modest ATF gains

(~200 °C) before large-scale melting

of the core begins in beyond design basis

events, such as a long-term station

blackout. Though it would not prevent

the contamination of the PWR primary

loop due to ballooning and bursting at

about 800 °C to 900 °C, the chromiumcoated

zirconium option could prevent

a TMI-2 type of accident from extending

into the fuel meltdown phase and

prevent extensive contamination of

the containment and perhaps preserve

the nuclear plant. This is because,

although the Cr-coated Zr may begin

to fail as the temperature exceeds

1,400 °C due to eutectic formation, it

does not rapidly oxidize as uncoated

zirconium alloys do, and does not provide

a rapid energy input spike into the

core (Figure 4). Note that, in this case,

Iron-chromium-aluminum (FeCrAl)

was used to model the performance

of chromium-coated zirconium since

the temperature and oxidation performance

is about the same. The

results for the station blackout

scenario (Figure 5) indicate that

­fission ­products can be contained

within SiC cladding for up to two

hours longer than current Zr-based

cladding due to its higher temperature

capability (~2,000 °C decomposition

temperature). These two hours can

be used to implement additional

responses by the operators. The lower

pressure in the system due to minimal

hydrogen production (Figure 6.) increases

the chances that alternate

means to feed cooling water to the

core at about 40 gpm can result in

avoidance of fuel melting, indefinitely

extending the coping time as long as

the water flow continues. The SiC

cladding, of course, prevents any

Fuel

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FUEL 445

| | Fig. 4.

Hottest core node for TMI-2 accident where coolant is restored

at ~9,900 seconds.

| | Fig. 5.

Hottest core node for PWR station blackout.

­leakage of fission products into the

primary loop since it will not balloon

and burst. Due to the short timespan

before coolant was re-introduced to

the system, the SiC cladding would

have had no adverse consequences

from a TMI-2 type accident (Figure 4).

5 Transition cycle analysis

for optimum ATF

implementation in

current PWRs

5.1 U 3 Si 2 fuel

As previously noted, one of the

­primary benefits of U 3 Si 2 is that it

increases the uranium density by up

to 17 percent as compared to UO 2 .

This yields an effective enrichment

of 0.84 weight percent U-235 as

compared to 0.71 weight percent

U-235 found in natural uranium. This

increase in density will support

improved fuel cycle economics and

reduce the total number of fuel

bundles that need to be inserted into

a reactor, resulting in significant

savings. Because of the increased

density, the use of U 3 Si 2 also extends

the energy output and cycle length

capability for PWR fuel assemblies,

while remaining below the 5 weight

percent enrichment limit for commercial

fuel. The Westinghouse ATF can

thus either decrease the fuel cycle cost

of 18-month cycles by reducing the

number of feed assemblies and increasing

fuel utilization, or it can

make 24-month cycles economical for

today’s uprated, high-power density

PWRs.

Economic analysis shows that the

Westinghouse EnCore Fuel has very

favorable economics, not only at the

ATF equilibrium cycle, but also during

the transition cycles from UO 2 to ATF.

This is especially applicable when

transitioning to a 24-month cycle

operational regime, which thus represents

the recommended path forward

for implementation. The higher

thermal conductivity of the U 3 Si 2 also

provides a very high tolerance for

transients while operating at higher

linear heat generation rates than is

possible for UO 2 – which will increase

plant operating margin. In addition,

the higher uranium density can

extend the core operating capability

compared to current fuels, while

maintaining the current 5 weight

percent 235U enrichment limit for

commercial fuel; yet enable economically

competitive fuel management

schemes for the longer cycles.

In particular, the introduction of

ATF in a current 18-month cycle

high-power density PWR to accomplish

a transition from UO 2 to ATF by

either maintaining the currently predominant

18-month cycle operational

regime, or extending it to a 24-month

cycle has been analyzed. Implementing

the Westinghouse ATF to achieve a

more cost effective 18-month cycle

will deliver fuel cost savings due to

fewer fresh assemblies per reload

and improved fuel utilization. Implementing

the Westinghouse ATF in

conjunction with a transition to

24-month cycle will yield economic

benefits due to the resulting reduced

number of outages and related

savings, which offset the slightly

higher fuel costs (as compared to

18-month cycle fuel costs). Analyses

have shown that the economic impact

of the transition cycles to implement a

24-month cycle operation with ATF is

significantly better than the economic

impact of transition cycles which implement

ATF and maintain an

18-month cycle operation.

It is anticipated that the fabrication

costs to make the U 3 Si 2 powder could

increase as compared to existing

UO 2 fabrication. However, after the

powder is made, only minor cost increases

are expected to occur in the

rest of the fuel manufacturing process.

Therefore, the overall cost increase

is anticipated to be offset by

the safety, economic and operational

benefits.

| | Fig. 6.

Total hydrogen generated for PWR station blackout.

5.2 Chromium-coated

zirconium cladding

Chromium-coated zirconium alloy

offers a higher accident temperature

capability, compared to uncoated

zirconium alloy cladding, of between

1,300 ˚C and 1,400 ˚C. The coated

cladding also reduces corrosion and

hydrogen pickup. Resistance to rod

wear is another benefit of this cladding

type. The potential for exothermic

reactions is greatly reduced

during LOCA or transient events

that lead to high-temperature fuel

transients. These attributes provide

both safety and economic benefits

that support licenseability and economically

viable transition scenarios.

5.3 SiC cladding

SiC cladding provides 25 percent lower

thermal neutron cross-sections

than current Zr cladding. This would

afford even greater neutron economy.

Additionally, the fuel and cladding

would be able to withstand temperatures

~2,000 ˚C in the event of a

beyond design basis accident. This

temperature increase could result in a

rise in design basis operating margins.

6 Licensing

To get EnCore Fuel licensed and

loaded into commercial reactor cores

in region quantities by 2027, Westinghouse

has initiated a program to

­significantly compress the licensing

timeframe from initial testing to

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446

OPERATION AND NEW BUILD

commercial delivery, while improving

the quality of the data and resulting

design models used to describe the

fuel. This new approach would be a

significant improvement compared

to the current, largely empirical

approach, which requires years to

obtain limited data from a very expensive

test reactor(s), as well as for the

fabricating, testing, cooling, transportation

and post-irradiation examination

of samples. To reduce the

licensing timeframe for EnCore Fuel,

Westinghouse plans to utilize:

• Atomic scale modeling:

• By utilizing first principles to

determine physical properties

of irradiated materials

• By leveraging Westinghouse

involvement in the Nuclear

Energy Advanced Modeling &

Simulation (NEAMS) Department

of Energy (DOE) program

on basic property prediction

• By leveraging Westinghouse

involvement in the Consortium

for Advanced Simulation of

Light Water Reactors (CASL) –

Virtual reactor design

• By continuing to utilize MedeA

and Thermo-Calc software

• Real-time data generation to verify

the atomic scale modeling:

• Poolside data generation

PP

Gamma emission tomography

based on gamma-ray spectroscopy

and tomographic reconstruction

can be used for

rod-wise characterization of

nuclear fuel assemblies without

dismantling the fuel to detect

pellet swelling, pellet- cladding

interaction and pellet cracking

PP

Potential use of a spectroscopic

detection system to select

different gamma-ray emitting

isotopes for analysis, enabling

nondestructive fuel characterization

with respect to a variety

of fuel parameters (fission gas

release)

• Wired or wireless transmission

technology for measuring

PP

Centerline temperature

PP

Fuel rod gas pressure

PP

Swelling of fuel

In addition to saving time and cost, with

this approach Westinghouse hopes to

achieve, an increased con­fidence by the

U.S. NRC due to the predictability of

performance that can be obtained since

the performance models will have a

theoretical basis in addition to an

empirical basis. There should also be

reduced time and effort due to the reduction

in the number of submissionreview-revision-

submission cycles. This

should remove the review process from

the critical path to commercialization.

Communication with the U.S. NRC

Commissioners, and coordination

between the DOE, NRC and industry

for licensing of ATF, are in progress

and continuing.

7 Conclusion

Westinghouse and its partners are

continuing to make good progress on

U 3 Si 2 fuel, SiC cladding, and chromium-coated

zirconium cladding. These

new designs will offer design-basisaltering

safety, greater uranium efficiency,

and significant economic

­benefits per reactor per year for PWRs.

While all testing and development to

date has been engineered for LWR

designs, Westinghouse believes the

technology could provide some of the

same safety and economic benefits to

CANDU and other reactor designs.

Fuel and accident modeling with

other types of reactor systems will be

required to evaluate the actual potential

for these benefits. This, together

with more beneficial power peaks,

lower impact of the transition cycles

and reduced dependence on uranium

price assumptions, make adoption of

the Westinghouse ATF, in conjunction

with a transition to 24-month cycle

operation, the recommended path

forward for implementation of the

Westinghouse ATF, EnCore Fuel.

References

[1] Gordon Kohse, MIT, 2016.

[2] Ed Lahoda, Sumit Ray, Frank Boylan,

Peng Xu and Richard Jacko, SiC Cladding

Corrosion and Mitigation, Top Fuel 2016,

Boise, ID, September 11, 2016, Paper

Number 17450, ANS, (2016).

[3] Jason Harp, Idaho National Laboratory

preliminary photographs.

[4] Lu Cai, Peng Xu, Andrew Atwood,

Frank Boylan and Edward J. Lahoda,

Thermal Analysis of ATF Fuel Materials

at Westinghouse, ICACC 2017, Daytona

Beach, FL, January 26, 2017.

[5] E. Sooby Wood, J.T. White and A.T.

Nelson, Oxidation behavior of U-Si

compounds in air from 25 to 1000 C,

Journal of Nuclear Materials, 484,

pages 245-257 (2017).

[6] Eugene van Heerden, Chan Y. Paik,

Sung Jin Lee and Martin G. Plys,

Modeling Of Accident Tolerant Fuel

for PWR and BWR Using MAAP5,

Proceedings of ICAPP 2017, Fukui and

Kyoto ,Japan, April 24-28, 2017.

Authors

Gilda Bocock

Robert Oelrich

Sumit Ray

Westinghouse Electric Company

5801 Bluff Road

Hopkins, SC 29061, USA

Analyses of Possible Explanations for the

Neutron Flux Fluctuations in German PWR

Joachim Herb, Christoph Bläsius, Yann Perin, Jürgen Sievers and Kiril Velkov

Revised version of a

paper presented at

the Annual Meeting

of Nuclear Technology

(AMNT 2017), Berlin.

During the last 15 years the neutron flux fluctuation levels in some of the German PWR changed significantly. During

a period of about ten years, the fluctuation levels increased, followed by about five years with decreasing levels after

taking actions like changing the design of the fuel elements [1, 2]. The increase in the neutron flux fluctuations resulted

in an increased number of triggering the reactor limitation system and in one case in a SCRAM [3].

There exist different possible explanations

how neutron flux oscillations are

caused by physical phenomena inside

a PWR. Possible explanations can be

based on complicated interactions

between thermo-hydraulical (TH),

structural-mechanical and neutron

physical processes (see Figure 1).

Yet, no comprehensive theory

exists, which can explain the neutron

flux fluctuation histories observed

in German PWR based on first ­

physical principles. Therefore, GRS

has started investigations to

explain the observed neutron flux

Operation and New Build

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

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| | Fig. 1.

Possible causes for neutron flux oscillations.

fluctuations and amplitude changes

[4].

Characteristics of neutron flux

fluctuations in German PWR

The neutron flux level during full

power operation is measured by inand

ex-core detectors sensitive to

thermal neutrons [5]. In German

Vorkonvoi and Konvoi type PWRs in

total 16 ex-core detectors (ionization

chambers) are located at four azimuthal

positions (see Figure 2) and at

four different axial heights outside the

RPV wall within the biological shield.

The signals of the upper two and lower

two are combined. They measure

the neutrons released from the core.

Inside the reactor core eight measurement

rods are located (see Figure 2).

Each measurement rod consists of six

self-powered neutron detectors (SP-

NDs) located at different elevations.

The neutron flux measurements used

in the following analyses have been

provided by an operator of a German

Vorkonvoi PWR. The data were sampled

with 250 Hz after they had been

low-pass filtered with a cutoff frequency

of 100 Hz.

Figure 3 (left) shows the power

spectral densities (per Hz) of the neutron

flux signals of ex-core detectors

measured in a Vorkonvoi PWR. The

power levels measured at the four

different azimuthal positions show no

significant differences. The highest

power spectral density of the neutron

flux is measured at low frequencies up

to 1 Hz.

Figure 3 (right) shows the distribution

of the time-dependent spectral

power density. For each time step,

a Fast-Fourier-Transformation was

calculated, using the following

parameters: sampling frequency =

250 Hz, numbers of samples = 4096,

Hanning window function. The time

steps in Figure 3 (right) are separated

by 14.3 s, which is 7/8 of the length of

a single FFT window (16.4 s). For each

point of time and frequency the

spectral power level is color coded.

The spectral power density changes

over time in a “chaotic” way. This

means that the frequency of the

maximal power density changes over

time. This observation does not

change if the number of samples used

for the FFT or the time resolution used

for the calculation spectrogram is

reduced or increased.

The top row of Figure 4 shows the

measured coherence and the phase

angles of two combinations of two different

ex-core detectors each. The coherence

was calculated by dividing

the absolute value squared of the cross

correlation of the corresponding two

detector signals by the autocorrelation

of the signals. Both detector

combinations show a strong coherence

at 1 Hz. The phase of the complex

valued frequency dependent

cross correlation was used to calculate

the frequency dependent phase shown

in Figure 4 (converted into units of

degree). The two detectors located at

perpendicular horizontal positions

relative to the core center (at 45° and

135°) exhibit constant phase difference

in the frequency range up to

1 Hz. In contrast, the two detectors

located at opposing sides of the core

center (at 45° and 225°) show a nearly

constant phase difference of 180° in

the frequency range up to 5 Hz. This

phase difference of 180° can be found

for all detector combinations calculated

by the cross correlation for two

detectors placed at opposing sides

relative to the reactor center.

The bottom row of Figure 4 shows

the relative signal strengths over time

of six in-core detectors located at different

axial heights on the Co4 measurement

rod (see Figure 2 for the positions

of the detectors). Even though

the amplitude at different elevations

shows some differences (higher amplitudes

at middle elevations than at

lower or higher ones) the temporal

OPERATION AND NEW BUILD 447

| | Fig. 2.

Horizontal positions of in-core (marked light

blue) and ex-core (marked red) detectors

within the core shroud and outside the reactor

pressure vessel.

| | Fig. 3.

Power spectral density (left) and spectrogram (right) of ex-core detector measurements in a Vorkonvoi PWR.

Operation and New Build

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

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OPERATION AND NEW BUILD 448

| | Fig. 4.

Coherence and phase angles between different ex-core detectors (top),

relative neutron flux measurements at six different elevations of the C04

in-core measurement rod (bottom); all measurements in a Vorkonvoi PWR.

progressions of the curves are identical

for all six elevations. It has to be

emphasized that no time lag can be

identified between measurements at

the bottom of the reactor core

compared with measurements at

the top. The same signal pattern can

be observed for all eight in-core

measurement positions.

All these observations are consistent

with different measurements

and analyses done during the last

decades [6, 7, 8]. Fiedler [8] compared

neutron flux fluctuation levels

in different plant types. He found that

the prominence of the 180° phase

difference between opposing detectors

(referred to as “beam mode”) is

special to KWU type PWRs.

Possible explanation based on

thermo-hydraulics effects

Already at the beginning of the

1970s, a model was published [9, 10]

coupling a point-kinetics neutron

physics model with a one-dimensional

TH model. It allows predicting neutron

flux fluctuation levels based

on coolant temperature or density

oscillations. Based on this model

it is already possible to understand

essential characteristics of the neutron

| | Fig. 5.

Simulated temperature fluctuations in frequency (top, left) and time (top, right) domain; layout of the coupled ATHLET-QUABOX/

CUBBOX model for a mini-core (bottom, left) and the resulting neutron flux fluctuations spectrum (bottom, right).

noise spectrum qualitatively, e. g. the

dependency of the neutron flux fluctuation

amplitude on the value of the

moderator temperature coefficient.

Following this approach and based

on some new simulations with the

CTF/PARCS codes [11, 12] a model of

the reactor core has been developed

using a coupled version of ATHLET

and QUABOX/CUBBOX [13]. In [12]

temperature fluctuations at the core

inlet were applied based on different

spectral properties. Temperature

oscil lations based on a white noise

spectrum resulted in much smaller

power/neutron flux oscillations than

temperature oscillations based on a

low-pass-filtered spectrum. A possible

explanation for that observation

might be alias-effects due to the limited

spatial and temporal resolution of

the coupled system. To avoid such

problems with the coupled system of

ATHLET and QUABOX/CUBBOX, a

Kolmogorov type spectrum [14] has

been applied for the temperature

­fluctuations at the inlet of the reactor

core. Figure 5 (top row, left) shows

the power spectral density of the

temperature oscillations over the

frequency. Such spectra were observed

in different reactors [15, 16,

17].

Based on the assumption that the

temperature fluctuations follow such

a Kolmogorov type spectrum the time

dependent temperature fluctuations

are calculated (Figure 5, top right).

The temperature fluctuations have the

same variance as a sine-wave with an

amplitude corresponding to 1 K.

The TH model layout is shown in

Figure 5 (bottom, left). It consists of

nine interconnected core channels

with common inlet and outlet thermofluid

elements. The mini core has a

typical neutron-physics characteristic

of an end of fuel cycle (EOC).

Figure 5 (bottom, right) shows the

power spectral density of the resulting

fluctuations in the reactor power

production, which is proportional to

the neutron flux amplitude. For frequencies

smaller than 3 Hz the calculated

power spectral density fits the

measured ex-core detector signals of a

Vorkonvoi PWR quite well over several

orders of magnitude. This suggests

that temperature fluctuations at the

inlets of the core channels are part of

the explanation. This model can also

explain the correlation between the

amplitude of the fluctuations and the

moderator temperature coefficient.

However, it is not possible to explain,

why no phase differences could be

observed between measurements of

Operation and New Build

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

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SPNDs at the same horizontal but

different axial locations. The transport

of temperature fluctuations

through the core channels should

result in a delay of measurements

between detectors in lower regions

and the upper regions of the core.

Furthermore, this approach cannot

explain the strict 180° phase differences

between measurement positions

at opposing sides of the reactor

(neither for ex-core nor for in-core

detector combinations). If this

approach should be continued sensitivity

studies on the parameters of

the applied Kolmogorov typed spectrum

will be necessary.

Possible explanations based

on mechanical motions

For decades, the analyses of neutron

flux fluctuations have been used for

the detection of mechanical oscillations

inside the reactor pressure

vessel, see e.g. [6, 8, 18]. However, the

mechanical oscillations considered in

these analyses are harmonic oscillations

with resonance frequencies

exceeding 2 Hz. Nevertheless, the

simultaneousness of detector signals

of one measurement rod as well as the

location of the maximum of the neutron

noise level in the middle of the

core height indicate that mechanical

motions in the reactor core, which

behave synchronous and without

phase differences over the full core

height, also contribute to the observed

fluctuations at low frequencies.

Point Source Model

To check whether the observed fluctuations

are consistent with a core wide

mechanical motion a model based on

a point source for the neutron flux has

been developed. The model is based

on the assumption that the signal

| | Fig. 6.

Moving point source model (yellow circles:

detectors used for trilateration, black star:

idle position of point source, blue star: point

position derived by trilateration, red circle:

estimation for position uncertainty).

| | Fig. 7.

Different detector combinations used for trilateration (left), estimated horizontal point source locations over time (right).

strengths at the detectors depend linearly

on the distances between the

point source and the detectors (see

Figure 6). Based on this assumption

the position of the point source can

be calculated by trilateration using

different detector combinations (see

Figure 7 left). Additionally, an estimation

of the uncertainty of the

assumed position of the point source

can be derived. The three combinations

considered here are the four

ex-core detectors (marked red), three

in-core detectors located at the left

side of the reactor core (marked

green), and three in-core detectors

located at the right side (marked

blue).

Figure 7 (right) shows for different

time steps the pathways of the

assumed point source calculated

by a combination of the four ex-core

detectors (red), three left in-core

detectors (green) and three right incore

detectors (blue). The position

calculated by the ex-core detectors is

scaled by a factor of 1/3 relative to the

center of the reactor core. Also shown

are the estimated uncertainties of the

point source position for the different

detector combinations.

The model results in consistent

point source location estimations for

the three detector combinations. Also

the estimated uncertainties are small

compared with the pathways of the

point source. If instead of the detectors

marked in Figure 7 (left) the two

inner-most detectors (J06, G10) are

included in the calculation of the

trajectories, no consistent trajectories

can be derived.

This indicates that a phenomenon

involving the full reactor core plays a

significant role for explaining the

­observed neutron flux fluctuations.

But it cannot explain the shape of the

measured power spectral density.

Structural-Mechanics

Considerations on Core-Wide

Motions of Fuel Assemblies

and further Core Internals

A synchronous excitation or synchronization

via mechanical coupling

can lead to core-wide correlated

mechanical motions of fuel assemblies,

which effect both in- and excore

neutron flux instrumentation.

This explanation is supported by

both the successful simulation of

the detector signals by an empirical

model of a moving point source and

the correlation between the neutron

fluctuation levels and the use of fuel

assemblies with reduced lateral

stiffness due to changes in the spacer

design. It also explains the simultaneity

of signals at different vertical

levels and the bow-shaped vertical

amplitude characteristic with a maximum

at or slightly below middle core

height.

Core barrel, grid plate and the

collective of fuel assemblies form

an enhanced system of coupled

mechanical oscillators. Core barrel

motions can have additional effects

on the neutron flux signal via

modulation of absorption and

­reflection in the ­water gap between

core barrel and reactor pressure

vessel. The fuel assemblies within

this coupled oscillator differ in type

and service time and thus mechanical

parameters, which can lead to chaotic

motions and interaction effects and

thus oscillations in a broad frequency

band. In a low-leakage loading pattern

the fuel assemblies with the longest

service time and lowest remaining

stiffness are located at the core

periphery, which can evoke additional

effects on the ex-core and outer

in-coresensors, e.g. via water gap

modulation or motion in a strong

flux gradient.

OPERATION AND NEW BUILD 449

Operation and New Build

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

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OPERATION AND NEW BUILD 450

There are three possibilities for an

excitation in general:

• Stochastic fluid forces from turbulent

flow would lead to oscillations

of components at their natural

frequencies. The lowest natural

frequency of the fuel assemblies is

reported around 2.6 to 4 Hz [6, 18],

which would not explain the coherence

maximum at 1 Hz. Calculations

with simplified finite element

models show that depending on

design and operational behavior,

i.e. lateral stiffness decrease due to

radiation induces spring relaxation

in the spacers, the lowest natural

frequency can be shifted significantly

towards lower values. In [6,

16] an additional mode of the fuel

assemblies around 1 Hz in form of

synchronously moving cantilevered

beams (fixed at the bottom) is supposed.

Nevertheless, regarding the

fixture of the fuel assemblies in the

grid plate, the manifestation of this

mode is questionable. A further

explanation is the excitation of the

coupled system of core barrel, grid

plate and fuel assemblies, which

might have additional natural

system frequencies below the natural

frequencies of the single fuel

assemblies.

• A second possibility would be the

existence of an excitation force,

which is oscillating at around 1 Hz

and evokes a subsequent transient

deflection of the fuel assemblies.

Pressure fluctuations from residual

imbalances of the coolant pump,

standing waves, cavity resonances

in the pressurizer or vibrations of

other components of the loop are

known to induce core barrel motions

which could propagate to the

fuel assemblies. Fluid mechanical

oscillating forces with direct effect

on the fuel assemblies, e.g. pressure

differences, are also possible.

• A third possibility would be a selfexcitation

of fuel assemblies in a

constant axial flow. Research on

fuel assembly bow gives hints that

in fluid-structure-interaction (FSI)

simulations local forces can arise

leading to instability of the zero

position of the fuel assembly [19].

To investigate and prove the mentioned

hypotheses, a coupled FSI

model of core components and the

surrounding fluid is essential.

Simulations of reflector

influence

Further, the reflector influence has

been studied by means of a simplified

2D core model, in which the reflector

Case description

Maximum (relative)

increase on the left side

cross-sections are manipulated in

order to simulate the effect of varying

water gap between core barrel and

reactor pressure vessel, which corresponds

to the reflector region. These

variations could be caused by mechanical

motions, e.g. of core barrel or

fuel assemblies at the core periphery,

and their effect increases with decreasing

boron concentration. In this

model the TH parameters are homogeneous

and representative of the

hot full power state at zero burnup.

Further assumptions are: fuel temperature

= 900 K, moderator density =

702 kg/m 3 and boron concentration

= 1,300 ppm.

Table 1 summarizes the results

obtained for different variations of

the thermal absorption and fast-tothermal

scattering crosssection. The

reflector is modified only in one half of

the core (the left side) to reproduce

the spatial oscillations observed in the

PWR. The results show that the effects

of thermal absorption and scattering

are additive. The amplitude of the

power variation can reach the same

order of magnitude as observed in the

PWR.

Additional study is necessary to

determine if actual mechanical motions

can cause such changes leading

to increase/decrease of the moderator

volume (coolant water) in the reflector

zone and in that way changing

the homogenized assembly crosssections.

In addition, time-dependent

simulations are needed to check if

the frequency observed in the PWR

can be reproduced. Nevertheless, this

preliminary result shows that this

hypothesis is very promising. The

recently published study [20] showed,

that a variation of the gap size

between fuel elements of about one

centimeter can result in changes

of the neutron flux amplitudes at

the ex-core detectors of up to the

order of magnitude of 10 %. Therefore,

the influence of mechanical

motions of the fuel elements relative

to each other and as an ensemble

­relative to the reflector cannot be

ruled out as explanation of the observed

neutron flux oscillations.

Summary and outlook

Several models based on single

­physical effects (TH fluctuations at

the core inlet, movement of a point

source, coupled oscillations of core

­internals, changes in the reflector

­coefficients) are used to simulate the

neutron flux. Each of these simple

models can reproduce some of the

characteristics of the observed neutron

flux fluctuations but does not

encompass all features observed in a

real reactor. This suggests that further

work on the combination of models

is needed. Thereby, the biggest challenges

will lie in FSI simulations of

fuel assemblies including further core

internals, neutron physics simulations

using time-dependent geometries,

and possibly the coupling of all three

physical models.

Acknowledgment

This work has been performed in the

framework of the German Reactor

Safety Research and was funded by

the German Federal Ministry for

Economic Affairs and Energy (BMWi,

project no. RS1533). The authors

would like to thank the operators of

one German Vorkonvoi PWR and one

Konvoi PWR for providing data of

­neutron flux measurements.

References

Maximum (relative)

decrease on the right side

-10 % thermal absorption 4 % -3 %

-10 % scattering 7 % -5 %

-10 % thermal absorption

-10 % scattering

10 % -8 %

-20 % thermal absorption 11 % -7 %

-20 % scattering 14 % -11 %

| | Tab. 1.

Summary of the reflector study results.

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.

Operation and New Build

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

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

4. C. Bläsius et al., Untersuchungen der

Ursachen für Neutronenflussschwankungen,

GRS-408, Gesellschaft

für Anlagen- und Reaktorsicherheit

(GRS) gGmbH, 2016.

5. G. Kaiser et al., Reaktorinstrumentierung.

Prozeßtechnik und Leistungsregelung im

Kernkraftwerk, VDE Verlag, 1983

6. J. Runkel, Rauschanalyse in

Druckwasserreaktoren, 1987.

7. L. J. Kostić, J. Runkel, D. Stegemann,

Thermohydraulics Surveillance of Pressurized

Water Reactors by Experimental

and Theoretical Investigations of the

Low Frequency Noise Field, Progress in

Nuclear Energy 21, pp. 421-430, 1988.

8. J. Fiedler, Schwingungsüberwachung

von Primärkreiskomponenten in Kernkraftwerken,

2002.

9. G. Kosaly, M. M. R. Williams, Point

theory of the neutron noise induced by

in-let temperature fluctuations and

random mechanical vibrations, Atomkernenergie

18(3) p. 203-208, 1971.

10. G. Kosaly., L. Mesko, Remarks on the

transfer function relating inlet temperature

fluctuations to neutron noise, Atomkernenergie

20(1), pp. 33-36, 1972.

11. A. Abarca et al., Analysis of Thermalhydraulic

Fluctuations in Trillo NPP with

CTF/PARCSv2.7 Coupled Code, 23 nd

International Conference Nuclear

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

OPERATION AND NEW BUILD 451

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

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OPERATION AND NEW BUILD 452

Revised version of a

paper presented at

the Annual Meeting

of Nuclear Technology

(AMNT 2017), Berlin.

Detailed Measurements and Analyses

of the Neutron Flux Oscillation

Phenomenology at Kernkraftwerk Gösgen

G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff

1 Introduction This paper summarises recent investigations [1], [2], [3] on measured neutron flux noise at

the Kernkraftwerk Gösgen-Däniken AG, who is operating since 1979 a German KWU pre-KONVOI, 3-Loop PWR with a

thermal power of 3,002 MWth (1,060 MWe). In a period of approx. 7 cycles from 2010 to 2016, an increase of the

­measured neutron noise amplitudes in the in- and out-core neutron detectors has been observed, although no ­significant

variations have being detected in global core, thermo-hydraulic circuits or instrumentation parameters. Verifications of

the instrumentation were performed and it was confirmed that the neutron flux instabilities increased from cycle to

cycle in this period. In the last two years, the level of neutron flux noise remains high but seems to have achieved a

saturation state.

In a power reactor, neutron noise is

the result of random fluctuations of

many parameters, primarily neutronic

ones such as the number of neutrons

emitted per fission, thermal-hydraulic

parameters such as the fluctuations of

the primary water inlet temperature,

and mechanical parameters as for

example main circulation pump vibrations

or core internal vibrations. In

a KWU-PWR as KKG, the significant

neutron noise is observed at a frequency

in the range of 0.1 Hz to about

10 Hz, with a peak close to 1 Hz. Each

component has a typical spectral

response in the frequency domain,

and such a spectrum analysis can be

used as a diagnostic tool for surveillance

[4]. A significant variation of

the measured spectrum during a cycle

can be potentially interpreted as of

relevance for the plant performance

or safety. For that reason the Reactor

Pressure Vessel (RPV) and main

| | Fig. 1.

Schematic representation of the 3002 MW 3-Loop KKG core and the radial

positions of the in-core (left white on the map) and ex-core neutron flux

detectors. The colour map shows the relative power map (Fq) at the

assembly level. The inner axial flux distribution is monitored via six axially

and uniformly distributed in-core Self-Powered Neutron Detectors, while

the four radial ex-core channels contain two compensated ionisation

chambers, i.e. for the upper and lower core regions.

cir culation pumps at KKG are

equipped with acceleration and absolute

position sensors.

To deepen the understanding of

this behaviour, neutron flux signals at

different core locations and burnup

have been newly measured at a

sampling rate up to 100 Hz in order to

analyse possible spatial correlations

between the measured signals. The

measurements corresponded to

Middle- of-Cycle (MOC) and End-of-

Cycle (EOC) conditions, for two

successive cycles aiming at analysing

noise evolution, additionally to the

known linear increase during the

cycle. During the cycle itself, the noise

amplitude increase is linearly correlated

to the decrease of the negative

moderator temperature reactivity

­coefficient (Γ T ), which is caused by

the decrease of the boron con centration

in the primary circuit; this

behaviour is well known and predictable.

The phenomena to be

investigated here is the variation from

cycle-to-cycle, which was unexpected.

Auto- and cross-correlations between

neutron signals in the time and

frequency domain were investigated

by means of signal analysis tools. In

this respect several hypotheses behind

the increase of neutron noise – e.g.

core loading pattern, fuel structure

design, variations of the core inlet

temperature, core asymmetry, etc. –

were identified and checked on

the measured high-frequency data.

Globally it was observed that the

highest neutron noise amplitudes

were to be found in one single core

quadrant, located between Loop 1 and

Loop 3 of the core. Radial correlations

were also identified between core

quadrants, but no measurable time

delays were found axially between

measurements from top and bottom

neutron signals.

Additional measurements of various

plant parameters were also performed,

in a second phase, to extend

the analysis not only to neutron flux

signals, but also temperature, pressure

or component vibrations. Correlations

between vibration signals and

neutron flux signals were analysed as

well.

A brief description of the KKG core

is provided in Section 2. The performed

measurements, neutron noise analysis

performed at KKG [3], along with the

results are described in Section 3.

Section 4 presents a summary of the

performed analysis and the current

model explaining its origin.

2 KKG Core design

The reactor is a Pressurized Water

Reactor (PWR) pre-KONVOI 3-Loop,

manufactured by KWU-Siemens with

a thermal power of 3002 MWth

(1060 MWe). The core contains 177

fuel assemblies with a 15 x 15 fuel

assembly layout and an active core

height of 352 cm.

Since 2014 (Cycle 36) the core is

for the first time fully loaded with

HTP fuel assemblies manufactured

by AREVA GmbH, whose fuel design

features Zircaloy/Duplex cladding

material, modern spacer grid geometries

and UO 2 fuel with 4.95%-wt

enrichment equivalent. The reactor is

typically operated at full power for

12-month cycles and has five different

radial burnup regions. The moderator

temperature coefficient of reactivity

Γ T is in the range of 30 pcm/K at BOC

to 70 pcm/K at EOC. The boron

concentration is typically 950 ppm at

BOC and is continuously decreasing

at a rate of ~ 3 ppm/day. The core

is operated at a maximal Linear

Heat Generation Rate (LHGR) of

525 W/cm, with an average power

density q’’’ of about 105 W/cm 3 [5].

Operation and New Build

Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

| | Fig. 2.

Illustration of four ex-core neutron flux signals (S1 at 100 Hz measured during 10 s.

The core features 48 Rod Cluster

Control Assemblies (RCCA), the

­absorber fingers being inserted within

20 guide tubes per fuel assembly. The

neutron flux within the reactor core is

monitored with six channels, each of

which contains six axial Self-Powered

Neutron Detectors (SPNDs) and a

3D aeroball system in 24 radial

positions. Four quadrants of the

core are equipped in the biological

shielding each with two ex-core

γ-compensated ionisation chambers

for the full power measurements:

one for the lower part of the core

and one for the upper part. The incore

and ex-core detectors allow a

detailed and continuous measurement

of the spatial distribution of

the neutron flux; an illustration of

the instrumentation's location is given

in Figure 1.

3 Neutron noise measurements

and analysis

In order to analyse the possible

reasons of the neutron noise increase

at KKG, the already existing neutron

flux measurements were complemented

in cycle 36 with two extensive

measurement campaigns using a

sampling rate of 100 Hz: one at

MOC and the other at EOC. Figure 2

depicts a typical ex-core neutron flux

measurement.

The in-core and ex-core neutron

signals, including signals from the

vibration monitoring system (“SÜS”)

of the RPV were measured for at least

two continuous hours. The large

amount of data were analysed with inhouse

MATLAB scripts in order to

determine and compare neutron noise

characteristics.

Figure 3 shows the Power Spectrum

Density (PSD) of two in-core

channels at core positions J14 and

G02. On the figure are depicted five

axial levels, the detector E01 is located

at the core top and E06 at the core

bottom. It can be observed that the

results have a non-white noise spectral

component and that position G02 has

lower neutron noise compared to J14,

although the core is symmetrically

loaded.

More specifically, auto- and cross-­

correlations between the neutron

signals in the time and frequency

domain were carefully investigated;

Figure 4 describes these correlations

in a graphic form. The analysis of

these results led to the interesting

observation that no time shifts were

found for the axial measurements

between top and bottom neutron

signals; suggesting that the origin of

the increased neutron noise amplitudes

are not primarily associated

with inlet temperature variations

that would propagate vertically at

flow ­velocity and thus requiring ca.

1 second to propagate.

Figure 5 shows the Probability

Density Functions (PDF) calculated for

two in-core detectors J14 and G02. It is

interesting to notice that, although

the two detectors are symmetrically

located in the core, the shape of the

PDF is highly asym metrical for position

J14. The curve features an upper

OPERATION AND NEW BUILD 453

a) b)

| | Fig. 3.

Power Spectrum Density (PSD) of SPND- J14 (a) and symmetric core position G02 (b), calculated from a sample of 4096 points measured on 18.12.2014. The instrumentation channel contains

axially 5 detectors at different heights starting with detector E01 on the top of the fuel assembly to E06 to the bottom. Higher intensities are measured at low frequency (< 1 Hz). A second

small peak at about 1.8 Hz (J14) is typically identified and corresponds approximately to the first eigenfrequency of HTP fuel assemblies.

Operation and New Build

Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff

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OPERATION AND NEW BUILD 454

tail distribution. High but short peaks

and low prob ability are mostly responsible

for the activation of the power

limitation function of the digital I&C

system. Probability density function

(e.g. Generalized Extreme Value GEV

a)

[6]) and fits of measured parameters

were calculated in an attempt to

predict maximal values and frequency

occurrences of the measured neutron

flux [3], which are of relevance for

operational core control.

Although ex-core raw signals

from the ionisation chambers are

electro­nically filtered in the signal

processing, high amplitude noise are

nevertheless registered with a certain

low residual probability of triggering

b)

| | Fig. 4.

Radial cross-correlations of the four ex-core detectors (a) and axial cross-correlations of in-core detector G02 (b) measured on 18.12.2014.

a)

| | Fig. 5.

Probability Density Functions (PDF) a) and probability distribution b) of in-core detectors at position J14 and position G02 axial level 5. The signals are fitted with

the Generalized Extreme Value (GEV) and Gauss functions (b). GEV fit is well suited for asymmetric distributions as observed at certain core positions in KKG.

b)

a)

| | Fig. 6.

Signal sample from vertical movement detector A1 located on the reactor pressure lid of the oscillation surveillance system (SÜS) recorded at MOC (a) and

probability distribution of two absolute position sensors at MOC (b).

b)

Operation and New Build

Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff

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an alarm. If two channels out of

four are simultaneously measuring a

reactor power PPKG.2.Max._Signal >

103 %, an alarm will be activated in

the control room and RCCA insertion

will be activated in order to reduce the

neutron flux. For this reason, probability

density functions of the in-core

and ex-core detectors were speci­fically

analysed (Figure 5).

Additional to physical measurements

of neutron flux and vibration

signals (Figure 6), special care was

given to the signal analysis of digitallybuilt

signals, for example the corrected

reactor thermal power, used into the

digital I&C system. Useful insights,

among other the evolution of the

positions with high neutron noise,

were obtained by comparing statistical

distributions at MOC and EOC of

those signals.

The neutron noise evolutions of

the in-core and ex-core detectors are

presented in Figure 7. The withincycle

evolutions of the neutron noise

amplitude are to be seen mostly as

linear; local trends are observed

and well coincide with the average

neutron flux trend within the cycle,

whose distribution is a result of boron

acid concentration, burnup of hot

spots in the core, decrease of the

radial peaking factors and RCCA

positions.

The signal correlations given in

Figure 4 revealed that the noise

signals at two opposite sides of the

core had strong negative correlations;

detectors of instrumentation channels

1 and 3 are strongly correlated. This

means that the measured flux increase

in one quadrant is at the same time

compensated by a flux reduction in

the opposite core quadrant. The

analysis has also shown, as illustrated

in Figure 7, that the largest noise

| | Fig. 7.

In-cycle evolution of neutron noise (1-σ standard deviation) measured during Cycle 36: ex-core ionization chambers (S1 – S4) and

in-core SPNDs at axial position 5 (close to fuel assembly inlet). The peak observed at ~20 EFPD is the result of a conducted power

level change.

amplitudes are located primarily in

one quadrant of the core centred on

core position J14 between Loop 1 and

3. The reason for the high neutron

noise in this region was analysed.

It is to note here that the core fuel

loading is 90° symmetric whereas the

RPV with the three loops is 120°

symmetric, implying that there is no

simple core symmetry; in addition,

the individual symmetries show

deviations from theory. To illustrate

this assumption, it can be mentioned

that the thermal loops have different

thermal powers, and their layout is no

perfectly 120° from one another.

Further thermo-hydraulic investigations

would be required to check the

impact of these asymmetries on the

neutron noise amplitudes. It can also

be mentioned that the 48 RCCA are

not positioned with a 90° symmetry in

the core.

Finally, the within-cycle evolution

of neutron noise was compared, at a

macroscopic level, to plant-specific

parameters such as the reactor power,

calibrated ex-core and in-core LHGRs,

and the calculated core flowrate

deduced from the pressure sensors

in the three loops. For illustration

­purposes, the neutron flux measured

by two different channels (Middle

range and SPNDs) and the primary

water temperature span are shown in

Figure 8.

4 Summary

The phenomena leading to an increase

of the neutron flux noise from

cycle to cycle since about 2010 have

been studied in detail through

detailed measurements performed in

the timeframe 2014 to 2015 over two

cycle at MOC and EOC states. The

results show that this increase can

OPERATION AND NEW BUILD 455

a) b)

| | Fig. 8.

Cycle Evolution during Cycle 36 at KKG of a) Measured neutron flux and b) Average core temperature difference (ΔT = T oulet – T inlet ).

Operation and New Build

Detailed Measurements and Analyses of the Neutron Flux Oscillation Phenomenology at Kernkraftwerk Gösgen ı G. Girardin, R. Meier, L. Meyer, A. Ålander and F. Jatuff

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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

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Effects of Airborne Volatile Organic

Compounds on the Performance of

Pi/TiO 2 Coated Ceramic Honeycomb

Type Passive Autocatalytic Recombiner

Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

1 Introduction Ensuring the containment integrity during a severe accident in nuclear power reactor by

maintaining the hydrogen concentration below an acceptable level has been recognized to be of critical importance

after Fukushima Daiichi accidents. Although there exist various hydrogen mitigation measures, a passive autocatalytic

recombiner (PAR) has been considered as a viable option for the mitigation of hydrogen risk under the extended station

blackout conditions because of its passive operation characteristics for the hydrogen removal [1]. As a post-Fukushima

action item, all Korean nuclear power plants were equipped with PARs of various suppliers. The capacity and locations

of PAR as a hydrogen mitigation system were determined through an extensive analysis for various severe accident

scenarios [2]. For some plants, dual hydrogen mitigation systems were equipped with a combination of newly installed

PARs and the existing igniters that each system has 100 % of full capacity for hydrogen control for postulated severe

accident conditions. Among a total of 24 operating units in Korea, a Pt/TiO 2 coated ceramic honeycomb type PAR

supplied by Ceracomb Co. Ltd. [10] has been installed in 18 operating plants and almost units have reached the second

or the third overhaul period since their first installation in 2013.

457

RESEARCH AND INNOVATION

The PAR makes use of a catalyst to

convert hydrogen (H 2 ) and oxygen

(O 2 ) into water vapor and heat. The

heat of reaction creates a natural

­convective flow through the recombiner,

eliminating the need of pumps

or fans to transport new hydrogen to

the surface of the catalyst. In spite of

an advantage of its passive operation,

there have been concerns about

adverse effects on the performance of

PARs by potential deactivators (chemical

poisons and physical inhibitors)

[3, 4, 5]. PARs are required to perform

their safety function not only after

exposure to potential contaminants

during operation, but also in an accident

environment that may contain

various gases or aerosols that are

potentially poisonous to the PAR

catalyst elements [6]. The Ceracomb

also has performed various tests and

demonstrated that its performance

degradation of hydrogen removal

capacity is within 25 % in severe

­accident conditions such as fission

product poisons, aerosols, cable

burns, carbon monoxide, etc. However,

its performance under the longterm

exposed condition to containment

air has not been fully investigated

because the Ceracomb PAR has

no operational experience in nuclear

power plants.

Under the long-term exposed condition

by airborne substances, it

has been known that the catalyst

shows a delayed response for hydrogen

removal [6]. These airborne substances

are known as volatile organic

compounds (VOC) that adsorb on

active sites of the catalyst surface thus

making them unavailable for catalytic

reaction to proceed. As a result, the

recombiner would require either a

higher hydrogen concentration, or a

higher temperature, or both, to start

the hydrogen recombination reaction,

compared with the catalyst in as-new

condition. The VOCs could be originated

from solvents, lubricants, oils,

insulations and paints, etc. which are

commonly used materials in the plant

maintenance. The key prameters of

catalyst performance under the longterm

exposed condition of VOCs

could be the start-up delay time for

catalyst reaction and its hydrogen

depletion (removal) rate because

these parameters directly affect the

results of hydrogen control analyses

in design basis and severe accident

conditions. The catalyst performance

should be verified up to sufficient

periods of plant operation and be

compared with the parameters on the

PAR performance used in the hydrogen

control analysis. Therefore, with

the exposure time to containment air,

the VOC effects will play a more

important role in PAR maintenance

during normal nuclear power plant

operation [7]. In comparison to the

performances under the accident conditions,

however, the performances

under the long term exposed condition

to containment air during normal

operation (i.e., effects of volatile organic

compounds) have not been fully

investigated becaue it requires long

time up to several overhaul periods in

the containment to obtain catalyst

samples and it includes the proprietary

information of PAR suppliers and

utilites.

This paper describes the test results

on the effect of airborne volatile

organic compounds in the containment

air on the performance of TiO 2

coated ceramic honeycomb type PAR

in Korean nuclear power plants

performed in 2014 ~ 2016 overhaul

periods. The test plants are extended

to seventeen (17) operating plants

compared to the previous eight (8)

operating plants [8]. A total of 152

tests have been performed with 680

catalyst samples to investigate the

effect of volatile organic compounds

(VOC) on the start-up performance on

the hydrogen removal. A total of 62

tests have been performed with 248

catalyst samples to identify the influence

on the hydrogen depletion rate

by the VOC effects. The analysis for

VOC components has been performed

for selected samples from seven (7)

plants to identify airborne substances

adsorbed on the surface of catalysts

using a qualitative GC/MS (gas

chromatograph/mass spectrometer)

method.

2 Test method

2.1 Pt/TiO 2 ceramic

honeycomb PAR

Figure 1 shows an illustrated view of

Pt/TiO 2 coated ceramic honeycomb

type PAR that has been installed in

eighteen (18) operating units. This

type of PAR has been developed and

Research and Innovation

Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

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RESEARCH AND INNOVATION 458

| | Fig. 1.

Pt/TiO 2 Coated Ceramic Honeycomb type Passive Autocatalytic Recombiner.

| | Tab. 1.

Specifications of Ceracomb PAR.

Small-Size Medium-Size Large-Size

Weight (kg) 42.1 75.8 144.3

Width (cm) 37.8 72.5 142.8

Depth (cm) 34.3 36.5 36.5

Height (am) 100 100 100

No. of Catalysts 4 8 16

H 2 Depletion Rate (g/sec) a

(4 %-H 2 , 60 °C, 1.5 bar)

> 0.2 g/sec > 0.4 g/sec > 0.9 g/sec

a) Required hydrogen depletion rate in the technical specification for PAR purchase of Korean NPPs.

2.2 Test facility

The VOC effect tests have been performed

using the PAR performance

test facility (PPTF). The PPTF comprises

a carbon steel pressure vessel

with the internal volume of 12.5 m 3 (a

cylindrical shape with 3.3 m in height

and 2.2 m in diameter). It was constructed

to perform performance tests

in various conditions of pressure,

temperature, humidity, hydrogen

concentration and chemical water

spray. Figure 3 shows types and locations

of measurements in the pressure

vessel of PPTF. Inside of the vessel,

mixing fans, spray nozzles and electrical

heaters are installed to maintain

a desired test condition. At the center

of the vessel a test PAR is located. A

small-sized PAR with four (4) catalysts

is used as a test PAR. Gates are

equipped at the PAR entrance and exit

to prevent air and hydrogen from

being in contact with the catalyst

surface before the test starts. The

hydrogen concentration is measured

with an accuracy of 2 % of full scale

sampling rate. The time lag of the

hydrogen concentration signal due to

the length of the gas sampling line is

estimated as below 50 sec.

| | Fig. 2.

Ceramic Honeycomb Catalyst.

supplied by Ceracomb Co. Ltd. [9, 10].

The Ceracomb PAR consists of a

stainless steel housing equipped with

catalysts inside the lower part of the

housing. The PARs are installed with

floor mount type or wall mount type

in the containment and its structures

are designed to meet the seismic

requirements of each plant. Air and

hydrogen mixture flows from bottom

of the PAR to the exit openings at the

upper part of PAR. The housing is

designed to have chimney effects so

that the heat generated in the catalytic

reaction in lower part of the housing

can promote a strong driving force for

natural convective flow and to protect

the catalyst from the direct impinge of

containment spray. There are three

different sizes of PAR according to the

number of the catalyst. The specifications

of the Ceracomb PAR are

summarized in Table 1.

Different types of catalytic recombiners

have been supplied by various

PAR suppliers such as AREVA, CANDU

Energy, NIS (formerly NUKEM), KNT

and Ceracomb. AREVA, CANDU Energy

and NIS utilized plate type catalysts

while original NUKEM invented a

specialized cartridge containing pellet

type catalysts. KNT and Ceracomb PAR

utilized ceramic honeycomb type

catalysts. In the present Pt/TiO 2

coated ceramic honeycomb type PAR,

a cubical catalyst with a honeycomb

microstructure has been used to

increase the surface area for the

reaction. The catalyst is manufactured

by coating a mixture of TiO 2 and Pt on

the supporting structure of the ceramic

honeycomb of 35 CPSI (cell per

square inch). Figure 2 shows an

illustrated view of ceramic honeycomb

catalyst. The dimensions of the

standard honeycomb catalyst are

15 cm by 15 cm with the height of

5 cm. A ­protected metal frame is

used to protect the catalyst because

the ceramic catalyst is fragile and

vulnerable to impact.

2.3 Test methods

Key parameters of catalyst performance

are considered as the start-up

delay time and hydrogen removal rate

which are directly related to PAR

modeling in the hydrogen control

analysis to determine the capacity and

locations of PAR system [2]. Under

the VOC-affected conditions, its performance

is hard to identify through

the perioic inspection method because

the start-up delayed time and

the hydrogen removal rate are defined

under the natural convection conditions.

Therefore, a number of catalysts

are withdrawn out of containment

during an overhaul period of each

plant and their performance is tested

in the PAR performance test facility

(PPTF) under the natural convection

conditions. Table 2 shows the number

of catalysts taken from various plants

for VOC effect tests performed during

2014 ~ 2016 outage periods in seventeen

(17) plants. Further tests for

other plants are scheduled according

to their outage schedules. The VOC

effect tests are performed into three

groups; (a) the measurement of startup

delay time for hydrogen removal,

(b) the measurement of hydrogen

depletion (removal) rate and (c) VOC

component analysis to identify airborne

substances adsorbed on the

Research and Innovation

Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

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| | Fig. 3.

PAR Performance Test Facility (PPTF) with Measurement Types and Locations.

Plant ID Plant Type Test Date

(yyyy/mm)

| | Tab. 2.

Number of Catalysts for VOC Effect Tests.

surface of catalysts. Four (4) catalysts

are withdrawn from one PAR considering

the installed location in the

containment so that the catalyst

samples be distributed uniformly

throughout containment area in order

to avoid local effects of the test results.

The exposure time of catalysts to the

containment air includes the normal

operation time of ~18 months and the

plant outage time that depends on the

outage schedule of each plant.

The start-up delay times for hydrogen

removal are measured in the PPTF

facility. Four (4) catalysts are mounted

in a small sized test PAR housing

No. of catalysts (No. of Tests)

Delay

Time

Depletion

Rate

VOC

Component

C1 PWR (W) a) 2014. 04 8 (2) 8 (2) 1

C2 PWR (W) 2015. 08 32 (8) 20 (5) -

D1 PWR (W) 2014. 11 28 (7) 12 (3) -

D2 PWR (W) 2016. 02 20 (5) 12 (3) -

F1 PWR (W) 2014. 11 28 (7) 20 (5) -

F2 PWR (W) 2016. 06 20 (5) 12 (3) -

G PWR (F) b) 2014. 12 84 (21) 20 (5) 1

H1 PWR (F) 2014. 11 84 (21) 20 (5) 4

H2 PWR (F) 2016.06 84 (21) 12 (3) 1

L PWR (O) c) 2014. 07 20 (5) 12 (3) -

M1 PWR (O) 2014. 07 40 (10) 12 (3) 1

M2 PWR (O) 2016. 04 20 (5) 12 (3) -

N PWR (O) 2015. 03 40 (10) 12 (3) -

O PWR (O) 2014. 12 20 (5) 12 (3) -

P PWR (O) 2014. 06 40 (10) 20 (5) 1

W PHWR d) 2015. 10 20 (5) 12 (3) -

Y PHWR 2014. 07 20 (5) 12 (3) 1

Total 608 (152) 248 (62) 10

Notes: a) PWR (W) : Westinghouse designed PWR b) PWR (F) : Framatome designed PWR

c) PWR (O) : Optimized Power Reactor (OPR) 1000 d) PHWR : CANDU6

* Each Data sets of C1/C2, D1/D2, F1/F2, H1/H2 and M1/M2 represent the same plants

but the tests are performed on different outage schedule.

that is the same model of the commercial

PAR so that four (4) catalyst samples

are used for a test. The test PAR is

installed at the center in the test vessel

of the PPTF. After the test vessel is

closed, mixing fans are turned on and

the hydrogen is injected to a desired

hydrogen concentration. Until desired

conditions are achieved, gates at the

PAR entrance and exit are closed in

order to prevent air and hydrogen

from being in contact with the catalyst

surface. The start-up delay tests are

performed at the initial conditions of

the hydrogen concentration of

3 vol. % and temperature of 60 °C

under the pressure of 1.5 bar (abs).

The start-up delay time is defined as

the required time for the hydrogen

concentration in the test vessel to start

to decrease by one percent (relative)

of the initial hydrogen concentration

after the hydrogen in the test vessel

starts to contact the catalysts in the

PAR (i.e., the gates at the PAR entrance

and outlet are opened).

The hydrogen depletion rates with

degraded catalysts under the normal

operation environments for an overhaul

period are measured using the

PPTF facility. The tests are performed

with the same procedure of the startup

delay time tests but with different

initial conditions. The hydrogen

depletion tests are performed with

the initial conditions with a hydrogen

concentration of 6.9 vol. % and temperature

of 60 °C under the pressure

of 1.5 bar (abs). The hydrogen

depletion rate is calculated from the

gradient of the hydrogen concentration

when the concentration at the

PAR entrance is 4 vol. %. The hydrogen

depletion rate from the present

tests are compared with the hydrogen

depletion rate required in the technical

specification of PAR purchase,

which is defined as above 0.2 g/s for

the small sized PAR at the conditions

of 4 vol. % of hydrogen, temperature

of 60 °C and pressure of 1.5 bar.

The composition adsorbed airborne

substances on the catalyst

surfaces is analyzed with GC/MS (gas

chromatograph/mass spectrometer)

method. Tests are performed by

Frontier Laboratories Co. Ltd. [11]

using Agilent 6890 GC/5973N MSD

and PT-2020D Pyrolyzer. Each catalyst

is heated up in an oven and the

temperature is raised up to 300 °C and

600 °C successively with a rate of

20 °C/min. The VOCs desorbed from

the catalyst surface were separated

continuously and their components

are analyzed qualitatively with GC/

MS method.

4 Results and Discussion

The performance of the catalyst

should be inspected periodically using

a specially designed device during

every plant outage period. In case of

the present ceramic honeycomb type

PAR, at least a quarter of the entire

catalysts are tested in every outage

period. The catalysts are tested in

single arrangement under the predetermined

flow and temperature

of air and hydrogen mixture by

measuring the temperature rise of

air-hydrogen mixture between inlet

and outlet of the test device. Figure 4

RESEARCH AND INNOVATION 459

Research and Innovation

Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

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RESEARCH AND INNOVATION 460

| | Fig. 4.

Tempearature and Hydrogen Concetration at the Exit of Test Device of Periodic Inspection

(New Catalyst: 3 % hydrogen and air mixture at 60 °C and 1 bar).

shows temperature rise behavior of

new catylists, which shows a similar

trend with time. Therefore, the PAR

supplier suggested the accepatance

criteria of the periodic inspection as

the temperature rise at a given time

(The exact values of temperature rise

and time are not described in this

paper because that information is a

supplier’s proprietary). Figure 5

shows temperature rise bebavior of

catylists that were exposed to containment

air during one overhaul period.

The behavior of temperature rise is

affected by the existence of VOC.

Some catalysts showed delayed startup

of hydrogen recombination and

others showed further increase of

temperature by combustion of VOC

itself. Figure 5 also shows the hydrogen

volume faction of air-hyrogen

mixture at the outlet of the test device.

It showed that the hydrogen recombination

already started although

the temperature does not reach the

required value. Therefore, there is a

possibility of unneccesary failure of

plant inspection with the current

method by temperature rise. This

method requires relatively long test

time because of larger heat capacity of

ceramic structure. In addition, it is

­difficult to correlate the hydrogen

recombination performance with the

amount of temperature rise and test

time. Threfore, we decided to change

the inspection method from the temperature

rise to the direct measurement

of hydrogen concentration with

new acceptance criterion.

Under the VOC-affected conditions,

the performance of PAR is hard

to identify through the current perioic

inspection method because the startup

delayed time and the hydrogen

­removal rate are defined under the

| | Fig. 5.

Tempearature and Hydrogen Concetration at the Exit of Test Device of Periodic Inspection

(After the Exposue of One Overhaul Period to Containment Air, 3 % hydrogen and

air mixture at 60 °C and 1 bar).

natural convection conditions. Therefore,

a number of catalysts are withdrawn

out of containment during an

overhaul period of each plant and

their performance is tested in the PAR

performance test facility (PPTF) under

the natural convection conditions.

A total of 152 tests are performed

with 608 catalyst samples to investigate

the effect of volatile organic

compounds (VOC) on the startup

performance on the hydrogen

removal. The catalyst samples are

taken from seventeen (17) plants with

four (4) different reactor types. For

plants C, D, F, H and M, the tests are

performed twice in the first and

second outage period to compare test

resuts between the first and the

second outages in the same plant.

Figure 6 shows the measured start-up

delay times in conditions of hydrogen

of 3 vol. %, temperature of 60 °C and

pressure of 1.5 bar. These test conditions

are selected because a start-up

delay time is considered after the

hydrogen concentration and the

temperature reached at both 3 vol. %

and 60 °C in the analysis of hydrogen

control to determine the capacity

and locations of PARs as a hydrogen

mitigation system [2]. Fifteen (15)

minutes of the start-up delay time are

assumed in severe accident analyses

while 12 hours of the start-up delay

time is assumed in design basis accident

analysis [12]. For new catalysts a

certain time is required until the flow

is fully developed by naural convection.

This time has been measured as

about 404 sec with a standard deviation

of 66.9 sec. As shown in Fig. 6,

the start-up delay times are well

within 15 minutes except the plants G

and H. The start-up delay times for

plant G and H1 show an average time

of 1,006 sec and 893 sec with a

standard deviation of 160 sec and

215 sec, respectively. The total averaged

start-up delay time for all plants

is estimated as 660.6 sec with a standard

deviation of 237.8 sec. For plants

C, D, F, H and M, the second tests does

not show a noticeable difference

­compared to its first tests.

In the design basis accident such as

a loss-of-coolant-accident (LOCA),

the hydrogen is generated gradually

and the hydrogen concentration could

be reached at 4 vol. % after several

days without a hydrogen mitigation

system after a LOCA takes places. In

the analysis of hydrogen concentration

in the LOCA, twelve (12) hours of

the start-up delay time were assumed

after the hydrogen concentration and

the catalysts temperature reach at

both 3 vol. % and 60 °C. Although the

start-up delays of 12 hours are considered,

there is a sufficient margin to

maintain the hydrogen concentration

below the regulatory limit of 4 vol. %.

However, in the severe accident conditions,

the hydrogen concentration in

the containment abruptly increases at

the timing of the reactor vessel failure

so that the margin for start-up delay

for hydrogen removal may not be

­sufficient compared to the situation of

a design basis accident. The regulatory

position in Korea is that the startup

delay times should be verified and

compared to the assumptions used in

the analysis of hydrogen control in

DBA and severe accident conditions.

In the case of plant G, H and N, the

analysis of hydrogen control in severe

accident conditions has been re-evaluated

with a longer delay time of

30 minutes in consideration of the

results of the start-up delay time

measurement tests in 2014. For the

Research and Innovation

Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

| | Fig. 6.

Start-up Delay Times after One Overhaul Period Exposure to VOC.

Plant ID

Compounds

other plants, the re-evaluation has

been performed in 2017.

Figure 7 shows the hydrogen

depletion rates after an overhaul

period of exposure to VOCs in containment

air. A total of 62 tests are

performed with 248 catalyst samples

from seventeen (17) plants as

described in Table 2. The test results

show that the hydrogen depletion

rates are much higher than the

required depletion rate of 0.2 g/sec

that is specified in technical specification

of PAR purchase in Koran

nuclear power plants. A total averaged

value is estimated as 0.270 g/sec with

C1 G H1 H2 M1 P Y Estimated Sources

of VOCs

Benzene ! ! ! ! ! ! ! Paint, Insulation, Glue

Docosane ! ! ! ! Oil

Eicosane ! ! ! ! ! Oil

Heptadecane ! ! ! ! ! ! ! Oil

Heptane, 3-methylene- ! ! ! Oil

Hexadecane ! ! ! ! ! ! Oil

Octadecane ! ! ! ! ! ! Oil

1-Propene, 2-methyl- ! ! ! Paint

Dibutylformamide ! ! ! Insulation

Diethyl phtalate ! ! ! Paint, Insulation

Heneicosane ! ! ! ! Oil

Methylstyrene ! ! ! ! Paint, Insulation

Nonadecane ! ! ! Oil

Tridecane ! ! ! ! Oil

Nonaneitrile ! ! ! Oil, Resin

Tetradecane ! ! ! Oil

Toluene ! ! Paint, Sealing

| | Tab. 3.

Major Compounds Adsorbed on the Sample Catalyst Surface.

a standard deviation of 0.03 sec. The

measured hydrogen depletion rates of

catalysts exposed to VOCs have no

difference with those of new catalysts

that is estimated as 0.2687 g/sec with

a standard deviation of 0.0108 sec.

The recombination reaction takes

place on some active sites on the

degraded catalyst releasing the heat

of reaction. This causes the catalyst

surface temperature to increase

creating a driving force for convective

flow. Increase convective flow

accelerates the reaction rate leading

to further increase in the catalyst

temperature until all the adsorbed

| | Fig. 5.

Hydrogen Depletion Rates after One Overhaul Period Exposure to VOC.

VOCs desorb and all the active sites

are free, i.e., the catalyst is fully

regenerated. The same conclusion

about the hydrogen depletion rate

has been reported in reference [6].

The adsorbed airborne substances

on the catalyst surface are analyzed

qualitatively using GC/MS (gas

chromatograph/mass spectrometer)

method for selected samples from

seven (7) plants. Various VOCs are

detected and their major compounds

are summarized in Table 3. It is

estimated that these compounds are

originated from paints, oils, lubricant,

insulation, glues, etc., which are commonly

used in the plant maintenance.

Although benzene, heptadecane etc.

are commonly detected, the detected

volaticle organic compounds differ

from each plants. In the previous

results, the plant H1 showed a relatively

longer start-up delay time compared

to other plants [8]. There was a

steam generator replacement in plant

G and H when the PARs were installed

in 2013. Further tests are performed

in next overhaul for plant H. The test

results of H 2 represents test results in

the second overhaul (2016) in plant H.

The detected VOCs are different from

the results of the first overhaul (2014)

but the start-up delay time still

remained in relatively larger value

than other plants. The common VOCs

detected in plant G, H1 and H2 are

benzene, hetadecane, octadecane etc.

(the plant G and H are the same type

plants). However, these materials are

also detected in other plants having a

relatively shorter start-up delay time.

From the present results, it is considerd

that the detected materials are

plant-specific and strongly dependent

on the maintenance activities. The

VOC materials presented in Table 3

are at least not strongly related to the

RESEARCH AND INNOVATION 461

Research and Innovation

Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

RESEARCH AND INNOVATION 462

start-up performance of PARs. Therefore

we could not identify which

materials of VOC could affect the

start-up performance using the

present GC/MS method.

The regulatory position in Korea

on the PAR is that the start-up delay

time and the hydrogen depletion rates

should be verified periodically and

compared to those assumed in the

hydrogen control analysis for design

basis and severe accidents because the

long-term operational experience of

PAR in the nuclear power plants has

not been fully insvestigated. Therefore,

the present paper have been

mainly focused on the start-up delay

time and hydrogen depletion rate in

a given condition to validate the

assumptions used in the hydrogen

control analysis. It is considered that

there is a sufficient margin to control

hydrogen below the regulatory limit

of 4 vol. % of hydrogen concentration

in design basis accidents. However, in

the severe accident conditions, the

hydrogen in the containment abruptly

increases at the timing of the reactor

vessel failure. There may not be

­sufficient margin for hydrogen control

in some severe accident scenarios if an

additional start-up delay time more

than 30 minutes is considered. However,

the capacity and locations of PAR

have been determed from very conservative

severe accident analyses [2]

and the temperature of containment

air is expected to be above or around

100 °C in severe accident conditions.

It could be postulated that the temperature

will be high enough to regenerate

the PAR catalyst that had resided

in the containment for a prolonged

time period so that the PAR will

promptly respond to hydrogen. Therefore,

it is important to identify in

which conditions the PAR will

promptly react with hydrogen in such

a long time exposed condition to

possible VOCs.

4 Conclusions

The hydrogen depletion rates and

the start-up delay time of a Pt/TiO 2

coated ceramic honeycomb PAR have

been measured using a total of 856

catalyst samples from seventeen (17)

operating nuclear power plants after

one overhaul period of normal operation

since its first installation in order

to investigate the effect of volatile

organic compounds (VOCs) on the

catalyst functionality. The measured

hydrogen depletion rate and start-up

delay time were compared to those

used in the hydrogen control analysis

because these are key parameters in

the determination of the capacity and

location of PARs. The tests showed

that the hydrogen depletion rates are

not affected by VOC accumulation on

the catalyst surface due to its volatile

nature at high temperature by exothermic

catalytic reaction. Through a

series of tests on the start-up delays

using VOC-affected catalysts, the VOC

delays the start-up for hydrogen

removal by poisoning or blocking of

the catalytic surface. Although the

measured delay times were well

within 30 minutes in the condition of

3 vol. % of hydrogen, 60 °C of temperature

and 1.5 bar of pressure, it is

expected that the delay time would

further increase in proportion to the

exposure time to containment air. The

type of airborne substances was

­identified through qualitative GC/MS

(gas chromatograph/mass spectrometer)

method from selected samples

from seven (7) plants. The volatile

organic substances adsorbed on the

catalyst surface were estimated

mainly from paints, lubricants, glues,

insulations and oils etc. It is expected

that the reduction of VOC in the

containment air may be a challenging

work. Therefore, it is important to

identify in which conditions the PAR

will promptly react with hydrogen in

such a long exposed condition of

possible VOCs. To this end, further

extensive tests on the catalyst performances

in various hydrogen concentrations

and temperatures will be

performed with catalysts that had

resided in various reactor containments

and for various exposure times

to containment air.

References

1. Status Report on Hydrogen Management

and Related Computer Codes,

NEA/CSNI/R(2014)8 (2014).

2. Kim, C. H. et al., Analysis Method for the

Design of a Hydrogen Mitigation

System with Passive Autocatalytic

Recombiners in OPR-1000, The 19 th

Pacific Basin Nuclear Conference (PBNC

2014), Vancouver, Canada, August 24–

28, 2014, Paper No. PBNC2014-072

(2014).

3. Effects of Inhibitors and Poisons on the

Performance of Passive Autocatalytic

Recombiners (PARs) for Combustible

Gas Control in ALWRs, EPRI ALWR

Program Report, Palo Alto CA (1997).

4. Studer, E. et al., Assessment of Hydrogen

Risk in PWR, 1 st IPSN/GRS EURSAFE

Meeting, Paris (1999).

5. OCED/NEA THAI Project: Hydrogen and

Fission Product Issues Relavent for

Containment Safety Assessment under

Severe Accident Conditions, NEA/

CSNI/R(2010)3 (2010).

6. Kelm, S. et al., Ensuring the Long-Term

Functionality of Passive Auto-Catalytic

Recombiners under Operational

Containment Atmosphere Conditions –

An Interdisciplinary Investigation,

Nuclear Engineering and Design,

Vol.239, pp. 274-280 (2009).

7. Reinecke, E-A. et al., Open Issues in the

Applicability of Recombiner

Experiments and Modeling to Reactor

Simulations, Progress in Nuclear Energy,

Vol.52, pp. 136-147 (2010).

8. Kim, C. H. et al., Operational Experience

of Ceramic Honeycomb Passvie Autocatalytic

Recombiner as a Hydrogen

Mitigation System, The 16 th International

Topcical Meeting on Nucear

Reactor Thermal Hydraulics

(NURETH-16), Chicago, IL, USA,

August 30 – September 4 (2015)

9. Kang, Y. S. et al., Hydrogen Recombination

over Pt/TiO2 Coated Ceramic

Honeycomb Recombiner, Appl. Chem.

Eng., Vol.22, No.6, pp. 648-652 (2011).

10. Ceracomb Co. Ltd., http://

www.ceracomb.co.kr/en/ (homepage).

11. Frontier Laboratories, Co. Ltd.,

http://frontier-lab.com (homepage).

12. Final Safety Analysis Report of Ulchin

Nuclear Power Plants, Units 3 & 4,

Section 6.2, Korea Hydro and Nuclear

Power Co., Ltd. (revised in 2013).

Authors

Chang Hyun Kim

Je Joong Sung

Sang Jun Ha

Central Research Institute

Korea Hydro and Nuclear Power

Co., Ltd.

25-1 Jang-dong, Yuseong-gu,

Deajeon, 305-343, Rep. of Korea

Phil Won Seo

Department of Research &

Development,

Ceracomb Co., Ltd.

312-25 Deuksan-dong, Asan-si,

Chungcheongnam-do, 336-120,

Rep. of Korea

Research and Innovation

Effects of Airborne Volatile Organic Compounds on the Performance of Pi/TiO 2 Coated Ceramic Honeycomb Type Passive Autocatalytic Recombiner ı Chang Hyun Kim, Je Joong Sung, Sang Jun Ha and Phil Won Seo

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

49 th Annual Meeting on Nuclear Technology (AMNT 2018)

Young Scientists' Workshop

Jörg Starflinger

During the Young Scientists' Workshop of the 49 th Annual Meeting on Nuclear Technology (AMNT 2018), 29 to 30

May 2018, Berlin, 13 young scientists presented results of their scientific research as part of their Master or Doctorate

theses covering a broad spectrum of technical areas.

This demonstrated again the strong

engagement of the younger generation

for nuclear technology and the

significant support of German institutions

involved.

Dr. Katharina Stummeyer (Gesellschaft

für Anlagen- und Reaktorsicherheit

gGmbH), Dr.-Ing. Wolfgang

Steinwarz (Founder and former jury

chairman of the Workshop „Preserving

Competence in Nuclear Technology”),

Prof. Dr.-Ing. Marco K. Koch (Ruhr-

Universität Bochum), and Prof. Dr.-Ing.

Jörg Starflinger (Universität Stuttgart)

as members of the jury assessed the

written compacts and the oral

presentations to award the prices

supported by GNS Gesellschaft für

Nuklear-Service mbH, Essen and

Forschungsinstitut für Kerntechnik und

Energiewandlung e.V., Stuttgart.

Vera Koppers (Gesellschaft für

Anlagen- und Reaktorsicherheit (GRS)

gGmbH, mentoring: Prof. Koch) as first

speaker reported on the present status

on Heuristic Methods in Modelling

Research Reactors for Deterministic

Safety Analysis. The goal is a

deeper understanding of modelling of

research reactors using the code

ATHLET. Good agreement of ATHLET

results with experiments from literature

has been achieved.

The presentation by Sebastian

Unger (Helmholtz-Zentrum Dresden-

Rossendorf, mentoring: Prof. Hampel)

described Experimental Investigation

on the Heat Transfer of Innovative

Finned Tubes for Passive

Cooling of Nuclear Spent Fuel Pool.

A single-phase cooling system for

spent fuel pools has been introduced.

The bottle neck in heat transfer lays

on the air-side heat exchange, which

is enhanced by innovative fins. The

potential of enhancement of heat

transfer has clearly been demonstrated

on small-scale.

Martin Arlit (Technische Universität

Dresden, mentoring: Prof. Hampel)

informed about Heat Transport from

Dried Surfaces of a Spent Fuel

Mock-up under Accident Conditions

with a Thermal Anemometry

Grid Sensor. A grid sensor has

been developed enabling the spatially

resolved measurement of fluid temperatures

and velocities within a rod

bundle. Small-scale experiments

showed that heat dissipation by convection

of the overall heating power is

below 10 %, but is of importance for

the cooling of the dried rod bundle

section above the water level.

Maria Freirìa López (Universität

Stuttgart, mentoring: Prof. Starflinger)

reported on Criticality Evaluation of

Debris Beds after a Severe Accident.

By means of Monte-Carlo-Code simulations,

a criticality map for debris

beds, forming during beyond-design

accidents, is currently being developed.

The first analyses indicates that

debris beds in fact might get critical,

but they also showed parameter

combinations (debris size, boration,

porosity, etc.), where criticality can be

intrinsically excluded.

Larissa Klaß (Forschungszentrum

Jülich GmbH, mentoring: Prof. Modolo)

described Modified Diglycolamides

for a Selective Separation of

Am(III): Complexation, Structural

Investigations and Process Applicability.

In her work the complexation

behaviour of new hydrophilic complexants

towards trivalent actinides

and lanthanides was investigated in

order to achieve a deeper understanding

of their coordination chemistry.

For the first time, the formation of

mixed complexes of hydrophilic and

lipophilic complexant in the organic

phase has been measured. Based on

this result, an innovative solvent

extraction procedure was developed,

which could simplify the existing

procedures.

Corbinian Nigbur (Universität

Stuttgart, mentoring: Prof. Starflinger)

introduced the Application of the

Integral Diffusion Approach to

the Modelling of the Oxidation of

Mixtures of Fuel and Zirconium.

The objective is to simulate the oxidation

process during accidents with

one integral approach replacing the

different numerical approaches within

thermal-hydraulic system codes.

Comparison of numerical simulations

with data from crucible experiments

showed good agreement.

Numerical implementation of

methods considering dynamic

soil-structure interaction was the

subject of the presentation given

Arthur Feldbusch, (Technische Universität

Kaiserslautern mentoring: Prof.

Sadegh-Azar). A dynamic model has

been developed for soil-structureinteraction.

Using the “Thin Layer

Method”, a tool is derived to evaluate

the soil-structure behaviour due to

mechanical loads. The model is

limited to linear calculations, but shall

be extended to non-linear capabilities.

Pascal Distler (Technische Universität

Kaiserslautern, mentoring: Prof.

Sadegh-Azar) reported about Airplane

Crash Analysis: Semi-hard and

hard Missile Impact on Reinforced

Concrete Structures, in which the

damage mechanisms were presented

and explained to determine the load

bearing capacity of the hard and the

soft impact of projectiles. A numerical

model has been set up and compared

with impact tests, which show reasonable

agreement. In the next step, the

actual model will be extended to

describe the interaction between the

reinforced concrete structure (target)

and the impacting projectile.

Danhong Shen (Karlsruhe Institute

of Technology (KIT), mentoring: Prof.

Cheng) gave an overview on An improved

turbulent mixing model in

sub-channel analysis code. Using

CFD simulations of two adjacent

sub-channels of a fuel assembly, an

improved numerical turbulent mixing

model has been derived to be used in

sub-channel codes. Three empirical

correlations are proposed to describe

the relationship between each turbulent

mixing coefficient and the

Reynolds number as well as the

geometry parameter. This investigation

will improve computation

capabilities of sub-channel codes.

The presentation of Dali Yu (Karlsruhe

Institute of Technology (KIT),

mentoring: Prof. Cheng) described

Modeling of post-Dryout Heat

Transfer. The aim of the work is to

predict the wall surface temperature

under Dryout conditions. The whole

post-dryout flow region is divided into

463

AMNT 2018 | YOUNG SCIENTISTS WORKSHOP

AMNT 2018 | Young Scientists' Workshop

Young Scientists' Workshop ı Jörg Starflinger

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

464

AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP

| | Award winners, sponsors and jury of the 49 th Annual Meeting on Nuclear Technology (AMNT 2018)

Young Scientists Workshop: (from left): Dr. Wolfgang Steinwarz, Prof. Dr. Marco K. Koch (Ruhr-Universität

Bochum), Dr. Katharina Stummeyer (Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH),

Vera Koppers, Winner of the 2018 Young Scientists' Award, Dr. Jens Schröder (GNS Gesellschaft für

Nuklear- Service mbH), Katharina Amend (2 nd ranked young scientist), Prof. Dr. Jörg Starflinger

( Universität Stuttgart), María Freiría López 3 rd ranked young scientist.

several different sections, each of

them modelled separately with

different correlations. A comparison

with experimental data showed fairly

reasonable results which are subject

for improvement as a next step.

Tobias Jankowski (Ruhr-Universität

Bochum, mentoring: Prof. Koch) reported

about Development and

Validation of a Correlation for

Droplet Re-Entrainment Estimation

from Liquid Pools. The correlation is

based on a dimensional analysis and

therefore considers thermohydraulic

boundary conditions by dimensionless

quantities, which are quantified

by empirical constants. These constants

are obtained by four nearly

steady test phases, taken from two

experimental facilities of different

scale. The correlation results are in a

good agreement with experimental

data.

The presentation entitled Comparison

of Different Wash-off

Models for Fission Products on

Containment Walls was given by

Katharina Amend (Universität der

Bundeswehr München, mentoring:

Prof. Klein). A parameter variation

was conducted with in the setting of a

simplified geometry and with the

geometry of the laboratory tests. One

key influencing parameter for the

resulting washed off mass is the

percentage of area covered by water

in each case, which differs with

­inclination and mass flow rate. First

simulations with the laboratory

geometry show satisfactory agreement,

when compared to the experiments.

Moritz Schenk (Karlsruhe Institute

of Technology (KIT), mentoring: Prof.

Cheng) gave a presentation about CFD

Analysis of centrifugal Liquid Metal

Pumps. Using the open-source software

OpenFOAM the influence of the

physical properties of liquid metals on

the performance of a pump impeller

and on the flow field is investigated.

In general, the simulations show a

­relatively strong negative influence on

head and efficiency for much higher

viscosities and nearly no effect for

lower viscosities compared to water.

This qualitative behaviour is in good

agreement with the literature. The

optimization of the liquid metal pump

is ongoing, focussing on the corrosion

potential of the liquid metal.

Summarizing, the scientific quality

of papers presented by the young

scientists in this year reached again

a very high level. Therefore, all participants

of the workshop should get

honourable recognition.

The jury awarded Vera Koppers

(Gesellschaft für Anlagen- und Reaktorsicherheit

(GRS) gGmbH) the 1 st price

of the 2018 competition. Her compact

as well as those of both the 2 nd ranked

author, Katharina Amend (Universität

der Bundeswehr München) and the 3 rd

ranked author Maria Freiria (Universität

Stuttgart) are published in this

issue of atw – nucmag.

Author

Prof. Dr.-Ing. Jörg Starflinger

Institute of Nuclear Technology

and Energy Systems (IKE)

Pfaffenwaldring 31

70569 Stuttgart, Germany

Young

Scientists'

Workshop

WINNER

Vera Koppers was

awarded with the

1 st price of the 49 th

Annual Meeting on

Nuclear Technology

(AMNT 2018) Young

Scientists' Workshop.

Heuristic Methods in Modelling

Research Reactors for Deterministic

Safety Analysis

Vera Koppers and Marco K. Koch

1 Introduction The national and international fundamental nuclear safety objective is to protect the public

from ionising radiation [IAEA2016]. Although research reactors may have a smaller risk potential to the public than

nuclear power plants, operators and researchers are at a higher risk [IAEA2016]. Deterministic safety analyses using

thermal-hydraulic system codes are a prevalent and important instrument to evaluate the safety of nuclear power plants

and research reactors. A wide range of safety analysis codes that are used for simulations of nuclear power plants are

applicable to simulations of research reactors. The application range of the thermal-hydraulic system code ATHLET

(Analysis of thermal-hydraulics of leaks and transients) – developed by GRS (Gesellschaft für Anlagen- und Reaktorsicherheit

gGmbH) – was extended to simulated subcooled nucleate boiling processes at low pressure in 1994 [GRS2009].

AMNT 2018 | Young Scientists' Workshop

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

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

After that, research reactor simulations using ATHLET were successfully performed at national and international

­research institutes. ATHLET uses the finite volume method and solves the partial differential equations matrix at

discrete meshed volumes. In order to simulate a plant system, the user has to build up a network of thermal hydraulic

volumes. This approach allows a wide range of code application due to free thermal-hydraulic nodalisation, but it takes

large amount of human resources and requires detailed plant descriptions. In ATHLET, the main modules are thermal

fluid dynamics (TFD), heat transfer and heat conduction (HECU), neutron kinetics (NEUKIN) as well as plant control

(GCSM). The user has to choose adequate input options out of a wide range of possibilities for each module. Analysing

foreign research reactors, technical support organisations and research institutes might be confronted with limited

available information of plant data. In case of emerging safety related questions, the complex input data structure of

safety analysis codes impede a fast response.

The present paper describes the development

of a new method for rapid

input deck development in the light of

limited available data. Due to high

diversity of research reactor designs,

a rule-based software system is

engineered to support the modelling

process for deterministic safety analysis

utilising the system code ATHLET.

The use of heuristic rules allows

an adequate input deck generation

despite limited data. The fundamental

elements of the input deck are generated

automatically by few input data

necessary. In the case of unavailable

data and urgently safety related questions,

the user is supported by this

software. In the following, the applied

heuristic rules realising the new

strategy of modelling are described.

After that, first functionality of the

new modelling system is demonstrated.

2 Heuristic methods

in modelling research

reactors

In this paper, heuristic methods are

defined as an approach to achieve an

appropriate modelling quality of

research reactors despite incomplete

data. For this purpose a new software

is developed that is structured in the

following main modules:

• process of user input

• build the research reactor model

• transform to ATHLET input format

• export as input deck

The required key data, which the user

has to provide to run the software, are

constricted to publicly available data.

Detail technical documentations, such

as safety analysis report, operating

manual, system descriptions and

schematics as well as technical

drawings are assumed to be not

accessible. The next text section

describes the main steps of the

strategy that are implemented in the

modelling software.

medicine, research reactors have a

wide range of designs and operation

modes. Realising a heuristically process

for research reactor modelling,

the number of reactor types considered

in this study was restricted.

To date, 241 research reactors

are operated around the world

[RRDB2018]. The TRIGA (Training,

Research, Isotopes, General Atomic)

and MTR (Material Testing Reactors)

reactors represent the most widely

installed research reactor types. About

25 % of the research reactors are

of MTR type and 21 % are of TRIGA

design [RRDB2018]. Consequently,

these types are selected as a model

design basis. The considered reactor

designs are abstracted to open core

and tank-in pool reactors as pictured

in Figure 2-1. The TRIGA design is

currently limited to reactors with

natural convection cooling.

To structure the research reactor

types, a modularisation approach is

used. The first level of modularisation

is also shown in Figure 2-1. On the

second level, the reactor components

are decomposed into their further

elements. Focusing on the central

component, the “reactor core”, typical

MTR research reactors have a cluster

of multiple assemblies installed at the

lower part of the reactor pool. The

assembly consists of several parallel

arranged fuel plates and the assembly

feet. The basic TRIGA core design

(Mark I and II) consists of a cylindrical

geometry and uses fuel moderator

rods. The TRIGA core is also located at

the lower part of the reactor pool.

The fuel elements of both types (MTR

or TRIGA) are made of a fuel meat

­section containing the fissile material

and outside cladding material. Within

this work, the fuel meat and cladding

material are the smallest units of

which a fuel element is made of. In

Figure 2-2 the modularization of an

MTR core is shown.

| | Fig. 2-1.

Generic design sense of TRIGA and MTR research reactors and modularisation of main components.

465

AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP

2.1 Abstraction and

modularisation of research

reactor designs

Due to their different applications in

the field of science, technology and

| | Fig. 2-2.

Modularisation of the reactor core (MTR example).

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| | 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

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pictured. Typically, four control rods

are required for reactivity control in

TRIGA reactors with thermal power

levels of less than 1 MW [IAEA2016B].

Further, graphite elements are at the

outer positions. For MTR research

reactors, there are often empty places

at the centre of the core grid for

radiation samples. If the input number

of fuel assemblies or elements

does not match the number of grid

positions, the implemented algorithm

considers these typical core characteristics.

The assemblies or elements are

positioned in respect of this information.

Furthermore, the free flow path

is calculated as a function of total core

area and number of fuel assemblies,

elements and other components

inside the research reactor core.

3 Generated input decks

of exemplary MTR and

TRIGA reactors

In this part, first functionality of

the new modelling system is demonstrated

by generating an exemplary

MTR and TRIGA research reactor

model. For this purpose, two reference

research reactors were chosen.

Providing technical details in

[ABD2008A] and comparative data in

[ABD2008B], the ETRR-2 was identified

as a MTR reference facility. The

ETRR-2 is a multipurpose research

reactor located in Inshas, the Arab

Republic of Egypt. It corresponds to

the rightmost research reactor design

in Figure 2-1. The ETRR-2 reactor

consists of 29 fuel assemblies of MTR

type with 19 fuel plates each and has

22 MW nominal power. Further

description is presented in [ABD2008].

The main nodalisation of the generated

ETRR-2 model in ATHLET is

pictured in Figure 3-1. On the left

side, the coolant loop is presented

in bright blue. The reactor pool is

modelled with two pipes interconnected

by cross-connections. The

inner pool pipe is connected to the reactor

chimney, which is marked in

brown, by a single junction pipe. The

reactor core is modelled with two

representative assemblies. Each is

composed of 18 core cooling channels.

One assembly is representing 28

grouped average assemblies. The

other assembly considers a hot channel

factor on the 19 fuel plates plus

one extra penalised fuel plate. The

nodalisation of both assemblies is

identically and shown in Figure 3-2.

To check the capability of the

nodalisation to reproduce the thermal

hydraulic plant conditions, steady

state calculations were performed.

| | Fig. 2-5.

MTR core layout (left) and TRIGA core layout (right), generated by software for input deck generation.

| | Fig. 3-1.

Overview of whole Nodalisation of the ETRR-2 (left) and one fuel assembly (right) with 18 core channels generated by the software

for input deck generation.

Power

[MW]

Loop mass

flow

[kg/s]

The initial conditions of the experiment

and the calculated parameters

are compared in Table 3-1. The

experiment was performed at 9.5 MW

thermal power. There is good agreement

between the calculated and

experimental stationary data.

As an exemplary TRIGA research

reactor, the IPR-R1 was identified.

The IPR-R1 is a TRIGA Mark I model,

installed in Belo Horizonte in Brazil

and operated since 1960. Several

analytic and experimental studies

were performed and published. As

reference data, experimental results

in [REI2009] were used. The IPR-R1

corresponds to the leftmost research

reactor design in Figure 2-1. It is

operating at 250 kW and consists of

63 fuel elements of TRIGA type.

Further description is presented in

[REI2009]. The main nodalisation of

the generated IPR-R1 model in

ATHLET is shown in Figure 3-2. On

the left side, the coolant loop is

Core mass

flow

[kg/s]

Core outlet

temperature

[°C]

Core pressure

drop

[bar]

| | Tab. 3-1.

Overview of whole Nodalisation of the ETRR-2 (left) and one fuel assembly (right) with 18 core channels generated by the software

for input deck generation.

presented in bright blue. The reactor

pool is modelled with two pipes interconnected

by cross-connections. The

inner pool pipe is connected to the

core entrance and core outlet. 13 core

channels, interconnected by crossconnections,

with 63 fuel elements

represent the reactor core (see Figure

3-2 right). The core nodalisation

based on the nodalisation presented

in Figure 2-4.

The experiment was performed at

50 kW thermal power. In Table 3-2,

the calculated steady state results are

compared to measured core inlet and

outlet temperatures. At different

positions, measuring devices were

installed (see [REI2009]). There are

small deviations but overall the results

are consistent.

Further, the ATHLET simulation is

compared to published RELAP steady

state calculation in [REI2009], which

reaches steady state conditions after

about 2000 s simulation time.

Reference

pressure

[bar]

Calculation 9.5 309.24 302.86 35.01 0.42 2.2

Reference

[ABD2015]

9.4 309.24 302.87 34.9 0.31* 2.0

*in [ABD2015] core pressure drop of 3.1 bar is mentioned, but in /IAEA2005/ 0.6 bar pressure drop at 100 % core power is referred

467

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References

468

AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP

| | Fig. 3-1.

Overview of whole Nodalisation of the IPR-R1 (left) and 13 core channels (right) generated by the software for input deck

generation.

Power

[kW]

| | Tab. 3-2.

Thermal hydraulic data IPR-R1.

Core inlet

temperature

(Position 3)

[°C]

Core outlet

temperature

(Position 3)

[°C]

There is good agreement between

the published RELAP calculations in

[REI2009] and the calculated ATHLET

data.

4 Summary

A new method based on a heuristic

approach for modelling selected

research reactor types in thermal

hydraulic analysis codes is presented.

This new approach allows a fast and

reliable generation of the input deck’s

fundamental elements despite limited

technical documentation. Focusing on

one MTR and one TRIGA design, the

main steps of developing process and

the characteristics of the new method

are highlighted. This includes the

Core inlet

temperature

(Position 8)

[°C]

Core outlet

temperature

(Position 3)

[°C]

Calculation 51 20.87 27.97 20.87 23.94

Reference

[REI2009]

50 20.95 26.95 22.95 24.95

abstraction and modularisation of

research reactor plant designs as well

as the conception of type-specific

nodalisation. At the end of this paper,

an exemplary MTR and TRIGA

research reactor is presented, generated

by the developed software.

Focusing on the stationary conditions,

there is a good agreement between

the calculated and experimental data.

This proves the basic functionality of

the developed modelling system by

generating a realistic plant model for

TRIGA and MTR type. In future work,

the nodalisation for both research reactor

designs will be reviewed and

tested against a range of safety transients

and accidents.

ABD2008A

ABD2008B

ABD2015

I.D. Abdelrazek, E.A. Villarino:

ETRR-2 Nuclear Reactor: Facility

Specification; Coordinated

Research Project on Innovative

Methods in Research Reactor

Analysis, organised by IAEA,

October 2008.

I.D. Abdelrazek, E.A. Villarino:

ETRR-2 Nuclear Reactor:

Experimental Results

Coordinated Research Project

on Innovative Methods in

Research Reactor Analysis, organised

by IAEA, October 2008.

I.D. Abdelrazek, et al.: Thermal

hydraulic analysis of ETRR-2

using RELAP5 code, Kerntechnik

80, 2015.

ATH2016 G. Lerchl et.al.: ATHLET 3.1A

User’s Manual, GRS-P-1/Vol.1,

Ref.7, March 2016.

IAEA2005

IAEA2016

IAEA2016B

REI2009

RRDB2018

Authors

IAEA: Research reactor

utilization, safety, decommissioning,

fuel and waste management,

ISBN 92-0-113904-7,

IAEA 2005.

IAEA: Safety of Research

Reactors, IAEA Safety Standards

Series No. SSR-3, Vienna

Austria, 2016, ISSN 1020-525X.

IAEA: History, development and

future of TRIGA research

reactors, Technical Report

Series No. 482, ISBN 978-92-0-

102016-1, IAEA 2016.

P. A. L. Resi, et al.: Assessment of

a RELAP5 model for the IPR-R1

TRIGA research reactor, International

Nuclear Atlantic

Conference – INAC 2009,

ISBN: 978-85-99141-03-8.

IAEA: Research Reactor

Database, Website URL:

https://nucleus.iaea.org/RRDB/

RR/ReactorSearch.aspx?rf=1

(01.02.2018).

Vera Koppers

Prof. Dr.-Ing. Marco K. Koch

Responsible Professor

Ruhr-Universität Bochum (RUB)

Universitätsstraße 150

44801 Bochum, Germany

| | Fig. 3-2.

Core inlet (left) and core outlet (right) temperature.

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Development and Validation of a CFD

Wash-Off Model for Fission Products

on Containment Walls

Katharina Amend and Markus Klein

The research project aims to develop a CFD model to describe the run down behavior of liquids (wall films, transition

of film flow into a discrete number of rivulets, droplets) and the resulting wash-down of fission products on surfaces in

the reactor containment. Numerical experiments allow for a deeper physical understanding, which is the basis for an

improved semi-empirical modeling.

This paper presents a three-dimensional

numerical simulation for water

running down inclined surfaces

coupled with an aerosol wash-off

model and the resulting particle transport

using the software package

OpenFOAM. The wash-off model is

based on the procedure used in AULA

(German: Abwaschmodell für unlösliche

Aerosole, wash-off of insoluble

aerosol particles) in the lumped

parameter code COCOSYS [1]. A

parameter variation was conducted

and the simulation results are compared

to the laboratory experiments

performed by Becker Technologies

[2].

1 Introduction

The desired goal is the prevention of

environmental contamination with

radioactive particles after a core

meltdown in a light water reactor. The

containment in a nuclear reactor

building prevents high pressure radioactive

steam from escaping in the

event of an emergency. During such a

critical accident in a light water

­reactor, most of the fission products

enter the containment building in the

form of soluble and insoluble aerosols.

These particles might deposit

on walls and installation surfaces.

Condensing steam that is also released

into the containment can wash down

even insoluble particles into the

containment sump.

In previous studies [3, 4] the understanding

of the run down behavior

of water, the formation of film flow,

rivulets or droplets, was the main

subject of interest. This study investigates

the wash-off of insoluble

particles based on the run down behavior

of water on inclined plates and

the developing flow patterns using

CFD simulations.

2 Laboratory experiments

The laboratory tests are part of the

THAI AW3 test program [5]. They

investigate the aerosol wash-down

behavior of non-soluble silver from

inclined walls by steam condensate.

Trapezoidal plates (plain stainless

steel or decontamination paint coating)

with different inclinations

are loaded with dry silver aerosol. At

the uppermost part water enters

the plate via a tubular distributor

with a given flow rate. The water

flows down the plate, washes off

part of the particles and is finally

collected in cups, which get exchanged

after a specified time period.

The samples are put into a cabinet

dryer and the remaining aerosol mass

is weighed to quantify the wash-off.

Pictures taken during the experiments

show the flow patterns and run

down behavior of the water on the

plates, see Figure 1.

Two kinds of silver aerosol particles

are used: a fine silver powder

and coarse silver powder. The fine

­silver powder is specified with a particle

diameter of 0.7-1.2 μm for 99.9 %

of the particles and as averaged

par ticle diameter of the undisturbed

powder d p = 1 μm can be assumed.

It has a bulk density of ρ bulk =

1.1 g/(cm 3 ) and a specific surface of

A sp = 2.5 m 2 /g. For the coarse silver

powder the specification of particle

diameter is 1.5 – 2.5 μm (99.9 %).

Here the averaged particle diameter is

d p =2 μm, the bulk density is also

ρ bulk = 1.1 g/(cm 3 ) and the specific

surface A sp = 1.21 m 2 /g.

3 Simulation of the water

field

In previous studies the simulation

of the flow field with three different

inclinations, namely 2°, 10° and 20°,

and with empirical contact angle field

and filtered randomized initial contact

angle distribution (FRICAF) were

presented [4, 6]. The computational

domain is a trapezoidal geometry

(length = 1.215 m, small base =

0.09 m, large base = 0.475 m, Figure

2), as used in the laboratory

experiments [5].

| | Fig. 1.

Pictures of lab tests [5] with 2° inclination and

a mass flow rate of 11 g/s after 2 min and

15 min.

| | Fig. 2.

Schematic of computational domain of

inclined trapezoidal plate, dimensions in mm.

The simulations are carried out

with the software package Open-

FOAM using the standard two-phase

solver InterFoam. The Navier-Stokes

equations for isothermal and incompressible

multiphase flow are solved

and the phase interface is captured

by the Volume-of-Fluid method. The

time step is adjusted such that the

maximum Courant number is below

0.4 to ensure a sufficient level of accuracy.

The time is discretized via Euler

implicit. The inlet is extending over

the entire upper boundary with a

Young

Scientists'

Workshop

Awarded

Katharina Amend

was awarded with the

2 nd price of the 49 th

Annual Meeting on

Nuclear Technology

(AMNT 2018) Young

Scientists' Workshop.

469

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AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP

| | Fig. 3.

Comparison of simulations with empirical contact angle model and the laboratory experiments by Becker

Technologies [5] (in false color representation) with mass flow rate ṁ = 11 g/s (ṁ =12 g/s for inclination

of 10°), three different inclinations (left 2°, middle 10°, right 20°) and without aerosol loading.

given water velocity parallel to the

surface such that a specified mass flow

rate is achieved. The flat plate is

bounded by vertical sidewalls and

has an inclination angle α. Material

properties of water and air are used.

For snapshots of the resulting flow

fields see Figure 3.

The simulations are conducted

with the empirical contact angle

­model and the filtered initial

­randomized contact angle field [6].

The contact angle is specified in the

boundary conditions of the water field

and is taken into account to calculate

the curvature of the water-air interface.

The contact angle has a huge

impact on the formation of rivulets

and their stability as shown in previous

studies [7]. The empirical contact

angle model accounts for the

wetted history and therefore enforces

a spatially and temporally stable

­rivulet flow.

4 Simplified geometry

This study also considers a simplified

geometry with dimensions of 6 cm

x 5 cm, 60° inclination and different

water loadings. As a first step the

­simplified geometry, for which additional

benchmark data from CFD

simulations and experiments are

available, is used for the parameter

variation to save computational effort

and time. Later the findings are

transferred to the larger laboratory

geometry. Also the experimental

data can be used to investigate the

empirical contact angle model [6] in

another scenario than the laboratory

geometry where it was developed. For

the simplified geometry Singh et al.

[8] provide results of CFD simulations,

as do Hoffmann [9] and Iso et.

al [10]. Experiments are conducted by

Ausner [11]. All of the latter use the

identical geometry, but different inlet

conditions (overflow weir and feed

tube) and various simulation tools

(Singh and Iso Fluent, Hoffmann

CFX). In the present study simulations

with constant contact angle and with

empirical contact angle model are

performed. The results for different

Weber numbers are evaluated and

compared to the results of the studies

mentioned above for validation. Five

different Weber numbers (We = 0.02,

We = 0.24, We = 0.47, We = 0.76 and

We = 1.10) are investigated, which

correspond to an increasing water

mass flow rate:

We =

with liquid density ρ l , inclination

angle α, volumetric mass flow rate Q,

surface tension σ, plate width W

and viscosity μ. As the water load

­increases, the flow pattern changes

from a thin rivulet to a more pronounced

rivulet to a fully wetting

­water film (see Figure 4).

The influence of the side walls

is also clearly visible and was also

observed by Hoffmann [9] and Ausner

[11]. With a constant contact angle of

70° (which is the value frequently

quoted in the literature for the material

combination water on steel) the

percentage of wetted area in the

present CFD calculations and in the

calculations of Hoffmann and Iso

tends to be underestimated, whereas

the similar setup of Singh yields, for

an unknown reason, larger values

of wetted area. In Figure 5 the

measurements are shown as blue

triangles, the results of Hoffmann, Iso

and Singh in purple, yellow and green,

respectively, and the current calculations

with the different contact angles

in red, gray and black. The simulations

with constant contact angle and

empirical contact angle model with

70° for a dry surface and 50° for a wet

one still slightly underestimate the

wetted surface. With the empirical

contact angle model 30°/70° the

results are very well within the variation

of the experiments.

5 Wash-off model

The particle wash-off consists of a

two-stage process. First the sedimented

particles on the plate floor are

| | Fig. 4.

Comparison of simulations with empirical contact angle model with

θ dry = 70° and θ wet = 30° for different Weber numbers We. The water height

is indicated by color.

| | Fig. 5.

Normalized wetted surface A wn for different Weber numbers. Blue triangles

indicate the experimental results; results of CFD simulations are displayed

with differently colored lines.

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resuspended into the water flow, and

then they are transported by the water

flow down the plate and through the

outlet. The model is based on the

approach suggested and investigated

in [1], which is also implemented in

AULA.

5.1 Shields criterion

In this section wash-off criteria, i.e.

the circumstances that have to be met

to resuspend settled particles, are

presented.

Many forces act upon a particle

lying in a sediment bed. The particle

starts to move, if the hydrodynamic

forces and the buoyancy exceed the

forces of gravity, friction, cohesion

and adhesion. Shields proposes a

criterion, which states, that the

incipient motion occurs, when the

shear velocity acting on the particle

exceeds a critical threshold, the critical

shear velocity. This critical shear

velocity u c can be approximated with

the help of the Shields-Rouse equation

[12]. Using the dimensionless

Rouse Reynolds number R *

(1)

(with the specific gravity of sediment

, the particle density ρ p , the

gravitational acceleration g and the

kinematic viscosity of water ν) the

critical dimensionless shear stress τ c

*

and further the critical shear velocity

u c can be calculated via

.

, (2)

(3)

In this relation the adhesion and

cohesion forces are neglected. Thus

according to this criterion all particles

with the same diameter and density

would erode exactly at the same time

as soon as u > u c holds. This leads to

the so called instantaneous total

wash-off. One way to also take the

adhesion and cohesion forces into

account is to model the wash-off as an

exponential decay of the sedimented

particle concentration c s (t) with a

mass erosion rate r e [13], defined as:

,(4)

(5)

and erosion constant (or wash-off

­coefficient) ~ r e [13] which has to be

estimated.

5.2 Particle transport

The second stage is the transport of

the volumetric particle concentration

c with [c] = kg/m 3 . It is based on the

OpenFOAM solver scalarTransport-

Foam, which solves a simple transport

equation for a scalar volume field

.

(6)

The resuspended aerosol concentration

is treated as massless particles

that follow the flow perfectly. The

­velocity field v, shared by the water

and air phase, is set to zero in cells

without water. Thus particles are

transported only within water and

not within air. The concentration of

eroded particles in each floor face at

each time step serves as the source

term S in the corresponding cell above

the floor. The result of the simulations

Name

is the time-resolved particle mass that

is transported through the outlet.

6 Results of the parameter

variation

In this section parameters such as

particle density and the wash-off

­coefficient are varied. Detailed correlations

or influences of the parameters

on the total washed-off mass are

analyzed. Table 1 summarizes the

constant particle properties, the initial

plate loading and the properties of

the water flow. To investigate the

­influence of the particle density ρ p

three different densities are used:

10 000 kg/m 3 which resembles the

density of silver, 5000 kg/m 3 and

2500 kg/m 3 which is the effective

density of the aerosol.

The erosion constant ~ r e is also

­varied with values of 0.027 s –1 ,

0.135 s –1 and 0.27 s –1 . Together with

the five different Weber numbers (see

Sec. 4) the parameter variation covers

a total number of 45 simulations that

are evaluated hereinafter.

The parameter variation is conducted

with the simplified geometry.

Three seconds of water field simulations

are calculated. The water field is

then in a pseudo-stationary state

and the water, the velocity and the

pressure fields are kept constant.

Overall the particle wash-off and

Value

α Inclination 60°

ρ l Density of water 1000 kg/m^3

c s Initial loading 27 g/m^2

θ dry Contact angle dry 70°

θ wet Contact angle wet 30°

d p Particle diameter 2 μm

| | Tab. 1.

Parameters for simulations with simplified geometry.

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| | 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 .

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

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| | Fig. 8.

Time resolved washed off mass for different wash-off coefficients.

Parameters according to Table 1 and ρ p = 5000 kg/m 2 .

transport simulations last for 30 s.

First the influence of the Weber

number is investigated, see Figure 6.

Increasing Weber numbers correspond

to larger water velocities and

an increasing percentage of wetted

surface. Consequently for larger

Weber numbers more particle mass is

washed off. Thus two effects manifest

in the results: first the larger velocities

are able to wash-off even particles

with larger density. And secondly the

enlarged percentage of wetted surface

enhances the particle wash-off, since

much more particles can be eroded

by the water.

Figure 7 shows the variation of the

particle density. Particles with larger

density cannot be eroded that easily

and hence the total washed off mass

decreases with increasing particle

density, as expected. In Figure 8 the

influence of the wash-off coefficient is

investigated. The total washed off

mass, which is to a large extent determined

by the area of wetted surface,

does not change with different values

of ~ r e but the temporal behavior does.

For a large value of ~ r e a large fraction

of particles erodes in a short timespan.

Asymptotically for t → ∞ the

total washed off mass converges

always to the same amount.

In order to compare the simulations

with experimental data a parameter

set based on Weber et. al [1]

is chosen. Figure 9 displays the results

of the simulation and the experimental

data of test 4. In the experiments

the particles are collected in intervals

of 10 s for a total duration of 130 s.

Due to this sampling strategy the time

resolved washed off particle mass in

the simulations is presented in the

same manner and for the same

duration. A good agreement for the

temporal course of the wash-off as

well as for the total washed off mass

can be achieved.

7 Conclusions and

discussion

This paper presents a CFD particle

wash-off model and particle transport

by gravity driven flows. A parameter

variation was conducted within the

setting of a simplified geometry and

with the geometry of the laboratory

tests. The particle wash-off model,

which is based on Shields criterion

[12] and Weber et. al [1], shows the

expected behavior for varying particle

properties such as particle density and

wash-off coefficient. One key influencing

parameter for the resulting

washed off mass is the percentage of

area covered by water in each case,

which differs with inclination and

mass flow rate. First simulations

with the laboratory geometry show

satisfactory agreement when compared

to the experiments. Nevertheless,

the prediction of particle

wash-off for a large variety of setups

as in the laboratory experiments

( different inclinations, particle and

surface properties and initial loadings)

remains a great challenge and

further comparisons for different

parameter sets are current work in

progress. This study contributes to

the development of a semi-empirical

model to quantify the aerosol washoff

and the wetted surface area during

an accident in a light water reactor.

Acknowledgment

The project underlying this report

is funded by the German Federal

Ministry of Economic Affairs and

Energy under grant number 1501519

on the basis of a decision by the

German Bundestag. The THAI project

was carried out on behalf of the

Federal Ministry for Economic Affairs

and Energy under grant number

1501455 on the basis of a decision by

the German Bundestag. We are also

grateful for the support from Becker

Technologies and the GRS.

References

| | Fig. 9.

Comparison of test 4 of the laboratory experiments with the simulations of particle wash-off

with inclination α = 20°, mass flow rate m = 11 g/s, initial loading c s = 12.5 g/m 2 ,

particle diameter d p = 2 μm, particle density ρ p = 5000 kg/m 3 and wash-off coefficient ~ r e = 0.025 s –1 .

[1] G. Weber, F. Funke, W. Klein-Hessling,

and S. Gupta. Iodine and silver washdown

modelling in COCOSYS-AIM by

use of THAI results. Proceedings of the

International OECD-NEA/NUGENIA-

SARNET Workshop on the Progress in

Iodine Behaviour for NPP Accident

Analysis and Management, 2015.

[2] S. Gupta, F. Funke, G. Langrock, G.

Weber, B. von Laufenberg, E. Schmidt,

M. Freitag, and G. Poss. THAI Experiments

on Volatility, Distribution and

Transport Behaviour of Iodine and

Fission Products in the Containment.

Proceedings of the International

OECD-NEA/NUGENIA-SARNET Workshop

on the Progress in Iodine

Behaviour for NPP Accident Analysis

and Management, p. 1-4, 2015.

[3] M. Freitag, B. von Laufenberg, M.

Colombet, K. Amend, and M. Klein.

Particulate fission product wash-down

from containment walls and installation

surfaces. Proceedings of the 47 th

Annual Meeting on Nuclear

Technology, Hamburg, 2016.

AMNT 2018 | Young Scientists' Workshop

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

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[4] K. Amend and M. Klein. Modeling and

Simulation of Water Flow on Containment

Walls with Inhomogeneous

Contact Angle Distribution. ATW

International Journal for Nuclear

Power, 62(7):477-481, 2017.

[5] B. von Laufenberg, M. Colombet, and

M. Freitag. Wash-down of insoluble

aerosols Results of the Laboratory Test

related to THAI AW3 Test. Technical

report, Becker Technologies, 2014.

[6] K. Amend and M. Klein. Simulation of

Water Flow down inclined Containment

Walls. 14 th Multiphase Flow

Conference, Dresden, 2016.

[7] K. Amend and M. Klein. Influence of the

contact angle model on gravity driven

water films. 13 th Multiphase Flow

Conference, Dresden, 2015.

[8] R. K. Singh, J. E. Galvin, and X. Sun.

Three-dimensional simulation of rivulet

and film flows over an inclined plate:

Effects of solvent properties and contact

angle. Chemical Engineering Science,

142:244–257, 2016.

[9] A. Hoffmann. Untersuchung mehrphasiger

Filmströmungen unter

Verwendung einer Volume-Of-Fluidähnlichen

Methode. PhD thesis,

Technische Universität Berlin, 2010.

[10] Y. Iso, X. Chen. Flow transition behavior

of the wetting flow between the film

flow and rivulet flow on an inclined

wall. Journal of Fluids Engineering

133.9:091101, 2011.

[11] I. Ausner. Experimentelle Untersuchungen

mehrphasiger Filmströmungen.

PhD thesis, Technische

Universität Berlin, 2006.

A Preliminary Conservative Criticality

Assessment of Fukushima Unit 1 Debris

Bed

María Freiría López, Michael Buck and Jörg Starflinger

[12] J. Guo. Hunter Rouse and Shields

diagram. Advances in Hydraulic and

Water Engineering, 2:1096–1098,

2002.

[13] R. Ariathurai. A finite element model of

cohesive sediment transportation. PhD

thesis, University of California, Davis,

California, 1974.

Authors

Katharina Amend

Prof. Dr.-Ing. habil. Markus Klein

Responsible Professor

Institute for Numerical Methods in

Aerospace Engineering Universität

der Bundeswehr München

Werner Heisenberg Weg 39

85577 Neubiberg, Germany

Young

Scientists'

Workshop

Awarded

473

AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP

1 Introduction On March 11, 2011, a big severe accident occurred at Fukushima Daiichi nuclear power plant

(NPP) in Japan resulting in largely melted cores of Units 1, 2 and 3. After the corium solidification, debris beds

were formed and they are considered to be distributed not only in the reactor pressure vessel but also in the primary

containment. If such debris enter in contact with water, recriticality becomes possible. To prevent recriticality, severe

accident mitigation measures prescribe the injection of borated water into the reactor core. However, some leakage of

cooling water and the inflow of groundwater into the reactor building make it very difficult to maintain the necessary

boron concentration to secure the subcritical condition. Currently, the subcriticality of the debris bed is being monitored

by measurements of short lifetime fission products gas (e.g. Xe 133 or Xe 135 ) and water temperature [1]. As no sign of

criticality has been detected until now, the fuel debris is estimated to be subcritical and no preventive measure against

a possible recriticality event is being taken [2]. Nonetheless, this apparently critical-stable condition can change at any

moment due to changes in debris conditions. During the retrieval operations, changes in the water level and debris

shape are expected to occur that will endanger this stability. Thus, using borated water is then planned to ensure the

subcriticality [3].

María Freiría López

was awarded with the

3 rd price of the 49 th

Annual Meeting on

Nuclear Technology

(AMNT 2018) Young

Scientists' Workshop.

A recriticality scenario would lead to a

power increase, new fission products

release and may have severe consequences

even causing a secondary

criticality accident. Prevention and

controlling core sub-criticality is

there fore one of the main accident

management objectives. A risk evaluation

of recriticality is necessary for

the safe preservation and handling of

fuel debris.

This study is part of a larger project,

which pursues to assess the

recriticality potential of fuel debris

after a severe accident taking

­Fukushima as reference. The final

aim is to develop a criticality map that

will be used to evaluate the potential

risk of criticality of a fuel debris

taking the debris conditions as input

parameters. The criticality situation of

Fukushima damaged reactors will be

assessed by placing onto the map the

fuel debris conditions revealed by

observations or sample analyses.

In this study, a conservative

criticality evaluation of the Fukushima

Daiichi Unit 1 debris bed was carried

out. Parameters, such as debris size,

porosity, particle size, fuel burnup

and the coolant conditions including

the water density and the content of

boron were considered. The effect of

these parameters on the neutron

multiplication factor was analysed

and safety parameter ranges, i.e.

zones where the recriticality can be

totally excluded, have been identified.

The objective is to fix some boundaries

for the selected parameters

and define the ranges in which the recriticality

could be an issue. This will

provide the starting point for a future

more detailed criticality evaluation.

The Monte Carlo code MCNP6.1

was used to model the hypothetical

debris bed and to calculate the

neutron multiplication factor (k eff )

[4]. The ENDF/B-VII.1 cross section

libraries were used to perform the

calculations.

2 Criticality of debris bed

after a severe accident

After a severe accident (SA), recriticality

occurs when the whole or part of the

reactor becomes unintentionally critical

after the reactor shutdown. This

study focuses on the analysis of recriticality

in debris beds that are formed

either at the bottom of the reactor

vessel (in-vessel debris bed) or in the

reactor containment (ex-vessel debris

bed) after the cool down of the reactor.

Debris beds are formed during a SA

after the solidification of the melted

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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

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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

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References

476

1. Tsuchiya A, Kondo T, Maruyama H.

Criticality calculation of fuel debris in

Fukushima Daiichi nuclear power station.

In: PHYSOR 2014. Kyoto, Japan; 2014.

AMNT 2018 | YOUNG SCIENTISTS' WORKSHOP

| | Fig. 4.

Debris size – Burnup Unit 1 Fukushima Daiichi criticality map.

critica lity cannot be reached. A boration

of 1,600 ppm B will ensure the

subcriticality independently of the

debris bed conditions.

Figure 4 provides criticality data

as function of the debris size and

burnup. It can be noticed how the k eff

decreases progressively with the

burnup of the core. If the SA happens

at the very end of a fuel cycle, when

the average burnup of the fuel is larger

than 53 GWd/t HM , recriticality will

not be reached under any conditions.

Additionally, the graph provides

the information about the criticality

condition of a debris bed depending of

its size. With these data, the critical

masses for the different burnups

can be calculated. The burnup of

Fukushima Unit 1 at the moment of

the accident was estimated to be

25.8 GWd/t HM . The minimum critical

size of a debris bed for this case is

about 55 cm. For these conditions, the

optimum porosity was calculated to

be 0.75. This results in critical mass of

226.5 kg, which represents only the

2.4 % of the core.

Conclusions

In this study, a conservative criticality

evaluation of the current debris bed

of Fukushima Daiichi Unit 1 was

performed. The lack of knowledge

regarding the debris bed properties

has compelled the use of very conservative

assumptions in the debris

bed models. Six of the most influencing

parameters on the k eff were considered:

debris size/mass, particle size,

porosity, water density and content of

boron in water. The effect of these parameters

on the criticality condition of

Fukushima Daiichi Unit 1 debris bed

was calculated and discussed. Finally,

it was concluded that recriticality can

be totally excluded if:

1. Porosity of the debris bed is lower

than 0.4 or

2. Void fraction of water is higher

than 78 % or

3. Debris mass is lower than 226.5 kg

or

4. Boration in water is equal or

greater than 1,600 ppm B

Additionally, for a reactor core with

UO 2 fuel and initial enrichment of

3.7 % wt 235 U it was found that if a

SA occurred at the very end of a fuel

cycle when the average burnup is

53 GWd/t HM or higher, recriticality is

not achievable under any conditions.

Taking severe accident scenarios

into account, the void fraction threshold

(2) and the debris mass threshold

(3) will be violated under almost all

circumstances. The molten mass

easily reaches values higher than

226 kg, which represents only 2 % of

the core mass, and the void fraction

does not stay at values higher than

78 % for the range of cool temperatures

considered. However, experiments

like DEFOR [11] or FARO [12]

indicate average porosities of about

38 %, which is slightly underneath

the “criticality safe” threshold (1) for

porosity.

As a next step, it is planned to

include new parameters, for example,

the presence of zirconium, control

rods or other reactor structural materials

in order to evaluate their

­influence on the criticality of debris

beds. Additionally, new debris bed

configurations will be also investigated.

The first samples and explorations

of debris beds in Fukushima are

planned for this year 2018. This

will provide more information

about the debris characteristics and

will allow a less conservative

and more accurate criticality evaluation.

Acknowledgments

The presented work was funded by

the German Ministry for Economic

Affairs and Energy (BMWi. Project no.

1501533) on basis of a decision by the

German Bundestag.

2. Kotaro Tonoike, Hiroki Sono, Miki Umeda,

Yuichi Yamane, Teruhiko Kugo, Kenya

Suyama. Options of Principles of Fuel Debris

Criticality Control in Fukushima Daiichi

Reactors. In: Ken Nakajima, editor. Nuclear

Back-end and Transmutation Technology

for Waste Disposal. Springer Open;

2015. p. 251–60.

3. Nuclear Damage Compensation and

Decommissioning Facilitation Corporation.

Technical Strategic Plan 2016 for

Decommissioning of the Fukushima

Daiichi Nuclear Power Station of Tokyo

Electric Power Company Holdings, Inc.

2016 Jul.

4. Goorley, John T., James, Michael R.,

Booth, Thomas E., Brown, Forrest B., Bull,

Jeffrey S., Cox, Lawrence J., et al. Initial

MCNP6 Release Overview – MCNP6 version

1.0. Los Alamos National Laboratory

(LA-UR-13-22934); 2013.

5. GPU NUCLEAR. Three Mile Island

Nuclear Station Unit II Defueling

Completion Report. 1990.

6. Freiría López M, Buck M, Starflinger J.

Neutronic Modelling of Fuel Debris for a

Criticality Evaluation. In: PHYSOR 2018.

Cancun, Mexico; 2018.

7. International Atomic Energy Agency

(IAEA). The Fukushima Daiichi Accident

Technical Volume 1/5 Description and

Context of the Accident Annexes.

Vienna (Austria): International Atomic

Energy Agency (IAEA); 2015.

8. Nishihara K, Iwamoto H, Suyama K.

Estimation of fuel compositions in

Fukushima-Daiichi nuclear power plant.

Japan Atomic Energy Agency; 2012.

9. Croff AG. ORIGEN 2.1. Oak Ridge

National Laboratory; 1991.

10. Nuclear Safety Standards Commission

(Kerntechnischer Ausschuss, KTA).

Storage and Handling of Fuel Assemblies

and Associated Items in Nuclear

Power Plants with Light Water Reactors.

2003 Nov. Report No.: KTA 3602.

11. Kudinov P, Karbojian A, Tran C-T,

Villanueva W. Agglomeration and size

distribution of debris in DEFOR-A

experiments with Bi2O3–WO3 corium

simulant melt. Nucl Eng Des.

2013;263(Supplement C):284–95.

12. Magallon D. Characteristics of corium

debris bed generated in large-scale

fuel-coolant interaction experiments.

Nucl Eng Des. 2006;236(19):1998–

2009.

Authors

María Freiría López

Dr.-Ing. Michael Buck

Prof. Dr.-Ing. Jörg Starflinger

Responsible Professor

Institute of Nuclear Technology

and Energy Systems (IKE)

University of Stuttgart

Pfaffenwaldring 31

70569 Stuttgart, Germany

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49 th Annual Meeting on Nuclear Technology (AMNT 2018)

Key Topic | Outstanding Know-How

& Sustainable Innovations

The following report summarises the presentations of the Focus Session International Regulation | Radiation

Protection: The Implementation of the EU Basic Safety Standards Directive 2013/59 and the Release of

Radioactive Material from Regulatory Control presented at the 49 th AMNT, Berlin, 29 to 30 May 2018.

The other Focus, Topical and Technical Sessions will be covered in further issues of atw.

477

AMNT 2018

Key Topic: Outstanding

Know-How & Sustainable

Innovations

Focus Session International

Regulation: Radiation Protection:

The Implementation of the EU

Basic Safety Standards Directive

2013/59 and the Release of

Radioactive Material from

Regulatory Control

Christian Raetzke

The topical session Radiation Protection:

The Implementation of the EU

Basic Safety Standards Directive

2013/59 and the Release of Radioactive

Material from Regulatory

Control was coordinated and chaired

by the author of this report.

As the chairman explained in his

short introductory statement, the implementation

of the Basic Safety

Standards (BSS) Directive 2013/59/

Euratom, which has introduced many

changes in radiation protection, has

posed considerable challenges to EU

Member States. In Germany, it became

the occasion for a major revision of the

legal framework and the creation of a

new Act on Radiation Protection. The

chairman expressed his delight that

two distinguished speakers had consented

to talk about implementation

in Germany and Sweden: Dr. Goli-

Schabnam Akbarian from the Federal

Ministry for Environment, Nature

Conservation and Nuclear Safety (BMU)

and Dr Jack Valentin from Sweden.

An important aspect in the regulation

of radiation protection is the

release of radioactive substances from

regulatory control. This is a topic

particularly discussed in Germany

where huge amounts of debris are

produced, and will continue to be produced,

by the dismantling of the fleet

of nuclear power plants. Two eminent

speakers had agreed to shed light on

this issue under a multinational, comparative

angle: Dr Edward Lazo from

the OECD Nuclear Energy Agency and

Dr. Jörg Feinhals from DMT.

As the first speaker, Dr. Goli-

Schabnam Akbarian (Head of Division

“Radiation Protection Law [ionising

radiation]” at the German Federal

Ministry for the Environment, Nature

Conservation and Nuclear Safety) outlined

The Implementation of the

New Euratom BSS in Germany. First,

Ms. Akbarian explained the genesis of

the new Act on Radiation Protection

(Strahlenschutzgesetz, StrlSchG). It

was triggered by the need to transpose

the BSS Directive 2013/59 into

national German law. However, there

were additional reasons for laying a

new foundation for German radiation

protection law which had hitherto

been regulated “merely” by a Government

ordinance (Strahlenschutzverordnung,

StrlSchV). For example, after

Fukushima a need was perceived to

revise the provisions on emergency

preparedness and response which

were scattered among different legal

texts and guidelines. The main body

of the German Strahlenschutzgesetz of

27 June 2017 was to enter into force

on 31 December 2018. It was to be

supplemented by a set of new

ordinances which, as Ms. Akbarian

explained, were currently under

preparation.

Next, she turned to the new structure

introduced by the Directive

2013/59, namely the three exposure

situations: planned, existing and

emergency exposure situations. The

Directive has a greatly enlarged scope

of application as compared to its

predecessor, the Directive 96/29/

Euratom, especially regarding NORM

(naturally occurring radioactive material)

and existing exposure situations.

However, Ms. Akbarian focused

on the category of planned exposure

situations which regards practices. i.e.

human activities that can increase the

exposure of individuals to radiation

from a radiation source. In this area of

particular importance to the nuclear

industry, she highlighted some areas

where meaningful changes had been

introduced. One example was exemption

values which were – though not

too substantially – adapted, which

may result in some activities to require

a licence which had hitherto been

exempted. New requirements were

also introduced concerning the

handling of high-activity sealed

­sources. Further modifications affected

the transport of radioactive substances,

a slight change in the dose

limits for occupational exposure and

the introduction of an inspection

programme. For all of these issues, the

new Act included transitional provisions

to allow smooth adaptation.

Ms. Akbarian concluded by

mentioning a host of other aspects

regulated by the new Act, such as

type approval, clearance, radon in

dwellings and at workplaces, and

many others. It became apparent that

the new Act is of fundamental importance,

laying a new foundation for

an area of nuclear law – the law of

radiation protection – which will

become even more important in the

future.

In the ensuing discussion, Ms.

Akbarian was asked about how the BSS

Directive's concept of radiation protection

expert (RPE) and radiation protection

officer (RPO) had been ­taken

into account in the German Act. She

replied that the traditional two roles

defined in German radiation protection

law, namely the person responsible

for radiation protection (Strahlenschutzverantwortlicher,

SSV) and the

expert entrusted with operational

radiation protection (Strahlenschutzbeauftragter,

SSB), had been retained

as they fulfil this concept. The SSB

basically performed both the role of

the RPE and the RPO. The new Act

strengthened his position, e.g. by introducing

protection against dismissal by

the employer. Another question from

the audience concerned the new

notion of dose constraints and how

stringent requirements for the operator

were. Ms. Akbarian explained

that dose constraints (Dosisrichtwerte)

were included in the new Act and in

supplementing ordinances but that

they were mainly an instrument of

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optimisation to be used by the regulator.

However, there were some requirements

on persons starting a new

practice to analyse whether dose

constraints were useful for this practice,

and to document this analysis

and, if asked, provide the analysis to

the authority.

Next, Dr Jack Valentin (independent

consultant, Sweden, former

­Scientific Secretary of the ICRP,

former senior radiation protection

regulator in Sweden) gave a presentation

on Implementation of the EU

BSS Directive in Sweden. First, Jack

Valentin outlined the genesis of the

new radiation protection requirements,

particularly the role of the

International Commission on Radiation

Protection (ICRP) and its

Recommendation no. 103 which was

the basis for the BSS Directive.

He highlighted four essential new

features of ICRP 103: The focus on the

exposure situation (planned/emergency/existing),

not the process

(practice/intervention); the optimisation

of radiological protection in all

exposure situations; the modulation

of optimisation using dose and risk

constraints, and finally, enhanced

protection of the environment by

maintaining biodiversity and ecosystems.

Interestingly, Jack Valentin

highlighted two issues where the BSS

Directive – and, as a consequence,

Swedish legislation and regulation –

was not fully in line with ICRP 103.

One concerned dose limits for occupational

exposure where the Directive

fixes an annual dose of 20 mSv with

no automatic averaging over five years

as had been the case before. As Jack

Valentin pointed out, averaging over a

5-year period facilitated the operator’s

optimisation of protection; for

example, in the case of rare major

jobs, the lowest collective dose was

achieved if a few specialists took

relatively high individual dose.

Concerning emergency worker dose

levels, whereas ICRP did not introduce

any dose limit for a life-saving

informed volunteer, relying instead on

an individual risk/benefit assessment,

the Directive featured a dose limit of

500 mSv. In Jack Valentin's view,

inexperienced rescue leaders might in

future be likely to omit life-saving for

fear of transgression (although doses

will rarely be higher than 500 mSv).

He next depicted the implementation

of the BSS Directive in Sweden

on three levels: the 2018 Radiation

Protection Act, the 2018 Radiation

Protection Ordinance and the 2018

Radiation Protection Regulations

(which have legal force and usually

also include a separate section giving

advice). Like in Germany, these new

or modified texts brought the law fully

into line with the Directive; in some

instances, they use a wording somewhat

different from that of the Directive

(e.g. Swedish law retained the

denomination “activities with ionising

radiation” for planned exposure

situations). And, like in Germany,

there were other reasons for the

legislative and regulatory overhaul

besides the BSS Directive.

When asked about why dose limits

in Sweden were contained in the regulations

rather than in the Act or the

Ordinance, Jack Valentin replied that

this provided some flexibility since

they could more easily be changed. Dr.

Akbarian noted that this was an interesting

viewpoint; she observed the

German view was rather to enshrine

them in legislation because of their

basic importance. Jack Valentin consented

that either view is perfectly

reasonable from its respective angle.

Responding to another comment, Jack

Valentin highlighted the importance

of participation of the public which

had always been a prominent feature

of Swedish nuclear and radiation

protection law and of more general

environmental law.

Next, Dr Edward (Ted) Lazo (Principal

Administrator, Division of Radiological

Protection and Human Aspects

of Nuclear Safety, OECD Nuclear

Energy Agency, Paris) spoke about

The NEA Report on Recycling and

Reuse of Materials Arising from

Decommissioning of Nuclear Facilities.

As Ted Lazo explained, significant

volumes of materials will be gen erated

from decommissioning of nuclear

facilities throughout the world. In

Europe, more than a third of currently

operating reactors were due to be shut

down by 2025. The importance of the

management of slightly contaminated

material was likely to grow and the

inherent value of these materials and

the need to reduce radioactive waste

to be disposed required attention.

However, the international community

was far from a complete

harmonization of the strategies and

regulations on this issue.

In order to rise to this challenge,

the NEA Cooperative Programme on

Decommissioning (CPD) Task Group

on Recycling and Reuse of Material

was created. The Task Group had produced

its first report in 1996; a new

report, updating and extending the

previous one, was released in 2016.

This recent report noted that in the

past two decades, international guidance

had been issued, notably the

IAEA guide RS-G1.7 and several

recommendations of the expert group

under article 31 of the Euratom Treaty.

Still, there was only a limited degree

of alignment of national regulations.

As the report noted, unconditional

clearance – which is normally preferred

to conditional clearance if

possible – is well-regulated in all

countries the report looked at, however

some differences between countries

remained, e.g. in the disposal of

rubble and concrete blocks from

dismantling. For conditional clearance,

in the absence of international

guidance, regulatory systems varied

greatly. As Ted Lazo pointed out, the

BSS Directive may help to achieve

greater consistency.

Generally, as he noted, since the

first report of 1996 a greater consolidation

and alignment of the requirements

to control dose and

exposure to workers, members of the

public and the environment had been

achieved; there was also an increase

in general public awareness but issues

over public acceptability remained.

Education, information sharing and

awareness-raising through direct

and public communications could be

utilized to alleviate many of the fears

surrounding recycling and reuse of

materials. Besides, a well-established

relationship between the nuclear

industry and the recycling industry

could have a considerably positive

effect to ensuring stakeholder and

public acceptance of materials. Ted-

Lazo concluded by saying that numerous

challenges to recycling and reuse

of materials persisted internationally

and that the Task Group felt that

success stories, such as those included

in its report, needed to be shared

internationally to help build consensus

for the safe recycling and reuse of

valuable materials.

Last not least, Dr. Jörg Feinhals

(Head of Project Group “Radiation

Protection and Disposal” at DMT,

Hamburg; Member of the Directorate

of the German-Swiss Association for

Radiation Protection) took the floor on

the topic Necessary Modifications

on Clearance Regulations in Germany

and Switzerland – Comparative

Analysis. Jörg Feinhals first

remarked that comparison between

the two countries is rendered more

difficult by the fact that sometimes

the same (German) word is used

with different meanings – a difficulty

which remarkably cannot arise with

English where there is a common

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understanding in the international

community. Next, Jörg Feinhals

depicted the Swiss situation. The

Swiss Radiation Protection Ordinance

(Strahlenschutzverordnung, StSV) was

revised with effect from 1 st January

2018 in order to keep up with the state

of the art (ICRP 103 and IAEA BSS)

and to be in compliance with EU BSS

Directive in most cases, however without

changing things being tried and

trusted. Besides, complementary

regulations were still in the making.

Jörg Feinhals analysed the criteria for

exemption and clearance in the Swiss

system, namely surface contamination,

net dose rate and activity. He

compared the new Swiss clearance

criteria to the German ones and concluded

that average parameters were

no longer more restrictive in Switzerland

than in Germany.

With a view to the revision of

German radiation protection law explained

by Goli-Schabnam Akbarian

in the first presentation, Jörg Feinhals

focussed on clearance. Clearance,

until now regulated in section 29 of

the existing Radiation Protection

Ordinance, was the object of section

68 of the new Act on Radiation Protection;

however, this section merely

empowered government to regulate

clearance in a new ordinance, which

was still in the making. Based on analysis

of a draft version of this new ordinance,

Jörg Feinhals concluded that

most values in a table appended to the

new ordinance were unchanged as

compared to the existing values in Appendix

3 Table 1 of the existing Ordinance.

However, there were some

changes in detail, most notably a new

term for specific clearance (Spezifische

Freigabe) and mass limits of 10.000

Mg/a for Cs-137 in concrete debris

and 10 Mg/a for scrap, if only one specific

nuclide is detected. As to the effects

of these differences

in terms of masses and cost, Jörg

Feinhals stated that there was a

tendency towards shifting between

clearance pathways (e.g. Cs-137) in a

direction from clearance to specific

clearance, from there to decay storage

and thence to long-term storage.

Besides, he expected in some cases

an increased time expenditure for

measurement or new equipment (e.g.

in the case of Eu-152/154). This

was somewhat offset by increase

of values for some nuclides (e.g. Pu-

238/39/40/41, Am-241). Whereas

the mass limit for concrete debris

Cs-137 was acceptable, the limit for

clearance of metal scrap in case of

single nuclides seemed to be out of

practice, Jörg Feinhals noted. Overall,

he predicted a (merely) moderate

increase of effort and costs, provided

however that the path of specific

clearance proved to be fully operational.

He concluded his presentation

by pointing out some constellations

where difficulties could arise due to a

lack of transitory or grace periods for

specific cases.

When asked about contaminated

soil after accidents, Jörg Feinhals

stated that from the view of emergency

preparedness and response it

was very necessary to have a plan for

the disposal of large amounts of contaminated

soil and of other materials.

This should not be based on the de

minimis concept but rather on the

basis of an existing exposure situation,

i.e. a 1 mSv/a dose limit for the

public.

The session closed with a panel discussion

with the four speakers and the

audience. The chairman opened the

discussion by sharing his impression

that while the BSS Directive and the

implementing legislation in EU

Member States introduced many new

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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

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30 ­institutes of 10 countries as well as representatives of the International Atomic Energy Agency (IAEA) attended the

event. The experts focused on scientific and technical aspects that may be important for extended storage. With 18 oral

contributions the science-focused agenda of the workshop reflected the broad diversity in current research projects.

The subjects ranged from cladding material behavior to the thermo-mechanical simulation of fuel rods and fuel

­assemblies. Furthermore, specific aspects were addressed such as non-destructive testing of casks or management

issues, as well as analysis of the still unresolved technical issues that need to be closed by further research programs.

The sessions started with a talk given

by Maik Stuke from GRS, Germany,

entitled “Current Research Activities

at GRS”. The presented activities focus

on the long-term behavior of drystored

fuel assemblies with special

emphasis on high burn-up values of

65 GWd/tHM UO2 and MOX fuel. The

presentation included detailed maps

of temperature fields of loaded casks.

The thermo-mechanical behavior of

the fuel rods was investigated using

the TESPA-ROD code. Furthermore,

research on the influence of hydride

behavior in cladding materials was

presented e.g. an in-depth analysis of

hydrogen terminal solid solubility.

Representing the IAEA, Alena

Zavazanova provided in her talk “IAEA

safety standards for dry storage of

SNF” an overview of the regulatory

considerations concerning nuclear fuel

management. Some of the IAEA Safety

Standards concerning the storage of

spent nuclear fuel were discussed in

greater detail, e.g. the “General Safety

Requirements” part 5 and 6 of the IAEA

Safety Standard “Predisposal Management

of Radio active Waste”, and the

“Specific Safety Guide 15: Storage of

Spent Nuclear Fuel”.

In their joint presentation

“ Response of Irradiated Nuclear Fuel

Rods to Quasi-Static and Dynamic

Loads” Efstathios Vlassopoulos and

Dimitri Papaioannou presented a

collaborative effort of the École

polytechnique fédérale de Lausanne

(EPFL) in Lausanne, Switzerland, the

Swiss National Cooperative for the

Disposal of the Radioactive Waste

(Nagra), the European Commission

Joint Research Center (JRC) in Karlsruhe,

Germany, and CADFEM (Suisse)

AG in, Aadorf, Switzerland. The group

investigates the response of spent

nuclear fuel in various loading conditions.

The focus lies on the determination

and the study of the

mechanical properties and rod failure

processes using experimental and

numerical techniques.

Jesus Ruiz-Hervias from the Technical

University of Madrid, Spain,

presented in his talk “Effect of Zirconium

Hydrides on the Mechanical

Behaviour of Cladding” investigations

on the effect of hydrogen embrittlement

on the mechanical behaviour

of un-irradiated cladding. One of the

objectives of the work was to develop

operative failure criteria to predict the

cladding behaviour during dry storage

and transport operations. He presented

experimental and numerical

results for ring compression tests of homogeneously

hydrogen loaded samples

and the derived failure criteria.

As chairperson of the Extended

Storage Collaboration Program

( ESCP) Steering Committee of the

Electric Power Research Institute

(EPRI), USA, Hatice Akkurt provided

in her talk “Extended Storage Collaboration

Program (ESCP) for Addressing

Long-Term Dry Storage

Issues” the actual ESCP Program. The

collaboration aims at enhancing the

technical bases to ensure a continued

safe long term used fuel storage and

transportability. It involves about

575 members from 19 countries and

is organized in 6 subcommittees:

Fuel Assembly, Thermal Modelling,

CISCC, Non-Destructive Examination,

Canister Mitigation/Repair, and International.

Amongst other topics she

discussed results from the Demo Project

in which a cask that has been

loaded in 2017 is investigated under

defined conditions.

In his capacity as Sub-Coordinator

Stefano Caruso of the Swiss NAGRA

presented the proposal for the Joint

Programme on Radioactive Waste

Management and Disposal in Europe

(RWMD-EJP). He discussed the aims

of this programme and its current

state of definition with focus on the

budgetary and time planning. As it

involves several authorities, it is

subjected to many constraints. The

proposal is currently undergoing the

second review; final submission is

planned for the end of September. The

first implementation phase will ­extend

over five years (EJP1 2019-2024),

with a maximum budget of 32.5 M€.

In his talk “Sensitivity Tests of

Several Factors Affecting Dynamic

Buckling Strength of Spacer Grids of a

Spent Nuclear Fuel”, Jae-Yong Kim

from the Korea Atomic Research

Institute (KAERI) reported about a

research program on spent nuclear

fuel. He discussed a pendulum impact

tester, installed in 2017 to improve

analytical skills of very limited impact

test results in hot cells. The tests were

established to assure consistency and

qualification of impact test results.

The functional verification tests are

performed to confirm the hammer’s

impact velocity, initial impact energy

and heating conditions of an electric

furnace. Finally, impact tests were

performed with simulated spacer

grids replacing the spent fuel spacer

grids by changing ambient temperature

and cell size.

Michel Herm from the Institute for

Nuclear Waste Disposal of Karlsruhe

Institute of Technology (KIT-INE), Germany,

presented “Research activities

on safety of extended dry storage of

spent nuclear fuel at KIT-INE”.

Using irradiated fuel rod segments

from the PWR Gösgen, Switzerland,

and Obrigheim, Germany, radionuclide

inventories of Zircaloy-4

samples were determined and compared

to theoretical predictions. UO 2

and MOX samples were used and

different methods applied according

to the nuclides. Nuclide inventories

were investigated in the fuel region, as

well as in the plenum. Separation

methods for Chlorine and Iodine are

currently under development.

A second talk from the KIT

was given by Mirko Grosse from

the Institute of Applied Materials

(KIT-IAM). He presented his work

entitled “Investigation of the hydrogen

diffusion and distribution in

Zirconium by means of Neutron

Imaging”. The work conducted in an

international effort described the

hydrogen diffusion and distribution

in zirconium, analysed by using

neutron imaging facilities CoNRad

(Berlin, Germany), ANTARES (Garching,

Germany), and ICON (Villigen,

Switzerland). Neutron imaging

enables generally in-situ measurements

with high accuracy. It was

especially used to study the hydrogen

diffusion and redistribution in case of

stressed samples. Delayed Hydride

Cracking (DHC) is of high interest and

will be further investigated.

Uwe Hampel from the Technical

University of Dresden, Germany presented

results from the Project

DSC-Monitor in his talk titled “Potential

Methods for the Long-term Monitoring

of the State of Fuel Elements in

Dry Storage Casks”. The fundamental

investigations aim on the feasibility

and applicability of potential methods

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for non-intrusive monitoring of the

state of fuel elements in dry storage

casks. In particular radiation-based

methods, thermography and acoustic

methods were discussed. The assessment

of the applicability, sensitivity,

and uncertainty of the proposed

methods are underway using numerical

and experimental techniques.

As a representative of the German

Federal Office for the Safety of Nuclear

Waste Management (BfE), Tobias

Zweiger briefly outlined the ­current

state of spent fuel storage in Germany,

the new structure of the BfE and the

work areas of the respective divisions.

A summary of ongoing work in the

spent fuel storage division and an outlook

on future research activities and

interests of the BfE was discussed.

Gerold Spykman from TÜV NORD

EnSys GmbH & Co. KG, Hannover,

Germany, provided in his talk “Dry

storage of high level waste in Germany

– Safety assessments for 40 years

and beyond” his view on the licensing

of cask inventories and on the

licensing of the storage facilities in

Germany. The formulation and the

ranking of the influencing factors on

storage, transportability and final

disposal were presented as a gap

analysis based on the experiences

from the licensing and surveillance

procedures in Germany from the TÜV

NORD EnSys point of view.

Francisco Feria from CIEMAT,

Spain, provided an overview entitled

“CIEMAT response to challenges on

fuel safety research during dry

storage”. The research focuses on

developing predictive capabilities on

fuel rod performance during dry

storage including extended storage. To

assess the spent nuclear fuel integrity

along dry storage and to determine its

characteristics prior to transport,

CIEMAT’s strategy consists of the

extension of the FRAPCON code

( FRAPCON-xt) to the dry storage and

thus to enable predictions of in-clad

hydrogen radial distribution and characterization

of the outward cladding

creep. The adoption of best- estimateplus-uncertainty

methodology (BEPU)

allows determining the code’s uncertainty.

The talk “Considerations on spent

fuel behaviour for transport after

extended storage” was given by

Konrad Linnemann from the Safety of

Transport Containers Division of the

German Bundesanstalt für Materialforschung

(BAM). His presentation

focused on the fuel rod failure in the

transport package safety assessment

and the assumptions for criticality

safety analysis, leading to the discussion

of aspects about transport after

extended storage. A stress limit was

determined, beyond which rod failure

is assumed to occur, leading to fissile

material release in the cask cavity.

As a conclusion, further experimental

investigations were described as

desirable.

A further talk entitled “R&D initiatives

at BAM concerning spent nuclear

fuel integrity during long term storage”

was given by Teresa Orellana

Pérez from the Safety of Storage

Containers Division of BAM. The

research project aims at developing

numerical methods that will enable

brittle failure probability assessments

of fuel claddings and the estimation of

boundary conditions to prevent cladding

failure. Experimental data

from ring compression tests will be

analysed in cooperation with the

University of Madrid. In addition, the

perspective to contribute to a comprehensive

fuel cladding characterization

in the frame of the EJP was discussed.

Julia Neles from the Öko-Institut

e.V., Germany, provided a talk entitled

“Organizational and management

aspects in extended storage”. One

focus was on the German Act on

Reorganization of Nuclear Waste

Responsibilities from 2017, which

regulates the transition of responsibilities

for the waste management

from the waste producers to the

public-owned operator BGZ (Gesellschaft

für Zwischenlagerung). Knowledge

management has to be applied at

authorities and the long-term preservation

of expert organisation

knowledge has to be clarified. It was

also pointed out, that the periodic

safety revisions should be strengthened

as an inspection tool for organizational

and management topics.

In his talk “Hydrides and Zr-

Cladding Mechanics”, Weija Gong of

the Swiss Paul Scherrer Institute (PSI)

presented an overview of ongoing research

topics at PSI. Using neutron

imaging, investigations were conducted

on hydrogen diffusion in

Zr-Materials under stress. Combining

experimental results and Finite-

Element-Modelling for the stress field,

a thermodynamic modelling was

achieved defining a stress-dependant

chemical potential. Liner claddings

were also carefully studied at PSI,

especially for hydrogen redistribution

issues during cooling and under stress

conditions. Also, some tests were

performed to determine the impact of

hydride reorientation on the fatigue

of the material.

The workshop was concluded with

the talk “Open questions on the road

to reliable predictions” presented

by Florian Rowold from GRS. He

discussed the large number of

parameters governing the cladding

hoop stress and their strong interdependencies.

Due to the latter it

seems indispensable to establish an

integrated calculation method which

covers the entire lifetime of a fuel

rod. It was shown that this is also

important with respect to conservatism

and predictability for long time

scenarios. An integrated approach

combines reliable end-of-life fuel

data, thermal modelling and fuel

performance code enhancement as

well as improved material behaviour

understanding and simulation.

Prior to the GRS workshop and in

conjunction to it, a two-day meeting

of the International Subcommittee of

the Extended Storage Collaboration

Program of the Electric Power

Research Institute was also hosted at

GRS in Garching. The objective of this

meeting was to further establish an

international network that shares

­essential information in the field of

long-term storage of spent fuel.

Both events showed the need of

intensive exchange of knowledge

with a clear focus on scientific and

technical aspects. The implications of

an extended storage of used nuclear

fuel cover a large variety of features,

phenomena and effects. Due to the

existing similarities in the international

context of the spent fuel characteristics,

it seems to be obvious to

involve experts from other countries.

This gives the opportunity for synergetic

effects, especially in the light of

large-scale experiments and limited

national research funding. The large

number of participants fortified the

general opinion that an exchange of

scientific and technical knowledge is

needed to identify and prioritize

the knowledge gaps for the German

situation. All participants valued the

workshop as a great success. The next

workshop “Safety of Extended Dry

Storage of Spent Nuclear Fuel” is

planned again as a three-day event at

GRS in Garching during the first week

of June 2019.

Authors

Klemens Hummelsheim

Florian Rowold

Maik Stuke

Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH

Boltzmannstr. 14

85748 Garching (München),

Germany

Report

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

ı Klemens Hummelsheim, Florian Rowold and Maik Stuke

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

Inside

483

KTG-Sektion NORD

Einladung: Erfolgreicher Nachweis

von kohärenten Neutrinos

im Kernkraftwerk Brokdorf

Neutrinos sind sogenannte „Geisterteilchen“, weil sie viele

Lichtjahre Flugweg Materie durchdringen können ehe sie

mit ihr wechselwirken. Sie entstehen in verschiedenen

Quellen, wie etwa im Herzen der Sonne bei Fusionsprozessen.

Kernreaktoren emittieren ebenfalls einige

Prozent der frei gesetzten Energie in Form von Neutrinos,

weswegen in unmittelbarer Nähe eines Reaktors sehr

interessante Experimente mit Neutrinos möglich sind.

Im Vortrag wird erklärt, wie man diese Neutrinos nachweisen

kann, welche spannenden Fragestellungen sich damit

verbinden und welche Rolle das Kernkraftwerk Brokdorf

dabei spielt.

Der Referent, Prof. Dr. Dr. h.c. Manfred Lindner ist

Direktor am Max-Planck-Institut für Kernphysik in Heidelberg.

Er forscht auf dem Gebiet der Teilchen- und Astroteilchenphysik

mit dem Ziel, die elementare Struktur und

Entstehung der Materie zu erklären. Dazu ist er führend

an internationalen Projekten aus dem Bereich der

Neutrino- Physik und der Suche nach Dunkler Materie

beteiligt. Im Anschluss an den etwa einstündigen Vortrag

wird Gelegenheit zur weiteren Diskussion sein.

Interessierte KTG-Mitglieder sowie Freunde und

Bekannte sind herzlich eingeladen am Mittwoch, den

17. Oktober 2018 um 13:00 Uhr, bei der PreussenElektra

GmbH, Tresckowstraße 5, Hannover, teilzunehmen.

Wir danken der PreussenElektra GmbH für die Initiative

zum und die Unterstützung des Vortrags.

Wir bitten um eine namentliche Anmeldung

der Teilnehmer bis zum 4. Oktober 2018 unter

Telefon 0511 439-2184 oder an

thomas.froehmel@preussenelektra.de

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

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

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)

Robert-Koch-Platz 4

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,

Eggenstein-Leopoldshafen

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,

Eggenstein-Leopoldshafen

30. Dr. Fritz Ruess, Forchheim

77 Jahre | 1941

21. Ing. Peter Schween,

Stutensee-Blankenloch

31. Dr. Eike Roth, Klagenfurt

76 Jahre | 1942

7. Dr. Klaus W. Stork, Bad Dürkheim

20. Dipl.-Ing. Norbert König, Baiersdorf

21. Dr. Enrique Horacio Toscano,

Stutensee

22. Dr. Alexander Alexas, Stutensee

75 Jahre | 1943

4. Klaus Günther, Bergisch Gladbach

9. Alfred Kapun, Obertshausen

70 Jahre | 1948

9. Bernd Müller-Kiemes, Bingen

14. Claus Fenzlein, Erlangen

65 Jahre | 1953

17. Edgar Albrecht, Beckedorf

20. Dieter Gaeckler, Lingen

Top

First Westinghouse AP1000

nuclear plant Sanmen 1

completes commissioning

(westinghouse) On 6 June 2018,

Westinghouse Electric Company,

China State Nuclear Power Technology

Corporation (SNPTC) announced

that the world’s first AP1000 nuclear

power plant located in Sanmen,

Zhejiang Province, China has successfully

completed initial criticality.

“Today we completed the final

major milestone before commercial

operation for Westinghouse’s AP1000

nuclear power plant technology,” said

José Emeterio Gutiérrez, Westinghouse

president and chief executive

officer. “We are one step closer to

­delivering the world’s first AP1000

plant to our customer and the world –

with our customers, we will provide

our customers in China with safe,

reliable and clean energy from

Sanmen 1.”

| | First Westinghouse AP1000 nuclear plant Sanmen 1 completes

commissioning (Photo: Westinghouse)

Following initial criticality will

be connection to the electrical grid.

Once plant operations begin at

­Sanmen 1, it will be the first AP1000

nuclear power plant in operation,

offering innovative passive safety

system technology, multiple layers of

defense and advanced controls for

unequaled reliability and safety.

Commenting on Westinghouse’s

strong partnership with the China

customer, Gavin Liu, president –

Asia Region stated, “Westinghouse’s

success in China is the joint effort

between Westinghouse and our China

customers.” He added, “This partnership

and cooperation model can help

to deploy a fleet of AP1000 units in the

world for many years to come.”

On 30 June 2018 the Sanmen

nuclear power plant has begun initial

connection to the electrical grid.

Sanmen 1’s turbine generator is now

initially connected to the electrical

grid and has begun generating

electricity.

Sanmen 1 is capable of generating

1,117 megawatts of electricity when at

full power. It’s also the first of a fleet of

four new AP1000 plants in eastern

China and will provide safe, reliable

and environmentally-friendly energy

for the next 60+ years.

Commenting on Westinghouse’s

recent successes in China, David

Durham, Westinghouse senior vice

president, New Projects Business

stated, “It’s such an exciting time for

Westinghouse, our China customer

and the nuclear industry, as we

proudly move closer and closer to

100 percent power and commercial

operation at Sanmen 1.”.

Westinghouse currently has six

AP1000 nuclear power plants progressing

through construction, testing

and start-up. These projects include

two units in Sanmen, Zhejiang

Province, China, two units in Haiyang,

Shandong Province, China, as well as

two units under construction at the

Alvin W. Vogtle Electric Generating

Plant near Waynesboro, Georgia, USA.

| | www.westinghousenuclear.com

World

Belarusian nuclear station

meets ‘Stress Test’ standards,

EU Peer Review concludes

(nucnet) EU regulators have concluded

that the Belarusian nuclear

power station under construction near

the town of Ostrovets complies with

the bloc’s risk and safety assessments

– so-called “stress tests” – but made a

number of recommendations to the

national regulator.

A European Nuclear Safety Regulators

Group (Ensreg) peer review gave

News

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

the Ostrovets nuclear power plant,

which is close to the Lithuanian

border, an “overall positive” review,

following a site investigation that took

place in March.

The stress tests are meant to ensure

nuclear power plants comply with

strict criteria established by the International

Atomic Energy Agency and

were established by the European

Commission and Ensreg as a direct

reaction to the earthquake and

tsunami that caused the shutdown of

the Fukushima-Daiichi nuclear station

in Japan in March 2011.

The peer review team, which

reviewed an earlier stress test report

prepared by Belarus, comprised of 17

members, two representatives from

the EC and three observers: one from

the IAEA, one from Russia and one

from Iran.

The team praised the Belarusian

authorities for complying with the

review, even though Belarus had no

obligation to do so because it is not an

EU member state.

Following the Fukushima-Daiichi

accident, the EU carried out stress

tests of all its nuclear power plants

and also invited interested non-EU

countries to take part in the exercise.

In a detailed report, Ensreg

addressed three main areas: the site’s

resilience to extreme natural events

like earthquakes and flooding; the

capacity of the plant to respond to

electric power outages and loss of

heat sink; and severe accident

management.

According to the findings, the site

is resistant to earthquakes, flooding

and extreme weather, although the

investigators warned that seismic data

was not fully available and called on

the regulator to make sure run-off water

cannot enter safety-related buildings.

There are two 1,109-MW Russian

VVER-1200 reactor units under construction

at the Belarusian nuclear

station. Construction of Unit 1 began

in November 2013 and of Unit 2 in

April 2014.

The final peer review report is

online: https://bit.ly/2NnOixf

| | europa.eu, www.ensreg.eu,

www.dsae.by

Japan: Approval of energy

plan paves way for reactor

restarts

(nucnet) Nuclear reactor restarts in

Japan have become more likely after

the government approved an energy

plan today confirming that nuclear

power will remain a key component of

Japan’s energy strategy.

The plan, known as the Basic

Energy Plan, calls for a nuclear

share of around 20-22% by 2030. The

nuclear industry group, the Japan

Atomc Industrial Forum (Jaif) has

said about 30 reactors must be

brought back online to meet the

target.

Japan shut down all 42 com mercial

nuclear reactors after the Fukushima-

Daiichi accident. According to the

International Atomic Energy Agency,

the country’s nuclear share in 2017

was about 3.6%. Before Fukushima,

Japan generated about 30% of its

electricity from nuclear and planned

to increase that to 40%

Nine units have been restarted in

Japan since the Fukushima accident.

They are: Ohi-3, Ohi-4, Genkai-3,

Genkai-4, Sendai-1, Sendai-2, Ikata-3,

Takahama-3 and Takahama-4.

The energy plan also strengthens

the government’s commitment to

giving renewables such as solar and

wind power a major role in energy

generation.

The plan, which charts the nation’s

mid- and long-term energy policy,

marks the fifth in a series that is

required by law to be reviewed about

every three years.

The plan also maintains a reliance

on coal-fired thermal power as a

485

NEWS

| | Editorial Advisory Board

Frank Apel

Erik Baumann

Dr. Maarten Becker

Dr. Erwin Fischer

Carsten George

Eckehard Göring

Florian Gremme

Dr. Ralf Güldner

Carsten Haferkamp

Dr. Petra-Britt Hoffmann

Christian Jurianz

Dr. Guido Knott

Prof. Dr. Marco K. Koch

Dr. Willibald Kohlpaintner

Ulf Kutscher

Herbert Lenz

Jan-Christian Lewitz

Andreas Loeb

Dr. Thomas Mull

Dr. Ingo Neuhaus

Dr. Joachim Ohnemus

Prof. Dr. Winfried Petry

Dr. Tatiana Salnikova

Dr. Andreas Schaffrath

Dr. Jens Schröder

Norbert Schröder

Prof. Dr. Jörg Starflinger

Prof. Dr. Bruno Thomauske

Dr. Brigitte Trolldenier

Dr. Walter Tromm

Dr. Hans-Georg Willschütz

Dr. Hannes Wimmer

Ernst Michael Züfle

Imprint

| | 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)

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ISSN 1431-5254

News

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

486

NEWS

baseload energy source despite high

emissions of carbon dioxide.

The administration of prime

minister Shinzo Abe decided to promote

nuclear energy when it revised

the plan in 2014, reversing the policy

of the previous government led by

the then-Democratic Party of Japan,

which pledged to phase out nuclear

power by 2039 in the face of public

concern over safety.

Under the latest plan, the ratio of

nuclear energy, renewables and coal

thermal power in the nation’s overall

energy as of fiscal 2030 will remain at

20-22%, 22-24% and 26%, respectively,

in line with the government’s

target set three years ago.

The plan doe not make any

mention of the need for building new

nuclear plants.

However, it re-endorses using the

nuclear fuel cycle, in which plutonium

extracted from spent nuclear fuel at

nuclear plants is used to generate

power.

But the plan, noting calls from the

US, says that Japan will make efforts

to cut its stockpile of plutonium,

which can be used in making nuclear

weapons.

Japan holds about 47 tonnes of

plutonium, a source of criticism from

the US and other countries. Spent

nuclear fuel containing plutonium

from nuclear power plants in Japan

is sent to the UK and France for

reprocessing and eventual fabrication

into uranium-plutonium mixed oxide

(MOX) fuel before being returned to

Japan.

| | www.japan.go.jp

Reactors

Kola-1 becomes first Russian

nuclear plant to get operating

extension

(rosatom, nucnet) Russia’s state

nuclear operator Rosenergoatom has

been granted a licence by regulator

Rostekhnadzor to operate the Kola-1

nuclear power unit in the north of the

country for an additional 15 years

until 2033.

In a statement on its website, state

nuclear corporation Rosatom said this

is the first time a nuclear power plant

in Russia has been given such an

extension.

In April 2018 Rosenergoatom said

it had begun an extensive refurbishment

and modernisation programme

at Kola-1, a 411-MW VVER which

Operating Results March 2018

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated. gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 743 670 174 1 971 306 256 625 492 100.00 100.00 98.15 99.28 99.12 100.34

OL2 Olkiluoto BWR FI 910 880 743 687 017 1 996 275 246 295 456 100.00 100.00 99.88 99.88 100.51 100.50

KCB Borssele PWR NL 512 484 743 381 342 1 107 719 159 314 638 99.81 99.83 99.81 99.83 100.59 100.54

KKB 1 Beznau 1,2,7) PWR CH 380 365 296 108 560 108 560 124 854 647 39.84 13.71 38.07 13.10 37.85 13.03

KKB 2 Beznau 7) PWR CH 380 365 743 285 428 829 214 131 994 087 100.00 100.00 100.00 100.00 101.14 101.09

KKG Gösgen 7) PWR CH 1060 1010 743 793 650 2 308 759 307 503 346 100.00 100.00 99.98 99.98 100.77 100.88

KKM Mühleberg BWR CH 390 373 724 276 060 823 500 125 161 645 97.44 99.12 96.02 98.58 95.27 97.80

CNT-I Trillo PWR ES 1066 1003 743 775 921 2 278 694 241 303 118 100.00 100.00 99.94 99.98 97.56 98.60

Dukovany B1 PWR CZ 500 473 743 371 963 1 083 856 109 714 339 100.00 100.00 99.97 99.95 100.13 100.40

Dukovany B2 1,2) PWR CZ 500 473 209 102 246 747 959 105 370 496 28.13 69.38 27.75 68.97 27.52 69.29

Dukovany B3 PWR CZ 500 473 743 369 915 1 075 918 103 698 345 100.00 100.00 100.00 100.00 99.57 99.67

Dukovany B4 PWR CZ 500 473 743 372 191 1 080 365 104 352 106 100.00 100.00 100.00 100.00 100.19 100.08

Temelin B1 PWR CZ 1080 1030 721 772 094 772 094 107 253 388 97.04 33.40 95.25 32.78 96.22 33.11

Temelin B2 PWR CZ 1080 1030 743 813 415 2 356 815 103 846 761 100.00 100.00 100.00 100.00 101.37 101.08

Doel 1 PWR BE 454 433 743 338 988 984 072 135 198 820 100.00 100.00 99.98 99.99 100.59 100.39

Doel 2 PWR BE 454 433 743 337 020 984 599 133 236 867 100.00 100.00 99.15 99.61 99.79 100.28

Doel 3 3) PWR BE 1056 1006 0 0 0 251 169 221 0 0 0 0 0 0

Doel 4 PWR BE 1084 1033 743 817 209 2 371 746 256 917 588 100.00 100.00 100.00 100.00 100.48 100.33

Tihange 1 PWR BE 1009 962 726 739 606 2 191 381 293 030 257 97.66 99.20 96.92 98.94 99.17 101.03

Tihange 2 PWR BE 1055 1008 743 794 003 2 296 556 251 246 094 100.00 100.00 100.00 99.68 101.96 101.44

Tihange 3 PWR BE 1089 1038 722 784 314 2 332 443 271 227 273 97.15 99.02 96.80 98.90 96.84 99.13

Operating Results May 2018

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Time availability

[%]

Energy availability Energy utilisation

[%] *) [%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf 1,2) DWR 1480 1410 641 848 394 3 707 680 343 899 739 86.10 77.42 79.28 72.34 76.74 68.88

KKE Emsland 2,4) DWR 1406 1335 596 777 929 4 790 392 340 113 675 80.09 95.91 79.82 95.86 74.19 94.04

KWG Grohnde DWR 1430 1360 744 1 001 747 4 022 931 370 650 510 100.00 82.75 99.53 80.46 93.53 77.17

KRB C Gundremmingen 1) SWR 1344 1288 136 153 154 3 540 364 324 120 256 18.31 76.80 15.43 75.98 15.18 72.25

KKI-2 Isar DWR 1485 1410 744 1 067 384 5 291 198 346 889 521 100.00 100.00 100.00 99.99 96.24 98.06

KKP-2 Philippsburg 1,2,4) DWR 1468 1402 256 300 335 4 349 845 359 517 361 34.41 86.53 33.94 86.36 26.80 80.44

GKN-II Neckarwestheim DWR 1400 1310 744 1 011 600 4 974 300 325 097 434 100.00 100.00 100.00 99.87 97.29 98.32

News

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

began commercial operation in

December 1973.

The work was scheduled to take

about six months, Rosatom said at the

time.

The Kola station, 200 km south of

the city of Murmansk on the shore of

Imandra Lake, generates about 60%

of electricity in the Murmansk region,

Rosatom said.

All four units at Kola are Sovietdesigned

pressurised water reactors.

Units 1 and 2, of the older V-230

model, began commercial operation

in the mid-1970s and Units 3 and 4, of

the newer V-213 model, in the

mid-1980s.

| | www.rosatom.ru

Company News

USA: Framatome completes

major refurbishment of 31

reactor coolant pump motors

(framatome) Framatome recently

completed the refurbishment of 31

reactor coolant pump motors for

three southeastern nuclear energy

facilities. From 2002 to May 2018, the

company modified and upgraded

these components, which resulted

in a 100 percent reliability and

zero- failure performance record since

being re-installed.

The motors in reactor coolant

pumps help move coolant around the

primary circuit of a nuclear reactor

core. This keeps the reactor from overheating

while ensuring the safe heat

transfer from a reactor core to steam

generators.

“The success of this refurbishment

campaign is a tribute to Framatome’s

dedicated and experienced employees,”

said Craig Ranson, senior vice president

of the Installed Base Business Unit at

Framatome in North America. “Their

unmatched expertise, bolstered by

access to world-class facilities, allows

us to provide our customers with solutions

that, in many cases, are

more innovative and cost effective

than their plant’s original equipment

manufacturer.”

Members of Framatome’s Installed

Base services team worked with the

plants’ personnel to remove each

motor. They then brought the motors

to the company’s 70,000-square-foot

Pump and Motor Service Center in

Lynchburg, Virginia. While at the

center, experts inspected the components,

completed necessary repairs

and replacements, and tested each

motor. Such refurbishments allow

these components, and thus their

nuclear facilities, to operate safely and

reliably for longer durations.

Following successful testing, pump

and motor specialists re-installed the

motors and assessed their performance

on-site.

| | www.framatome.com

URENCO to supply EDF with

new uranium enrichment

services

(urenco) URENCO and EDF have

signed a new enrichment contract to

serve EDF’s French reactor fleet.

The high value and long-term

contract supports the recycling of

nuclear fuel by enriching uranium

recovered from fuel which has been

previously used and reprocessed.

The technical complexities of

enriching this material will involve

expertise from across URENCO and

upgrading our facilities.

Dominic Kieran, URENCO’s Chief

Commercial Officer, said: “URENCO is

proud to be part of EDF’s endeavour to

recycle spent nuclear fuel. It is a

­significant step in further proving the

sustainability of nuclear energy and a

testimony to URENCO’s technical

capabilities.”

| | www.urenco.com

Full core of Westinghouse fuel

achieved at South-Ukraine

nuclear power plant unit 3

(westinghouse) Westinghouse Electric

Company announced that Ukraine’s

State Enterprise National Nuclear

Energy Generation Company (SE

*)

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

1)

Refueling

2)

Inspection

3)

Repair

4)

Stretch-out-operation

5)

Stretch-in-operation

6)

Hereof traction supply

7)

Incl. steam supply

8)

New nominal

capacity since

January 2016

9)

Data for the Leibstadt

(CH) NPP will

be published in a

further issue of atw

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB

487

NEWS

Operating Results April 2018

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated. gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 720 651 030 2 622 336 257 276 522 100.00 100.00 98.66 99.13 99.36 100.09

OL2 Olkiluoto BWR FI 910 880 522 480 632 2 476 907 246 776 088 72.50 93.12 72.18 92.95 72.56 93.51

KCB Borssele PWR NL 512 484 720 361 216 1 468 935 159 675 854 97.83 99.33 97.81 99.33 98.13 99.94

KKB 1 Beznau 1,2,7) PWR CH 380 365 720 276 656 385 216 125 131 303 100.00 35.29 100.00 34.83 101.19 35.07

KKB 2 Beznau 7) PWR CH 380 365 720 275 430 1 104 644 132 269 517 100.00 100.00 100.00 100.00 100.72 101.00

KKG Gösgen 7) PWR CH 1060 1010 720 759 700 3 068 459 308 263 046 100.00 100.00 99.91 99.96 99.54 100.55

KKM Mühleberg BWR CH 390 373 720 277 490 1 100 990 125 439 135 100.00 99.34 99.89 98.91 98.82 98.06

CNT-I Trillo PWR ES 1066 1003 720 762 241 3 040 935 242 065 359 100.00 100.00 100.00 99.98 98.85 98.66

Dukovany B1 PWR CZ 500 473 720 357 871 1 441 727 110 072 210 100.00 100.00 99.43 99.82 99.41 100.15

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

Dukovany B3 PWR CZ 500 473 529 258 528 1 334 447 103 956 874 73.47 93.37 72.49 93.12 71.81 92.70

Dukovany B4 PWR CZ 500 473 496 240 156 1 320 521 104 592 262 68.89 92.22 66.94 91.73 66.71 91.73

Temelin B1 PWR CZ 1080 1030 720 777 874 1 549 968 108 031 262 100.00 50.05 99.96 49.60 99.85 49.83

Temelin B2 PWR CZ 1080 1030 720 783 901 3 140 716 104 630 662 100.00 100.00 100.00 100.00 100.81 101.01

Doel 1 PWR BE 454 433 541 245 643 1 229 715 135 444 462 75.19 93.80 75.00 93.74 75.06 94.05

Doel 2 PWR BE 454 433 720 328 895 1 313 494 133 565 762 100.00 100.00 99.98 99.70 100.44 100.32

Doel 3 3) PWR BE 1056 1006 0 0 0 251 169 221 0 0 0 0 0 0

Doel 4 PWR BE 1084 1033 720 787 926 3 159 672 257 705 513 100.00 100.00 99.86 99.97 99.90 100.23

Tihange 1 PWR BE 1009 962 720 732 505 2 923 886 293 762 762 100.00 99.40 99.94 99.19 101.25 101.09

Tihange 2 PWR BE 1055 1008 720 753 721 3 050 277 251 999 815 100.00 100.00 98.60 99.41 99.80 101.03

Tihange 3 PWR BE 1089 1038 0 0 2 332 443 271 227 273 0 74.25 0 74.16 0 74.34

News

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

488

NEWS

NNEGC) Energoatom’s South-Ukraine

NPP Unit 3 near Yuzhnoukrainsk

in Mykolaiv province was loaded with

a full core of Westinghouse VVER-1000

fuel. This is the first unit in Ukraine to

operate with Westinghouse VVER-

1000 fuel assemblies as the sole fuel

source.

“Westinghouse began supplying

fuel to Ukraine in 2005, when the first

lead test assemblies were delivered to

South-Ukraine Unit 3,” said Aziz Dag,

vice president and managing director,

Northern Europe. “We are proud

to continue supporting Ukraine

with their energy diversification by

supplying a full core of Westinghouse

VVER-1000 fuel to our customer,

Energoatom.”

Westinghouse currently supplies

fuel to six of Ukraine’s 15 nuclear

power reactors. Beginning in 2021,

the number of reactors with Westinghouse

fuel will increase to seven.

“Westinghouse has made significant

investments over the last several

years in order to further enhance our

fuel delivery support to Energoatom,”

said Michele DeWitt, senior vice

president, Nuclear Fuel. “We have

dedicated production lines for

VVER-1000 fuel and stand ready to

supply fuel for further contract

expansions.”

The nuclear fuel delivered by

Westinghouse is manufactured in its

fuel fabrication facility in Västerås,

Sweden. Nuclear power continues to

be an important energy source for the

country of Ukraine, accounting for

approximately 50% of its electricity

production.

| | www.westinghousenuclear.com

Forum

GRS-IRSN Workshop zu

Sicherheitskriterien von

Brennelementen

(grs) In den vergangenen Jahren

wurde in Frankreich aufgrund neuerer

experimenteller Erkenntnisse das

kerntechnische Regelwerk hinsichtlich

der Sicherheitskriterien für

Brennelemente und deren Verhalten

bei Betrieb und in Störfällen überarbeitet

und aktualisiert. Da ähnliche

Fragestellungen in der Vergangenheit

auch Thema in Deutschland waren

und zu Regelwerksänderungen geführt

hatten, veranstalteten die

Gesellschaft für Anlagen- und Reaktorsicherheit

(GRS) gGmbH und das

Institut de Radioprotection et de

Súreté Nucléaire (IRSN) einen

gemein samen Workshop zum Thema

„Fuel Safety Criteria“, welcher am

20./21. Juni 2018 in Paris in den

Räumen des IRSN stattfand. Neben

Experten der GRS und des IRSN nahmen

Vertreter aus Belgien, Tschechien

und Litauen, sowie der deutschen Reaktorsicherheitskommission

und des

Betreibers PreussenElektra an der Veranstaltung

teil.

In fünf Sitzungen wurden Informationen

und Erfahrungen zu Sicherheitskriterien

und zugehörigen

Nachweisverfahren hinsichtlich betrieblicher

und störfallbedingter

Phänomene wie Hüllrohrkorrosion,

-oxidation, Wasserstoffversprödung,

Reaktivitäts- und Kühlmittelverluststörfälle,

mechanische Pellet-Hüllrohr-

Wechselwirkungen (Pellet Cladding

Mechanical Interaction, PCMI),

Brennstoff-Verlagerung und -Auswurf

bei Hochabbrand sowie Brennelementverbiegungen

ausgetauscht.

Es wurde deutlich, dass beide Länder

trotz mitunter unterschiedlicher

Sicherheitsphilosophien, regulatorischer

Anforderungen und Brennelement-Ausführungen

mit weitgehend

übereinstimmenden Problemstellungen

konfrontiert waren und

entsprechende Änderungen in den

ihren einschlägigen Regelwerken

umgesetzt haben. Der Workshop ist

daher als Startpunkt für ein gemeinsames

Verständnis brennstoffbezogener

Sicherheitskriterien und

zugehöriger Nachweisverfahren zu

verstehen. Weitere Detail-Diskussionen

zu ausgewählten Teilaspekten

sind geplant. Das nächste Expertentreffen

in diesem Themenfeld wird

voraussichtlich in Berlin stattfinden.

| | www.grs.de

People

Dipl.-Ing. Christoph Michel

wird zum 1. Januar 2019

Nachfolger von Dr.-Ing. Hans

Fechner als Sprecher der

Geschäftsführung der

Siempelkamp Gruppe

(siempelkamp) Mit Wirkung zum

1. August 2018 ist Dipl.-Ing. Christoph

Michel zum weiteren Mitglied der

Geschäftsführung der G. Siempelkamp

GmbH & Co. KG bestellt worden.

Er wird ab dem 1. Januar 2019 als

Sprecher der Geschäftsführung der

Siempelkamp Gruppe die Nachfolge

von Dr.-Ing. Hans Fechner übernehmen,

der nach vielen Jahren

erfolgreicher Tätigkeit in den Ruhestand

geht.

Christoph Michel hat Luft- und

Raumfahrttechnik an der Universität

Stuttgart studiert und später berufsbegleitend

einen MBA an der Duke

University, USA abgeschlossen. Er

blickt auf eine 18-jährige erfolgreiche

Karriere im Maschinen- und Großanlagenbau

zurück.

| | www.siempelkamp.com

Market data

(All information is supplied without

guarantee.)

Nuclear Fuel Supply

Market Data

Information in current (nominal)

­U.S.-$. No inflation adjustment of

prices on a base year. Separative work

data for the formerly “secondary

market”. Uranium prices [US-$/lb

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

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

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

(Separative work unit)].

2014

• Uranium: 28.10–42.00

• Conversion: 7.25–11.00

• Separative work: 86.00–98.00

2015

• Uranium: 35.00–39.75

• Conversion: 6.25–9.50

• Separative work: 58.00–92.00

2016

• Uranium: 18.75–35.25

• Conversion: 5.50–6.75

• Separative work: 47.00–62.00

2017

• Uranium: 19.25–26.50

• Conversion: 4.50–6.75

• Separative work: 39.00–50.00

2018

January 2018

• Uranium: 21.75–24.00

• Conversion: 6.00–7.00

• Separative work: 38.00–42.00

February 2018

• Uranium: 21.25–22.50

• Conversion: 6.25–7.25

• Separative work: 37.00–40.00

March 2018

• Uranium: 20.50–22.25

• Conversion: 6.50–7.50

• Separative work: 36.00–39.00

April 2018

• Uranium: 20.00–21.75

• Conversion: 7.50–8.50

• Separative work: 36.00–39.00

May 2018

• Uranium: 21.75–22.80

• Conversion: 8.00–8.75

• Separative work: 36.00–39.00

June 2018

• Uranium: 22.50–23.75

News

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

• Conversion: 8.50–9.50

• Separative work: 35.00–38.00

| | Source: Energy Intelligence

www.energyintel.com

Cross-border Price

for Hard Coal

Cross-border price for hard coal in

[€/t TCE] and orders in [t TCE] for

use in power plants (TCE: tonnes of

coal equivalent, German border):

2012: 93.02; 27,453,635

2013: 79.12, 31,637,166

2014: 72.94, 30,591,663

2015: 67.90; 28,919,230

2016: 67.07; 29,787,178

2017: 91.28, 25,739,010

| | Uranium spot market prices from 1980 to 2018 and from 2008 to 2018. The price range is shown.

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

489

NEWS

2018

I. quarter: 89.88; 5.838.003

| | Source: BAFA, some data provisional

www.bafa.de

EEX Trading Results

June 2018

(eex) In June 2018, the European

Energy Exchange (EEX) increased

volumes on its power derivatives

markets by by 28% to 231.1 TWh

(June 2017: 181.2 TWh). On the

Dutch power market, volumes increased

by 141% to 3.2 TWh (June

2017: 1.3 TWh). EEX achieved strong

double-digit growth in the markets for

France (22.0 TWh, +22%), Italy

(44.5 TWh, +46%) as well as in

power options (9.3 TWh, +45%).

Volumes in Phelix-DE Futures increased

to 132.7 TWh.

On the EEX markets for emission

allowances, the total trading volume

almost tripled to 297.4 million tonnes

of CO 2 in June (June 2017:

105.1 ­million tonnes of CO 2 ). On the

EUA secondary market (including

options), volumes increased sixfold to

217.8 million tonnes of CO 2 (June

2017: 30.6 million tonnes of CO 2 ).

Primary market auctions contributed

79.6 million tonnes of CO 2 to the total

volume.

The Settlement Price for base load

contract (Phelix Futures) with

delivery in 2019 amounted to

43.14 €/MWh. The Settlement

Price for peak load contract (Phelix

Futures) with delivery in 2019

amounted to 53.55 €/MWh.

The EUA price with delivery in

December 2018 amounted to

14.24/16.14 €/ EUA (min./max.).

July 2018

(eex) In July 2018, the European

Energy Exchange (EEX) increased

volumes on its power derivatives

| | Separative work and conversion market price ranges from 2008 to 2018. The price range is shown.

)1

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

markets by 46% to 213.8 TWh (July

2017: 146,2 TWh). On the Spanish

power market, volumes exceeded

the mark of 10 TWh for the first time,

doubling last year’s volume

(10.6 TWh, July 2017: 4.3 TWh).

Furthermore, the markets for France

(18.2 TWh, +18%) and Italy

(37.3 TWh, +70%), in particular,

developed positively. In Phelix-DE

Futures, trading volumes amounted

to 128.7 TWh which is clearly above

the total July volume in 2017 in the

products for the German market

( Phelix-DE and Phelix-DE/AT in July

2017: 98.1 TWh).

The Settlement Price for base load

contract (Phelix Futures) with

delivery in 2019 amounted to

43.79 €/MWh. The Settlement Price

for peak load contract (Phelix

Futures) with delivery in 2019

amounted to 53.93 €/MWh.

The EUA price with delivery

in December 2018 amounted to

15.08/17.40 €/ EUA (min./max.).

| | www.eex.com

MWV Crude Oil/Product Prices

May 2018

(mwv) According to information and

calculations by the Association of the

German Petroleum Industry MWV e.V.

in May 2018 the prices for super fuel,

fuel oil and heating oil noted higher

compared with the pre vious month

April 2018. The average gas station

prices for Euro super consisted of

145.62 €Cent ( April 2018:

138.96 €Cent, ­approx. +6.6 % in

brackets: each information for previous

month or rather previous month

comparison), for diesel fuel of

126.22 €Cent (121.09; +5.13 %) and

for heating oil (HEL) of 67.93 €Cent

(63.12 €Cent, +4.81 %).

Worldwide crude oil prices

(monthly average price OPEC/Brent/

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

approx. +4.21 % (+7.39 %) in May

2018 compared to April 2018.

The market showed a stable

development with slightly higher

prices; each in US-$/bbl: OPEC

basket: 73.22 (68.43); UK-Brent:

74.40 (72.11); West Texas Intermediate

(WTI): 67.87 (66.25).

June 2018

In June 2018 the prices for super

fuel, fuel oil and heating oil noted

inconsistent compared with the

pre vious month May 2018. The

average gas station prices for Euro

super consisted of 147.60 €Cent (May

2018: 145.62 €Cent, ­approx. +1.98 %

in brackets: each information for previous

month or rather previous month

comparison), for diesel fuel of

129.41 €Cent (126.2; +3.19 %) and

for heating oil (HEL) of 67.67 €Cent

(67.937 €Cent, -0.38 %).

| | www.mwv.de

News

---

atw Vol. 63 (2018) | Issue 8/9 ı August/September

Why do We Allow Nuclear to Take the ‘Silly Season’ Media Heat?

490

NUCLEAR TODAY

John Shepherd is a

UK-based energy

writer and editor- inchief

of Energy

Storage Publishing.

Links to reference

sources:

ICL briefing paper:

https://bit.ly/

2vkIM6Y

Rick Perry’s remarks:

https://bit.ly/

2Kz3WTB

GMB union statement

on UK nuclear:

https://bit.ly/2OPk3jd

John Shepherd

As I started to write this article, we were approaching the end of what is often referred to in the UK as the ‘silly season’

– the main summer holiday period when hard news is hard to come by.

The time of year always means all manner of weird and

wonderful stories finding their way into newspapers and

broadcast news, stories that would most probably never

see the light of day outside the silly season. This year

has been slightly different, because the lengthy spell of

hot weather that many of us across Europe experienced

generated much of the journalistic ‘heat’.

But for the nuclear industry, the hot spell fanned the

media flames of an old anti-nuclear favourite, as it became

clear operations at some nuclear power plants were being

halted temporarily to comply with restrictions that prevent

cooling water further heating local rivers and waterways.

Some media outlets preferred the alarmist over the

factual. I was dismayed to hear one BBC report claim “one

ageing (European) nuclear power plant” had been “forced”

by the heat wave to cut back on production “to keep vital

equipment cool”. That statement was misleading – albeit

probably more out of ignorance than malice.

I don’t recall hearing from any of our industry representatives

early on in the summer, communicating the

facts on the cooling issue to the public and journalists. If

there was no industry-wide effort on this there should

have been. It’s not a new situation for our industry and

every opportunity should be taken to head off misinformation

that experience tells us is just around the

corner. PR directors should be making a note in their

diaries for next year just in case – because forewarned is

forearmed.

But there was more refreshing news out of the UK over

the summer in the form of a briefing paper by researchers

at Imperial College London (ICL). According to the paper by

ICL’s Grantham Institute – Climate Change and the

Environment, nuclear power “will be essential for meeting

the UK’s greenhouse gas emissions reduction target, unless

we can adapt to depend largely on variable wind and solar,

or there is a breakthrough in the commercialisation of

carbon capture and storage”.

The paper acknowledged the difficulties involved in

attracting private investment to build new nuclear projects,

but said the UK government’s decision to procure the

3.2 gigawatt Hinkley Point C nuclear plant “represents a

crucial opportunity for the conventional nuclear industry,

which is under significant financial stress, to rebuild itself”.

There certainly does appear to be a new realism in the

UK about the urgent need to turn talk about investments in

nuclear into real action. One of the country’s major trade

unions, the GMB, put new nuclear firmly on the agenda.

The union was quick to respond to reports that the UK’s

planned Moorside nuclear plant in Cumbria, northwest

England, could be scrapped unless a buyer is found.

Moorside is being developed by NuGen, which is owned

by Toshiba. NuGen has been put up for sale as Toshiba

­restructures its operations in the aftermath of financial

issues triggered by losses in its US nuclear business,

Westinghouse. The three AP1000 reactor units proposed for

Moorside were to have come from Westinghouse.

Now the GMB has reiterated its call for the UK government

to take a stake in the financing of the Moorside

project, “rather than leaving this vital project at the mercy

of foreign companies”.

GMB national secretary Justin Bowden said: “As well

as eradicating the uncertainty, by the government taking a

stake and taking control at Moorside, the price to consumers

will be greatly reduced making good all round

sense, not just the obvious benefits to bill payers but

because the government is ‘the lender of last resort’ when

it comes to guaranteeing the country’s energy supply and

so direct public funding of the construction does away

with the nonsensical pretence that this is some other

country or company’s responsibility.”

And the union cautioned the UK against an over reliance

on renewables in energy policy. According to the GMB, “for

the 12 months from 7 March 2017, every one in 5.6 days

was a low wind day (65 days in total) when the output of

the installed and connected wind turbines in the UK

produced less than 10 % of their installed and connected

capacity for more than half of the day”.

“For 341 days in the year, solar output was below 10 %

of installed capacity for more than half of the day,” the

union said.

Such championing of public investment in nuclear from

the union is welcome as the UK struggles to advance its

civil nuclear ambitions.

However, it’s a different story for one of the world’s

nuclear newcomer nations – the United Arab Emirates –

where nuclear development continues apace. In August,

the Emirates Nuclear Energy Corporation (ENEC) announced

the successful completion of hot functional testing

at unit 2 of the Barakah nuclear plant, which is under

construction around 240 kilometres west of Abu Dhabi.

ENEC said that as of June 2018, the construction progress

rate of unit 2 was 93 % and overall construction progress

rate for the four Barakah units is now more than 89 %.

Meanwhile, in the US, energy secretary Rick Perry made

his first visit to a nuclear power plant since his ­appointment

16 months earlier. Speaking at the James A FitzPatrick

plant, Perry gave a ringing endorsement of nuclear on

behalf of the Trump administration.

Perry said: “Nuclear provides approximately 20 % of

the electricity generated in the United States. It is one of

our most reliable sources of baseload power, and it is also

one of our cleanest sources of power, providing about 60 %

of our carbon-free energy output.”

And a day after Perry’s visit, the Department of Energy

announced $ 36.4 million (€ 31.5 m) in funding for 37

research awards at universities, national laboratories, and

private industry on a range of topics in fusion energy

sciences. The Department said the research “is designed to

help lay the groundwork for the development of nuclear

fusion as a future practical energy source”.

Investment in nuclear construction and research should

be welcomed wherever it comes and our industry should

not be afraid to campaign for public investment. The

renewables lobby has been doing this successfully for some

time. Nuclear should not shy away from speaking up too.

Author

John Shepherd

Shepherd Communications

3 Brooklands

West Sussex

BN43 5FE

Nuclear Today

Why do We Allow Nuclear to Take the ‘Silly Season’ Media Heat? ı John Shepherd

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ı

ı

Gestaltung der Übergänge zwischen der Betriebs- und der Stilllegungsgenehmigung

Gestaltung der Übergänge zwischen Genehmigungs- und Aufsichtsphase

Rechtliche Gestaltung des Genehmigungsverfahrens

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Elemente des Umweltbereiches

UVP und Erörterungstermin

Anfechtung von Genehmigungen

Stellungnahmen der EU-Kommission

Reststoffe und Abfälle

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Vorstellung von Reststoffkonzepten

Bewertung der Reststoffkonzepte

Neue Regelungen zum Übergang der Entsorgungsverantwortung

Entsorgungsfragen rund um Abfälle aus dem Rückbau

Zielgruppe

Die 2-tägige Schulung wendet sich an Fach- und Führungskräfte, Mitarbeiterinnen und Mitarbeiter

von Betreibern, Industrie und Dienstleistern, die sich mit der Thematik aktuell bereits beschäftigen

oder sich künftig damit auseinander setzen werden.

Maximale Teilnehmerzahl: 12 Personen

Referenten

Dr. Matthias Bauerfeind

Dr. Christian Raetzke

Wir freuen uns auf Ihre Teilnahme!

ı Abteilung Stilllegung, Entsorgung, Reaktorphysik, TÜV SÜD Energietechnik

GmbH Baden-Württemberg

ı Rechtsanwalt, Leipzig

Bei Fragen zur Anmeldung rufen Sie uns bitte an oder senden uns eine E-Mail.

Termin

2 Tage

24. bis 25. September 2018

Tag 1: 10:30 bis 17:45 Uhr

Tag 2: 09:00 bis 16:45 Uhr

Veranstaltungsort

Geschäftsstelle der INFORUM

Robert-Koch-Platz 4

10115 Berlin

Teilnahmegebühr

1.598,– € ı zzgl. 19 % USt.

Im Preis inbegriffen sind:

ı Seminarunterlagen

ı Teilnahmebescheinigung

ı Pausenverpflegung

inkl. Mittagessen

Kontakt

INFORUM

Verlags- und Verwaltungsgesellschaft

mbH

Robert-Koch-Platz 4

10115 Berlin

Petra Dinter-Tumtzak

Fon +49 30 498555-30

Fax +49 30 498555-18

seminare@kernenergie.de

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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