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nucmag.com<br />

<strong>2018</strong><br />

4<br />

217<br />

Heat Transfer Systems<br />

for Novel Nuclear<br />

Power Plant Designs<br />

221 ı Operation and New Build<br />

Safety Research for GEN IV Reactors<br />

226 ı Operation and New Build<br />

Numerical Analysis for the MYRRHA Project<br />

ISSN · 1431-5254<br />

24.– €<br />

238 ı Environment and Safety<br />

Passive Heat Removal Systems Research<br />

270 ı Nuclear Today<br />

‘Ugly’ Nuclear Deserves a Political Makeover


The International Expert Conference on Nuclear Technology<br />

Who will attend<br />

AECOM • AiNT • Alpiq • ANSTO • August Alborn • <strong>atw</strong> – International Journal for Nuclear Power<br />

• Axpo • BAM • Becker Technologies • BGE • BKW Energie • Brenk Systemplanung • Canadian<br />

Nuclear Safety Commission • CIS • CONLAR • Czech Technical University in Prague • DAHER<br />

NUCLEAR TECHNOLOGIES • DAtF • DBE • Deggendorf Institute of Technology • Department for<br />

International Trade • DGZfP • DIAMO • DMT • Eisenwerk Bassum • Embassy of the Czech Republic<br />

• EnBW Kernkraft • Energus • ENSI • E.ON • EPZ N.V. • EPRI • EUROfusion • European Commission<br />

• EWN • Fachverband für Strahlenschutz • Federal Ministry for Economic Affairs and Energy •<br />

Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety •<br />

Fennovoima • FH Aachen • Fortum • Framatome • Forschungszentrum Jülich • German Waste<br />

Management Commission • GNS • GRS • HALFEN • Helmholtz Zentrum Dresden Rossendorf •<br />

Hochschule Magdeburg-Stendal • Hochschule Mannheim • Hochschule Zittau/Görlitz • HTW<br />

Dresden • IAEA • IAF – Radioökologie • ICRP • IEM FörderTechnik • IGN consult • IKE • INFORUM •<br />

Innogy • Istanbul Technical University • iUS • Jiangsu CASHH Nuclear Material Technology • JRC •<br />

KCCA • Kerntechnische Entsorgung Karlsruhe • KIT • Kraftanlagen Heidelberg • Krantz • KROHNE<br />

Messtechnik • KSR COLLEGE OF TECHNOLOGY • KTE • KTG • Leibniz Universität Hannover • Liese •<br />

Mammoet Deutschland • Max Planck Institute • Ministry of Energy, Agriculture, the Environment<br />

and Digitalization of Schleswig Holstein • Mirion Technologies • MIT • Nagaoka University of<br />

Technology • Nagra • National Centre for Nuclear Research • Nawah Energy Company • NPP<br />

Brunsbüttel • NPP Gundremmingen • NPP Isar • NPP Krümmel • NRC Kurchatov Institute •<br />

NRG • Nuclear Decommissioning Authority • Nuclear Engineering International • NUKEM •<br />

Nuklearforum Schweiz • Nuvia • OECD • ONR • Orano • PreussenElektra • REEL-NKMNOELL<br />

Special Cranes • Röhr + Stolberg • RST • Ruhr-Universität Bochum • RWE • RWTH Aachen •<br />

Safetec Entsorgungs- und Sicherheitstechnik • Schminke Krantechnik • Sellafield • Siempelkamp<br />

• Simulatorzentrum KSG/ GfS • Skoda • Southern Medical University • STEAG Energy Services<br />

• swissnuclear • Technische Hochschule Deggendorf • Technische Universität Dresden •<br />

The University of Manchester • TÜV Nord • TÜV SÜD • TÜV Thüringen • Tyrolit Hydrostress •<br />

ÚJV Řež • Uniper • Université du Luxembourg • Universität Siegen • Universität Stuttgart<br />

• Universitätsklinikum Hamburg-Eppendorf • University of Copenhagen • University of<br />

Pisa • URENCO • Vattenfall • VGB PowerTech • VKTA • VPC • Westinghouse • Women in Nuclear •<br />

WTI • Young Generation Network<br />

In alphabetical order. Subject to change.<br />

Register now online<br />

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www.nucleartech-meeting.com


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Security of Supply ...<br />

and the Clock is Ticking ...<br />

Deal reader, More than one hundred years ago, around 1890, a conflict flared up between the two well-known<br />

protagonists of electricity supply, Thomas Alva Edison and George Westinghouse, on the large-scale power supply and the<br />

construction of power grids in the United States of America. While Edison technically preferred D.C. voltage, Westinghouse<br />

counted instead on alternating voltage. In the end it was not a matter of the most suitable technique but of the anticipated<br />

market shares of each company General Electric or Westinghouse Electric and the patents behind. At a breath taking pace,<br />

the most important developments for the use of electricity were preceding: In the year 1866 Werner Von Siemens<br />

discovered the dynamo- electrical principle, which enabled larger performance. The development of alternating voltage<br />

in the year 1881 enabled generally technically and cost-effectively the transportation of electricity over long distances<br />

– we are talking back then about distances of some ten kilometres. Alternating voltage enforced itself at that time due to<br />

possible further transportation length enabled through higher trans mission voltage.<br />

207<br />

EDITORIAL<br />

Both current types have something in common: generation<br />

and use need to take place simultaneously. The grid fails if<br />

both do not fit together. Neither alternating current grids<br />

nor direct current grids offer storage possibilities. Thus, a<br />

stable power system also requires a stable and reliable<br />

generation, because if a larger system “fails”, the system<br />

restoration is, from its task and process, a large-scale<br />

project.<br />

Different believes e.g. from politics or other interest<br />

groups are simply wrong, power systems are – without any<br />

further active establishments and plants- no accumulators.<br />

A reliable power supplying system needs at any time<br />

reliable generation. “Surpluses”, meaning potentials for a<br />

higher generation than demand, when so ever, cannot be<br />

shifted or stored “electrically” in the system at a later time.<br />

It was not an inconspicuous message, which appeared<br />

multiple times in the press at the end of February,<br />

beginning of March <strong>2018</strong>. Headlines such as “Time<br />

synchronisation per power system: Energy shortages make<br />

watches lose time”, described a phenomenon, of which,<br />

according to the media “one became aware of – only<br />

( editor’s note) - after weeks”: What happened?<br />

As an indicator for the stability of alternating power<br />

systems stand supply voltage a well as system frequency.<br />

For the system frequency applies that she needs to be<br />

identical at any point of the system. If generation and<br />

consumption do not fit, deviations occur, leaking generation<br />

leads among others to a perceived frequency decrease<br />

among the entire connected system. As the system<br />

frequency is defined for our alternating electricity net with<br />

constant 50 Hertz, it is also qualified for watches, which<br />

use the frequency as direct clock indicator.<br />

We can for example – due to cost reasons – renounce to<br />

a frequency stabilising quartz oscillator. Nevertheless, this<br />

technical simplification is bought with failures in time, if<br />

the frequency deviates from the standard over a longer<br />

period. Only a few hundred Hertz is enough for days and<br />

weeks in order to, as in the current case, generate a time<br />

deviation of minus 360 seconds, 6 minutes, and those<br />

inside the entire affected system of 25 West, Middle- and<br />

South European countries.<br />

The cause for this incident was later communicated by<br />

the European Network of Transmission System Operators for<br />

Electricity (short ENTSO-E) and the Swiss net operator<br />

swissgrid, that in the control zone Serbia, Macedonia,<br />

Montenegro (the so called SMM rule block), especially in<br />

Kosovo and Serbia less energy was fed into the system. A<br />

deficit of 113 gigawatt hours was shown, not much, in view<br />

of a European daily production of around 8,000 gigawatt<br />

hours. But especially this shows how delicate our power<br />

system is and how sensitive it reacts to the smallest<br />

malfunctions.<br />

Reliable measures in power generation – meaning<br />

currently only for conventional techniques, thus need,<br />

with all considerations on the reconstruction of electricity<br />

supply, to be reconsidered. Additionally and almost<br />

simultaneously another alarming “availability message”<br />

came in: At the beginning of March <strong>2018</strong> European gas<br />

storage tanks were only filled with a quantity of 26.2 per<br />

cent, Germany even on average only with 23.8 per cent.<br />

Thus, according to an EU-conform proceeding an early<br />

warning level was reached, because the filling level of<br />

storage tanks may not be lower than around 20 % due to<br />

reasons of guaranteeing mechanical stability. On top came<br />

the message that more natural gas was imported to Europa<br />

than in the previous years. All first hints, that there might<br />

not be enough natural gas in Europe for dispose filling in as<br />

a “reserve”?<br />

In all, these are all important references that any,<br />

especially neither direct market- nor technically driven,<br />

interventions – where compensation factors can con tribute<br />

– need to be well thought in our power system. Furthermore,<br />

does the availability of a broad basis of conventional<br />

generation not only gain more importance, she is even<br />

more important than it is conceded for “conventionals”<br />

vision wise in many places in terms of an „energy<br />

transition“. To what extend “the clock” might tick on<br />

possible severe supply shortfalls or even large-scale loss of<br />

off-site power… one does not know…<br />

Christopher Weßelmann<br />

– Editor in Chief –<br />

Editorial<br />

Security of Supply ... and the Clock is Ticking ...


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

EDITORIAL 208<br />

Versorgungssicherheit und die Uhr tickt ...<br />

Liebe Leserin, lieber Leser, vor mehr als hundert Jahren, um 1890, entbrannte eine Auseinandersetzung<br />

zwischen den beiden bekannten Protagonisten der Elektrizitätsversorgung, Thomas Alva Edison und George Westinghouse,<br />

zur weiträumigen Versorgung der Vereinigten Staaten von Amerika mit Strom und dem Aufbau geeigneter Stromnetze.<br />

Während Edison technisch die Gleichspannung favorisierte, setze Westinghouse die Wechselspannung dagegen.<br />

Letztendlich ging es aber nicht wesentlich um die Frage der geeigneteren Technik, sondern um die avisierten Marktanteile<br />

der jeweiligen Unternehmen General Electric bzw. Westinghouse Electric und die dahinter stehenden Patente. Vorangegangen<br />

waren in atemberaubendem Tempo die wichtigsten Entwicklungen für die Nutzung der Elektrizität: Im Jahr<br />

1866 entdeckte Werner von Siemens das dynamoelektrische Prinzip, das größere Leistungen ermöglichte. Die<br />

Entwicklung des Wechselstromtransformators im Jahr 1881 ermöglichte technisch grundsätzlich und kostengünstiger<br />

den Transport von Strom über längere Strecken – wir sprechen hier zu jener Zeit über Strecken im Bereich von einigen<br />

zehn Kilometern. Durchgesetzt hatte sich aufgrund der durch höhere Übertragungsspannungen möglichen weiteren<br />

Transportlängen zu jener Zeit die Wechselspannung.<br />

Beiden Stromarten ist eines gemeinsam: Erzeugung und<br />

Nutzung müssen exakt zeitgleich erfolgen. Sind Erzeugung<br />

und Gebrauch nicht im Einklang, bricht das Netz<br />

zusammen. Weder Wechsel- noch Gleichspannungsnetz<br />

bieten „Speichermöglichkeiten“. Für ein stabiles Stromnetz<br />

ist daher auch eine stabile und verlässliche Erzeugung<br />

erforderlich, denn wenn einmal ein größeres Stromnetz<br />

„zusammenbricht“, ist der Netzwiederaufbau ein von der<br />

Aufgabe und dem zeitlichen Ablauf her aufwendiges<br />

Vorhaben. Anderslautende Stimmen z.B. aus der Politik<br />

oder von Interessengruppen sind schlichtweg falsch,<br />

Strom netze sind – ohne weitere aktive Einrichtungen und<br />

Anlagen – keine Speicher. Ein verlässliches Stromversorgungsnetz<br />

benötigt eine jederzeit verlässliche Erzeugung.<br />

„Überschüsse“, also Potenziale für eine höhere Erzeugung<br />

als die vorhandene Nachfrage, wann und warum auch<br />

immer, lassen sich „elektrisch“ im Netz nicht auf spätere<br />

Zeiten verschieben, also speichern.<br />

Es war eine nicht unscheinbare Nachricht, die Ende<br />

Februar, Anfang März <strong>2018</strong> mehrfach durch die<br />

Presse ging. Überschriften wie „Zeit-Synchronisation per<br />

Stromnetz: Energieknappheit lässt Uhren nachgehen“,<br />

beschrieben ein Phänomen, dessen man sich nach Angaben<br />

in der Presse „nach Wochen – erst (Anm. der Red.) –<br />

bewusst wurde“: Was war geschehen?<br />

Für die Stabilität bzw. als Indikator für die Stabilität<br />

von Wechselstromnetzen stehen die Netzspannung sowie<br />

die Netzfrequenz. Für die Netzfrequenz gilt dabei, dass<br />

diese an jedem Punkt in einem Netz identisch ist. Stimmen<br />

Erzeugung und Verbrauch nicht überein, kommt es zu<br />

Abweichungen, fehlende Erzeugung führt u.a. zu einer<br />

im gesamten angebundenen Netz fühlbaren Frequenzabnahme.<br />

Da die Netzfrequenz für unser Wechselstromnetz<br />

mit konstant 50 Hertz vereinbart ist, eignet sich diese<br />

auch für Uhren, die die Frequenz als direkten Taktgeber<br />

nutzen. Auf z.B. einen frequenzstabilisierenden Quarzoszillator<br />

kann – aus Kostengründen – verzichtet werden.<br />

Diese technische Vereinfachung erkauft man sich allerdings<br />

mit Fehlern in der Uhrzeit, wenn die Frequenz über<br />

einen längeren Zeitraum vom Standard abweicht. Schon<br />

wenige hundertstel Hertz reichen über Tage und Wochen<br />

aus, um, wie im aktuellen Fall, eine kumulierte Zeitabweichung<br />

von Minus 360 Sekunden, also 6 Minuten,<br />

hervorzurufen; und dies im ganzen betroffenen Netz von<br />

25 West-, mittel- und südosteuropäischen Ländern.<br />

Als Ursache für dieses Ereignis wurde später vom<br />

Verband Europäischer Übertragungsnetzbetreiber (kurz<br />

ENTSO-E, European Network of Transmission System<br />

Operators for Electricity) und dem Schweizer Netzbetreiber<br />

swissgrid kommuniziert, dass in der Kontrollzone Serbien,<br />

Mazedonien, Montenegro (dem sogenannten SMM Regelblock),<br />

insbesondere in Kosovo und Serbien zu wenig<br />

Energie ins Netz eingespeist wurde. Ein Fehlbetrag von<br />

113 Gigawattstunden wurde ausgewiesen, nicht viel,<br />

angesichts einer europaweiten Tagesproduktion von rund<br />

8.000 Gigawattstunden. Aber gerade diese zeigt, wie<br />

filigran unser Stromnetz ist und wie empfindlich es doch<br />

auf kleinste Störungen reagiert.<br />

Verlässliche Größen in der Stromerzeugung, sprich<br />

derzeit letztendlich nur die konventionellen Techniken,<br />

müssten von daher in allen Überlegungen zum Umbau der<br />

Stromversorgung neu überdacht werden. Hinzu kam fast<br />

zeitgleich eine weitere bedenkliche energiewirtschaftliche<br />

„Verfügbarkeitsmeldung“: Europas Gasspeicher waren zu<br />

Anfang März <strong>2018</strong> nur noch zu 26,2 Prozent gefüllt,<br />

Deutschland gar im Schnitt nur zu 23,8 Prozent. Damit<br />

war nach einem EU-einheitlichen Verfahren eine Frühwarnstufe<br />

erreicht, denn die Speicher dürfen ihren Füllgrad<br />

aus Gründen der Gewährleistung ihrer mechanischen<br />

Stabilität nicht unter rund 20 % absenken. Hinzu kam die<br />

Mitteilung, dass mehr Erdgas nach Europa importiert<br />

wurde, als in den Vorjahren. Alles erste Hinweise darauf,<br />

dass vielleicht in Zukunft doch nicht ausreichend Erdgas<br />

in Europa zur Verfügung stehen wird, um als „Reserve“<br />

einzuspringen?<br />

In Summe sind dies alles wichtige Hinweise darauf,<br />

dass jegliche, vor allem weder direkt markt- noch technisch<br />

getriebenen Eingriffe – wo ausgleichende Faktoren wirken<br />

können – in unser Stromversorgungssystem wohl überdacht<br />

sein müssen. Zudem gewinnt die Verfügbarkeit einer<br />

breiten Basis konventioneller Erzeugung damit nicht nur<br />

an Bedeutung, sie ist bedeutungsvoller als vielerorts in<br />

Visionen einer „Energiewende“ den Konventionellen<br />

zugestanden wird.<br />

Inwieweit „die Uhr“ möglicher schwerwiegender Versorgungsengpässe<br />

oder gar großflächiger Netzausfälle<br />

tickt ... man weis es nicht ...<br />

Christopher Weßelmann<br />

– Chefredakteur –<br />

Editorial<br />

Security of Supply... and the Clock is Ticking ...


Kommunikation und<br />

Training für Kerntechnik<br />

Suchen Sie die passende Weiter bildungs maßnahme<br />

im Bereich Kerntechnik?<br />

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

3 Atomrecht<br />

Ihr Weg durch Genehmigungs- und Aufsichtsverfahren RA Dr. Christian Raetzke 24.04.<strong>2018</strong><br />

18.09.<strong>2018</strong><br />

Navigation im internationalen nuklearen Vertragsrecht Akos Frank LL. M. 25.04.<strong>2018</strong> Berlin<br />

Atomrecht – Was Sie wissen müssen RA Dr. Christian Raetzke 12.06.<strong>2018</strong> Berlin<br />

Das Recht der radioaktiven Abfälle RA Dr. Christian Raetzke 23.10.<strong>2018</strong> Berlin<br />

3 Energie, Politik und Kommunikation<br />

Berlin<br />

Schlüsselfaktor Interkulturelle Kompetenz –<br />

International verstehen und verstanden werden<br />

Public Hearing Workshop –<br />

Öffentliche Anhörungen erfolgreich meistern<br />

Kerntechnik und Energiepolitik im gesellschaftlichen Diskurs<br />

– Themen und Formate<br />

Angela Lloyd 26.09.<strong>2018</strong> Berlin<br />

Dr. Nikolai A. Behr 16.10. - 17.10.<strong>2018</strong> Berlin<br />

N.N. 12.11. - 13.11.<strong>2018</strong> Gronau/Lingen<br />

3 Kerntechnik, Rückbau und Strahlenschutz<br />

Export kerntechnischer Produkte und Dienstleistungen –<br />

Chancen und Regularien<br />

In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:<br />

3 Nuclear English<br />

Das neue Strahlenschutzgesetz –<br />

Folgen für Recht und Praxis<br />

Stilllegung, Rückbau und Entsorgung –<br />

Recht und Praxis<br />

RA Kay Höft, M.A.,<br />

RA Olaf L. Kreuzer,<br />

Dr. Wolfgang Steinwarz<br />

RA Dr. Christian Raetzke,<br />

Maria Poetsch<br />

RA Dr. Christian Raetzke,<br />

Dr. Matthias Bauerfeind<br />

20.06. - 21.06.<strong>2018</strong> Berlin<br />

05.06. - 06.06.<strong>2018</strong><br />

27.06. - 28.06.<strong>2018</strong><br />

05.11. - 06.11.<strong>2018</strong><br />

Berlin<br />

24.09. - 25.09.<strong>2018</strong> Berlin<br />

Advancing Your Nuclear English (Aufbaukurs) Devika Kataja 11.04. - 12.04.<strong>2018</strong><br />

10.10. - 11.10.<strong>2018</strong><br />

Enhancing Your Nuclear English Devika Kataja 04.07. - 05.07.<strong>2018</strong> Berlin<br />

3 Wissenstransfer und Veränderungsmanagement<br />

Berlin<br />

Veränderungsprozesse gestalten – Heraus forderungen<br />

meistern, Beteiligte gewinnen<br />

Erfolgreicher Wissenstransfer in der Kern technik –<br />

Methoden und praktische Anwendung<br />

Dr. Christien Zedler,<br />

Dr. Tanja-Vera Herking<br />

Dr. Christien Zedler,<br />

Dr. Tanja-Vera Herking<br />

28.11. - 29.11.<strong>2018</strong> Berlin<br />

26.03. - 27.03.2019 Berlin<br />

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

Kontakt<br />

INFORUM Verlags- und Verwaltungs gesellschaft mbH ı Robert-Koch-Platz 4 ı 10115 Berlin<br />

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

Die INFORUM-Seminare können je nach<br />

Inhalt ggf. als Beitrag zur Aktualisierung<br />

der Fachkunde geeignet sein.


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

210<br />

Issue 4<br />

April<br />

CONTENTS<br />

217<br />

Heat Transfer Systems<br />

for Novel Nuclear<br />

Power Plant Designs<br />

| | The Swiss nuclear power plants generate up to 40 % of the country’s electricity production. At the Beznau site, two pressurised water<br />

reactors are in operation with a gross capacity of 380 MW each and a net capacity of 365 MW. Switzerland’s Federal Nuclear Safety<br />

Inspectorate, ENSI, gave the go-ahead for the restart of Beznau-1 after approving the safety case presented by operator Axpo following<br />

the discovery in 2015 of flaw indications in the reactor pressure vessel. (Courtesy: Axpo)<br />

Editorial<br />

Security of Supply ... and the Clock is Ticking ... . . 207<br />

Versorgungssicherheit und die Uhr tickt ... . . . . 208<br />

Abstracts | English . . . . . . . . . . . . . . . . . . . 212<br />

Abstracts | German . . . . . . . . . . . . . . . . . . . 213<br />

Inside Nuclear with NucNet<br />

Euratom: Industry Softens Stance<br />

as Government Lays Out Plans for Transition . . . 214<br />

NucNet<br />

Calendar . . . . . . . . . . . . . . . . . . . . . . . 216<br />

Operation and New Build<br />

Heat Transfer Systems for Novel<br />

Nuclear Power Plant Designs . . . . . . . . . . . . . 217<br />

Sebastian Vlach, Christoph Fischer and Herman van Antwerpen<br />

Experimental and Analytical Tools<br />

for Safety Research of GEN IV Reactors . . . . . . . 221<br />

G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak<br />

DAtF Notes. . . . . . . . . . . . . . . . . . . . . .215<br />

221<br />

| | Centrum Výzkumu Řež facilities list.<br />

217<br />

Numerical Analysis of MYRRHA<br />

Inter- wrapper Flow Experiment at KALLA . . . . . 226<br />

| | Koeberg PWR steam generator and simulation model.<br />

Abdalla Batta and Andreas G. Class<br />

Contents


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

226<br />

CONTENTS<br />

211<br />

| | Velocity magnitude within bundle showing flow distribution.<br />

Heat Balance Analysis for<br />

Energy Conversion Systems of VHTR . . . . . . . . 230<br />

SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon<br />

and Soyoung Park<br />

Spotlight on Nuclear Law<br />

Information Requirements Versus<br />

Confidentiality Obligations – Extension of<br />

the In-Camera Procedure Planned . . . . . . . . . . 235<br />

Informationsbedarf versus<br />

Geheimhaltungspflichten – Erweiterung<br />

des In camera-Verfahrens geplant . . . . . . . . . . 235<br />

Tobias Leidinger<br />

Environment and Safety<br />

CFD Modeling and Simulation of Heat and Mass<br />

Transfer in Passive Heat Removal Systems . . . . . 238<br />

Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas<br />

|241<br />

249<br />

| | Collecting soil samples.<br />

Research and Innovation<br />

Irradiation Tests of a Flat Vanadium Self- Powered<br />

Detector with 14 MeV Neutrons . . . . . . . . . . . 246<br />

Prasoon Raj and Axel Klix<br />

Nanofluid Applied Thermo-hydro dynamic<br />

Performance Analysis of Square Array<br />

Subchannel Under PWR Condition. . . . . . . . . . 249<br />

Jubair Ahmed Shamim and Kune Yull Suh<br />

| Computational domain created in Star-CCM+.<br />

KTG Inside . . . . . . . . . . . . . . . . . . . . . . 257<br />

238<br />

| | Liquid Volume fraction distribution.<br />

Decommissioning and Waste Management<br />

The Decommissioning of the ENEA RB3<br />

Research Reactor in Montecuccolino . . . . . . . . 241<br />

F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi,<br />

R. Lorenzelli and A. Rizzo<br />

News . . . . . . . . . . . . . . . . . . . . . . . . . 260<br />

Nuclear Today<br />

Czechs and Balances and Why ‘Ugly’<br />

Nuclear Deserves a Political Makeover . . . . . . . 270<br />

Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 236<br />

AiNT. . . . . . . . . . . . . . . . . . . . . . . . . . . .Insert<br />

AMNT <strong>2018</strong>: Registration Form . . . . . . . . . . . Insert<br />

Contents


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

212<br />

ABSTRACTS | ENGLISH<br />

Euratom: Industry Softens Stance as<br />

Government Lays Out Plans for Transition<br />

NucNet | Page 214<br />

The UK’s nuclear industry has welcomed a government<br />

commitment to continuity with existing<br />

arrangements with Euratom, Europe’s nuclear safety<br />

and research watchdog, a softening of its earlier<br />

stance that the UK needed to stay in the group to<br />

protect vital nuclear research and new-build projects,<br />

and to make sure access to nuclear fuel and<br />

medical isotopes is not disrupted. The next phase of<br />

discussions will focus on the UK’s future relationship<br />

with Euratom. Specific objectives include a close<br />

association with the Euratom Research and Training<br />

Programme, including the Joint European Torus<br />

(JET) and the International Thermonuclear Experimental<br />

Reactor (ITER) projects.<br />

Heat Transfer Systems for Novel<br />

Nuclear Power Plant Designs<br />

Sebastian Vlach, Christoph Fischer and<br />

Herman van Antwerpen | Page 217<br />

This article focuses on designing or modifying heat<br />

exchangers found in the auxiliary systems of any<br />

power plant. The basic premise is to show that the<br />

software provides a one-stop solution for designing<br />

many types of heat transfer systems, where the<br />

interaction bet ween various loops connected by<br />

heat exchangers can be assessed. The nuclear power<br />

plant industry is addressed as the quality control in<br />

the development of the software makes it most<br />

suitable for nuclear related applications. Moreover,<br />

the software discussed has the capability to do<br />

contaminant tracing, which could be very useful<br />

for nuclear contamination studies in designing<br />

specialized ventilation systems. To highlight the<br />

versatility of the software network approach it will<br />

be shown how to model any setup and kind of heat<br />

exchanger such as plate, tube-in-tube, liquid/gas,<br />

finned tube etc. Additionally, the Koeberg pressurized<br />

water reactor steam generator comparison and<br />

the THTR steam generator comparison are shown<br />

as examples.<br />

Experimental and Analytical Tools for<br />

Safety Research of GEN IV Reactors<br />

G. Mazzini, M. Kyncl, Alis Musa and<br />

M. Ruscak | Page221<br />

Current research on nuclear safety in the world, in<br />

addition to supporting existing nuclear power<br />

plants is focused on the more detailed aspects of the<br />

new reactors. The new generation reactors are<br />

expected inter alia to use innovative types of fuel<br />

and new types of coolants, such as e.g. Super-<br />

Critical Water (SCW), supercritical CO 2 , liquid<br />

metals, fluoride salts or high-temperature Helium.<br />

The paper will describe new experimental infrastructure<br />

build recently in Research Centre Řež<br />

under the SUSEN (Sustainable Energy) project and<br />

available analytical tools for supporting safety<br />

research of GEN IV reactors. Two experimental<br />

loops – SCWL (Supercritical Water Loop) and HTHL<br />

(High Temperature Helium Loop) will serve as<br />

in-pile loops in the active core of the research<br />

reactor LVR-15. The paper provides examples of<br />

analyses made using codes ATHLET (supercritical<br />

water) and TRACE (high temperature He) illustrating<br />

process of their assessment and practical use.<br />

Numerical Analysis of MYRRHA Inter-wrapper<br />

Flow Experiment at KALLA<br />

Abdalla Batta and Andreas G. Class | Page 226<br />

The MYRRHA reactor, which is developed at<br />

SCK-SCN in Belgium, represents a multi-purpose<br />

irradiation facility. Its prominent feature is a pool<br />

design with the nuclear core submerged in liquid<br />

metal lead bismuth. During transients between<br />

normal operation and accident conditions decay<br />

heat removal is ensured by forced and natural<br />

convection, respectively. The flow in the gap<br />

between the fuel assemblies plays an important role<br />

in limiting maximum temperatures which should<br />

not be exceeded to avoid core damage. Due to the<br />

scarce database, within the Horizon 2020 – research<br />

and innovation framework program of the EU, the<br />

SESAME project was established to develop and<br />

validate advanced numerical approaches, to<br />

achieve a new or extended validation base and to<br />

establish best practice guidelines including verification<br />

& validation and uncertainty quantification.<br />

In particular the current work supports the<br />

inter-wrapper flow experiment at KALLA.<br />

Heat Balance Analysis for Energy<br />

Conversion Systems of VHTR<br />

SangIL Lee, YeonJae Yoo, Deok Hoon Kye,<br />

Gyunyoung Heo, Eojin Jeon<br />

and Soyoung Park | Page 230<br />

VHTR(Very High Temperature Gas Reactor) with<br />

helium used as a coolant can easily produce heat<br />

required in high-temperature thermochemical process,<br />

and because of low heat output density, the<br />

possibility of core melting is low. In this study, provided<br />

that VHTR is located in the primary system,<br />

the heat conversion system will be discussed in<br />

which hydrogen production and power supply are<br />

possible. In order to control the ratio between power<br />

and hydrogen production, the helium flowing<br />

through nuclear reactor is made to pass through<br />

heat exchanger for hydrogen production and steam<br />

generator or heat exchanger. This study proposes<br />

the whole heat conversion system model, and<br />

carries out thermodynamic feasibility calculation<br />

according to major design variable at each point<br />

and sensitivity analysis for efficiency optimization.<br />

Information Requirements Versus<br />

Confidentiality Obligations – Extension<br />

of the In-Camera Procedure Planned<br />

Tobias Leidinger | Page 235<br />

The justified right of the public to detailed information<br />

on a project requiring nuclear licensing is<br />

opposed by the state’s interest in effective protection<br />

of sensitive data. This conflict is manifested<br />

in licensing procedures but also at court. The differentiated<br />

legal provisions that regulate the balancing<br />

of these conflicting interests are now to be supplemented<br />

by a further facet: An expanded in-camera<br />

trial at court. According to the coalition agreement<br />

of 7 February <strong>2018</strong>, the regulation is to take place in<br />

the current 18th legislative period.<br />

CFD Modeling and Simulation of Heat<br />

and Mass Transfer in Passive Heat<br />

Removal Systems<br />

Amirhosein Moonesi, Shabestary,<br />

Eckhard Krepper and Dirk Lucas | Page 238<br />

The CFD-modelling and simulation of condensation<br />

inside passive heat removal systems are presented.<br />

Designs of future nuclear boiling water reactor concepts<br />

are equipped with emergency cooling systems<br />

which are passive systems for heat removal. The<br />

emergency cooling system consists of slightly<br />

inclined horizontal pipes which are immersed in a<br />

tank of subcooled water. The focus of the project is<br />

on detection of different morphologies such as<br />

annular flow, stratified flow, slug flow and plug flow<br />

and also modeling of the laminar film which is<br />

occurring during the condensation near the wall.<br />

The Decommissioning of the ENEA RB3<br />

Research Reactor in Montecuccolino<br />

F. Rocchi, C. M. Castellani, A. Compagno,<br />

I. Vilardi, R. Lorenzelli and A. Rizzo | Page 241<br />

The ENEA RB3 reactor was a 100 Wth research<br />

installation owned and operated by ENEA, in its<br />

center of Montecuccolino near Bologna, from 1971<br />

to 1989. In 1989, the RB3 reactor was shut down,<br />

and in the late 2010 ENEA received by ministerial<br />

decree the authorization to its dismantling, with the<br />

aim of reaching the “green field” status. This paper<br />

presents the three main pillars of the decommissioning<br />

of RB3, namely the strategy and methods<br />

for the dismantling, the strategy and methods for<br />

the radiological characterization of the building,<br />

and finally the strategy and methods for the radiological<br />

characterization of the site.<br />

Irradiation Tests of a Flat Vanadium<br />

Self-Powered Detector with<br />

14 MeV Neutrons<br />

Prasoon Raj and Axel Klix | Page 246<br />

Self-powered detector (SPD) represents a class of<br />

neutron and gamma monitoring instruments used<br />

in the fission reactor cores worldwide. This detector<br />

has inherent advantages of functioning without a<br />

bias voltage, simple measurement scheme, compactness,<br />

ease of maintenance, and high reliability.<br />

We are studying SPD for application as flux monitors<br />

in the European test blanket modules (TBM) of<br />

ITER, fusion reactor under construction in southern<br />

France.<br />

Nanofluid Applied Thermo-hydrodynamic<br />

Performance Analysis of Square Array<br />

Subchannel Under PWR Condition<br />

Jubair Ahmed Shamim and Kune Yull Suh | Page 249<br />

Efficient engineered design of heat transfer and<br />

fluid flow with enhanced heating or cooling requires<br />

two pivotal aspects that must be taken into consideration<br />

for extracting thermal energy from<br />

nuclear fission reactions in order to save energy,<br />

reduce process time, raise thermal rating and<br />

increase the operating life of a reactor pressure<br />

vessel. Hence, one of the major challenges in<br />

designing a new nuclear power plant is the quantification<br />

of the optimal flow of coolant and distribution<br />

of pressure drop across the reactor core.<br />

Recently, nanofluid has gained much renewed<br />

attention as a promising coolant for pressurized<br />

water reactors (PWRs) due to its enhanced thermal<br />

capabilities with least penalty in pressure drop.<br />

Czechs and Balances and Why ‘Ugly’<br />

Nuclear Deserves a Political Makeover<br />

John Shepherd | Page 270<br />

As if Europe does not have enough on its plate<br />

to deal with at the moment – politically and<br />

economically just for starters – could Brussels be on<br />

a collision course with the Czech government over<br />

the countries plans to expand nuclear energy?<br />

There is certainly friction over the issue between<br />

Prague and the European Commission (EC), to put<br />

it mildly. But why?<br />

The veteran head of the Czech Republic’s State<br />

Office for Nuclear Safety, Dana Drábová, last month<br />

accused other EU member states of “pressurising”<br />

Prague over the early closure of its oldest nuclear<br />

reactor units.<br />

Abstracts | English


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Euratom: Britische Industrie zufrieden<br />

mit Übergangsplänen der Regierung<br />

NucNet | Seite 214<br />

Die britische Nuklearindustrie hat die Zusage der<br />

Regierung begrüßt, die bestehenden Vereinbarungen<br />

mit Euratom, dem europäischen Rahmen<br />

für nukleare Sicherheit und Forschung, aufrechtzuerhalten<br />

und ihren früheren Standpunkt, dass<br />

das Vereinigte Königreich im Euratom-Vertrag<br />

verbleiben müsse, um wichtige Forschungs- und<br />

Neubauprojekte sicherzustellen, und den Zugang<br />

zu Kernbrennstoffen und medizinischen Isotopen<br />

zu gewährleisten, zu relativieren. Die nächste Phase<br />

der Gespräche im Rahmen des Brexit wird sich auf<br />

die künftigen Beziehungen des Vereinigten Königreichs<br />

zu Euratom konzentrieren. Zu den spezifischen<br />

Zielen gehört eine enge Zusammenarbeit<br />

mit den Euratom-Forschungs- und Ausbildungsprogrammen,<br />

einschließlich der Projekte Joint<br />

European Torus (JET) und International Thermonuclear<br />

Experimental Reactor (ITER).<br />

Fortgeschrittene Wärmeübertragungssysteme<br />

für zukünftige Kernkraftwerkskonzepte<br />

Sebastian Vlach, Christoph Fischer und<br />

Herman van Antwerpen | Seite 217<br />

CFD-Systemsimulation mit FlownexSE ermöglicht<br />

es Ingenieuren, einfache und komplexe strömungstechnische<br />

und thermische Netzwerke schnell und<br />

effizient aufzubauen und zu analysieren. Die<br />

Simulation ermöglicht es Ingenieuren, Systeme<br />

aufzubauen, effizient auszulegen und bereits frühzeitig<br />

Schwachstellen in Entwürfen zu finden<br />

sowie geeignete Änderungen und Maßnahmen zu<br />

entwickeln und im Netzwerkmodell zu testen.<br />

Besondere Aufmerksamkeit wird in diesem Artikel<br />

den vielseitigen Möglichkeiten gewidmet, einfache<br />

und komplexe Wärmetauschersysteme der verschiedensten<br />

Arten (Plattenwärmetauscher, Rohrbündel<br />

etc.) für moderne Kernkraftwerke anzuwenden. Als<br />

praktische Beispiele werden gemessene Daten von<br />

den Kraftwerken Koeberg und Hamm-Uentrop mit<br />

den Ergebnissen aus der Simulation verglichen.<br />

Experimentelle und analytische<br />

Werkzeuge für die Sicherheitsforschung<br />

zu GEN-IV-Reaktoren<br />

G. Mazzini, M. Kyncl, Alis Musa und<br />

M. Ruscak | Seite 221<br />

Die aktuelle Forschung zur Sicherheit von<br />

Kernkraftwerken konzentriert sich neben den<br />

Aktivitäten für bestehende Kernkraftwerke auf die<br />

detaillierteren Aspekte neuer Reaktorkonzepte.<br />

Hier werden u.a. innovative Brennstoffe und Kühlmittel<br />

wie z.B. überkritisches Wasser, überkritisches<br />

CO 2 , Flüssigmetalle, Salzschmelzen oder Helium<br />

eingesetzt. Vorgestellt wird dazu die neue experimentelle<br />

Infrastruktur, die im Forschungszentrum<br />

Řež im Rahmen des SUSEN-Projekts (Sustainable<br />

Energy) aufgebaut wurde, sowie die verfügbaren<br />

Analyseinstrumente zur Unterstützung der Sicherheitsforschung<br />

zu GEN IV-Reaktoren.<br />

Numerische Analyse der Zwischenspaltströmung<br />

im MYRRHA-Reaktor mit Ergebnissen<br />

des Strömungsexperiment KALLA<br />

Abdalla Batta und Andreas G. Class | Seite 226<br />

Der am SCK-SCN in Belgien entwickelte MYRRHA-<br />

Reaktor ist eine Mehrzweck-Bestrahlungsanlage.<br />

Sein herausragendes Merkmal ist eine Reaktorkonstruktion<br />

mit einer Kernkühlung aus flüssigem Blei-<br />

Wismut. Bei Transienten zwischen Normalbetrieb<br />

und Unfallbedingungen wird die Wärmeabfuhr<br />

durch erzwungene bzw. natürliche Konvektion<br />

sichergestellt. Die Strömung im Spalt zwischen den<br />

Brennelementen spielt eine wichtige Rolle bei der<br />

Begrenzung von Maximaltemperaturen, die zur<br />

Vermeidung von Kernschäden nicht überschritten<br />

werden sollten. Im Rahmenprogramm Horizon<br />

2020 – Forschung und Innovation der EU wurde<br />

dazu das Projekt SESAME initiiert, um fortgeschrittene<br />

numerische Ansätze zu entwickeln und<br />

zu validieren, die eine neue oder erweiterte<br />

Validierungsbasis für damit verbundene Fragestellungen<br />

zur Verfügung stellen.<br />

Wärmebilanzanalyse für<br />

Energieumwandlungssysteme von VHTR<br />

SangIL Lee, YeonJae Yoo, Deok Hoon Kye,<br />

Gyunyoung Heo, Eojin Jeon und<br />

Soyoung Park | Seite 230<br />

VHTR (Very High Temperature Gas Reactor) mit<br />

Helium als Kühlmittel können Wärme bereit stellen,<br />

die bei thermochemischen Hochtemperaturprozessen<br />

benötigt wird. In Bezug auf die Sicherheit ist<br />

aufgrund der geringen Wärmeleistungsdichte das<br />

Risiko einer Kernschmelze minimiert. Diskutiert<br />

werden Voraussetzungen für die Nutzung von<br />

VHTR für eine Wasserstofferzeugung und Stromversorgung.<br />

Vorgestellt wird ein Gesamtmodell des<br />

Wärmeumwandlungssystems mit einer thermodynamischen<br />

Machbarkeitsberechnung.<br />

Informationsbedarf versus<br />

Geheimhaltungspflichten – Erweiterung<br />

des In camera-Verfahrens geplant<br />

Tobias Leidinger | Seite 235<br />

Dem berechtigten Anspruch der Öffentlichkeit auf<br />

detaillierte Informationen über ein atomrechtlich<br />

genehmigungsbedürftiges Vorhaben steht das<br />

staatliche Interesse an einem effektiven Geheimnisschutz<br />

sensibler Daten gegenüber. Dieser Konflikt<br />

tritt regelmäßig im Genehmigungsverfahren aber<br />

auch vor Gericht zu Tage. Die differenzierten<br />

Gesetzesbestimmungen, die den Ausgleich dieser<br />

widerstreitenden Interessen regeln, sollen nun<br />

durch eine weitere Facette ergänzt werden: Ein<br />

erweitertes In-camera-Verfahren bei Gericht. Nach<br />

dem Koalitionsvertrag vom 7. Februar <strong>2018</strong> soll die<br />

Regelung in der schon laufenden 18. Legislaturperiode<br />

erfolgen.<br />

CFD-Modellierung und Simulation<br />

von Wärme- und Stoffaustausch<br />

in passiven Wärmeabfuhrsystemen<br />

Amirhosein Moonesi, Shabestary,<br />

Eckhard Krepper und Dirk Lucas | Seite 238<br />

Die CFD-Modellierung und Simulation der Kondensation<br />

in passiven Wärmeabfuhrsystemen wird vorgestellt.<br />

Zukünftige Siedewasserreaktorkonzepte<br />

werden mit Notkühlsystemen ausgestattet, die eine<br />

passive Wärmeabfuhr gewährleisten. Das Notkühlsystem<br />

besteht aus leicht geneigten horizontalen<br />

Rohren in einem Wasserbehälter. Der Schwerpunkt<br />

des vorgestellten Projektes liegt auf der Identifikation<br />

verschiedener Morphologien wie Ringströmung,<br />

Schichtenströmung, Schwallströmung<br />

und Pfropfenströmung sowie der Modellierung des<br />

laminaren Films, der bei der Kondensation in<br />

Wandnähe auftritt.<br />

Die Stilllegung der ENEA RB3<br />

Forschungsreaktor in Montecuccolino<br />

F. Rocchi, C. M. Castellani, A. Compagno,<br />

I. Vilardi, R. Lorenzelli und A. Rizzo | Seite 241<br />

Der ENEA RB3-Reaktor war eine 100-Watt-Forschungsanlage,<br />

die von 1971 bis 1989 im Zentrum<br />

von Montecuccolino bei Bologna, Italien betrieben<br />

wurde. 1989 wurde der RB3-Reaktor abgeschaltet<br />

und Ende 2010 erhielt ENEA per Ministerialerlass<br />

die Genehmigung zu seinem Rückbau mit dem Ziel,<br />

den Status „Grünen Wiese“ zu erreichen. Vorgestellt<br />

werden die drei wesentlichen Fragestellungen für<br />

die Stilllegung des RB3: Strategie und Methoden<br />

für den Rückbau, Strategie und Methoden für die<br />

radiologische Charakterisierung des Gebäudes und<br />

schließlich die Strategie und Methoden für die<br />

radiologische Charakterisierung des Standortes.<br />

Bestrahlungstests eines Vanadium-<br />

Detektors mit 14 MeV Neutronen<br />

Prasoon Raj und Axel Klix| | Seite 246<br />

Self-powered Detektoren (SPD) sind eine Klasse<br />

von Neutronen- und Gamma-Überwachungsgeräten,<br />

die weltweit in Kernreaktoren eingesetzt<br />

werden. Diese Detektoren besitzen die Vorteile,<br />

dass keine Spannungsversorgung erforderlich ist,<br />

das Messverfahren einfach und die Detektoreinheit<br />

kompakt, wartungsfreundlich und zuverlässig ist.<br />

SPDs werden im Rahmen des vorgestellten Projektes<br />

für den Einsatz als Flussmonitor in den Blanketmodulen<br />

des in Bau befindlichen Fusionsreaktors<br />

ITER .<br />

Einsatz von Nanofluiden und<br />

thermohydraulische Analyse<br />

für Druckwasserreaktoren<br />

Jubair Ahmed Shamim und Kune Yull Suh | Seite 249<br />

Eine effiziente Auslegung von Wärmeübertragung<br />

und Flüssigkeitsströmung mit verbessertem<br />

Wärme übergang, -transport oder Kühlung bedingt<br />

zwei zentrale Aspekte, die in Kernkraftwerken<br />

berücksichtigt werden müssen: Leistungsdichte<br />

und technische Lebensdauer des Reaktordruckbehälters.<br />

Eine Herausforderung für die Auslegung<br />

neuer Kernkraftwerkskonzepte ist daher die<br />

Quantifizierung einer optimalen Kühlmittelverteilung<br />

und die Verteilung des Druckverlustes über<br />

den Reaktorkern. In jüngster Zeit werden „Nanofluide“<br />

als vielversprechendes Kühlmittel für Druckwasserreaktoren<br />

(DWR) aufgrund verbesserter<br />

thermischer Eigenschaften mit geringst möglichem<br />

Druckabfall diskutiert, die auch Thema dieser<br />

Arbeit sind..<br />

Tschechien und Ausgewogenheit und<br />

warum es die „hässliche“ Kernenergie verdient,<br />

politisch neu bewertet zu werden<br />

John Shepherd | Seite 270<br />

Als ob Europa derzeit nicht genug zu tun hätte, mit<br />

sich selbst – politisch und wirtschaftlich, nur um<br />

zwei Themenbereiche zu nennen – ... könnte jetzt<br />

Brüssel auf Kollisionskurs mit der tschechischen<br />

Regierung zu den Plänen des Landes zum Ausbau<br />

der Kernenergie gehen?<br />

In der Frage zwischen Prag und der Europäischen<br />

Kommission (EC) geht es, um es milde auszudrücken,<br />

sicherlich um Differenzen. Aber warum?<br />

Die langjährige Leiterin der tschechischen<br />

Aufsichtsbehörde für nukleare Sicherheit, Dana<br />

Drábová, warf zudem im vergangenen Monat<br />

anderen EU-Mitgliedstaaten vor, die Regierung in<br />

Prag unter inakzeptablem Druck zu setzen hinsichtlich<br />

der Forderung einer vorzeitigen Stilllegung<br />

ihrer ältesten Kernkraftwerke.<br />

213<br />

ABSTRACTS | GERMAN<br />

Abstracts | German


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

214<br />

INSIDE NUCLEAR WITH NUCNET<br />

Euratom: Industry Softens Stance as<br />

Government Lays Out Plans for Transition<br />

NucNet<br />

The UK’s nuclear industry has welcomed a government commitment to continuity with existing<br />

arrangements with Euratom, Europe’s nuclear safety and research watchdog, a softening of its earlier stance<br />

that the UK needed to stay in the group to protect vital nuclear research and new-build projects, and to make<br />

sure access to nuclear fuel and medical isotopes is not disrupted.<br />

Energy secretary Greg Clark said in a written statement to<br />

parliament on 11 January <strong>2018</strong> that the government wants<br />

to include Euratom in any implementation period agreed<br />

as part of wider discussions on Brexit and plans to put in<br />

place “all the necessary measures” to ensure that the UK<br />

can operate as an independent and responsible nuclear<br />

state from day one of Brexit and its separation from the<br />

Euratom Treaty, which regulates the nuclear industry and<br />

the movement of nuclear material across Europe.<br />

According to Mr Clark’s statement, the government has<br />

made good progress on separation issues in the last few<br />

months as part of phase one of negotiations with the EU.<br />

Negotiations have covered a set of legal and technical<br />

issues related to nuclear material and waste, and safeguards<br />

obligations and equipment.<br />

The next phase of discussions will focus on the UK’s<br />

future relationship with Euratom. Specific objectives<br />

include a close association with the Euratom Research and<br />

Training Programme, including the Joint European Torus<br />

(JET) and the International Thermonuclear Experimental<br />

Reactor (ITER) projects.<br />

For the nuclear industry, rapid departure from Euratom<br />

without a clear replacement spells disaster. Scientists have<br />

warned that British nuclear stations may not be able to<br />

source nuclear fuel if it cannot be legally transported<br />

across borders. The shipment of medical isotopes used in<br />

scans and cancer treatment might be jeopardised.<br />

European workers on shared research projects, such as<br />

experimental fusion reactors, face an equally uncertain<br />

future without Euratom’s separate guarantees of freedom<br />

of movement.<br />

But the London-based Nuclear Industry Association<br />

(NIA), which represents more than 260 nuclear companies,<br />

cautiously welcomed Mr Clark’s statement, calling it a<br />

“useful and welcome step” in setting out the government’s<br />

approach in seeking to secure equivalent arrangements to<br />

those the UK benefits from as a member of Euratom. The<br />

NIA also welcomed clarity on the government’s intention<br />

to negotiate an implementation period to ensure a smooth<br />

transition from the current to new arrangements.<br />

It warned, however, that there is much still to do in<br />

equipping the UK’s regulator to take on Euratom’s safeguarding<br />

activities. The UK needs to reach post-Euratom<br />

agreements with the International Atomic Energy Agency,<br />

the US, Canada, Australia, Japan and others. It needs to<br />

agree new trading arrangements with the Euratom<br />

community and conclude a new funding agreement for the<br />

UK to continue its work in Euratom’s fusion R&D activities.<br />

“It is vital government continues to prioritise these issues<br />

in the period ahead if there is to be a successful outcome,”<br />

the NIA said.<br />

Unlike the dozens of other regulatory arrangements for<br />

industries such as aviation or pharmaceuticals, Euratom<br />

has been singled out for special treatment through the<br />

Brexit process because it is not technically part of the EU.<br />

Instead, the treaty that established this body to coordinate<br />

Europe’s civil nuclear energy industry was born in parallel<br />

with the birth of the European economic community in<br />

1957. The UK’s participation in Euratom therefore required<br />

a separate legal relationship with the European court of<br />

justice to enforce it.<br />

The nuclear industry had been hoping that because of<br />

this separation from the “mainstream Brexit,” the UK<br />

might decide to remain part of Euratom.<br />

The NIA and the Brussels-based trade body Foratom<br />

both said the UK should maintain its membership. They<br />

argued that the nuclear industry is global, and the ease of<br />

movement of nuclear goods, people and services enables<br />

new build, decommissioning, R&D and other programmes<br />

of work to continue without interruption.<br />

The government insists that leaving Euratom is an<br />

inevitable consequence of Brexit – a position shared by the<br />

European negotiators. But is says it wants continuity of<br />

open trade arrangements for nuclear goods and products<br />

to ensure the nuclear industry is able to continue to trade<br />

across EU borders without disruption.<br />

Support for remaining in Euratom had come not only<br />

from within the industry, but also from politicians.<br />

Conservative MPs said they would for the government to<br />

fight harder for the UK to stay in Euratom. The opposition<br />

Labour Party said Britain should remain in Euratom,<br />

adding it is increasingly clear that the government acted<br />

“recklessly” by giving up on membership.<br />

Scientists said leaving Euratom will cause widespread<br />

confusion and have a potentially devastating impact<br />

on the nuclear industry. They warned of potential problems<br />

related to the transportation of nuclear materials, including<br />

nuclear fuel; research, especially fusion research; and<br />

overseas investment in development of British nuclear<br />

power stations.<br />

Mr Clark’s statement addressed another concern for the<br />

industry – the issue of accessing a skilled pan-European<br />

workforce for the sector once Brexit is complete.<br />

Mr Clark said the nuclear sector needs the workforce for<br />

decommissioning, operation of existing facilities and<br />

new-build projects. He said proposals for the UK’s future<br />

immigration system will be set out shortly and “we will<br />

ensure that those businesses and communities, and<br />

parliament have the opportunity to contribute their views<br />

before making any decisions about the future system”.<br />

Whatever the outcome of negotiations with the EU,<br />

it is vital that the civil nuclear industry has a safeguards<br />

regime that meets international standards. But this<br />

is not dependent on the EU negotiations and the UK<br />

government is well advanced in delivering this plan, the<br />

statement said.<br />

Inside Nuclear with NucNet<br />

Euratom: Industry Softens Stance as Government Lays Out Plans for Transition ı NucNet


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Advertisement<br />

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DAtF Notes


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

216<br />

CALENDAR<br />

Calendar<br />

<strong>2018</strong><br />

08.04.-11.04.<strong>2018</strong><br />

International Congress on Advances in Nuclear<br />

Power Plants – ICAPP 18. Charlotte, NC, USA,<br />

American Nuclear Society (ANS), www.ans.org<br />

08.04.-13.04.<strong>2018</strong><br />

11 th International Conference on Methods and<br />

Applications of Radioanalytical Chemistry –<br />

MARC XI. Kailua-Kona, HI, USA, American Nuclear<br />

Society (ANS), www.ans.org<br />

12.04.<strong>2018</strong><br />

Desalination Powered by Nuclear Energy. Essen,<br />

Germany, Deutsche Meerwasser Entsalzung GmbH<br />

in cooperation with International Atomic Energy<br />

Agency (IAEA) and PowerTech Training Center<br />

( Kraftwerksschule, KWS), www.dme-gmbh.de,<br />

www.iaea.org, www.kraftwerksschule.de<br />

16.04.-19.04.<strong>2018</strong><br />

Einführung in die Kerntechnik. Mannheim,<br />

Germany, TÜV SÜD, nucleartraining@tuev-sued.de<br />

16.04.-17.04.<strong>2018</strong><br />

VdTÜV Forum Kerntechnik – Sicherheit im Fokus.<br />

Berlin, Germany, VdTÜV mit Unterstützung des<br />

TÜV NORD, des TÜV SÜD und des TÜV Rheinland,<br />

www.tuev-sued.de/tagungen<br />

17.04.-19.04.<strong>2018</strong><br />

World Nuclear Fuel Cycle <strong>2018</strong>. Madrid, Spain,<br />

World Nuclear Association (WNA),<br />

www.world-nuclear.org<br />

18.04.-19.04.<strong>2018</strong><br />

9. Symposium zur Endlagerung radioaktiver<br />

Abfälle. Vorbereitung auf KONRAD – Wege zum<br />

G2-Gebinde. Hanover, Germany, TÜV NORD<br />

Akademie, www.tuev-nord.de/tk-era<br />

22.04.-26.04.<strong>2018</strong><br />

Reactor Physics Paving the Way Towards More<br />

Efficient Systems – PHYSOR <strong>2018</strong>. Cancun, Mexico,<br />

www.physor<strong>2018</strong>.mx<br />

24.04.-25.04.<strong>2018</strong><br />

Integrated Waste Management Conference.<br />

Penrith, Cumbria, United Kingdom, The Nuclear<br />

Institute, www.iwmeurope.com<br />

08.05.-10.05.<strong>2018</strong><br />

29 th Conference of the Nuclear Societies in Israel.<br />

Herzliya, Israel. Israel Nuclear Society and Israel<br />

Society for Radiation Protection, ins-conference.com<br />

13.05.-19.05.<strong>2018</strong><br />

BEPU-<strong>2018</strong> – ANS International Conference on<br />

Best-Estimate Plus Uncertainties Methods. Lucca,<br />

Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,<br />

ANS, IAEA, NEA, www.nineeng.com/bepu/<br />

13.05.-18.05.<strong>2018</strong><br />

RadChem <strong>2018</strong> – 18th Radiochemical Conference.<br />

Marianske Lazne, Czech Republic,<br />

www.radchem.cz<br />

14.05.-16.05.<strong>2018</strong><br />

ATOMEXPO <strong>2018</strong>. Sochi, Russia,<br />

atomexpo.ru<br />

15.05.-17.05.<strong>2018</strong><br />

11 th International Conference on the Transport,<br />

Storage, and Disposal of Radioactive Materials.<br />

London, United Kingdom, Nuclear Institute,<br />

www.nuclearinst.com<br />

20.05.-23.05.<strong>2018</strong><br />

5 th Asian and Oceanic IRPA Regional Congress<br />

on Radiation Protection – AOCRP5. Melbourne,<br />

Australia, Australian Radiation Protection Society<br />

(ARPS) and International Radiation Protection<br />

Association (IRPA), www.aocrp-5.org<br />

29.05.-30.05.<strong>2018</strong><br />

49 th Annual Meeting on Nuclear Technology<br />

AMNT <strong>2018</strong> | 49. Jahrestagung Kerntechnik.<br />

Berlin, Germany, DAtF and KTG,<br />

www.nucleartech-meeting.com<br />

03.06.-07.06.<strong>2018</strong><br />

38 th CNS Annual Conference and 42nd CNS-CNA<br />

Student Conference. Saskotoon, SK, Canada,<br />

Candian Nuclear Society CNS, www.cns-snc.ca<br />

03.06.-06.06.<strong>2018</strong><br />

HND<strong>2018</strong> 12 th International Conference of the<br />

Croatian Nuclear Society. Zadar, Croatia, Croatian<br />

Nuclear Society, www.nuklearno-drustvo.hr<br />

04.06.-05.06.<strong>2018</strong><br />

13 th European Nuclear Energy Forum. Bratislava,<br />

Slova Republic, European Commission, ec.europa.eu<br />

04.06.-07.06.<strong>2018</strong><br />

10 th Symposium on CBRNE Threats. Rovaniemi,<br />

Finland, Finnish Nuclear Society, ats-fns.fi<br />

04.06.-08.06.<strong>2018</strong><br />

5 th European IRPA Congress – Encouraging<br />

Sustainability in Radiation Protection.<br />

The Hague, The Netherlands, Dutch Society<br />

for Radiation Protection (NVS), local organiser,<br />

irpa<strong>2018</strong>europe.com<br />

06.06.-08.06.<strong>2018</strong><br />

2 nd Workshop on Safety of Extended Dry Storage<br />

of Spent Nuclear Fuel. Garching near Munich,<br />

Germany, GRS, www.grs.de<br />

25.06.-26.06.<strong>2018</strong><br />

index<strong>2018</strong> – International Nuclear Digital<br />

Experience. Paris, France, Société Française d’Energie<br />

Nucléaire, www.sfen.org, www.sfen-index<strong>2018</strong>.org<br />

27.06.-29.06.<strong>2018</strong><br />

EEM – <strong>2018</strong> 15 th International Conference on the<br />

European Energy Market. Lodz, Poland, Lodz<br />

University of Technology, Institute of Electrical Power<br />

Engineering, Association of Polish Electrical<br />

Engineers (SEP), www.eem18.eu<br />

24.06.-30.06.<strong>2018</strong><br />

ANNETTE Summer School on Nuclear Technology,<br />

Nuclear Waste Management and Radiation<br />

Protection. Turku, Finland, Advanced Networking<br />

for Nuclear Education, Training and Transfer of<br />

Expertise, annettesummerschool.org, www.enen.eu<br />

29.07.-02.08.<strong>2018</strong><br />

International Nuclear Physics Conference 2019.<br />

Glasgow, United Kingdom, www.iop.org<br />

22.08.-31.08.<strong>2018</strong><br />

Frédéric Joliot/Otto Hahn (FJOH) Summer School<br />

FJOH-<strong>2018</strong> – Maximizing the Benefits of<br />

Experiments for the Simulation, Design and<br />

Analysis of Reactors. Aix-en-Provence, France,<br />

Nuclear Energy Division of Commissariat à l’énergie<br />

atomique et aux énergies alternatives (CEA)<br />

and Karlsruher Institut für Technologie (KIT),<br />

www.fjohss.eu<br />

28.08.-31.08.<strong>2018</strong><br />

TINCE <strong>2018</strong> – Technological Innovations in<br />

Nuclear Civil Engineering. Paris Saclay, France,<br />

Société Française d’Energie Nucléaire, www.sfen.org,<br />

www.sfen-tince<strong>2018</strong>.org<br />

05.09.-07.09.<strong>2018</strong><br />

World Nuclear Association Symposium <strong>2018</strong>.<br />

London, United Kingdom, World Nuclear Association<br />

(WNA), www.world-nuclear.org<br />

09.09.-14.09.<strong>2018</strong><br />

21 st International Conference on Water<br />

Chemistry in Nuclear Reactor Systems.<br />

San Francisco, CA, USA, EPRI – Electric Power<br />

Research Institute, www.epri.com<br />

17.09.-21.09.<strong>2018</strong><br />

62 nd IAEA General Conference. Vienna, Austria.<br />

International Atomic Energy Agency (IAEA),<br />

www.iaea.org<br />

17.09.-20.09.<strong>2018</strong><br />

FONTEVRAUD 9. Avignon, France,<br />

Société Française d’Energie Nucléaire (SFEN),<br />

www.sfen-fontevraud9.org<br />

17.09.-19.09.<strong>2018</strong><br />

4 th International Conference on Physics and<br />

Technology of Reactors and Applications –<br />

PHYTRA4. Marrakech, Morocco, Moroccan<br />

Association for Nuclear Engineering and Reactor<br />

Technology (GMTR), National Center for Energy,<br />

Sciences and Nuclear Techniques (CNESTEN) and<br />

Moroccan Agency for Nuclear and Radiological<br />

Safety and Security (AMSSNuR), phytra4.gmtr.ma<br />

26.09.-28.09.<strong>2018</strong><br />

44 th Annual Meeting of the Spanish Nuclear<br />

Society. Avila, Spain, Sociedad Nuclear Española,<br />

www.sne.es<br />

30.09.-04.10.<strong>2018</strong><br />

TopFuel <strong>2018</strong>. Prague, Czech Republic, European<br />

Nuclear Society (ENS), American Nuclear Society<br />

(ANS). Atomic Energy Society of Japan, Chinese<br />

Nuclear Society and Korean Nuclear Society,<br />

www.euronuclear.org<br />

02.10.-04.10.<strong>2018</strong><br />

7 th EU Nuclear Power Plant Simulation ENPPS<br />

Forum. Birmingham, United Kingdom, Nuclear<br />

Training & Simulation Group, www.enpps.tech<br />

14.10.-18.10.<strong>2018</strong><br />

12 th International Topical Meeting on Nuclear<br />

Reactor Thermal-Hydraulics, Operation and<br />

Safety – NUTHOS-12. Qingdao, China, Elsevier,<br />

www.nuthos-12.org<br />

14.10.-18.10.<strong>2018</strong><br />

NuMat <strong>2018</strong>. Seattle, United States,<br />

www.elsevier.com<br />

16.10.-17.10.<strong>2018</strong><br />

4 th GIF Symposium at the 8 th edition of Atoms<br />

for the Future. Paris, France, www.gen-4.org<br />

22.10.-24.10.<strong>2018</strong><br />

DEM <strong>2018</strong> Dismantling Challenges: Industrial<br />

Reality, Prospects and Feedback Experience. Paris<br />

Saclay, France, Société Française d’Energie Nucléaire,<br />

www.sfen.org, www.sfen-dem<strong>2018</strong>.org<br />

22.10.-26.10.<strong>2018</strong><br />

NUWCEM <strong>2018</strong> Cement-based Materials for<br />

Nuclear Waste. Avignon, France, French<br />

Commission for Atomic and Alternative Energies<br />

and Société Française d’Energie Nucléaire,<br />

www.sfen-nuwcem<strong>2018</strong>.org<br />

24.10.-25.10.<strong>2018</strong><br />

Chemistry in Power Plant. Magdeburg, Germany,<br />

VGB PowerTech e.V., www.vgb.org<br />

05.11.-08.11.<strong>2018</strong><br />

International Conference on Nuclear<br />

Decom missioning – ICOND <strong>2018</strong>. Aachen,<br />

Eurogress, Germany, achen Institute for Nuclear<br />

Training GmbH, www.icond.de<br />

2019<br />

07.05.-08.05.2019<br />

50 th Annual Meeting on Nuclear Technology<br />

AMNT 2019 | 50. Jahrestagung Kerntechnik.<br />

Berlin, Germany, DAtF and KTG,<br />

www.nucleartech-meeting.com<br />

Calendar


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Heat Transfer Systems for Novel Nuclear<br />

Power Plant Designs<br />

Sebastian Vlach, Christoph Fischer and Herman van Antwerpen<br />

This article focuses on work that involves designing or modifying heat exchangers that usually can be found in the auxiliary<br />

systems of any power plant. The basic premise of the article is to show that the software provides a one-stop solution for<br />

designing many types of heat transfer systems, where the interaction between various loops connected by heat exchangers can<br />

be assessed. This article especially addresses the audience among nuclear power plants as the quality control in the development<br />

of the software makes it most suitable for nuclear related work. Moreover, the software discussed in this article has the<br />

capability to do contaminant tracing, which could be very useful for nuclear contamination studies in designing specialized<br />

ventilation systems. To highlight the versatility of the software network approach it will be shown how to model any setup and<br />

kind of heat exchanger such as plate, tube-in-tube, liquid/gas, finned tube etc. Additionally, the Koeberg pressurized water<br />

reactor (PWR) steam generator comparison and the Hamm-Uentrop thorium high temperature reactor (THTR) steam<br />

generator comparison are shown as practical examples.<br />

Introduction “Every type of technology benefits from advances inspired by new knowledge and understanding.<br />

Although nuclear energy has operated mostly safely in the past, nuclear engineers do continue to devise new ideas for<br />

making nuclear energy even safer and more secure. The future of reliable nuclear energy requires scientific research to<br />

verify that new types of advanced nuclear fuels and materials are robust enough to withstand the conditions inside a<br />

nuclear reactor during normal and abnormal conditions.” (Idaho National Laboratory).<br />

217<br />

OPERATION AND NEW BUILD<br />

Based on the laws of thermodynamics 1D system<br />

simulation is extremely robust, fast, and reliable. One<br />

software package for 1D system simulation that gains<br />

more and more attention recently was developed in the<br />

early 1990ies by a South African company, namely M-Tech<br />

Industrial. Initially, Flownex® Simulation Environment<br />

was developed for aerospace applications and the energy<br />

sector. Moreover, nuclear validation and verification were<br />

supervised by the governmental ESKOM institution<br />

through its subsidiary PBMR Ltd., who developed a<br />

high-temperature gas-cooled (pebble-bed) reactor in<br />

cooperation with Jülich Research Centre at that time.<br />

Specifically for the nuclear safety analyses required by<br />

PBMR, the software has Nuclear Quality Assurance<br />

( NQA-1) Certification and its development process is<br />

based on ISO 9001.<br />

System simulation programmes provide engineers and<br />

designers a fast and efficient way to set up simulation<br />

models for simple as well as complex fluid dynamic<br />

networks. Such networks commonly contain several<br />

components such as fans, pumps, heat exchangers etc. that<br />

can be computed almost instantly. Furthermore, dynamics<br />

and the control of such networks can be investigated by<br />

running different operation scenarios, such as start-up,<br />

shut down, and various loading conditions, where steady<br />

state and transient effects are taken into account. Thus,<br />

weak spots within a system can be eliminated during the<br />

design process prior to manufacturing as literally any<br />

modification can be tested virtually.<br />

Subsequently, the user is able to analyse the results very<br />

quickly.<br />

Material data that the software supports can be<br />

gaseous, gas mixtures, as well as incompressible pure<br />

fluids and two-phase pure fluids. The user is able to access<br />

a vast library based on the NIST data base. Hence, complex<br />

flows can be modelled using temperature and pressure<br />

dependent material data as well as multiphase effects like<br />

conden sation, evaporation, and cavitation.<br />

The software is equipped with a vast array of components<br />

that cover most required simulation scenarios.<br />

Those components can be used as single components or as<br />

building blocks of components found in thermal fluid<br />

systems or subsystems.<br />

Building blocks, with various levels of detail are<br />

available to model heat transfer phenomena as shown in<br />

Figure 1. Some of the simple heat exchanger models<br />

utilises the Number of Transfer Units (NTU) Method while<br />

other more complex versions employ a fully discretised<br />

approach to heat exchanger modelling. The heat exchanger<br />

types range from tube to plate heat exchangers that can be<br />

modelled as parallel, counter, or cross flow types. Other<br />

components can be vessels, reactors, tube systems, valves,<br />

pumps, fans, compressors, seals etc. Moreover, a whole<br />

library of com ponents for dynamics and control is available<br />

within the software.<br />

1D System Simulation<br />

Flownex® Simulation Environment includes all the<br />

necessary numerical formulations for solving all important<br />

thermo-fluid physical phenomena and moreover, a modern<br />

Windows-GUI that enables an intuitive and easy interaction<br />

for the user. Therefore, the user can concentrate on<br />

design and optimisation rather than on the complexities<br />

usually associated with operating such calculation software.<br />

Typical simulations are run in real time or in the<br />

order of seconds, which makes parameter studies and<br />

optimisation loops extremely fast and very efficient.<br />

| | Fig. 1.<br />

Library for heat exchangers [1].<br />

| | Fig. 2.<br />

Heat transfer library [1].<br />

Operation and New Build<br />

Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 218<br />

| | Fig. 3.<br />

Plate heat exchanger model with a two-pass hot side and a single-pass cold side (left) and<br />

schematic (right) [2].<br />

If one is looking into thermo dynamic analyses, simple<br />

components would be used to represent radiation,<br />

conduction, or convection as shown in Figure 2. Thus,<br />

heat exchangers can be custom-built to answer the<br />

question at hand.<br />

Figure 3 shows a simple custom made plate heat<br />

exchanger consisting of composite heat transfer components<br />

and pipe components. The flow path is represented<br />

with a hydraulic diameter and the flow area. The plates are<br />

represented with heat transfer area and actual metal<br />

thickness. User-specified correlations according to the<br />

plate corrugation profiles are defined allowing for full<br />

discretisation along the flow path that results in accurate<br />

pinch-point calculation and transient response.<br />

Another heat exchanger example is shown in Figure 4<br />

where a finned tube air-water heat exchanger can be seen.<br />

The air-side is modelled as a straight-through flow path<br />

(left to right) whereas the water-side is modelled as an<br />

up-down overall counter flow (right to left) configuration<br />

according to the design of the header box plates. The fully<br />

discretised flow path provides an accurate transient<br />

response. The fin-side pressure drop and heat transfer<br />

correlations can be specified with Chilton-Colburn J-factor<br />

tables.<br />

Heat exchangers are crucial for any power plant design.<br />

Figure 5 shows the schematic of the Koeberg PWR steam<br />

generator and the equivalent model built in the software.<br />

For the dryer/separator a complete phase separation is<br />

assumed. The recirculation flow rate is calculated from<br />

buoyancy-driven flow (red circle) that is dependent on<br />

heat transfer coefficient and flow resistance. The model<br />

also assumes a homogeneous two-phase flow. The Chen<br />

correlation for the shell side was implemented with<br />

scripting. Specific material properties can be implemented<br />

via Engineering Equation Solver (EES) coupling or<br />

scripting if necessary.<br />

Table 1 shows the comparison of measured and<br />

simulated data of the Koeberg PWR (South Africa) steam<br />

generator at 60 % and 100 % power load. The software<br />

shows reasonably good agreement to the measured data,<br />

especially when looking at the recirculation ratio R circ<br />

which is a good indication of the overall calculation<br />

accuracy.<br />

Another power plant example is the Hamm-Uentrop<br />

thorium high temperature reactor (THTR-300, Germany)<br />

power plant. One challenge in modelling the THTR is that<br />

at certain combinations of flow rate and heat input, the<br />

flow could be oscillatory. Several types of oscillation are<br />

possible: density wave, pressure wave, and critical heat<br />

flux (dryout)-related oscillations. Fundamental fluiddynamic<br />

modelling is crucial to detect this, which is<br />

provided in the software. Furthermore, this capability is<br />

critical to determine the minimum flow through a steam<br />

generator because it is typically at low power levels that<br />

the steam flow becomes oscillatory. Figure 6 shows the<br />

schematic of the THTR-300 and an equivalent model built<br />

in the software.<br />

The THTR steam generator plant was modelled to<br />

verify the steady-state performance of the assembled<br />

steam generator model. Figure 7 shows the comparison of<br />

measured and simulated data of the THTR-300 at 40 %<br />

and 100 % power load. The software shows very good<br />

agreement to the measured data. The simulation revealed<br />

| | Fig. 4.<br />

Model of a finned tube air-water heat exchanger with multiple water-side passes and a single air pass (top) and schematic (bottom) [2].<br />

T pi<br />

[C]<br />

T po<br />

p so<br />

[kPa]<br />

p si T si x so ṁ s<br />

[kg/s]<br />

Q boiler<br />

[MW]<br />

R circ<br />

60 % power Koeberg 294 273 4889 5055 195 1.0 341 670 7.0<br />

Simulation 294 273 4889 4919 195 1.0 341 666 6.4<br />

100 % power Koeberg 312 279 4911 5277 220 1.0 618 1143 3.8<br />

Simulation 312 280 4911 4951 220 1.0 618 1092 3.8<br />

| | Tab. 1.<br />

Koeberg PWR steam generator comparison.<br />

Operation and New Build<br />

Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

that the helium-side heat transfer correlation needed to<br />

have an appropriate Reynolds-number dependence as the<br />

error became quite large at lower power or flow levels<br />

neglecting this.<br />

As aforementioned, heat exchangers are crucial for<br />

any power plant design, especially when designing new<br />

power plants. In addition to the heat transfer modelling<br />

capabilities and with respect to nuclear power generation<br />

the software has recently expanded the Generic Nuclear<br />

Reactor model to simulate the latest nuclear reactor<br />

designs of any geometry. Novel nuclear reactor designs<br />

include liquid fuel reactors, liquid-metal-cooled reactors,<br />

and high temperature gas-cooled reactors (HTGR). In<br />

more detail, there are six reactor types that have gained<br />

researches interest all over the world:<br />

• Very High Temperature Reactor,<br />

• Molten Salt Reactor,<br />

• Sodium-Cooled Fast Reactor,<br />

• Supercritical-Water-Cooled Reactor,<br />

• Gas-Cooled Fast Reactor, and<br />

• Lead-Cooled Fast Reactor.<br />

The new “generalized fuel zone” in the GNR model that is<br />

shown in Figure 8 is capable of handling any fuel geometry<br />

and any fluid type. It expands the geometry capability to<br />

plate fuel, cylindrical fuel rods, spherical fuel elements,<br />

irregular cross-section fuel (like the four-lobe cross-shape<br />

produced by the Lightbridge Corporation), as well as<br />

prismatic block fuel used in some HTGRs.<br />

Appropriate pressure drop and heat transfer correlations<br />

can be selected from the built-in library or defined by<br />

the user. For neutronic calculations, the generalized fuel<br />

zone can provide temperature feedback, as well as heat<br />

generation in all solids and in the core coolant.<br />

The default neutronics model that is supplied with the<br />

software is the point kinetic model which requires the<br />

following inputs:<br />

• Temperature feedback coefficients,<br />

• Heat distribution map, and<br />

• Control rod worth vs. position.<br />

This point kinetic model is provided in a user-editable C#<br />

script, which makes it possible to replace the point kinetic<br />

model by linking the simulation model to an external<br />

neutronics code. The scripted neutronics model also makes<br />

it possible for the user to define one’s own feedback<br />

mechanisms based on the design of the specific reactor.<br />

| | Fig. 5.<br />

Koeberg PWR steam generator schematic (left) [2] and simulation model (right).<br />

| | Fig. 6.<br />

Hamm-Uentrop THTR schematic (left) [2] and simulation model (right).<br />

OPERATION AND NEW BUILD 219<br />

| | Fig. 7.<br />

Hamm-Uentrop THTR-300 steam generator comparison experiment (Exp) [3] and simulation (FNX).<br />

Operation and New Build<br />

Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 220<br />

| | Fig. 8.<br />

Schematic of the General Nuclear Reactor (GNR) model [1].<br />

Being able to model all kinds of heat transfer accurately<br />

and to include fission physics makes the software a<br />

valuable tool for every nuclear engineer and power cycle<br />

developer. Figure 9 shows an integrated simulation model<br />

that includes a reactor, steam generator, heat exchange,<br />

and some turbomachinery.<br />

Summary<br />

In order to size control valves or determine the control<br />

strategy for a loop, it is necessary to have the pump<br />

performance curve, the heat exchanger pressure drop and<br />

heat transfer characteristics as well as reactor dynamic<br />

behaviour in one simulation model. In this article, a fast<br />

and efficient solution for designing many types of heat<br />

transfer systems is presented. It was shown how to model<br />

any setup and kind of heat exchanger such as plate, tubein-tube,<br />

liquid/gas, finned tube etc. Flownex® Simulation<br />

Environment offers a straight-forward workflow for<br />

engineers who are involved in designing auxiliary systems<br />

that usually contain one or more heat exchangers, such as<br />

in the power plant industry. The software is a specialized<br />

software (e.g. used by ITER, X Energy, BATAN, Hyundai<br />

Heavy Industries) for sizing specific types of heat<br />

exchangers or for doing basic steady-state and transient<br />

mass-and-energy balances. The value of the software in<br />

this area is that one can really integrate the information<br />

from all available sources into a single representative<br />

model, where one can size all kind of devices, test control<br />

strategies, and do integrated system-level analysis and<br />

design. Furthermore, examples from the nuclear power<br />

plant industry, namely the Koeberg PWR steam generator<br />

and the Hamm-Uentrop THTR-300 steam generator which<br />

demonstrated the software’s usability for nuclear related<br />

work were shown. In addition, the lately incorporated<br />

Generic Nuclear Reactor model was introduced.<br />

Further Reading<br />

| | Flownex® SE: www.flownex.de<br />

| | M-Tech Industrial: www.mtechindustrial.com<br />

| | Idaho National Laboratory: www.inl.gov<br />

References<br />

[1] Flownex (2017) User Manual.<br />

[2] Van Antwerpen, H.: Design and Optimization of Advanced<br />

Nuclear Technologies with 1-d Simulation. 7 th Annual<br />

International SMR and Advanced Reactor Summit 2017,<br />

30-31 March, Atlanta, GA, USA.<br />

[3] Esch, M., Hurtado, A., Knoche, D., and Tietsch, W.: Analysis of the<br />

Influence of Different Heat Transfer Correlations for HTR Helical<br />

Coil Tube Bundle Steam Generators with the System Code TRACE.<br />

Nuclear Engineering and Design, 251, 374-380, 2012.<br />

[4] Van Antwerpen, H., Chi, H., Brits, Y., and Botha, F.: Plant-Wide<br />

Simulation Model for Transient Studies on the Xe-100. 2016 ANS<br />

Winter Meeting and Nuclear Technology Expo, 6-10 November<br />

2016, Las Vegas, NV, USA.<br />

Authors<br />

Sebastian Vlach<br />

Leiter Marketing & Vertrieb<br />

Christoph Fischer (PhD)<br />

CFX Berlin Software GmbH<br />

Berlin, Germany<br />

Herman van Antwerpen (PhD)<br />

M-Tech Industrial (Pty) Ltd<br />

South Africa<br />

| | Fig. 9.<br />

Layout of a complete plant power cycle with an example reactor geometry input map (left) [4].<br />

Operation and New Build<br />

Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Experimental and Analytical Tools<br />

for Safety Research of GEN IV Reactors<br />

G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak<br />

Current research on nuclear safety in the world, in addition to supporting existing nuclear power plants (PLEX,<br />

mitigation of severe accidents, the development of accident tolerant fuel, decommissioning, etc.), is focused on the<br />

more detailed aspects of the new reactors. The new generation reactors are expected inter alia to use innovative types<br />

of fuel and new types of coolants, such as e.g. Super-Critical Water (SCW), supercritical CO 2 , liquid metals, fluoride<br />

salts or high-temperature Helium. The paper will describe new experimental infrastructure build recently in Research<br />

Centre Řež under the SUSEN (Sustainable Energy) project and available analytical tools for supporting safety research<br />

of GEN IV reactors. Two experimental loops - SCWL (Supercritical Water Loop) and HTHL (High Temperature Helium<br />

Loop) will serve as in-pile loops in the active core of the research reactor LVR-15. The loops insertion in the reactor<br />

LVR-15 requires performing additional safety analyses studying the mutual interference of the loops and the reactor,<br />

especially in conditions of abnormal operation or accident conditions of the loops. The paper will provide examples of<br />

these analyses made using codes ATHLET (supercritical water) and TRACE (high temperature He) illustrating process<br />

of their assessment and practical use. These activities provide significant opportunity for TSO team in building its new<br />

competencies.<br />

Revised version<br />

of a paper presented<br />

at the Eurosafe,<br />

Paris, France, 6 and<br />

7 November 2017.<br />

OPERATION AND NEW BUILD 221<br />

1 Introduction<br />

The Centrum Výzkumu Řež (CVŘ) and<br />

its partners in the Czech Republic and<br />

abroad are supporting the development<br />

[1] of the Generation IV and<br />

Fusion concepts as well as demonstrators<br />

of these technologies such<br />

as ALLEGRO, ALFRED, DEMO and<br />

others. For this reason, the CVŘ has<br />

had a large R&D program financed<br />

from SUStainable Energy (SUSEN)<br />

project and from its continuation<br />

Research 4 Sustenibility (R4S) [2].<br />

The construction and the operation of<br />

the new SUSEN infrastructure was<br />

supported by the grant of the Ministry<br />

of Education, Youth and Sports as the<br />

part of state help for the large research<br />

infrastructure in the Czech Republic<br />

dedicated to the period 2011–2019.<br />

The SUSEN project consists of 4<br />

programs:<br />

1. Technological Experimental Circuits<br />

(TEO)<br />

2. Structural and System Diagnostics<br />

(SSD)<br />

3. Nuclear Fuel Cycle (NFC)<br />

4. Material Research (MAT)<br />

Within this program, several facilities<br />

were designed and built in order to<br />

study and to address new challenges<br />

of such new technologies. In particular,<br />

the paper focuses on two new<br />

loops which are going to be inserted<br />

inside the LVR-15 research reactor<br />

existing in Řež. The LVR-15 is a light<br />

water tank-type research reactor in<br />

operation since 1957. It is placed in a<br />

stainless steel vessel under a shielding<br />

cover, has forced cooling, uses IRT-4M<br />

type fuel and an has an operational<br />

power level of 10 MWt. The reactor<br />

operations run in campaigns that<br />

usually last for 3 weeks, followed by<br />

an outage lasting for 10 to 14 days<br />

necessary for maintenance and fuel<br />

reloading. There can be also other<br />

campaigns which can operate for<br />

‘short-time’ experiments. Some of the<br />

LVR-15 applications are in the field of<br />

material irradiation research and services,<br />

neutron physics, development<br />

and production of new radiopharmaceuticals<br />

[3]. The loops in concern are<br />

the High Temperature Helium Loop<br />

(HTHL) and the Super Critical Water<br />

Loop (SCWL) and their main scope<br />

are to analyse the cladding behaviour<br />

and structural materials under different<br />

pressure, temperature and coolant<br />

media conditions different from the<br />

standard Light Water Reactors (LWR)<br />

technology [2].<br />

In order to get the regulatory<br />

permit for in-pile operation of these<br />

loops in LVR-15, CVŘ has to prepare<br />

an amendment to the Final Safety<br />

Analyses Report (FSAR) containing<br />

safety analyse of the loops under<br />

| | Fig. 1.<br />

CVR Facilities list.<br />

operational and accidental conditions.<br />

Aim of this paper is to present<br />

the methodology and the analyses<br />

done in support of this process, starting<br />

from code benchmarking/assessment<br />

and the methods adopted in preparing<br />

the safety case.<br />

2 Facilities description<br />

The map of experimental facilities put<br />

into operation in 2016 and those<br />

under preparation to be finalized in<br />

2017 is shown in Figure 1 in the<br />

technology – knowledge map.<br />

In particular, the SCWL and HTHL<br />

represent a pioneer and unique experimental<br />

facility for Gen. IV and Fusion.<br />

2.1 SCWL<br />

The SCWL is going to be a part of a<br />

research facility dedicated to GIV<br />

technologies which will focus on<br />

obtaining data in several areas of the<br />

supercritical fields like: corrosion<br />

processes of construction materials in<br />

supercritical water, with influence of<br />

Operation and New Build<br />

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 222<br />

radiation field, supercritical water<br />

radiolysis and its influence on materials<br />

and water chemistry, development<br />

and testing of sensors, mostly for<br />

measuring of electrochemical potential<br />

(ECP), testing and optimization of<br />

supercritical water regimes [2]. The<br />

specimens being tested will be placed<br />

into the test chamber located in the<br />

active channel where high pressure/<br />

temperature of SCW flow parameters<br />

will be reached.<br />

The SCWL heart is the active<br />

channel, where water reaches required<br />

parameters (pressure of 25 MPa;<br />

temperature of 600 °C; very clean<br />

demineralised water. After successful<br />

out-of-pile (i.e. non active, without<br />

presence of radiation field) operation,<br />

the active channel will be inserted into<br />

the LVR-15 research reactor core. The<br />

bottom part of the active channel is<br />

then submerged between the core’s<br />

fuel assemblies and will face a neutron<br />

flux of up to 1.5 × 10 18 n/m 2 s (thermal<br />

neutrons) and 3 × 10 18 n/m 2 s (fast<br />

neutrons).<br />

The fluid flows in the SCWL is<br />

shown in Figure 2a while the CAD<br />

sketches is shown in Figure 2b.<br />

The active channel has been<br />

modelled with the use of ATHLET<br />

code in two different configurations<br />

see S3.1:<br />

• the out-of-pile configuration that<br />

takes into consideration only pressure<br />

and temperature conditions;<br />

• the in-pile configuration, with the<br />

channel placed inside the LVR-15<br />

active core, that takes into account<br />

also the gamma heating.<br />

2.2 HTHL<br />

HTHL test facility is designed for the<br />

material testing under the simulation<br />

of Gas-cooled Fast Reactor (GFR)<br />

and/or Very High Temperature Reactor<br />

(VHTR) operational conditions.<br />

The specimens being tested will be<br />

placed into the test chamber located<br />

in the active channel where high<br />

pressure/temperature helium flow<br />

parameters will be reached. In addition<br />

to that exposure, during the<br />

in-pile operation, with the active<br />

channel placed into predefined position<br />

of LVR-15 active core rectangular<br />

grid the irradiation effects on the<br />

samples will be studied. The scheme<br />

of the flows in the HTHL is shown in<br />

Figure 3a while the CAD sketch can<br />

be seen Figure 3b.<br />

The active channel has been<br />

modelled with the use of TRACE<br />

code in two different configurations<br />

see S3.2:<br />

• the out-of-pile configuration that<br />

takes into consideration only pressure<br />

and temperature conditions;<br />

• the in-pile configuration, with the<br />

channel placed inside the LVR-15<br />

active core, that takes into account<br />

also the gamma heating.<br />

Views of the channel and of the<br />

coolant flow pattern can be seen in<br />

Figure 3a and Figure 3b.<br />

The temperature inside the channel<br />

is reached through electrical<br />

heater and the coolant flow circulation<br />

is maintained using a two stages<br />

compressor.<br />

3 Methodology<br />

The methodology used to select the<br />

codes and to perform the analyses for<br />

the amendment for the LVR-15 FSAR<br />

consisted in 3 – independent steps:<br />

• Searching and assessing the codes<br />

ability to simulate helium and SCW<br />

during steady-state and transients<br />

conditions.<br />

• Creating the loops model to be<br />

used for the TH analyses and<br />

developing it based on the steadystate<br />

thermohydraulic parameters<br />

• Performing analyses of the selected<br />

scenarios in order to verify the<br />

safety criteria and obtaining the<br />

necessary data for the structural<br />

analyses.<br />

The present methodology complies<br />

with the [4] IAEA standard in introducing<br />

new research facilities inside<br />

nuclear research installations such as<br />

the LVR-15 reactor.<br />

3.1 ATHLET 3.1A code<br />

ATHLET 3.1 patch A code [5] is a<br />

thermal hydraulic system code<br />

developed by the GRS for simulating<br />

time-dependent phenomena in the<br />

PWRs and BWRs. Furthermore, the<br />

code can also simulate GEN IV working<br />

fluids like helium, liquid metals<br />

and supercritical water.<br />

The heat transfer behaviour in<br />

supercritical water represents a<br />

challenging task mainly connected<br />

with ensuring safety and reliable<br />

operation. Nowadays, the understanding<br />

of the supercritical water<br />

regimes is rather limited, specifically<br />

regarding the close proximity of the<br />

critical point.<br />

For the simulation of supercritical<br />

water, a range of properties approximation<br />

has been extended up to a<br />

pressure of 100 MPa. An additional<br />

module cover the pressure range from<br />

22.5 to 100 MPa. The transition<br />

between subcritical and the supercritical<br />

properties is performed by a<br />

suitable interpolation between these<br />

packages for pressures between 22.0<br />

and 22.5 MPa [5].<br />

In ATHLET 3.1A the selection of<br />

correlations for supercritical water is<br />

performed by switching a built in flag<br />

found in the heat structure module.<br />

A number of six correlations are<br />

available which were tested against<br />

the results obtained by IAEA-benchmark<br />

exercise [6] and three of them<br />

were selected for the purpose of the<br />

certification and further use.<br />

3.2 TRACE 5 Patch 4 codes<br />

The TRACE code has been used as an<br />

alternative to the RELAP5/Mod3.3<br />

code, since US NRC decided to stop<br />

| | Fig. 2a.<br />

SCWL Flow.<br />

| | Fig. 2b.<br />

CAD Sketches.<br />

| | Fig. 3a.<br />

Flow in HTHL.<br />

| | Fig. 3b.<br />

HTHL CAD Sketches.<br />

Operation and New Build<br />

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 4.<br />

HTHL TRACE Nodalization.<br />

the development of RELAP starting<br />

with next year.<br />

TRACE has been designed to perform<br />

best-estimate analyses of loss- ofcoolant<br />

accidents (LOCAs), operational<br />

transients, and other accident<br />

scenarios in pressurized light-water<br />

reactors (PWRs) and boiling lightwater<br />

reactors (BWRs). It can also<br />

model phenomena occurring in<br />

experimental facilities designed to<br />

simulate transients in reactor systems.<br />

Models used include multidimensional<br />

two-phase flow, none quilibrium<br />

thermo-dynamics, generalized heat<br />

transfer, reflood, level tracking, and<br />

reactor kinetics. In addition, TRACE is<br />

able to simulate several other coolants<br />

such as helium and water in subcooled<br />

condition and atmospheric pressure<br />

(LVR-15 conditions). [7], [8]<br />

For this reason, TRACE code was<br />

selected and used for the simulation in<br />

the Helium at 7 MPa with a temperature<br />

rise from 200 °C up to 900 °C<br />

(nominal parameters for HTHL). The<br />

correlation adopted for simulating the<br />

heat transfer from heat structures to<br />

the helium coolant and vice versa<br />

implemented in TRACE are Gnielinsky<br />

and El Genk [7-9].<br />

3.3 Codes assessment<br />

The code assessment was done by<br />

benchmarking of the codes with<br />

available experimental results done in<br />

different facilities around the world.<br />

One of the most important steps<br />

was selecting the code that can<br />

perform the heat transfer calculation<br />

under the high temperature He or<br />

SCW conditions along with adequate<br />

correlations [10], [11]. In the case of<br />

ATHLET, the code was carefully assed<br />

and benchmarked with experimental<br />

results of a project coordinated by<br />

IAEA [6] for steady state and with<br />

Chinese SWAMUP facility [12] for<br />

| | Fig. 5.<br />

SCWL ATHLET Nodalization.<br />

the transition from supercritical to<br />

subcritical condition.<br />

The aim of this analyses was to<br />

simulate the deterioration phenomenon<br />

[9] of heat transfer with fluid<br />

transiting between subcritical and<br />

supercritical condition. According to<br />

Ref. [6], Mokry, Gupta and Watts-<br />

Chou correlations show acceptable<br />

prediction capabilities of the Heat<br />

Transfer Coefficient (HTC). Both our<br />

analyses and IAEA CRP program<br />

concluded that an uncertainty for<br />

calculating HTC is about ±25% while<br />

the calculating wall temperature was<br />

between ±10 to 15 %. As a result of<br />

this exercise, the code certification<br />

was obtained from SONS (State of<br />

Office for Nuclear Safety) in March<br />

2017 for using the code in simulating<br />

the CVŘ SCWL.<br />

The TRACE assessment was done<br />

with the data available from the<br />

project GoFastR [13] financed by the<br />

EC in the Framework Program 7, in<br />

particular with data related to the<br />

HE-FUS3 facility [14], [15]. The<br />

facility operational parameters are<br />

similar to the HTHL.<br />

The TRACE HE-FUS3 thermal hydraulic<br />

model was developed and<br />

compared with experimental data<br />

from steady state loop operation and<br />

selected transients. The comparison<br />

showed that the TRACE T/H model<br />

can simulate the helium temperatures<br />

as well as the piping wall temperatures<br />

along the different sections<br />

of the facility accurately. After a sensitivity<br />

analysis, the electrical heater<br />

power has been lowered to 10.76 kW.<br />

The certification for TRACE code was<br />

obtained from SONS in December<br />

2016 by CVŘ for simulating water in<br />

PWR condition, sub-cooled water at<br />

atmospheric pressure (such as LVR-15<br />

operational condition) and helium<br />

behaviour in the range of 7 MPa for a<br />

temperature range between 200 to<br />

900 °C. [16]<br />

3.4 Model description for<br />

HTHL and SCWL<br />

The HTHL and SCWL are similar<br />

experimental facilities characterized<br />

by 2 steps upward and downward<br />

flows, although some major differences<br />

exist in the design. In particular,<br />

the HTHL active channel contains<br />

all necessary components for heat<br />

transfer inside except of the compressor<br />

and the main compensator, which<br />

are located in the chemical control<br />

system. The Figure 4 shows the<br />

TRACE nodalization containing simulated<br />

components.<br />

The SCWL is different in such way<br />

that it needs some extra components<br />

larger than the HTHL to help the sub<br />

critical water to become gas. For this<br />

reason additional axillary facilities,<br />

such as a recuperator, cooler, pump,<br />

compensator and other 4 sections of<br />

electrical heater are located in a<br />

different building along with the<br />

chemical control system.<br />

The ATHET SCWL loop model<br />

shown in Figure 5 is focused mainly<br />

on the active channel from inlet to<br />

outlet, although all the previous<br />

components are also simulated as a<br />

part of the primary and the secondary<br />

circuits. In addition to the primary<br />

and the secondary circuits of the<br />

SCWL, there is the third open loop<br />

representing the active channel position<br />

into the LVR-15 core and providing<br />

additional heat transfer between<br />

active channel and reactor coolant.<br />

3.5 Analysed Scenarios<br />

The planned in-pile operation of both<br />

loops requires an amendment of the<br />

LVR-15 Final Safety Report providing<br />

thermohydraulic and structural integrity<br />

analyses during normal operation<br />

OPERATION AND NEW BUILD 223<br />

Operation and New Build<br />

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 224<br />

Normal operating<br />

conditions<br />

Steady State<br />

LVR-15 Start up<br />

LVR-15 Shutdown<br />

Loops Start up<br />

Loops Shutdown<br />

and during Loss of Flow Accident<br />

( LOFA) and Loss of Coolant Accident<br />

(LOCA) accident conditions. In particular,<br />

the structural integrity analyses<br />

required the temperature profile<br />

inside the Pressure Envelope (PE) as<br />

boundary condition. For this reason<br />

the normal operation and abnormal<br />

operation conditions were calculated<br />

using TRACE and ATHLET codes with<br />

very narrow mesh nodal distribution<br />

in the PE. For structural integrity<br />

following criteria and limitation due<br />

to the non-boiling condition in LVR-15<br />

were used:<br />

1. PE maximum temperature during<br />

normal/abnormal transients is less<br />

than 450 °C.<br />

2. PE maximum temperature during<br />

accident conditions is less than<br />

500 °C.<br />

3. Aluminium surface of the Receiver<br />

maximum temperature in contact<br />

with LVR-15 coolant less than 45 °C<br />

during normal/abnormal conditions.<br />

4. Aluminium surface of the Receiver<br />

maximum temperature in contact<br />

with LVR-15 coolant less than 60 °C<br />

during accident conditions.<br />

In the case of accident conditions,<br />

both active channels of HTHL and<br />

SCWL will have to be replaced. The<br />

analysed scenarios are described in<br />

the Table 1.<br />

4 Illustrative results<br />

The results described in the paper<br />

refer to the simulations of SCWL and<br />

Pressure<br />

tests<br />

(not simulated)<br />

| | Tab. 1.<br />

Operational and Accident Scenarios Description.<br />

Abnormal<br />

conditions<br />

Switch off Loops Electrical<br />

Heater for 1 min.<br />

LVR-15 SCRAM and switch off<br />

of Loops Electrical Heater<br />

at t = 0 s + pump trip after 1 min.<br />

Switch off Loops Electrical Heater<br />

at t = 0 s + LVR15 SCRAM and<br />

Pump Trip after 3 min.<br />

Parameter Value Unit<br />

Pressure 25 MPa<br />

Inlet Flow<br />

Temperature<br />

Outlet Flow<br />

Temperature<br />

Max Flow<br />

Temperature<br />

Sample Area<br />

Mass flow<br />

| | Tab. 2.<br />

SCWL main parameters calculated during<br />

steady state.<br />

HTHL during the steady state operation<br />

with continuing in LOFA condition.<br />

The results represent an extract<br />

of the large number of calculations of<br />

various combinations of operational<br />

transients with the aim to demonstrate<br />

the capabilities of the codes to<br />

simulate behaviour the loops.<br />

4.1 SCWL steady state and<br />

LOFA analyses<br />

The main parameters for the steady<br />

state are shown in Table 2. The whole<br />

steady state calculation was rather<br />

long due to some inertia of the system.<br />

The computer model simulated<br />

behaviour during the transient of all<br />

heat structures representing the<br />

complete piping system. In the calculation<br />

some numerical instability<br />

complicated the steady state due to<br />

Accident<br />

conditions<br />

385 ºC<br />

406 ºC<br />

600 ºC<br />

35 %<br />

By pass flow 65 %<br />

Mass flow 200 kg/h<br />

Loss of Flow Accident<br />

(LOFA)<br />

Loss of Coolant Accident<br />

(LOCA)<br />

small dimensions of the component<br />

facing the deterioration flow phenomenon<br />

during the heating up process.<br />

For these reasons, the whole steady<br />

state was completed in 25,000 s,<br />

where 15,000 to 20,000 s were needed<br />

to adjust the steady state and the<br />

rest 5,000 s were used to verify the<br />

steady behaviour of the main parameters.<br />

After this period the model simulated<br />

the accident scenario – LOFA<br />

without the reactor SCRAM in order<br />

to maximize the consequences and to<br />

calculate the time to reach temperature<br />

of PE (AC) 500 °C.<br />

The scenario is described in the<br />

following steps:<br />

1. Pump stops in 1 s after the initialization<br />

event (25,001 s)<br />

2. Active channel internal electrical<br />

heaters shut down to 0 % on the<br />

nominal power in 7s (25,007 s)<br />

3. The LVR-15 SCRAM starts at 40 s<br />

when the maximum temperature<br />

in the PE rises above the 500 °C.<br />

(25,040 s)<br />

4. The whole transient is completed<br />

in 15,000 s (40,000 s), when the<br />

SCWL and LVR-15 are in the<br />

controlled cold state.<br />

The Figure 6 and Figure 7 represent<br />

the SCW maximum temperature<br />

calculated in the sample area and the<br />

outlet temperature from the active<br />

channel, while the Figure 8 shows the<br />

maximum temperature of the PE,<br />

where there is the neutron flux peak<br />

in the Boltzmann distribution.<br />

4.2 HTHL steady state and<br />

LOFA analyses<br />

The design conditions calculated for<br />

the active channel are described in<br />

Table 3. And they are mainly summarized<br />

as reported:<br />

1. Mass flow rate of 0.0105 kg/s<br />

2. Design pressure of 7 MPa<br />

3. Design electrical heater power of<br />

11.85 kW<br />

4. Cold helium temperature of 210 °C<br />

| | Fig. 6.<br />

SCWL Coolant Maximum Temperature in LOFA.<br />

| | Fig. 7.<br />

SCWL Active Channel Outlet Temperature in LOFA.<br />

Operation and New Build<br />

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Location of the Thermocouples<br />

Thermocouple<br />

| | Fig. 8.<br />

SCWL Maximum EP Temperature in LOFA.<br />

Parameter Value Unit<br />

Pressure 7 MPa<br />

Inlet Flow<br />

Temperature<br />

Max Flow<br />

Temperature<br />

Maximum AC<br />

Pressure Envelop (PE)<br />

Temperature<br />

210 ºC<br />

900 ºC<br />

450 ºC<br />

Mass flow 40 kg/h<br />

Inlet into the interpiping space of the reheater<br />

Output from the interpiping space of the reheater<br />

Entry into the test chamber<br />

Inlet to the reheater piping space<br />

Output from the reheater piping space<br />

Output from the primary side of the heat exchanger<br />

Maximum helium temperature<br />

| | Tab. 4.<br />

Thermocouples position and description.<br />

T1<br />

T2<br />

T3<br />

T4<br />

T5<br />

T6<br />

Tmax<br />

OPERATION AND NEW BUILD 225<br />

| | Tab. 3.<br />

HTHL main parameters calculated during<br />

steady state.<br />

The steady state simulation was<br />

run in null transient mode for 5,000 s<br />

and the stabilized conditions were<br />

reached after 3,500 s. The LOFA<br />

transients was characterized by an<br />

immediate safety shutdown of the<br />

reactor due to the loss of power. As a<br />

result of the SCRAM, the temperature<br />

went immediately down following the<br />

heat generated by decay gamma flux.<br />

Figure 9 shows the calculated<br />

temperature for various thermocouples<br />

positions (according to<br />

Table 4), while Figure 10 represents<br />

the maximum temperatures in the<br />

HTHL PE.<br />

| | Fig. 9.<br />

HTHL Helium temperatures during LOFA.<br />

5 Conclusions<br />

The article provides a brief introduction<br />

about the SUSEN project and the<br />

experimental facilities built in CVŘ in<br />

the Czech Republic for research and<br />

development in support of the safe,<br />

reliable and long‐term sustainable<br />

operation of existing energy facilities<br />

and in development of GIF IV and<br />

fusion technologies. The SUSEN<br />

R&D activities include four complementary<br />

programmes, mentioned in<br />

the introduction, which are focused<br />

on material science, thermal hydraulics,<br />

neutronics, radiation protection,<br />

nuclear chemistry, waste management<br />

and environmental studies. A<br />

significant part of the research programme<br />

is devoted to HTH and SCW<br />

experimental loops, which are going<br />

to be installed into the active core of<br />

the research reactor LVR-15. Both of<br />

| | Fig. 10.<br />

HTHL PE temperature during LOFA.<br />

these unique facilities are challenging<br />

to model and the selection of appropriate<br />

codes was a demanding process.<br />

A special methodology was used for<br />

assessing the abilities of the codes to<br />

simulate these advanced coolants and<br />

to obtain regulatory certificate/ permit<br />

for their use in operational and accident<br />

conditions and for preparation of<br />

the amendment of the LVR-15 FSAR.<br />

These presented activities represent<br />

only starting steps for the further<br />

codes validation which will be based<br />

on benchmarking of the codes with<br />

experimental data provided by the<br />

SCWL and HTHL loops in their<br />

experimental campaigns.<br />

Aknoledgment<br />

The authors would like to thank<br />

Mr. Miroslav Hrehor and Dr. Vincenzo<br />

Romanello for their kind revisions and<br />

suggestions.<br />

The presented work was financially<br />

supported by the Project CZ.02.1.01/<br />

0.0/0.0/15_008/0000293: Sustainable<br />

energy (SUSEN) – 2 nd phase,<br />

realized in the framework of the<br />

Operation and New Build<br />

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 226<br />

European Structural and Investment<br />

Funds.<br />

This work has been supported<br />

by the Project CZ.02.1.01/0.0/0.0/<br />

15_008/0000293: Sustainable energy<br />

(SUSEN) – 2 nd phase realized in the<br />

framework of the European Structural<br />

and Investment Funds.<br />

References<br />

[1] CVR Annual Report 2016.<br />

[2] http://susen2020.cz/<br />

[3] http://cvrez.cz/en/infrastructure/<br />

research-reactor-lvr-15<br />

[4] IAEA, Standards Safety in the Utilization<br />

and Modification of Research Reactors”,<br />

Safety Standard n° SSG-24, VIENNA,<br />

2012.<br />

[5] ATHLET 3.1A, 2016 User manual:<br />

ATHLET Mod 3.1 Cycle a, G. Lerchl,<br />

H. Austregesilo, P. Schoffel, D. von<br />

der Cron, F. Weyermann, March 2016.<br />

[6] Heat Transfer Behaviour and Thermohydraulics<br />

Code Testing for Supercritical<br />

Water Cooled Reactors (SCWRs),<br />

IAEA. http://www-pub.iaea.org/<br />

books/IAEABooks/10731/Heat-<br />

Transfer-Behaviour-and-Thermo-<br />

hydraulics-Code-Testing-for-<br />

Supercritical-Water-Cooled-R<br />

[7] TRACE V5.840 Theory Manual,<br />

U.S. Nuclear Regulatory Commission,<br />

Washington DC, March 2013.<br />

[8] TRACE V5.840 User’s Manual, Volume 1:<br />

Input Specification, U.S. Nuclear<br />

Regulatory Commission, Washington<br />

DC, February 2014.<br />

[9] TRACE V5.840 User’s Manual, Volume 2:<br />

Modelling Guidelines, U.S. Nuclear<br />

Regulatory Commission, Washington<br />

DC, February 2014.<br />

[10] G. Mazzini et al., ATHLET 3.1A<br />

SIMULATION CAPABILITIES FOR SUPER-<br />

CRITICAL STATE, CVR 1581, 1.1.2017.<br />

[11] G. Mazzini et al., ATHLET 3.1A HEAT<br />

TRANSFER ASSESMENT FOR SUPER-<br />

CRITICAL WATER, CVR 1582, 1.1.2017.<br />

[12] G. Mazzini et al., ATHLET 3.1A<br />

CAPABILITIES IN SIMULATING SWAMUP<br />

FACILITY IN SCW CONDITIONS, CVR<br />

1583, 1.1.2017.<br />

[13] M. Polidori, HE-FUS3 Benchmark<br />

Specifications, GoFastR-DEL-1.5-01,<br />

Rev. 0, ENEA, July 2011.<br />

[14] M. Polidori, HE-FUS3 Experimental<br />

Campaign for the Assessment of<br />

Thermal-Hydraulic Codes: Pre-Test<br />

Analysis and Test Specifications,<br />

Report RSE/2009/88.<br />

[15] M. Polidori et al, HE-FUS3 Benchmark<br />

Results, GoFastR-DEL-1.5-6, Rev. 0,<br />

November 2012.<br />

[16] Miloš Kynčl, Development and Assessment<br />

of TRACE HTHL-2 Facility Thermal<br />

Hydraulic Model, Internal Project Status<br />

Report, CVŘ 1334, March 2017.<br />

Authors<br />

G. MazziniM. Kyncl<br />

Alis Musa<br />

M. Ruscak<br />

Centrum Vyzkumu Rez (CVŘRez)<br />

Hlavní 130<br />

250 68 Husinec – Řež,<br />

Czech Republic<br />

Numerical Analysis of MYRRHA Interwrapper<br />

Flow Experiment at KALLA<br />

Abdalla Batta and Andreas G. Class<br />

Introduction The MYRRHA reactor, which is developed at SCK-SCN in Belgium, represents a multi-purpose<br />

irradiation facility. Its prominent feature is a pool design with the nuclear core submerged in liquid metal lead bismuth.<br />

During transients between normal operation and accident conditions decay heat removal is ensured by forced and<br />

natural convection, respectively. The flow in the gap between the fuel assemblies plays an important role in limiting<br />

maximum temperatures which should not be exceeded to avoid core damage. The term inter-wrapper flow (IWF)<br />

describes the convection in the small gap between the wrapper tubes of neighbouring fuel assemblies (FAs). It plays an<br />

important role for passive decay heat removal (DHR).<br />

Based on numerous experiments<br />

several correlations have been proposed<br />

for the flow within wirewrapped<br />

rod bundles. However, for<br />

the flow within the gap between<br />

neighbouring bundles only few<br />

studies are reported. Recently [1]<br />

reviewed the existing correlations by<br />

Rheme [2], Baxi & Dalle Donne [3]<br />

Cheng and Tordreras [4], and Kirillov<br />

[5] for the pressure-drop in wirewrapped<br />

rod bundles. The existing<br />

correlations were compared to all the<br />

available experimental data and<br />

showed that agreement of approximately<br />

±20 % can be expected. For<br />

the inter-wrapper flow within the<br />

gap only few studies exist, see [6].<br />

Due to the scarce database, within the<br />

Horizon 2020 – research and innovation<br />

framework program of the EU,<br />

the SESAME project was established<br />

to develop and validate advanced<br />

numerical approaches, to achieve a<br />

new or extended validation base and<br />

to establish best practice guidelines<br />

including verification & validation<br />

and uncertainty quantification, see<br />

[7]. In particular the current work<br />

supports the inter-wrapper flow<br />

experiment at KALLA. Three fuel<br />

assemblies including the gap flow are<br />

studied covering the full range of<br />

thermo- hydraulic conditions expected<br />

in the reactor application. For this<br />

purpose, an experimental test matrix<br />

has been established which covers<br />

relevant scenarios. The aim of our<br />

numerical pre-test study is to help the<br />

design of the experiment. The current<br />

study applied RANS-CFD methods for<br />

design support of the experiment. In<br />

the body of this compact the experiment,<br />

the corresponding numerical<br />

model, and preliminary numerical<br />

results are provided.<br />

1 Experimental setup<br />

The KALLA experiment investigates<br />

IWF between three bundles which<br />

are thermally connected by a gap.<br />

Figure 1 shows a cross-sectional view<br />

of the test section which consists of<br />

three ducts representing the fuel<br />

assemblies. Each duct contains 7 wirewrapped<br />

electrically-heated pins<br />

representing the fuel rods. The gap<br />

between the channels, i.e. assemblies,<br />

is filled with liquid metal, so<br />

that strong thermal coupling exists<br />

between neighbouring assemblies.<br />

The test matrix covers independent<br />

variation of flow and thermal conditions<br />

in both the gap and the bundles.<br />

Detailed description of the experiment<br />

is reported in [8]. The geometrical<br />

parameters of the bundle and the<br />

nomenclature are also shown in<br />

Figure 1. The experimental loop<br />

facility THESYS at KALLA and the<br />

Operation and New Build<br />

Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 1.<br />

SCWL Coolant Maximum Temperature in LOFA.<br />

| | Fig. 2.<br />

Left: Experimental loop facility THESYS at KALLA showing location where the inter wrapper flow<br />

experiment (see Figure 3) will be installed; right: flow diagram for the IWF tests with four parallel<br />

channels; the valves V2.1-V2.3 control the flow through the assemblies Q1-Q3. V.2.4 controls the<br />

flow in the gap [8].<br />

location where the IWF experiment<br />

will be installed is shown in Figure 2<br />

left. Figure 2 right shows the flow<br />

diagram of the IWF tests with four<br />

parallel channels representing the<br />

three assemblies (Q1-Q3) and the gap<br />

( illustrated by the box containing<br />

Q1-Q3). The flow and temperature<br />

within each assembly and the gap can<br />

be set individually by choosing valve<br />

openings (V2.1-V2.4) and heating<br />

rates according to the KALLA test<br />

matrix. Figure 3 shows the geometry<br />

of the IWF test section.<br />

and mesh resolution for the thermoshydraulic<br />

investigation of the gap and<br />

the bundle. In particular, we include<br />

the upstream components to verify<br />

their influence on the flow field within<br />

the test section. We employ the k-ε<br />

turbulence model and the commercial<br />

CFD-code Star CCM+. Our first<br />

studied case (i) focuses on the gap<br />

| | Fig. 3.<br />

Geometry of the IWF test section, dimensions are in mm, the heated part<br />

of the bundle is marked red on the left side of the figure, 600 mm, [8].<br />

flow and our second case (ii) on the<br />

fuel assembly. For the study of case (i)<br />

a computational domain including<br />

the lower flow distributer, riser pipe<br />

( including venture tube), upper flow<br />

vessel, and the gap are considered (for<br />

corresponding technical drawings of<br />

components refer to Figure 3). For the<br />

study of case (ii) the computational<br />

domain includes the lower flow distributer,<br />

riser pipe (including venture<br />

tube), one inlet expansion and a single<br />

7-pin bundle. Flow properties of the<br />

liquid metal Lead-Bismuth eutectic at<br />

200 °C are employed. Note that corresponding<br />

upstream pipes and flow<br />

conditioners are modelled so that<br />

all relevant geometric details are<br />

captured. Quantifying the effect of<br />

the flow conditioning sections is<br />

important for future simulations, as it<br />

would enable the use of a simpler<br />

computational domain, which still<br />

provides accurate results. In the future<br />

post-test analysis, the smallest representative<br />

computational domain (e.g.,<br />

potentially without flow conditioner<br />

etc.) will be used to compose a fully<br />

coupled thermos-hydraulic simulation<br />

of the three bundles including<br />

the IWF in the gap. Figures 4 left<br />

and right show the computational<br />

domains for the pre-test studies<br />

OPERATION AND NEW BUILD 227<br />

2 Numerical study<br />

A comprehensive analysis of the<br />

experiment requires efficient simulations.<br />

In the pre-test analysis of the<br />

hydraulics separate simulations of the<br />

gap region and the fuel assembly are<br />

performed. In a first step, we determine<br />

suitable computational domains<br />

| | Fig. 4.<br />

Computational domain for IWF-gap (left) and bundle (right) including the upstream domains.<br />

Operation and New Build<br />

Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 228<br />

of cases (i) and (ii), respectively.<br />

Obviously, a substantial effort was<br />

undertaken to include the upstream<br />

flow domain, so that the inflow into<br />

the fuel assembly and the gap are<br />

properly represented in the flow<br />

simulations.<br />

Since we have less experience with<br />

the gap region, and in particular, the<br />

applicable turbulence regime we have<br />

considered 3 cases corresponding to<br />

laminar flow, transitional flow, and<br />

fully developed turbulence, respectively.<br />

This covers the flow range 0.17<br />

to 0.86 kg/s (Re = 1,250 to 6,250),<br />

proposed by the test matrix. For the<br />

investigation of the fuel assembly, we<br />

consider the nominal flow rate, i.e. the<br />

maximum flow rate planned in the<br />

test matrix. This corresponds to a flow<br />

rate of 3.58 kg/s and Re = 8,910<br />

where Re is based on the bundle<br />

hydraulic diameter. All cases considered<br />

in the experimental test matrix<br />

are within the range of transitional<br />

flow according to Cheng and Todreas<br />

[4] (see next section on correlations)<br />

so that no distinction of various flow<br />

regimes is needed for the comparison<br />

to correlations.<br />

2.1 Inter-wrapper flow gap<br />

region<br />

The objective of case study (i) is to<br />

investigate the effects of all upstream<br />

components on the flow distribution<br />

entering the gap, i.e. the inter wrapper<br />

flow region. This study employs the<br />

computational domain shown in<br />

Figure 4 left. For the simulation, a<br />

mesh with approximately 0.72 million<br />

cells has been generated. The investigated<br />

range of flow rates results in<br />

turbulent flow in all components<br />

upstream of the gap, since the<br />

Reynolds- numbers based on pipe<br />

diameter varies between 5,200 and<br />

26,000. However, within the gap<br />

the Reynolds-number based on gapwidth<br />

equates to 1,250 to 6,250 corresponding<br />

to the transitional regime of<br />

turbulence. The pressure drop along<br />

the gap accounts for about 20 % of the<br />

total pressure drop. Since we are<br />

interested in accurately predicting the<br />

upstream flow in the gap region the<br />

use of a turbulent model is mandatory.<br />

Moreover, in order to judge the uniformity<br />

of the flow entering the gap<br />

there is no need to use a very accurate<br />

result within the gap. Thus, a high-<br />

Reynolds-number turbulence model<br />

using automatic wall functions is<br />

used. Figure 5 shows the velocity<br />

vectors in the gap entrance region for<br />

the case where the Reynolds-number<br />

is 5,200 based on pipe diameter<br />

(Reynolds-number is 1,250 based on<br />

the gap hydraulic diameter). We<br />

observe that the flow within the gap<br />

becomes near uniform after a short<br />

length, which does not exceed 10 %<br />

of the length of the gap region. The<br />

heated zone starts further downstream<br />

approximately at half the<br />

length of the gap region. For higher<br />

Reynolds-number a qualitative similar<br />

result is obtained. For future simulations<br />

aiming at accurately simulating<br />

the temperature field, we conclude<br />

that the effect of upstream components<br />

is negligible. In Table 1 the<br />

pressure drop across the simulated<br />

region is compared to design values<br />

for three selected cases covering<br />

the full range of flow rates. Design<br />

values are calculated using lumped<br />

parameter models. Both results agree<br />

reasonably well, indicating that<br />

lumped parameter models well<br />

describe the flow in the gap.<br />

2.2 Flow within a single<br />

wire-wrapped rod bundle<br />

As in the previous study, we aim at<br />

investigating whether the upstream<br />

region that conditions the flow<br />

entering the wire-wrapped bundle<br />

influences the flow in the heated<br />

section of the bundle. Here, i.e. in case<br />

study (ii), a single Reynolds-number<br />

of 8,900 based on the bundle hydraulic<br />

diameter is considered. This<br />

corresponds to the nominal flow<br />

rate as well as the maximum flow<br />

rate intended in the experimental<br />

tests. The computational domain of<br />

Figure 4 right uses approximately<br />

1 million cells. Figure 6 shows the<br />

velocity magnitude within the bundle.<br />

At the entrance, we still observe pronounced<br />

non-uniformities of the flow<br />

distribution. These quickly equilibrate<br />

so that a more-uniformly distributed<br />

flow is observed well before the<br />

heated section of the bundle is<br />

reached (for the location of the heated<br />

region refer to Figure 3).<br />

This result suggests that inflow<br />

effects are negligible for the intended<br />

thermal analysis of the bundle. Thus<br />

in a second simulation we remove the<br />

flow-conditioning region to reduce<br />

the size of the considered flow<br />

domain. To validate our simulation<br />

results we use higher mesh resolution<br />

within the smaller domain. Figure 7<br />

shows the pressure along two selected<br />

axial lines, which are depicted in the<br />

small inset. The influence of the wirewrap<br />

manifests in the periodical<br />

modulation of the pressure profile.<br />

Obviously, development effects have<br />

decayed at a length of approximately<br />

100 mm. We compute the pressure<br />

| | Fig. 5.<br />

Velocity vectors within the gap upstream region.<br />

Flow rate<br />

[kg/s]<br />

design Δp tot ,<br />

[Pa]<br />

CFD Δp tot , [Pa]<br />

1. 0.86 15350 13500<br />

2. 0.688 9964 -<br />

3. 0.516 5667 5500<br />

4. 0.344 2586 -<br />

5. 0.172 663 850<br />

| | Tab. 1.<br />

Comparison of design values evaluated by lumped parameter model<br />

versus computed pressure drop across the test section including the flowconditioning<br />

components.<br />

| | Fig. 6.<br />

Velocity magnitude within bundle showing non-uniformities of flow distribution at leftmost plane and<br />

more-uniformly distributed flow in subsequent planes.<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 7.<br />

Pressure along two selected axial lines in the wire-wrapped rod bundle.<br />

The inset specifies location of lines.<br />

drop using data at corresponding<br />

wire-wrap positions, i.e. from axial<br />

positions 0.065 m to 1.268 m. The<br />

mean pressure drop is 946 Pa/m.<br />

2.3 Model validation<br />

In this subsection, results of our<br />

numerical study are compared to the<br />

simplified Cheng and Todreas [1986]<br />

correlation. The correlation was<br />

recently recommended in (1) to<br />

predict pressure drop (Δp) in bundles<br />

with an accuracy of ±20 %. It applies<br />

for a wide range of Reynolds- numbers.<br />

The friction factor (f) is defined in<br />

eq. 1, where d h,bdl , L, and u b<br />

2<br />

are<br />

hydraulic diameter, length, and<br />

average axial bundle velocity, respectively.<br />

(1)<br />

The correlation for f reads<br />

for Re < Re L<br />

for Re L ≤ Re ≤ Re T<br />

for Re > Re T (2)<br />

where<br />

Re L = 300 x 10 1.7(P/D−1.0) (3)<br />

Re T = 10,000 x 10 0.7(P/D−1.0) (4)<br />

ψ = log(Re/Re L ) / log(Re T /Re L ) (5)<br />

C fL = (-974.6 + 1612.0(P/D) −<br />

598.5(P/D) 2 )(H/D) .06-0.085(P/D)<br />

(6)<br />

C fT = (0.8063 − 0.9022(log(H/D)) +<br />

0.3526(log(H/D)) 2 ) ×<br />

(P/D) 9.7 (H/D) 1.78-2.0(P/D) (7)<br />

We compare the nominal flow case<br />

of 3.580 kg/s which corresponds to a<br />

velocity of 0.2 m/s and Re is 8910,<br />

which is in the transient region.<br />

According to eqns (3) and (4), Re L and<br />

Re T are 902 and 15735, respectively.<br />

The calculated friction factor f<br />

equates to 0.0557. This corresponds<br />

to a pressure drop in the bundle of<br />

1407.2 Pa. The predicted pressure<br />

drop resulting from the CFD study is<br />

1,138 Pa. The difference is near 19 %,<br />

which lays within the accuracy limits.<br />

In future thermos-hydraulic simulations,<br />

the current model can be<br />

applied. For posttest analysis, additional<br />

sensitive studies might be<br />

necessary to further reduce the<br />

uncertainty.<br />

Conclusions<br />

The flow in the gap between neighbouring<br />

fuel assemblies plays an<br />

important role in transients between<br />

forced and natural convection. At<br />

KALLA an experiment on the interwrapper<br />

flow is currently setup and<br />

accompanied by pre-test numerical<br />

CFD studies. These proof that both<br />

the flow in the gap region and the<br />

fuel bundle are not influenced by the<br />

upstream flow-conditioning region.<br />

Moreover, development length are<br />

much shorter than the unheated<br />

length of the test section, so that<br />

the thermal field is uninfluenced by<br />

flow non-uniformities. Preliminary<br />

comparison of pressure losses computed<br />

by CFD and correlation provide<br />

reasonable agreement for both the<br />

gap and bundle. The result of our<br />

study enters pre-test studies of the<br />

thermal field within the EU-H2020<br />

SESAME project. There complete<br />

simulation of the test section consisting<br />

of three bundles connected<br />

by the gap region including conjugate<br />

heat transfer is performed.<br />

Acknowledgement<br />

This project has received funding from<br />

the Euratom research and training<br />

programme 2014-<strong>2018</strong> under grant<br />

agreement No 654935 and from the<br />

AREVA Nuclear Professional School.<br />

References:<br />

[1] Chen, S.; Todreas, N.; Nguyan, N.<br />

(2014). Evaluation of existing correlations<br />

for the prediction of pressure drop<br />

in wire-wrapped hexagonal array pin<br />

bundles. Nuclear Engineering and<br />

Design 267, pp. 109 – 131<br />

[2] Rehme, K. (1973). Pressure drop<br />

correla tions for fuel element spacers.<br />

Nuclear Technology 17, 15–23.<br />

[3] Baxi, C.B., Dalle Donne, M., (1981).<br />

Helium cooled systems, the gas cooled<br />

fast breeder reactor. In: Fenech, H. (Ed.),<br />

Heat Transfer and Fluid Flow in Nuclear<br />

Systems. Pergamon Press Inc.,<br />

pp. 410–462.<br />

[4] Cheng, S.-K.; Todreas, N. (1986). Hydrodynamic<br />

models and correlations for<br />

bare and wire-wrapped hexagonal rod<br />

bundles - Bundle friction factors,<br />

subchannel friction factors and mixing<br />

parameters. Nuclear Engineering and<br />

Design 92 (2), 227 – 251.<br />

[5] Kirillov, P.L., Bobkov, V.P., Zhukov, A.V.,<br />

Yuriev, Y.S., (2010). Handbook on<br />

Thermo hydraulic Calculations in<br />

Nuclear Engineering. Thermohydraulic<br />

Processes in Nuclear Power Facilities,<br />

vol. 1. Energoatomizdat, Moscow.<br />

[6] Kamide, H.; Hayashi, K.; Toda, S. (1998).<br />

An experimental study of intersubassembly<br />

heat transfer during<br />

natural circulation decay heat removal<br />

in fast breeder reactors. Nuclear<br />

Engineering and Design 183, 97 – 106.<br />

[7] http://sesame-h2020.eu/<br />

[8] Pacio, J, et. al. (2016), Deliverable 2.10 –<br />

KALLA Inter- wrapper flow setup for<br />

SESAME (thermal hydraulics Simulations<br />

and Experiments for the Safety<br />

Assessment of MEtal cooled reactors)<br />

project, activity: NFRP-01-2014<br />

Improved safety design and operation<br />

of fission reactors, H2020 Grant<br />

Agreement Number: 654935.<br />

Authors<br />

Abdalla Batta<br />

Andreas G. Class<br />

AREVA Nuclear Professional School<br />

Karlsruhe Institute of Technology<br />

Karlsruhe, Germany<br />

OPERATION AND NEW BUILD 229<br />

Operation and New Build<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 230<br />

Heat Balance Analysis for Energy<br />

Conversion Systems of VHTR<br />

SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park<br />

VHTR (Very High Temperature gas Reactor) which helium is used as a coolant can easily produce heat required in<br />

high-temperature thermochemical process, and because of low heat output density, the possibility of core melting is<br />

low. Helium has the advantage of safety, and the coolant can become super high temperature, thereby power production<br />

as well as hydrogen production application is possible. In this study, provided that VHTR is located in the primary<br />

system, the heat conversion system will be discussed in which hydrogen production and power supply are possible. In<br />

order to control the ratio between power and hydrogen production, the helium flowing through nuclear reactor is made<br />

to pass through heat exchanger for hydrogen production and steam generator or heat exchanger. Power production was<br />

made to be composed of ultra-super critical steam cycle (USC) and supercritical CO 2 (SCO 2 ) cycle so that efficient<br />

operation condition can be selected. This study proposed the whole heat conversion system model, and carried out<br />

thermodynamic feasibility calculation according to major design variable at each point and sensitivity analysis for<br />

efficiency optimization.<br />

1 Introduction<br />

Recently, an interest on hydrogen as a<br />

clean energy source and a fossil fuel<br />

substitute has been increasing. From<br />

the viewpoint that hydrogen utilizes<br />

the energy system which uses the<br />

existing fossil fuel without the<br />

emission of environmental pollution<br />

material, contrary to fossil fuels,<br />

hydrogen is emerging as a promising<br />

future clean energy. Among hydrogen<br />

production methods, high-temperature<br />

pyrolysis hydrogen production<br />

method using heat chemical process is<br />

considered as a proper method for<br />

mass hydrogen production. Heat is<br />

required much for high-temperature<br />

heat chemical process, and lightwater<br />

reactor that uses water as coolant<br />

does not produce heat required<br />

for high-temperature heat chemical<br />

process. VHTR (Very High Temperature<br />

gas Reactor) which uses helium<br />

as coolant can easily produce heat<br />

required for high-temperature thermochemical<br />

process, so recently the<br />

study of the use of high temperature<br />

gas for hydrogen production has been<br />

the research trend [1, 2].<br />

VHTR has no possibility of core<br />

melting due to low heat output<br />

density, and it does not use water, so<br />

there is no risk of explosion danger<br />

due to hydrogen generation in the<br />

case of coolant loss accident. Besides,<br />

it has the advantage that high-temperature<br />

coolant can be made compared<br />

to water-cooled reactor, so it has the<br />

advantage of power production and<br />

process heat supply [3]. Nuclear<br />

reactor is in charge of heat supply, and<br />

this can be converted variously to be<br />

used as the production of hydrogen or<br />

power. In this study, by borrowing<br />

general name in the atomic power<br />

field, VHTR is called as a primary<br />

system, the part which hydrogen production<br />

and power supply are possible<br />

through heat conversion, is defined as<br />

the secondary system. Helium flowing<br />

in nuclear reactor delivers the heat of<br />

the primary system to the secondary<br />

system through HX (Heat Exchanger).<br />

Helium flowing through the secondary<br />

system passes first through heat<br />

exchanger where hydrogen production<br />

occurs, and secondly and thirdly<br />

passes through steam generator and<br />

heat exchanger composed of ultrasuper<br />

critical cycle (Ultra- supercritical<br />

steam cycle: USC) and super critical<br />

carbon dioxide (Supercritical CO 2 :<br />

SCO 2 ) cycle, respectively, producing<br />

process heat and power. In this study,<br />

the authors proposed the overall heat<br />

conversion system model, and performed<br />

the thermodynamic feasibility<br />

calculation in accordance with major<br />

design variable at each point and<br />

sensitivity analysis for efficiency<br />

optimization.<br />

2 Research methodology<br />

2.1 Concept and methodology<br />

of hydrogen production<br />

equipment<br />

As a method of hydrogen production<br />

which uses water as a raw material by<br />

using 900 °C heat, high temperature<br />

electrolysis using heat energy simultaneously<br />

and the mixed method of<br />

using thermochemistry process method<br />

and electrolytic method. Recently,<br />

research has been focused on Sulfur-<br />

Iodine thermochemical cycle where<br />

iodide and sulfuric acid were used to<br />

break down water. This is because the<br />

required equipment can be scaled up<br />

and process handling material is only<br />

composed of gas and liquid so that<br />

continuous operation is possible.<br />

Besides, it is advantageous to use<br />

nuclear reactor where the safety of<br />

load change is demanded as heat<br />

source [2].<br />

In the hydrogen production equipment<br />

where high temperature heat is<br />

used, according to the Reaction 1<br />

below, sulfuric acid (H 2 SO 4 ) can be<br />

broken down into water vapor<br />

(H 2 O(g)), oxygen (O 2 (g)), and sulfur<br />

dioxide (SO 2 (g)).<br />

Reaction 1:<br />

2H 2 SO 4 + Heat 2H 2 O + 2SO 2 + O 2<br />

After decomposition, oxygen(O 2 (g))<br />

is removed, and water vapor(H 2 O(g))<br />

and sulfur dioxide (SO 2 (g)) are cooled<br />

down, reacting with iodide (I).<br />

According to Reaction 2 below,<br />

sulfuric acid (H 2 SO 4 ) and hydrogen<br />

iodide (HI) are formed.<br />

Reaction 2:<br />

4H 2 O + 2SO 2 + 2I 2 2H 2 SO 4 + 4HI<br />

+ Heat<br />

Finally, by using high temperature<br />

heat, hydrogen Iodide (HI) can be<br />

separated into hydrogen (H 2 ) and<br />

iodide (I) according to the reaction 3<br />

below.<br />

Reaction 3: 4HI + heat 2I 2 + 2H 2<br />

2.2 The concept and status<br />

of USC and S-CO 2 cycle<br />

USC power plant means the power<br />

plant where vapor pressure is 254 kg/<br />

cm 2 or higher, and main vapor’s<br />

or reheated vapor’s temperature is<br />

593 °C or higher. The reasons why<br />

pressure and temperature of the<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

evaporator are elevated are that the<br />

efficiency of power plant is increased.<br />

When the temperature of main evaporator<br />

and reheating steam increases<br />

by 10 °C, the efficiency increases<br />

by 0.5 %; and pressure increases by<br />

10 kg/cm 2 , the efficiency increases by<br />

about 0.2 %. Domestically, in 1990’s,<br />

500 MW-grade standard coal thermal<br />

power plant was designed and built,<br />

and its operation condition was pressure<br />

246 kg/cm 2 and temperature<br />

538 °C.<br />

In the case of Dangjin Thermal<br />

Power No. 9, No. 10 and Samcheok<br />

Thermal Power No. 1, No .2 that have<br />

been being built, the pressure of<br />

250 kg/cm 2 , temperature of 600 °C<br />

were accomplished [4].<br />

SCO 2 cycle is the power generation<br />

technology of the Gas Brayton Cycle<br />

method where pressurized carbon<br />

dioxide is heated by the pressure<br />

greater than critical condition to high<br />

temperature and turbine is driven.<br />

Presently, CO 2 power generation cycle<br />

can be applied to most heat sources<br />

used, and also it can be used for large<br />

power plant, small scale distribution<br />

power supply, or power supply for<br />

marine plant.<br />

Super critical condition means the<br />

conditions for temperature and pressure<br />

greater than critical point in the<br />

general material state where liquid-gas<br />

phase change occurs, and the<br />

temperature and pressure at the lower<br />

pressure part is greater than 32 °C, 74<br />

atm, and all parts of cycle are maintained<br />

over critical condition. While<br />

operation is carried out at high<br />

pressure, volumetric flow decreases,<br />

so the size of overall heat conversion<br />

cycle can be decreased; accordingly,<br />

construction period and production<br />

unit price can be lowered to secure<br />

high economic feasibility.<br />

Besides, compared to water vapor,<br />

the compatibility with existing material<br />

is excellent, so it can be supplied<br />

to turbine at the temperature higher<br />

than evaporator cycle. From this, the<br />

increase of additional power generation<br />

efficiency can be possible [5].<br />

2.3 Heat Conversion Model<br />

Design<br />

IHX loop of VHTR that is studied in<br />

the present study is the system where<br />

the high temperature heat generated<br />

in the reactor by connecting hydrogen<br />

generation equipment and power<br />

generation equipment in series can be<br />

supplied in the same manner.<br />

IHX loop nuclear reactor shown in<br />

Figure 1 provides 350 MWt heat output,<br />

and the heat generated from<br />

| | Fig. 1.<br />

IHX Loop Modelling.<br />

nuclear fission is supplied to helium<br />

fluid. For heat transfer to produce<br />

hydrogen, heat exchanger, steam generator<br />

for the power generation via<br />

USC cycle, and in the power generation<br />

via SCO 2 , one heat exchanger is<br />

provided. In order to utilize the result<br />

of the study regarding the existing<br />

VHTR, the major principle and<br />

variable if heat conversion model<br />

were set as follows. Temperature and<br />

pressure at No. 1, 2, 3, 4, 10 were<br />

presumed by reference literature [8].<br />

Temperature and pressure of ultrasuper<br />

critical cycle No. 5, 6 and SCO 2<br />

cycle, No. 8 were assumed by using<br />

reference literature [9]. The model to<br />

be explained below was defined as<br />

reference model, and then the present<br />

authors will plan to develop a model<br />

that considers a variety of heat<br />

efficiency improvement method. In<br />

the present study, in the concept<br />

similar to general Rankine cycle’s<br />

reheating cycle, bypass mode was<br />

proposed.<br />

To begin with, the reference model<br />

is as follows. After 910 °C helium fluid<br />

discharging from VHTR carries out<br />

heat exchange with heat exchanger 1,<br />

hydrogen is produced by receiving<br />

heat from high temperature helium<br />

fluid in the heat exchanger 1. 846 °C<br />

helium fluid passing heat exchanger 1<br />

enters into steam generator 2 and go<br />

through heat exchange. The fluid of<br />

this steam generator is ultra-super<br />

critical state water, and produces<br />

power. The temperature of helium<br />

fluid that passes through steam<br />

generator 2 is 614.8 °C, this helium<br />

fluid enters into heat exchanger 3<br />

where heat exchange is carried out.<br />

The fluid of this heat exchanger is<br />

super critical-state carbon dioxide,<br />

and it produces power by the heat<br />

supplied. The temperature of helium<br />

fluid coming out of heat exchanger 3<br />

is 450 °C. The heat output that is produced<br />

in heat exchanger 1 producing<br />

hydrogen is 37.37 MWt. The mass flow<br />

of helium from IHX is m 1 , and the<br />

mass flow of water flowing in heat<br />

exchanger 1 is m 2 , the mass flow of<br />

water flowing in steam generator 2 is<br />

m 3 , and the mass flow of CO 2 flowing<br />

in heat exchanger 3 is m 4 . In this<br />

study, the temperatures and pressures<br />

from No.1 to No.10 in Figure 1 were<br />

assumed, and m 1 and m 2 were calculated<br />

by using the Equation (1), and<br />

m 3 and m 4 were calculated by using<br />

the Equation (2). Besides, considering<br />

the characteristics of general longitudinal<br />

temperature difference of heat<br />

exchanger, the temperature at No. 6<br />

and No. 9 was assumed to decrease by<br />

10 °C compared to the temperature at<br />

No. 4 and No. 7 of the steam generator<br />

inlet.<br />

Major equation or relationship for<br />

heat equilibrium analysis is as follows:<br />

• Equation used for calculating m 1<br />

and m 2<br />

: W = m∆h = m(h out – h in )... (1)<br />

Here,W : Thermal power (MWt)<br />

m : Mass flow (kg/hr)<br />

h : Enthalpy (kJ/kg)<br />

in : Entrance of the equipment<br />

out : Outlet of equipment<br />

• Equation used for calculating m 5<br />

and m 8<br />

∑m in h in = ∑m out h out ... (2)<br />

In the case of hydrogen production, it<br />

was assumed that all heat was<br />

converted to work required, and in<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 232<br />

the case of power production, it was<br />

assumed that only a part of the heat<br />

delivered was converted to electricity.<br />

Besides, it was considered that the<br />

pumping power was consumed due to<br />

the flow in the power generation,<br />

therefore it was considered in the<br />

calculation of efficiency.<br />

The general efficiency of USC cycle<br />

and SCO 2 cycle was 43 % and 45 %,<br />

respectively. Using Equation (3),<br />

efficiency was corrected, and more<br />

realistic calculation was carried out<br />

[6].<br />

• efficiency correction equation<br />

η Oper = [1.0+{(T h,oper – T h,des ) × C}]<br />

× η Des … (3)<br />

Here, η Des : standard efficiency<br />

according to reference<br />

literature<br />

η Oper : Standard efficiency’s<br />

correction efficiency according<br />

to high temperature<br />

T h,des : Standard exit temperature in<br />

Steam Generator tube according<br />

to reference literature<br />

T h,oper : Exit temperature within<br />

specified range at Steam<br />

Generator or Heat exchanger<br />

tube<br />

C : Efficiency correction factor;<br />

USC : 0.3 % / 5 °C [6],<br />

SCO 2 : 1.0 % / 5 °C applied<br />

(assumption)<br />

The output of steam generator 2 and<br />

heat exchanger 3 is as follows, and<br />

total output W gross is the sum of all the<br />

values.<br />

W 1 = m 2 × (h 3 – h 2 ) kJ/hr<br />

W 2 = η 2,Operator × m 3 ×(h 4 – h 7 ) kJ/hr<br />

W 3 = η 3,Operator × m 4 ×(h 7 – h 10 ) kJ/hr<br />

Here, w pump : Work used<br />

in the pump (MWt)<br />

η pump : Pump efficiency<br />

v : Specific volume (m 3 /kg)<br />

P : pressure (kPa)<br />

In order to simulate the above model,<br />

the flow of helium gas, water, and<br />

carbon dioxide was calculated by<br />

using thermodynamic system analysis<br />

software, EES (Engineering Equation<br />

Solver).<br />

The following is regarding IHX<br />

loop model to which Bypass mode was<br />

added. Bypass mode was added to the<br />

existing IHX loop, and the efficiency<br />

improvement of overall heat conversion<br />

cycle was studied. The temperature<br />

of the entrance of evaporator 2<br />

and heat exchanger 3 was reheated<br />

by using high temperature helium<br />

coming from IHX, and the output<br />

change was studied.<br />

In the same way as the existing IHX<br />

loop, VHTR supplies heat generated<br />

by nuclear fission in 350 MWt nuclear<br />

reactor. The fluid coming from IHX is<br />

helium, and the fluid flowing in heat<br />

exchanger 1 is water, the fluid flowing<br />

in steam generator 2 is ultra-supercritical-state<br />

water, and the fluid<br />

flowing in heat exchanger 3 is super<br />

critical-state carbon dioxide.<br />

The mass flow of helium from<br />

IHX is m 1 , and m 1 is divided into m 2<br />

and m 3 , and m 3 enters into heat<br />

exchanger 1, and do heat exchange<br />

with the water flowing in heat exchanger<br />

1. At this time, the mass flow<br />

of water flowing in heat exchanger 1<br />

is m 5 . and m 2 is divided into m 4 and<br />

m 9 , and m 4 enters into No. 7 in order<br />

to reheat helium that went through<br />

heat exchange in heat exchanger 1,<br />

and the reheated temperature is that<br />

of No. 8. m 9 enters No. 12 in order to<br />

reheat helium that went through heat<br />

exchange in the steam generator 2,<br />

and the reheated temperature is the<br />

temperature of No. 13. The mass flow<br />

in steam generator 2 is m 10 , and mass<br />

flow of CO 2 in heat exchanger 3 flowing<br />

through heat exchanger 3 is m 14 .<br />

The temperature and pressure at<br />

every point except No. 12 were assumed,<br />

and m 1 value was obtained by<br />

using Equation (1) in the same as m 1<br />

of the existing IHX Loop.<br />

At this time, the temperature of<br />

No.7 and No.12 were assumed to be<br />

that of No. 4 and No. 7 of the existing<br />

IHX loop. Besides, it was assumed that<br />

the temperatures of No. 8 and No. 13<br />

increased to 860 °C and 620 °C,<br />

respectively due to m 4 and m 9 . By<br />

this, the change of the output and<br />

efficiency on the cycle of steam generator<br />

2 and heat exchanger 3. Heat<br />

exchanger 1 in accordance with the<br />

addition of Bypass mode was assumed<br />

to produce the same output, 37.37<br />

MWt, as the existing IHX loop, and<br />

fixed m 5 value.<br />

Considering the characteristics of<br />

the general longitudinal temperature<br />

difference of the heat exchanger as<br />

the existing IHX loop, it was assumed<br />

that the temperatures of No. 11 and<br />

No. 15 decrease by 10 °C compared to<br />

that of No. 8 (at Steam Generator<br />

entrance) and No. 13 (at Heat<br />

Exchanger). With the obtained m 1<br />

value, m 14 value was calculated by<br />

using Equation (2). After that, m 5<br />

value was fixed to the value which can<br />

make the output as the existing IHX<br />

loop, then m 3 value was obtained<br />

by Equation (2). m 2 was calculated by<br />

m 2 = m 1 – m 3 , and m 4 was obtained by<br />

using Equation (2). m 9 was obtained<br />

Here, W 1 : heat exchanger 1<br />

heat output (MWt)<br />

W 2 : Steam generator 2<br />

heat output (MWt)<br />

W 3 : Heat exchanger 3<br />

heat output (MWt)<br />

In steam generator 2 and heat<br />

exchanger 3 in order to consider<br />

pumping power in accordance with<br />

mass flow, pump’s efficiency (η pump )<br />

was assumed to be 0.9, and Equation<br />

(4) was used.<br />

w pump = η pump × m × v out<br />

× (P out – P in ) kJ/hr … (4)<br />

W net = W gross – w pump<br />

| | Fig. 2.<br />

Bypass mode-added IHX loop Modelling.<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

by m 9 = m 2 – m 4 , and the temperature<br />

of No. 12 can be calculated by using<br />

Equation (2). Finally, m 10 was also<br />

obtained by using Equation (2). Like<br />

the existing IHX loop, in the bypass<br />

mode-added IHX loop, correction<br />

efficiency and pumping power in<br />

accordance with mass flow were<br />

considered, and pumping power used<br />

the above Equation (4). For the<br />

simulation for this, thermodynamic<br />

system analysis software, EES, was<br />

used, in the same way with the<br />

existing IHX loop model obtained<br />

before, and the flow of helium gas<br />

and fluid was analyzed.<br />

3 Result<br />

Table 1 shows the result of physical<br />

value at each point by simulating the<br />

existing IHX loop EES [5]. Physical<br />

value of each point was assumed in<br />

accordance with reference [8], [9]<br />

literature, and the assumed values<br />

were colored.<br />

The temperature of helium fluid<br />

that leaves from the first heat exchanger<br />

after producing the hydrogen<br />

decreases to 846 °C from 910 °C, and<br />

the temperature of helium fluid that<br />

leaves from the second steam generator<br />

is 614.8 °C, and the temperature<br />

of helium fluid that leaves from the<br />

last heat exchanger is 450 °C. The<br />

temperature of helium fluid decreases<br />

steadily, but because the fluid flowing<br />

each steam generator and heat<br />

exchanger is different, efficient electricity<br />

can be produced by using each<br />

characteristics. The existing IHX<br />

Loop’s m 5 and m 8 are in inverse<br />

proportion, as more mass flow moves<br />

toward high efficiency, the amount of<br />

overall electricity output increases.<br />

Although the efficiency of heat<br />

conversion cycle connected to each<br />

steam generator may be influenced by<br />

various causes, but in the present<br />

study, correction factor presumed<br />

about high temperature was used, so<br />

the detailed design for this part would<br />

be needed.<br />

If heat conversion cycle connected<br />

to each steam generator should be<br />

operated simultaneously by a specific<br />

objective, considering the inverse proportion<br />

relationship between m 5 and<br />

m 8 , the output must be distributed.<br />

Besides, the exit temperature at the<br />

tube part of steam generator 2 is in<br />

inverse proportion with the exit<br />

temperature at the shell part. This will<br />

eventually influence on the exit<br />

temperature of the tube part of the<br />

heat exchanger 3. When operating<br />

heat conversion cycle connected to<br />

each steam generator, it is necessary<br />

to find balanced point on the temperature<br />

between steam generators.<br />

In the steam generator 2 and heat<br />

exchanger 3, exit temperature and<br />

mass flow are in inverse proportion.<br />

This is because if high exit enthalpy is<br />

maintained in order to deliver the<br />

same heat energy, less mass flow is<br />

needed, and if a large amount of mass<br />

flow is needed, exit enthalpy should<br />

be maintained low. Maximum output<br />

would be in the parabolic form as exit<br />

enthalpy and temperature change, so<br />

if maximum output is needed, proper<br />

exit temperature must be selected. Or<br />

in case there is a requirement for exit<br />

temperature, it is possible that output<br />

would be determined according to<br />

that.<br />

Table 2 shows the physical value at<br />

each point where IHX loop added by<br />

bypass mode is simulated with EES.<br />

No. Fluid Temperature<br />

(°C)<br />

| | Tab. 1.<br />

IHX loop Simulation Result.<br />

In the case of IHX loop to which<br />

bypass mode was added, the helium<br />

fluid that passed through the first<br />

hydrogen-producing heat exchanger<br />

is 910 °C~ 846 °C, which is the same<br />

as the existing IHX loop, but here by<br />

reheating high-temperature helium<br />

fluid, the temperature increases to<br />

860 °C. The temperature of the helium<br />

fluid that passed through the second<br />

steam generator is 614.8 °C, which<br />

is the same as that of helium fluid<br />

that passed through the second evaporator.<br />

However, since the temperature at<br />

the entrance reheated, and returned,<br />

the amount of electricity output produced<br />

increases. When helium fluid<br />

enters the third heat exchanger, it is<br />

reheated from 614.8 °C to 620 °C, the<br />

temperature of helium fluid is 450 °C,<br />

and the amount of electricity output<br />

Pressure<br />

(kPa)<br />

| | Tab. 2.<br />

Result of IHX Loop to which Bypass Mode was added.<br />

Enthalpy<br />

(kJ/kg)<br />

Mass flow<br />

(kg/hr)<br />

1 Helium 910.0 4000 6,161.00 527,662<br />

2 Water 193.0 18,000 828.70 49,839<br />

3 Water 585.0 16,500 3,528.00 49,839<br />

4 Helium 846.0 4,000 5,829.00 527,662<br />

5 Water 260.2 20,790 1,134.00 208,359<br />

6 Water 836.0 16,475 4,174.00 208,359<br />

7 Helium 614.8 4,000 4,628.00 527,662<br />

8 CO 2 203.5 19,760 96.59 902,043<br />

9 CO 2 604.8 19,290 597.00 902,043<br />

10 Helium 450.0 4,000 3,773.00 527,662<br />

No. Fluid Temperature<br />

(°C)<br />

Pressure<br />

(kPa)<br />

Enthalpy<br />

(kJ/kg)<br />

Mass flow<br />

(kg/hr)<br />

1 Helium 910.0 4,000 6,161.00 527,662<br />

2 Helium 910.0 4,000 6,161.00 122,749<br />

3 Helium 910.0 4,000 6,161.00 404,913<br />

4 Helium 910.0 4,000 6,161.00 113,375<br />

5 Water 195.0 18,000 837.50 50,000<br />

6 Water 585.0 16,500 3,528.00 50,000<br />

7 Helium 846.0 4,000 5,829.00 404,913<br />

8 Helium 860.0 4,000 5,901.00 518,288<br />

9 Helium 910.0 4,000 6,161.00 9,374<br />

10 Water 260.2 20,790 1,134.00 214,580<br />

11 Water 850.0 16,475 4,209.00 214,580<br />

12 Helium 614.8 4,000 4,628.00 518,288<br />

13 Helium 620.0 4,000 4,655.00 527,662<br />

14 SCO 2 203.5 19,760 96.59 918,581<br />

15 SCO 2 610.0 19,290 603.50 918,581<br />

16 Helium 450.0 4,000 3,773.00 527,662<br />

OPERATION AND NEW BUILD 233<br />

Operation and New Build<br />

Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

OPERATION AND NEW BUILD 234<br />

increases compared to the existing<br />

IHX loop.<br />

The following is the major comparison<br />

of the result of reference<br />

model and bypass mode model.<br />

Mass flow IHX loop Bypass Mode Loop<br />

Heat Exchanger 1 49839 50000<br />

Steam Generator 2 208359 214580<br />

Heat Exchanger 3 902043 918581<br />

| | Tab. 3.<br />

IHX Loop and Bypass Mode IHX Loop: Mass Flow Comparison.<br />

Pumping power<br />

When IHX loop and bypass modeadded<br />

IHX loop were compared, mass<br />

flow of m 10 and m 14 in the bypass<br />

mode-added loop was greater compared<br />

to the mass flow of m 5 and m 8 in<br />

the existing loop, as shown in Table 3.<br />

IHX loop<br />

(MWt)<br />

| | Tab. 4.<br />

IHX Loop and Bypass Mode IHX loop: Pumping Power Comparison.<br />

As shown in Table 4, depending on<br />

the difference of mass flow value,<br />

pumping power used in the pump also<br />

can be high in steam generator 2 and<br />

heat exchanger 3. However, although<br />

pumping power is higher in the bypass<br />

IHX loop, by reheating, efficiency of<br />

steam generator 2 increased from<br />

53.79 % to 54.4 %, and that of heat<br />

exchanger 3 increased from 45.83 %<br />

to 45.87 %; accordingly, it is seen that<br />

the value of Power increased. As a<br />

result, Net Power that considered<br />

pumping power in Total Power was<br />

178.6 MWt in the present IHX<br />

loop, but the IHX loop to which<br />

bypass mode was added increased<br />

to 185.3 MWt, as shown in Table 5.<br />

| | Tab. 5.<br />

IHX loop and Bypass mode IHX loop: Power Comparison.<br />

Bypass mode loop<br />

(MWt)<br />

Steam Generator 2 1.091 1.123<br />

Heat Exchanger 3 9.82 10<br />

Power<br />

IHX loop<br />

(MWt)<br />

Bypass mode loop<br />

(MWt)<br />

Heat Exchanger 1 37.37 37.37<br />

Steam Generator 2 94.63 99.7<br />

Heat Exchanger 3 57.46 59.34<br />

Net Power 178.6 185.3<br />

Since the assumption was that<br />

constant heat was supplied from the<br />

primary system, it is seen that the<br />

efficiency of the bypass model where<br />

net power is high, and it is judged that<br />

efficiency optimization model can be<br />

formulated by detailed design.<br />

4 Conclusion<br />

In this study, VHTR system was<br />

modelled for supplying high temperature<br />

heat, by distribution, produced in<br />

the high temperature gas furnace to<br />

hydrogen producing equipment and<br />

power generation equipment.<br />

Provided that high temperature<br />

gas- cooled reactor is located in<br />

primary system, the secondary system<br />

where hydrogen production and<br />

power supply are possible were<br />

explained. The helium that flows in<br />

the nuclear reactor first passes<br />

through the HX (heat exchanger)<br />

whose purpose is the production of<br />

hydrogen, and secondly and thirdly<br />

pass through the steam generator<br />

composed of super critical carbon<br />

dioxide cycle, and heat exchanger,<br />

respectively, producing the process<br />

heat and power. In order to analyze<br />

existing IHX loop model and bypass<br />

mode-added IHX loop model, the<br />

present authors studied the input &<br />

output conditions and output change<br />

of each steam generator and heat<br />

exchanger, and based on this result,<br />

by designing IHX loop in the power<br />

production part in detail, the authors<br />

performed the calculation of thermodynamic<br />

physical value and efficiency<br />

at each point. Additionally, the<br />

authors studied the change regarding<br />

electricity output and efficiency<br />

according to bypass mode, when<br />

reheating cycle is added, the possibility<br />

on the efficiency optimization<br />

was proposed.<br />

References<br />

[1] Kim. Y. W., 2015, Nuclear Hydrogen<br />

Production Technology development<br />

Using Very High Temperature Reactor,<br />

Trans. Korean Soc. Mech. Eng. C, Vol. 3,<br />

No. 4, pp. 299~305.<br />

[2] Chang. J. H., 2006, Current Status of<br />

Nuclear Hydrogen Development,<br />

Journal of Energy Engineering, Vol.15,<br />

No.2, pp. 127~137.<br />

[3] Lee. S. I., 2015, Heat Balance Study<br />

on Integrated Cycles for Hydrogen<br />

and Electricity Generation in VHTR,<br />

Transaction of the KNS Spring Meeting.<br />

[4] Sung. H. C., 2012, Development of<br />

Ultra-Supercritical (USC) Power Plant,<br />

Trans. Korean Soc. Mech. Eng. B,<br />

Vol. 36, No.2, pp.205~210.<br />

[5] Yeom Chung-seop, Im Dong-ryeol,<br />

Lee Jung-ik, 2014, Trend of Electricity<br />

Generation Technology using supercritical<br />

CO 2 , Institute for Advanced<br />

Engineering, KIC News, Volume 17,<br />

No.1.<br />

[6] K.C.Cotton,1998, Evaluating and<br />

Improving Steam Turbine Performance,<br />

2 nd edition, Cotton Fact Inc.<br />

[7] F-Chart Software, 2016,<br />

Engineering Equation Solver,<br />

http://www.fchart.com/ees/<br />

[8] NGNP Conceptual Design Report/Steam<br />

Cycle Modular Helium Reactor<br />

(SC-MHR) Demonstration Plant,<br />

Table 3-6 SC-MHR Conceptual Design<br />

Point Design Parameter.<br />

[9] SangIL Lee, Yeon Jae Yoo, Gyunyoung<br />

Heo, Soyoung Park, Yeon Kwan Kang,<br />

Heat Balance Study on Integrated<br />

Cycles for Hydrogen and Electricity<br />

Generation in VHTR-Part 2, Korean<br />

Nuclear Society Autumn Meeting,<br />

Oct 28-30, 2015.<br />

Authors<br />

SangIL Lee<br />

YeonJae Yoo<br />

Deok Hoon Kye<br />

Department of Nuclear Team<br />

Power & Energy Plant Division<br />

Hyundai Engineering Company<br />

Seoul, Korea<br />

Gyunyoung Heo<br />

Eojin Jeon<br />

Soyoung Park<br />

Department of Nuclear<br />

Engineering<br />

Kyung Hee University<br />

Yongin Korea<br />

Operation and New Build<br />

Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Informationsbedarf versus Geheimhaltungspflichten –<br />

Erweiterung des In-camera-Verfahrens geplant<br />

235<br />

Tobias Leidinger<br />

Dem berechtigten Anspruch der Öffentlichkeit auf detaillierte Informationen über ein atomrechtlich genehmigungsbedürftiges<br />

Vorhaben steht das staatliche Interesse an einem effektiven Geheimnisschutz sensibler Daten<br />

gegenüber. Dieser Konflikt tritt regelmäßig im Genehmigungsverfahren aber auch vor Gericht zu Tage. Die differenzierten<br />

Gesetzesbestimmungen, die den Ausgleich dieser widerstreitenden Interessen regeln, sollen nun durch eine<br />

weitere Facette ergänzt werden: Ein erweitertes In-camera-Verfahren bei Gericht. Nach dem Koalitionsvertrag vom<br />

12. März <strong>2018</strong> soll die Regelung in der schon laufenden 18. Legislaturperiode erfolgen.<br />

I Grundkonflikt Informationsbedarf vs.<br />

Geheimhaltungspflicht<br />

In atomrechtlichen Genehmigungsverfahren zeigt sich<br />

regelmäßig ein Grundkonflikt: Dem Interesse der Öffentlichkeit<br />

an möglichst vertieften Informationen über<br />

alle sicherheits- und sicherungsrelevanten Aspekte des<br />

Vorhabens steht das Erfordernis eines effektiven Geheimnisschutzes<br />

in Bezug auf sensible Daten gegenüber.<br />

Genauer betrachtet lassen sich für beide Pole Grundrechtspositionen<br />

anführen: Einerseits ist Information Voraussetzung<br />

für Transparenz und Teilhabe der Öffentlichkeit<br />

am Genehmigungsverfahren. Das Verfahren dient der<br />

Gewährleistung materieller Schutzansprüche Dritter.<br />

Ohne Information ist Kontrolle gegenüber der Verwaltung<br />

kaum realisierbar. Information ist die Grundlage für<br />

Partizipation und Teilhabe der Öffentlichkeit an einem<br />

Verfahren. Das BVerfG bringt dies mit der Formel „Grundrechtsschutz<br />

durch Verfahren und Teilhabe an Information“<br />

auf den Punkt.<br />

Für die andere Seite, dem Interesse an Geheimhaltung<br />

sensibler Daten, lassen sich aber nicht minder gewichtige<br />

Grundrechtsinteressen anführen: Die Geheimhaltung<br />

dient ebenfalls zum Schutz der Grundrechtsträger: Ist der<br />

Staat zum Schutz der Grundrechte („Leben, Gesundheit“)<br />

seiner Bürger verpflichtet, bedarf es des Geheimnisschutzes<br />

in Bezug auf sensible Daten, damit eine effektive<br />

Terrorabwehr – gerade zum Schutz der Bürger – gewährleistet<br />

bleibt. Die Nicht-Preisgabe sicherheits- und<br />

sicherungsrelevanter Informationen ist mithin nicht<br />

minder essentielle Voraussetzung für einen effektiven<br />

Grundrechtsschutz der Bürger.<br />

II Interessenausgleich durch differenzierte<br />

Gesetzesregelungen<br />

Der Gesetzgeber trägt zur Lösung dieser widerstreitenden<br />

Interessen im atomrechtlichen Genehmigungsverfahren<br />

bereits heute durch eine ganze Reihe differenzierter<br />

Regelungen bei. Nach § 6 der Atomrechtlichen Verfahrensordnung<br />

(AtVfV) sind nicht nur der Antrag, der Sicherheitsbericht<br />

und eine Kurzbeschreibung des jeweils zu<br />

genehmigenden Vorhabens für die Öffentlichkeit auszulegen,<br />

sondern es besteht nach § 6a Abs. 2 Satz 1 und<br />

Abs. 3 AtVfV die Möglichkeit, in Bezug auf das Vorhaben –<br />

im Interesse der Sicherheit und Sicherung – geheimhaltungsbedürftige<br />

Informationen durch eine Beschreibung<br />

oder Inhaltsdarstellung zu ersetzen. Anstelle einer<br />

„Schwärzung“ von Unterlagen – die letztlich eine „Verweigerung“<br />

von Information bedeutete –, tritt so die<br />

Möglichkeit, geheimhaltungsbedürftige Informationen zu<br />

umschreiben, so dass der Dritte in der Lage bleibt, seine<br />

Betroffenheit durch das Vorhaben gleichwohl erkennen<br />

und beurteilen zu können.<br />

Eine Einschränkung von Informationsansprüchen ist<br />

auch jenseits dieser Regelung möglich: Während eines<br />

atomrechtlichen Verfahrens besteht der Anspruch auf<br />

Akteneinsicht gemäß § 6 Abs. 4 AtVfV i.V.m. § 29 Abs. 1<br />

S. 3, Abs. 2 und 3 des Verwaltungsverfahrensgesetzes<br />

(VwVfG) nur nach Ermessen der Behörde (also nicht<br />

„ unbedingt“). Informationen, die sicherheits- oder<br />

sicherungsrelevant sind, weil sie den Ansatz für die Ausschaltung<br />

von Sicherheits- und Sicherungsmaßnahmen<br />

oder für die Identifizierung/Lokalisierung von Schwachstellen<br />

eröffnen könnten, können – soweit durch ihre<br />

Preisgabe ein „Nachteil zum Wohl des Bundes oder<br />

Landes“ zu befürchten ist – von der Offenlegung ausgeschlossen<br />

werden. Spezialgesetzlich ist die Geheimhaltung<br />

von sensiblen Informationen im Sicherheitsüberprüfungsgesetz<br />

(SÜG) geregelt. Besteht danach die Gefahr<br />

eines „Nachteils“ oder wäre die Preisgabe der Information<br />

sogar „schädlich“ für Bund oder Land, kann sie<br />

nach Maßgabe der Verschlusssachen-Anweisung (VS-<br />

Anweisung) durch den Geheimschutzbeauftragten der<br />

Behörde als „Verschlusssache – Nur für den Dienstgebrauch“<br />

oder sogar als „Verschlusssache – Vertraulich“<br />

eingestuft und ihre Offenlegung verweigert werden. Was<br />

nach Maßgabe des SÜG i.V.m. VS-Anweisung geheim zu<br />

halten ist, darf auch nicht in anderem Zusammenhang<br />

preisgegeben werden: So bestehen – auch außerhalb eines<br />

atomrechtlichen Verfahrens – Informationsansprüche<br />

Dritter, z.B. auf Herausgabe von umweltrelevanten<br />

Informationen gegen die Genehmigungsbehörde nach<br />

Umweltinformationsgesetz (UIG) oder – soweit die<br />

Informationen nicht umweltrelevant sind – nach Maßgabe<br />

des Informationsfreiheitsgesetzes (IFG). Pressevertreter<br />

können sich darüber hinaus auch auf das jeweilige<br />

Landes-Pressegesetz stützen. In all diesen Fällen besteht<br />

indes die Möglichkeit – mit oder ohne ausdrücklichen<br />

Bezug auf das SÜG –, dass sicherheits- und sicherungsrelevante<br />

Informationen im Ergebnis nicht offenbart<br />

werden müssen, wenn die materiellen Schutzvoraussetzungen<br />

nach SÜG i.V.m. der VS-Anweisung vorliegen.<br />

III In-camera-Verfahren de lege lata und<br />

de lege ferenda<br />

Verweigert die atomrechtliche Genehmigungsbehörde die<br />

Herausgabe sensibler Informationen unter Verweis auf<br />

den Geheimschutz auch im Gerichtsverfahren, – in dem<br />

z.B. über die Rechtmäßigkeit einer atomrechtlichen<br />

Genehmigung gestritten wird – so sieht die bislang<br />

existierende Gesetzesregelung zum sog. In-camera-<br />

Verfahren in § 99 Verwaltungsgerichtsordnung (VwGO)<br />

vor, dass über die Frage der Geheimhaltungsbedürftigkeit<br />

ein speziell besetzter Fachsenat vorab entscheidet. Ihm<br />

sind ausschließlich die geheimhaltungsbedürftigen Akten<br />

vorzulegen („in camera“), um zu prüfen, ob die Einstufung<br />

als „geheim“ zurecht erfolgt ist und daher die Verweigerung<br />

der Aktenvorlage durch die Behörde Bestand hat<br />

oder nicht. Nur wenn die Geheimhaltungsbedürftigkeit<br />

verneint wird, ist die vorenthaltene Information dem<br />

Verwaltungsgericht zugänglich zu machen. Nur dann<br />

kann es darauf zugreifen und seine Entscheidung darauf<br />

stützen.<br />

SPOTLIGHT ON NUCLEAR LAW<br />

Spotlight on Nuclear Law<br />

Information Requirements Versus Confidentiality Obligations – Extension of the In-Camera Procedure Planned ı Tobias Leidinger


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

SPOTLIGHT ON NUCLEAR LAW 236<br />

Der Koalitionsvertrag vom 12. März <strong>2018</strong> (vgl. Seite<br />

141) sieht nun vor, dass die Regelungen für das In-camera-<br />

Verfahren für das Atomrecht dahingehend erweitert<br />

werden sollen, dass geheimhaltungsbedürftige Unter lagen<br />

auch zum Zwecke des Nachweises der Genehmigungsvoraussetzungen<br />

in ein verwaltungsgerichtliches Hauptsacheverfahren<br />

– bei gleichzeitiger Wahrung des Geheimschutzes<br />

– eingeführt werden können. Das In-camera-<br />

Verfahren dient dann nicht (mehr allein) zur Klärung der<br />

Frage der Geheimhaltungsbedürftigkeit einer Unterlage<br />

(wie bisher), sondern ermöglicht darüber hinaus eine<br />

weitergehende Prüfung in der Sache durch das Gericht.<br />

Das Gericht prüft dann auch, ob der erforderliche Schutz<br />

gegen Störmaßnahmen Dritter (SEWD) als gegeben unterstellt<br />

werden darf oder nicht. Die Gewährleistung des<br />

SEWD-Schutzes ist eine wesentliche Tatbestandsvoraussetzung,<br />

die erfüllt sein muss, damit eine atomrechtliche<br />

Genehmigung erteilt werden kann. Dabei ist aber auch in<br />

einem erweiterten In-camera- Verfahren sicherzustellen,<br />

dass die behördliche Ein schätzungsprärogative in Bezug<br />

auf genehmigungs relevante Wertungen bei Sicherheit und<br />

Sicherung beachtet werden. Das bedeutet, dass das Gericht<br />

sich nicht an die Stelle der Behörde setzen darf, also eine<br />

eigene Entscheidung anstelle der Behörde trifft, sondern<br />

bei seiner Nachprüfung auf eine Vertretbarkeitskontrolle<br />

beschränkt bleibt.<br />

die Frage, ob im Ergebnis davon ausgegangen werden darf,<br />

dass die erforderliche Schadensvorsorge und der gebotene<br />

Schutz gegen SEWD-Ereignisse gewährleistet ist oder<br />

nicht, könnte auf diese Weise weitergehend als bisher entschärft<br />

werden. Idealerweise bliebe der Geheimnisschutz<br />

auch so gewahrt, zugleich aber wäre dem Interesse<br />

der Drittbetroffenen an einer Überprüfung essentieller<br />

Genehmigungsvoraussetzungen unter Berücksichtigung<br />

geheimhaltungsbedürftiger Informationen weitergehend<br />

als bisher entsprochen. Das wäre als konstruktiver Beitrag<br />

zur Stärkung eines effektiven Grundrechtsschutzes zu<br />

bewerten: Ein erweitertes In-camera-Verfahren diente so<br />

in besonderer Weise zur Gewährleistung der dem Dritten<br />

zustehenden Schutzansprüche und wahrte dabei<br />

gleichwohl den erforderlichen Geheimschutz, der nicht<br />

minder einem effektiven Grundrechtsschutz der Bürger<br />

geschuldet ist.<br />

Allerdings bleiben die konkrete Ausgestaltung und der<br />

Vollzug dieser Regelung in der Praxis abzuwarten: Folgt<br />

einer guten Absicht des Gesetzgebers eine in der Praxis<br />

tatsächlich und rechtlich brauchbare Lösung? Ziel müsste<br />

sein, dadurch nicht neue Fragen zur Anwendung und<br />

Reichweite eines erweiterten In-camera-Verfahrens<br />

aufzuwerfen, sondern eine inhaltlich klare und hinreichend<br />

bestimmte Norm zu schaffen, die das Versprechen<br />

des Koalitionsvertrages vollzugsfähig einlöst.<br />

IV Erweiterung des In-camera-Verfahrens:<br />

Bedenkenswerter Schritt<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

238<br />

ENVIRONMENT AND SAFETY<br />

CFD Modeling and Simulation<br />

of Heat and Mass Transfer in<br />

Passive Heat Removal Systems<br />

Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas<br />

This paper is presenting the CFD-modelling and simulation of condensation inside passive heat removal systems.<br />

Designs of future nuclear boiling water reactor concepts are equipped with emergency cooling systems which are<br />

passive systems for heat removal. The emergency cooling system consists of slightly inclined horizontal pipes which are<br />

immersed in a tank of subcooled water. At normal operation conditions, the pipes are filled with water and no heat<br />

transfer to the secondary side of the condenser occurs. In the case of some accident scenarios the water level may<br />

decrease in the core, steam enters the emergency pipes and due to the subcooled water around the pipe, this steam<br />

condenses. The emergency condenser acts as a strong heat sink which is responsible for a quick depressurization of the<br />

reactor core. This procedure acts passive i.e. without any additional external measures. The actual project is defined to<br />

model the phenomena which are occurring inside the emergency condensers. The focus of the project is on detection of<br />

different morphologies such as annular flow, stratified flow, slug flow and plug flow and also modeling of the laminar<br />

film which is occurring during the condensation near the wall.<br />

The condensation procedure inside the<br />

pipe is determined by two important<br />

phenomena. The first one is wall<br />

condensation and the second one is the<br />

direct contact condensation (DCC).<br />

The Algebraic Interfacial Area Density<br />

(AIAD) concept is used in order to<br />

model the interface between liquid<br />

and steam. In the next steps the Generalized<br />

Two-Phase Flow ( GENTOP)<br />

model will be used to model also the<br />

dispersed phases which are occurring<br />

inside the pipe. Finally, the results of<br />

the simulations will be validated by<br />

experimental data which will be available<br />

in HZDR. In this paper the results<br />

of the first part are presented.<br />

1 Introduction<br />

Condensation plays a crucial role in<br />

the emergency condenser of passive<br />

heat removal systems of nuclear power<br />

plants. Passive safety systems do not<br />

need any external power supplies and<br />

they mostly depend on physical phenomena<br />

such as natural circulation<br />

and gravity driven flows. In order to<br />

assess the performance of passive safety<br />

systems and their efficiency mostly<br />

one-dimensional codes are used such<br />

as ATHLET, RELAP and TRACE. These<br />

codes are able to calculate most of the<br />

phe nomena in power plants; however,<br />

they cannot reflect the 3D phenomena.<br />

Therefore, Computational Fluid<br />

Dynamics (CFD) methods should be<br />

used to simulate and predict the<br />

complex multiphase flow structure.<br />

Despite the previous research being<br />

done on the two-phase flow behavior,<br />

this phenomenon needs much more<br />

investigations. The two-phase flow<br />

patterns and transition between vapor<br />

and liquid are studied by Thome and<br />

Hajal et al. [1, 2]. They introduced a<br />

logarithmic mean void fraction (LMe)<br />

method in order to calculate the vapor<br />

void fractions which change from the<br />

low pressure up to the critical pressure<br />

point. Moreover, they proposed a new<br />

heat transfer model based on the same<br />

simplified flow structures that have<br />

been used in the flow boiling model<br />

of Kattan et al. [3]. The model can<br />

predict the local condensation heat<br />

transfer coefficient for different flow<br />

regimes such as annular, intermittent,<br />

stratified-wavy fully stratified and<br />

wavy flow.<br />

Many attempts have been done to<br />

investigate the mass transfer between<br />

liquid and gas phase in condensation.<br />

Lee et al. [4] introduced a model for<br />

prediction of the mass transfer. They<br />

assumed that the interface between<br />

liquid and steam is on saturation<br />

temperature and introduced an<br />

iterative technique in order to reach to<br />

desired boundary condition inside<br />

each cell. This model depends on a<br />

relaxation factor which needs to be<br />

tuned. The tuning needs many trial<br />

and error simulations which is<br />

time-consuming and doesn’t have any<br />

predictive capabilities.<br />

Moreover, there are empirical or<br />

semi-empirical methods to calculate<br />

the mass transfer in the interface.<br />

Strubelj et al. [5] by using ANSYS CFX<br />

and NEPTUNE_CFD [6] code tried to<br />

simulate Direct Contact Condensation<br />

(DCC) in stratified flows. In DCC the<br />

phase change occurs due to the direct<br />

contact interaction of subcooled water<br />

and saturated steam. The defined<br />

phase change mass flux depends on<br />

thermal conductivity of the liquid and<br />

Nusselt number of the liquid. The<br />

Nusselt number was calculated<br />

by Coste et al. [7] based on Surface<br />

Renewal Theory (SRT) [8]. The SRT<br />

theory calculates the mass transfer<br />

according to the renewal period of<br />

eddies and the liquid turbulent<br />

properties. Hughes and Duffey [9]<br />

used the surface renewal theory and<br />

the Kolmogorov turbulent length<br />

scale theory to define a correlation for<br />

the heat transfer coefficient. They<br />

considered that the heat removal from<br />

interface occurs by smallest turbulent<br />

scales. This model will be introduced<br />

more detailed in the next sections.<br />

This correlation is validated for<br />

Pressurized Thermal Schock (PTS)<br />

phenomenon by Egorov [10] and<br />

Apanasevich [11]. Further to Hughes<br />

correlation, Shen et al. [12] developed<br />

another correlation for calculation of<br />

heat transfer coefficient based on the<br />

surface renewal theory. Ceuca et al.<br />

[13] used both of these correlations<br />

in order to simulate the direct contact<br />

condensation for the LAOKOON<br />

facility [14]. By comparison of Hughes<br />

and Duffey correlation with Shen<br />

correlation, Ceuca et al. [13] concluded<br />

that both of the models provide<br />

accurate results for the horizontal<br />

stratified quasi-steady state.<br />

Evidently, many attempts have been<br />

done in the modeling of con densation<br />

inside the pipes. The goal of the current<br />

work is modeling of the transition<br />

between different mor phologies which<br />

are occurring during the condensation<br />

inside the pipe ( Figure 1). In order to<br />

do that, several CFD models such as<br />

IMUSIG, AIAD and GENTOP which<br />

have been developed in HZDR in cooperation<br />

with ANSYS are available. The<br />

Inhomo geneous MUSIG model considers<br />

the bubble size distribution and<br />

is used for modeling the small-scaled<br />

dispersed gas phase [15]. The AIAD<br />

Environment and Safety<br />

CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems<br />

ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 1.<br />

Schematic representation of horizontal flow and different morphologies.<br />

(Algebraic Interfacial Area Density<br />

Model) is developed for detection of<br />

the local morphology and corresponding<br />

switch between them [16]. The<br />

recently developed GENTOP- model<br />

combines both concepts. GENTOP<br />

(Generalized Two-Phase Flow) approach<br />

is able to simulate co-existing<br />

large-scaled (continuous) and smallscaled<br />

(polydispersed) structures [17].<br />

All these models are validated for adiabatic<br />

cases without any phase change.<br />

Therefore, the start point of the current<br />

work project is using the available<br />

models and integrating phase transition<br />

and con densation models into<br />

them. In the current work as initial<br />

stages the AIAD model has been used<br />

since in this model 2 continues phases<br />

should be considered and it is less complicated<br />

compare to GENTOP model<br />

which also considers a poly- dispersed<br />

phase. In the proceeding sections a<br />

more detail explanation of AIAD model<br />

will be given.<br />

2 CFD model formulation<br />

In the current work a multi-field twophase<br />

CFD approach is used with<br />

ANSYS CFX 17.2 in order to simulate<br />

the condensation inside horizontal<br />

pipe flows. The mass, momentum and<br />

energy equations can be defined,<br />

respectively, as follow:<br />

• Mass conservation equation:<br />

(1)<br />

where S Mi describes user specified<br />

mass source.<br />

χ iβ the mass flow rate per unit volume<br />

from phase β to phase i.<br />

• Momentum conservation equation:<br />

(2)<br />

where S mi is the momentum source<br />

caused by external body forces<br />

and user defined momentum<br />

sources.<br />

M i is the interfacial forces acting<br />

on phase i due to the presence<br />

of other phases.<br />

χ + iβ v β – χ + βi v i is the momentum<br />

transfer induced<br />

by mass transfer.<br />

• The total energy equation:<br />

(3)<br />

where: h tot is the total enthalpy<br />

related to static enthalpy by:<br />

(4)<br />

<br />

T i , λ i represents the temperature<br />

and the thermal conductivity<br />

of phase i.<br />

S Ei describes external heat sources.<br />

Q i is interphase heat transfer<br />

to phase i across interfaces<br />

with the other phase.<br />

χ + iβ h βs – χ + βi h is denotes the interphase<br />

mass transfer.<br />

In ANSYS CFX in order to describe the<br />

phase change which occurs due to the<br />

interphase heat transfer, the Thermal<br />

Phase Change Model has been introduced<br />

[30]. This model is particularly<br />

useful in simulation of the condensation<br />

of saturated vapor. The heat<br />

flux from the interface to phase i and<br />

phase β is:<br />

q i = h i (T sat – T i ) (5)<br />

q β = h β (T sat – T β ) (6)<br />

where h i , h β and T sat are heat transfer<br />

coefficients of the phase i and phase<br />

β and the saturation temperature,<br />

respectively. ṁ iβ is the mas flux from<br />

phase β to phase i. H is and H βs are the<br />

interfacial enthalpy values which<br />

come into and out of the phase due<br />

to phase change which occurs. By<br />

usage of the total heat balance<br />

equation the interphase mas flux can<br />

be determined as follow:<br />

<br />

| | Fig. 2.<br />

3D geometry of the pipe and mesh of the cross section.<br />

(7)<br />

ṁ iβ > 0 → H is = H i,sat , H βs = H β (8)<br />

ṁ iβ < 0 → H is = H i , H βs = H β,sat (9)<br />

In the current work, the steam<br />

con sidered to be in saturation temperature.<br />

Therefore, the heat flux<br />

from the steam to the interface equals<br />

zero since both are in saturation<br />

temperature. As a result, the interphase<br />

mass flux formula can be<br />

written as:<br />

<br />

(10)<br />

In this work in order to model the heat<br />

transfer coefficient the Hughes and<br />

Duffy model has been used which is<br />

based on the SRT model [9]. They<br />

used the Surface Renewal Theory<br />

(SRT) and the Kolmogorov turbulent<br />

length scale theory to find a correlation<br />

for heat transfer coefficient.<br />

Therefore, the heat transfer coefficient<br />

was derived as:<br />

<br />

(11)<br />

where ε is the turbulent dissipation, v l<br />

is the kinematic viscosity and λ is the<br />

thermal conductivity.<br />

3 Computational grid and<br />

boundary conditions<br />

In Figure 2 the pipe and the boundary<br />

conditions are shown. The pipe is<br />

horizontal and has 1 m length<br />

and 0.043 m diameter. In order to<br />

define a mesh for the pipe ANSYS<br />

ICEM software is used. Due to the<br />

higher importance of the wall region<br />

compare to the middle of the pipe,<br />

the mesh near the wall needs to be<br />

finer than the mesh in the pipe<br />

center. The number of nodes is<br />

1,250,000.<br />

Mass flow rate<br />

[Kg/s]<br />

Temperature<br />

(k)<br />

Inlet 0.5 537.1<br />

Wall - 312.18<br />

outlet outflow -<br />

ENVIRONMENT AND SAFETY 239<br />

CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems<br />

Environment and Safety<br />

ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

ENVIRONMENT AND SAFETY 240<br />

| | Fig. 3.<br />

(a) Area averaged liquid volume fraction in different cross sections over the pipe length,<br />

(b) Temperature distribution in the outlet of the pipe for 5 different radial lines.<br />

4 Results<br />

The results are obtained with the AIAD<br />

approach for modeling the free surface<br />

and morphologies. Moreover, the<br />

Hughes correlation is used for the heat<br />

transfer coefficient. Figure 3 represents<br />

the qualitative profiles of liquid<br />

volume fraction and tem perature. In<br />

Figure 3 (a) the volume fraction profile<br />

in the vertical cross section in the<br />

middle of the pipe and in the streamwise<br />

direction is represented. As it can<br />

be seen at the inlet the pure steam<br />

exists and by going further in the pipe,<br />

due to the film condensation a liquid<br />

film starts to generate near the wall.<br />

The liquid film is growing and leads to<br />

the thicker film. In a cross section<br />

500 mm far from the inlet the liquid<br />

film is falling down gradually and<br />

gathering at lower part of the pipe.<br />

The liquid film always exists near the<br />

wall because as soon as the liquid is<br />

falling down the steam becomes in the<br />

direct contact with the wall and condenses<br />

and again new film generates.<br />

Moreover, in Figure 3 (d) the temperature<br />

profile is shown for different<br />

cross sections along the pipe. As<br />

mentioned before, the steam is fixed<br />

at the satu ration temperature, but<br />

further along the pipe by generating<br />

the liquid the temperature of the liquid<br />

is decreasing because of the heat<br />

flux to the wall. Moreover, the wall<br />

heat flux is cooling the liquid which<br />

causes the direct contact condensation<br />

between liquid and steam interface. As<br />

the steam is on the saturation temperature<br />

there is no heat flux between<br />

the interface which is also on the saturation<br />

temperature and the steam.<br />

Therefore, just the phase is changing<br />

and the steam turns into the liquid.<br />

| | Fig. 4.<br />

(a) Liquid Volume fraction distribution on a cross section along the pipe, (b) temperature distribution on a cross section along the pipe,<br />

(c) Volume fraction distribution on different cross sections, (d) temperature distribution on different cross sections.<br />

Figure 4 (a) shows the change of<br />

cross section averaged liquid volume<br />

fraction along the pipe. According to<br />

the figure the average liquid volume<br />

fraction at the inlet is 0.0 and due to<br />

the mass transfer it’s increasing along<br />

the pipe and it reaches to around 0.1<br />

at end of the pipe. Therefore, in a<br />

horizontal pipe with one meter length<br />

the total condensation rate is around<br />

10 percent. In Figure 4 (b) the temperature<br />

distribution for the five<br />

radial lines on the outlet of the pipe<br />

is presented. This plot shows the<br />

temperature difference from the<br />

center of the pipe towards the wall.<br />

As the plot shows, in the center the<br />

temperature is equal to the saturation<br />

temperature. As far as getting closer to<br />

the wall which is in subcooled<br />

tem perature, the temperature gradient<br />

is increasing. In other words, in<br />

the region near the wall the temperature<br />

difference from the saturation<br />

temperature is higher. Moreover,<br />

slope of the plot for L5 is higher than<br />

L1. The reason is in lower part of the<br />

pipe (which is showed by L5) the<br />

amount of cooled liquid is higher<br />

which causes higher temperature<br />

gradient in the lower parts of the pipe.<br />

As the pipe is symmetric and the<br />

boundary conditions for both sides of<br />

the pipe are same, there is no need to<br />

plot the temperature distribution in<br />

another half of the cross section.<br />

5 Conclusion<br />

The ANSYS CFX 17.2 has been used in<br />

order to simulate the condensation<br />

inside horizontal tubes. In order to<br />

model the two phase flow, heat<br />

transfer and phase change are included<br />

in the available AIAD concept<br />

which was developed for adiabatic<br />

cases. Moreover, the Hughes heat<br />

transfer coefficient correlation is<br />

implemented for the modeling of the<br />

direct contact condensation in the<br />

interface. The changes of the flow<br />

structure inside the pipe and the<br />

volume fraction and the temperature<br />

profiles have been studied in detail.<br />

The liquid film which is generated<br />

near the wall due to the wall condensation<br />

is modeled and it can be seen in<br />

the volume fraction profiles. By generating<br />

the liquid film near the wall both<br />

wall condensation and direct contact<br />

condensation are occurring inside<br />

the pipe at the same time. Whereas in<br />

the actual paper only the test for<br />

plausibility of the AIAD model was<br />

done, in the near future the comparison<br />

to the experiment is planned.<br />

The next step which is an ongoing<br />

part of the project is simulation of the<br />

Environment and Safety<br />

CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems<br />

ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

whole condensation phenomena and<br />

flow morphologies by using GENTOP<br />

concept. Further to the AIAD concept<br />

which considers two continuous<br />

fluids, the GENTOP approach is a<br />

three field two fluid model and considers<br />

also a poly dispersed phase.<br />

Acknowledgments<br />

This project is an ongoing project in<br />

Helmholtz-Zentrum Dresden Rossendorf<br />

(HZDR), which is funded by Bundesministerium<br />

für Bildung und Forschung<br />

(BMBF) under grant number<br />

02NUK041B in Germany.<br />

References<br />

[1] Hajal, J.El.; Thome, J.; Cavallini, A.<br />

Condensation inside horizontal tubes,<br />

part 1: two phase flow pattern map.<br />

International Journal of Heat and Mass<br />

Transfer 46: 3349-3363(2003).<br />

[2] Thome, J.; Hajal, J.El; Cavallini, A.<br />

Condensation inside horizontal tubes,<br />

part 2: New heat transfer model based<br />

on flow regimes. International Journal<br />

of Heat and Mass Transfer 46: 3365-<br />

3387(2003).<br />

[3] Kattan, N.; Thome, J.R.; Favrat, D. Flow<br />

boiling in horizontal tubes:part2-New<br />

heat transfer data for five refrigerants.<br />

J. Heat Transfer 120: 148-155 (1998).<br />

[4] Lee, W. H. A Pressure Iteration Scheme<br />

for Two-Phase Flow Modeling. Multiphase<br />

Transport Fundamentals,<br />

Reactor Safety, Applications: 407–432,<br />

(1980).<br />

[5] Štrubelj, L.; Ézsöl, Gy. ; Tiselj, I. Direct<br />

Contact Condensation Induced<br />

Transition from Stratified to Slug Flow.<br />

Nuclear Engineering and Design 240:<br />

266–274 (2010).<br />

[6] Lavieville, J.; Quemerais, E.; Boucker, M.;<br />

Maas, L., NEPTUNE CFD V1.0 User Guide<br />

(2005).<br />

[7] Coste, P. ; Pouvreau, J. ; Lavieville, J.;<br />

Boucker, M. A Two-phase CFD approach<br />

to the PTS problem evaluated on COSI<br />

experiment. Proceedings of the 16 th<br />

International Conference on Nuclear<br />

Engineering ICONE16, USA, (2008).<br />

[8] Banerjee, S.; A surface renewal model<br />

for interfacial heat and mass transfer in<br />

transient two-phase flow. International<br />

Journal of Multiphase Flow, Vol.4:<br />

571-573 (1978).<br />

[9] Hughes, E. D.; Duffey, R. B. Direct<br />

Contact Condensation and Momentum<br />

Transfer in Turbulent Separated Flows.<br />

Internal Journal of Multiphase Flow 17:<br />

599–619 (1991).<br />

[10] Egorov, Y. Validation of CFD codes with<br />

PTS relevant test cases. Technical Report<br />

EVOL-ECORA-D07, ANSYS, Germany<br />

(2004).<br />

[11] Apanasevich, P. ; Lucas, D.; Beyer, M.;<br />

Szalinski, L. CFD based approach for<br />

modeling direct contact condensation<br />

heat transfer in two-phase turbulent<br />

stratified flows. International Journal of<br />

Thermal Sciences 95: 123-135(2015).<br />

[12] Shen, L.; Triantafyllou, G.S.; Yue. D.K.P.<br />

Turbulent diffusion near a free surface<br />

Journal of Fluid Mechanics 407:<br />

145–166 (2000).<br />

[13] Ceuca, S. C. ; Macián-Juan R. CFD<br />

Simulation of Direct contact Condensation<br />

with ANSYS CFX using Locally<br />

defined Heat Transfer Coefficient.<br />

In ICONE-20, Anaheim, California, USA,<br />

No. 54347 (2012).<br />

[14] Goldbrunner, M.; Karl, J. ; Hein, D.<br />

Experimental Investigation of Heat<br />

Transfer Phenomena During Direct<br />

Contact Condensation in the Presence<br />

of Noncondensable gas by means of<br />

Linear Raman Spectroscopy. In 10 th Int.<br />

Symp. on Laser Techniques Applied to<br />

Fluid Mechanics, Lisbon (2000).<br />

[15] Krepper, E.; Frank, Th.; Lucas, D.; Prasser,<br />

H.-M.; Zwart, P.J. The Inhomogeneous<br />

MUSIG model for the simulation of<br />

poly-dispersed flow. Nuclear Engineering<br />

Design 238: 1690-1702 (2008).<br />

[16] Höhne, T.; Deendarlianto; Lucas, D.<br />

Numerical simulations of countercurrent<br />

two-phase flow experiments in<br />

a PWR hot leg model using an area<br />

density model. International Journal<br />

of Heat and Fluid Flow 31 (5):<br />

1047-1056 (2011).<br />

[17] Hänsch, S.; Lucas, D.; Krepper, E.;<br />

Höhne, T. A multi-field two-fluid<br />

concept for transitions between<br />

different scales of interfacial structures.<br />

International Journal of Multiphase<br />

Flow 47:171-182(2012).<br />

Authors<br />

Amirhosein Moonesi Shabestary,<br />

Eckhard Krepper,<br />

Dirk Lucas<br />

Helmholtz-Zentrum<br />

Dresden-Rossendorf<br />

P.O.Box 510119<br />

01314 Dresden, Germany<br />

241<br />

DECOMMISSIONING AND WASTE MANAGEMENT<br />

The Decommissioning of the ENEA RB3<br />

Research Reactor in Montecuccolino<br />

F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo<br />

The ENEA RB3 reactor was a 100 Wth research installation owned and operated by ENEA, in its center of Montecuccolino<br />

near Bologna, from 1971 to 1989. It consisted of a cylindrical aluminium vessel, about 4.3 m high and 2.9 m in diameter,<br />

which could host various types of fuel elements suspended from the top of a special adjustable rack and submerged into<br />

moderating and cooling heavy water. Principal aim of the reactor was to provide neutronics data for the CIRENE NPP, a<br />

SGHWR that was being designed and then partially built in Latina starting from 1979. The specific RB3 core, surrounded<br />

by a graphite reflector and housed inside a concrete biological shielding, allowed to test easily very different fuel<br />

element configurations by changing their pitches and by regulating the heavy water level inside the vessel. The reactor<br />

design, similar to that of the ZED-II Canadian research facility, was originally developed by CEA for its Aquilon facility<br />

in Saclay in 1956; in fact, through a special arrangement between ENEA and CEA, parts of the Aquilon facility were<br />

ultimately donated to ENEA at the end of the 60s for the construction of RB3. In 1989, the RB3 reactor was shut down,<br />

and in the late 2010 ENEA received by ministerial decree the authorization to its dismantling, with the aim of reaching<br />

the “green field” status and with the unconditional release of its building, which is actually owned by the University of<br />

Bologna. The dismantling activities started in May 2013 and were concluded at the end of 2014; after that, a campaign<br />

for the radiological characterization of the building was initiated and concluded in June 2015. Now, all the necessary<br />

site characterization activities are being conducted with the aim to present the results declaring the “green field” status<br />

before the end of 2017. This paper will present the three main pillars of the decommissioning of RB3, namely the<br />

strategy and methods for the dismantling, the strategy and methods for the radiological characterization of the building,<br />

and finally the strategy and methods for the radiological characterization of the site. The radionuclide limits imposed<br />

by the Italian Regulatory Body, together with the challenges encountered so far will be likewise shown and described.<br />

Revised version of<br />

a paper presented<br />

at the Eurosafe,<br />

Paris, France, 6 and 7<br />

November 2017.<br />

Decommissioning and Waste Management<br />

The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

DECOMMISSIONING AND WASTE MANAGEMENT 242<br />

1 Introduction<br />

The ENEA RB3 (Reattore Bologna 3)<br />

reactor was a 100 Wth research installation<br />

owned and operated by ENEA<br />

in its center of Montecuccolino, near<br />

Bologna, from 1971 to 1989. It consisted<br />

of a cylindrical aluminium vessel,<br />

about 4.3 m high and 2.9 m in diameter,<br />

which could host various types of<br />

fuel elements suspended from the top<br />

of a special adjustable rack, and submerged<br />

into heavy water serving both<br />

as moderator and coolant. Principal<br />

aim of the reactor was to provide<br />

neutronics data for the CIRENE NPP, a<br />

SGHWR that was being designed, and<br />

then partially built in Latina, starting<br />

from 1979. The specific RB3 core, surrounded<br />

by a graphite reflector and<br />

housed inside a concrete biological<br />

shielding, allowed to test easily very<br />

different fuel element configurations<br />

by changing their pitches and by<br />

regulating the heavy water level inside<br />

the vessel. The reactor design, similar<br />

to that of the ZED-II Canadian<br />

research facility, was originally developed<br />

by CEA for its Aquilon facility in<br />

Saclay in 1956; in fact, through a<br />

special arrangement between ENEA<br />

and CEA, parts of the Aquilon facility<br />

were ultimately donated to ENEA at<br />

the end of the 60s for the construction<br />

of RB3. In 1989, after more than 18<br />

years of operation, the RB3 reactor<br />

was shut down, and in the late 2010,<br />

after waiting for the entry into force of<br />

Legislative Decree (L.D.) 230/1995<br />

[1], which introduced new laws for<br />

the decommissioning of NPPs, ENEA<br />

received by ministerial decree the<br />

authorization to its dismantling, with<br />

the aim of reaching the “green field”<br />

status and with the unconditional<br />

release of its building, including the<br />

reactor concrete biological shielding,<br />

which is actually owned by the<br />

University of Bologna. In fact the site<br />

of Montecuccolino, some 3.5 km to<br />

the South of downtown Bologna,<br />

hosted three research reactors: RB1,<br />

owned and operated by the University<br />

of Bologna, RB2, owned and operated<br />

by AGIP Nucleare, and RB3, owned<br />

and operated by ENEA. RB1 and RB2<br />

were decommissioned up to the green<br />

field status well before the entry into<br />

force of L.D. 230/1995.<br />

Figure 1 shows an aerial view of<br />

the Montecuccolino research center,<br />

with the area hosting RB3 contoured<br />

in red. Figure 2 shows a plan of the<br />

main reactor hall, with in red the<br />

area once occupied by the reactor<br />

vessel, surrounded by the hectagonal<br />

graphite reflector and encased within<br />

a thick concrete biological shielding.<br />

Figure 3 shows a vertical section of<br />

the RB3 building; the lowermost floor<br />

hosted 4 large tanks for a total of<br />

20,000 L (in red) for the storage of the<br />

heavy water which was daily pumped<br />

up into the vessel to reach criticality<br />

and then drained after the conclusion<br />

of the experiments. Three floors are<br />

present in the building: floor +6.0 m<br />

corresponding to the ground level,<br />

floor +0.0 m, corresponding to the<br />

level of the reactor vessel, and floor<br />

-3.0 m, with the heavy water storage<br />

tanks, heating and cooling systems,<br />

and other auxiliaries. The control<br />

| | Fig. 2.<br />

Plan of main hall of RB3.<br />

room was located at floor +0.0 m.<br />

While allowed to operate up to 100<br />

Wth, operations at RB3 were always<br />

conducted at 50 Wth.<br />

Between 1991 and 1992, all the<br />

fuel elements used at RB3 were either<br />

restituted at their owner (JRC Euratom<br />

Ispra) or sent to the ENEA Research<br />

Center of Saluggia or to the fuel fabrication<br />

plant of Fabbricazioni Nucleari<br />

at Bosco Marengo. Between 1992 and<br />

1993 all the heavy water was transferred<br />

to the ENEA Research Center of<br />

Borgo Sabotino, and before the end of<br />

1997 all the sealed radioactive sources<br />

used at the plant were disposed of.<br />

2 Regulatory Requirements<br />

and Classification<br />

of Components and<br />

Materials<br />

In the late 2010, ENEA received, by<br />

decree of the Italian Ministry of<br />

Economic Development, the authorization<br />

[2] to proceed with the dismantling<br />

of RB3; included in the<br />

| | Fig. 1.<br />

Aerial view of the Montecuccolino site; the RB3 building is inside the red square.<br />

| | Fig. 3.<br />

Section of the RB3 building.<br />

Decommissioning and Waste Management<br />

The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Radionuclide Metals Concrete Other<br />

materials<br />

Reused Recycle Both reuse and recycle Demolition<br />

Surface<br />

(Bq/cm 2 )<br />

decree were also the requirements<br />

imposed by the Italian Nuclear Regulatory<br />

Body ISPRA. According to these<br />

requirements, the components and<br />

materials of RB3 were classified by<br />

ENEA, taking into account the various<br />

areas of the plant and the history of<br />

its functioning, into four main<br />

categories:<br />

A) materials and components which<br />

were both in contact with possibly<br />

contaminated or activated process<br />

fluids and subject to neutron flux;<br />

B) materials and components which<br />

were in contact with possibly<br />

contaminated or activated process<br />

fluids but not directly irradiated by<br />

neutrons;<br />

C) materials and components which<br />

were irradiated by the neutron flux<br />

but which never went into contact<br />

with possibly contaminated or<br />

activated process fluids;<br />

D) s.c. “exempt” materials, which<br />

were never irradiated and never<br />

went into contact with possibly<br />

contaminated or activated process<br />

fluids.<br />

The only component classified in the<br />

A category was the aluminium vessel;<br />

the only components in the B category<br />

were the heavy water distribution<br />

pipings. Exempt materials, given their<br />

unirradiated and uncontaminated<br />

status, were subject only to a general<br />

screening through CANBERRA In Situ<br />

Object Counting Systems (ISOCS) to<br />

estimate any possible level of presence<br />

of 60Co and 137Cs; if the measured<br />

levels were below the decision threshold<br />

of the measuring system in terms<br />

of mass concentration levels, then<br />

Surface<br />

(Bq/cm 2 )<br />

| | Tab. 1.<br />

Surface or mass activity concentration levels for clearance.<br />

Mass<br />

(Bq/g)<br />

these materials were automatically<br />

discarded from the plant without any<br />

further radiological analysis. This<br />

demonstrates the “instrumental” zero<br />

of this category of materials hence the<br />

“exempt” classification. All materials<br />

which had been classified as “exempt”<br />

were released unconditionately, for a<br />

total mass of about 30 tons, between<br />

March 2013 and May 2015. For all the<br />

other three categories, the clearance<br />

levels imposed by the Regulatory<br />

Authority are summarized in Table 1.<br />

These were derived either from the<br />

Italian L.D. n. 230/95 or from RP 89<br />

[3] and RP 113 [4] publications. In<br />

presence of more than one radionuclide,<br />

the sum of the ratios of<br />

the measured concentrations to the<br />

respective levels must be lower than 1.<br />

The components and materials<br />

were further grouped by ENEA into 12<br />

s.c. “homogeneous groups” using<br />

material and historic criteria; homogeneous<br />

groups are therefore constituted<br />

by components (or parts of<br />

them) made by the same material and<br />

possibly with a homogeneous and<br />

uniform activity content.<br />

3 Radiological Characterization<br />

of Homogeneous<br />

Groups<br />

Before the radiological characterization<br />

of the batches of materials from<br />

the various homogeneous groups<br />

started, a preliminary, special campaign<br />

was conducted to exclude the<br />

presence of various isotopes among<br />

those given in Table 1, expecially in<br />

the most potentially activated or<br />

contaminated materials (category A).<br />

Surface<br />

(Bq/cm 2 )<br />

Mass<br />

(Bq/g)<br />

3 H 10,000 100,000 1 10,000 1 1<br />

14 C 1,000 1,000 1 10,000 1 1<br />

Mass<br />

(Bq/g)<br />

54 Mn 10 10 1 10 0.1 0.1<br />

55 Fe 1,000 10,000 1 10,000 1 1<br />

59 Ni 10,000 10,000 1 100,000 1 1<br />

60 Co 1 10 1 1 0.1 0.1<br />

63 Ni 1,000 10,000 1 100,000 1 1<br />

90 Sr 10 10 1 100 1 1<br />

125 Sb 10 100 1 10 1 1<br />

134 Cs 1 10 0.1 10 0.1 0.1<br />

137 Cs 10 100 1 10 1 1<br />

152 Eu 1 10 1 10 0.1 0.1<br />

154 Eu 1 10 1 10 0.1 0.1<br />

Generic Alfa 0.1 0.1 0.1 0.1 0.1 0.01<br />

241 Pu 10 10 1 100 1 1<br />

In particular 54Mn, 59Ni, 90Sr,<br />

125Sb, 134Cs, 137Cs, 239Pu, 240Pu<br />

and 241Pu were excluded from<br />

further analyses finalized to the unconditional<br />

release of materials. Then,<br />

for each homogeneous group, a precharacterization<br />

measurement campaign<br />

was con ducted with a three-fold<br />

aim: 1) to verify if the hypothesis on<br />

the homogeneity of activity for that<br />

given group held; 2) to evaluate the<br />

minimum number of samples to be<br />

analized<br />

for the subsequent characterization<br />

phase; 3) to evaluate the value of<br />

isotopic ratios of 55Fe to 60Co and<br />

of 63Ni to 60Co, so to limit the next<br />

analyses only to the research of 60Co<br />

contents. After that, and using typically<br />

13 multiple measurements for<br />

each batch of each homogeneous<br />

group, summations of the ratios<br />

between measured activity concentrations<br />

and limits (Table 1) over all<br />

the relevant isotopes were carried out.<br />

If these summations resulted


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

DECOMMISSIONING AND WASTE MANAGEMENT 244<br />

| | Fig. 4.<br />

Dismantling of the lower layers of the graphite reflector.<br />

in a 1:10 ratio with other similar<br />

metals of warranted non-nuclear<br />

provenance in order to be used again<br />

for various purposes. All the homogeneous<br />

groups were pre-characterized,<br />

characterized and released before the<br />

end of 2014. All the measurements<br />

were performed by trained ENEA staff<br />

and within qualified ENEA laboratories,<br />

with the exception of some 14C<br />

measurements of a small lot of rubbers<br />

which were performed, under special<br />

contract, by the LASE Laboratory<br />

of CEA in Saclay. Workmen for heavy<br />

or peculiar technological operations<br />

were hired from the Modena Fallimenti<br />

SaS, a private Italian company specialized<br />

in the dismantling of special<br />

plants. Further details about the plan<br />

for the characterization of materials<br />

and components can be found in<br />

[5,6].<br />

4 Radiological Characterization<br />

of the Building<br />

After the completion of all the dismantling<br />

activities, and after the release<br />

of all the batches of materials, a<br />

radiological characterization of the<br />

building of RB3 has been made. This<br />

consisted of two main steps. The first<br />

was the characterization of the activation<br />

status of the baritic concrete<br />

biological shielding of the core. This<br />

consisted in seven core drillings, (see<br />

Figure 5) each 16 cm long, so distributed:<br />

1 on the floor, 1 on the<br />

northern wall, 1 on the western wall,<br />

1 on the eastern wall, and 3 (at<br />

different heights) on the souther wall.<br />

All the drilling points were at positions<br />

where the neutron flux during<br />

operation was maximum. From each<br />

drilling, four aliquots, 4 cm long,<br />

where taken, so to cover the depth<br />

profile of any activation distribution<br />

inside the biological shielding. Each<br />

aliquot was subject to gamma spectrometry<br />

to search for the presence of<br />

60Co, 134Cs, 152Eu and 154Eu. All<br />

the 28 samples yielded results for all<br />

the four isotopes lower than a few<br />

mBq/g. Then, all the samples were<br />

subject to thorough statistical analysis,<br />

based on several Bartlett tests, to<br />

verify if they were all and altogether<br />

representative of the same statistical<br />

distribution of activity and therefore<br />

representative of a same “homogeneous<br />

group” constituted of the whole<br />

biological shielding. Once this condition<br />

has been verified, a Noether test,<br />

using 10 randomly chosen measurements,<br />

was put in place to verify the<br />

minimum number of samples to be<br />

used for the final characterization of<br />

the biological shielding. This resulted<br />

in 13 samples, randomly extracted<br />

from the complete set of all the 28<br />

available samples. However, ENEA decided<br />

to use all the 28 samples to verify<br />

the free release condition for the<br />

shielding, and for all the 28 samples<br />

the condition resulted verified, meaning<br />

that no significative activation of<br />

the shielding had been realized. As a<br />

further consequence, it could be<br />

proven that no activation of walls<br />

outside the biological shielding was<br />

in place, just because, due to its<br />

screening effect, the neutron flux<br />

outside the shielding itself was 6 to 7<br />

orders of magnitude lower.<br />

The second step of the characterization<br />

consisted in the assessment of<br />

the surface contamination of the various<br />

areas of the building. These were<br />

separated into three main surfaces:<br />

1) ceiling; 2) surfaces over +6.0 m<br />

level; 3) surfaces below +6.0 m level.<br />

The ceiling was indeed a false ceiling<br />

made of thin aluminium plates; these<br />

could have been contaminated by<br />

tritiated water vapours emerging<br />

from the core once open for refueling<br />

or fuel reshuffling. To investigate this,<br />

the aluminium plates were dismantled,<br />

taken to ground, and analyzed. It<br />

was assumed that, if no contamination<br />

was found, then also the real<br />

ceiling behind it was not contaminated.<br />

This proved indeed to be the<br />

case. Surfaces over +6.0 m were<br />

investigated randomly (Figure 6), by<br />

sampling a given number of points,<br />

quantified basing on statistical considerations.<br />

All surfaces below +6.0 m<br />

were completely measured, both walls<br />

and floors. The measurement technique<br />

consisted in using surface<br />

contamination meters (Berthold<br />

LB165 and LB124), properly cali brated<br />

with large area reference sources, to<br />

sum up count rates over 14C, 60Co,<br />

134Cs, 152Eu and 154Eu. A similar<br />

measurement methodology was successfully<br />

applied for the decommissioning<br />

of the ASTRA research reactor<br />

in Vienna [7]. Background contributions<br />

due to natural radionuclides in<br />

the different materials were subtracted<br />

after having made suitable averages<br />

from surely clean, similar materials<br />

to those which were to be measured<br />

inside the building. As a further, conservative<br />

penalization, it was decided<br />

to attribute to each of the 5 abovementioned<br />

nuclides the whole net<br />

counting over each surface portion<br />

being measured, counting time per<br />

surface element being about 30 seconds<br />

to reach a desired minimum<br />

detectable activity. LB124 hand held<br />

monitor was used for surfaces over<br />

+6.0 m, while LB165 (wheeled monitor<br />

as in Figure 7) was used over all<br />

other surface portions. A special automated<br />

vertical translational sledge<br />

(Figure 8) was used to carry LB165<br />

over the portions of the walls. In case<br />

a given measurement yielded values<br />

above the clearance limits, special<br />

cleaning procedures were to be<br />

adopted until subsequent measurements<br />

proved to be below the limits<br />

| | Fig. 5.<br />

Core drilling of the biological shielding.<br />

| | Fig. 6.<br />

LB124 measurements of selected portions of walls above +6.0 m level.<br />

Decommissioning and Waste Management<br />

The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 7.<br />

LB165 measurement of floor portions.<br />

themselves. No cleaning procedures<br />

were ever necessary and all the surface<br />

portions could be considered not<br />

contaminated and so able to be freely<br />

released.<br />

5 Radiological Characterization<br />

of the Site<br />

The third and last pillar of the decommissioning<br />

of RB3 is the radiological<br />

characterization of the areas of the<br />

site surrounding the building. In this<br />

regard, it is important to mention that<br />

during the operational life of the<br />

plant, given its low power and its<br />

construction features, no radiocativity<br />

discharges were in place and therefore<br />

no environmental analyses were prescribed<br />

by the Regulatory Authority.<br />

Another point worth of mentioning is<br />

that no radiological status of the site<br />

prior to the construction and exercise<br />

of RB3 is known. However, in light of<br />

the graded approach which is going to<br />

be taken for this third pillar by the<br />

Regulatory Authority, given the fact<br />

that no activated materials have been<br />

found and that no activation or contamination<br />

of the building has been<br />

detected, it is decided to base this<br />

characterization upon the measurement<br />

of some selected nuclides in<br />

certain terrain samples (soil) taken<br />

around the area of the RB3 site.<br />

In particular, 12 measurements of<br />

239+240Pu through alpha spectrometry<br />

will be done, together with<br />

25 gamma spectroscopy assessments<br />

for 54Mn, 60Co, 125Sb, 134Cs, 137Cs,<br />

| | Fig. 9.<br />

Collecting soil samples from the RB3 site.<br />

Radionuclide<br />

152Eu and 154Eu. Each terrain<br />

sample will be a parallelepiped of<br />

25x20x10 cm 3 corresponding roughly<br />

to 5 liters of humid soil (Figure 9).<br />

The site will be sampled considering<br />

both near-range and far-range positions<br />

in order to find patterns of radioactivity<br />

correlated with the distance<br />

from the RB3 building, if any at all.<br />

The obtained values will be confronted,<br />

through proper summations,<br />

with the limits for the free release<br />

of nuclear sites prescribed by the<br />

German national law, which correspond<br />

to the radiological nonrelevance<br />

value of 10 microSv/year<br />

to the public [8,9]. The limits for<br />

the above-mentioned isotopes are<br />

reported in Table 2.<br />

References<br />

[1] D.Lgs. 17 marzo 1995, n. 230,<br />

Attuazione delle direttive Euratom<br />

80/836, 84/467, 84/466, 89/618,<br />

90/64, 92/3, 96/29.<br />

[2] D. M. 29 Novembre 2010 Ministero<br />

dello Sviluppo Economico di<br />

Autorizzazione alla Disattivazione<br />

Impianto Nucleare di Ricerca Reattore<br />

RB-3 di Montecuccolino (BO) dell’ENEA.<br />

[3] Radiation Protection 89, Recommended<br />

radiological protection criteria for the<br />

recycling of metals from the<br />

dismantling of nuclear installations,<br />

European Commission, 1998.<br />

| | Fig. 8.<br />

LB165 and its translational sledge to measure wall portions.<br />

Concentration Limit<br />

(Bq/g)<br />

54 Mn 0.09<br />

60 Co 0.03<br />

125 Sb 0.08<br />

134 Cs 0.05<br />

137 Cs 0.06<br />

152 Eu 0.07<br />

154 Eu 0.06<br />

239 Pu 0.04<br />

240 Pu 0.04<br />

| | Tab. 2.<br />

Proposed clearance limits for the free release<br />

of the RB3 site.<br />

[4] Radiation Protection 113,<br />

Recommended radiological protection<br />

criteria for the clearance of buildings<br />

and building rubble from the<br />

dismantling of nuclear installations,<br />

European Commission, 2000.<br />

[5] I. Vilardi, C. M. Castellani, D. M.<br />

Castelluccio, F. Rocchi, Piano di<br />

Caratterizzazione Radiologica di Materiali<br />

provenienti dalla Disattivazione<br />

dell’impianto Nucleare di Ricerca Rb-3<br />

dell’enea sito in Bologna – Montecuccolino<br />

ai Fini del loro Allontanamento,<br />

Convegno Nazionale AIRP 2014, Aosta.<br />

[6] M. Capone, N. Cherubini, A. Compagno,<br />

A. Dodaro, F. Rocchi, The Dismantling of<br />

the Montecuccolino RB3 Research<br />

Reactor: Radiological Characterisation of<br />

Materials for Free Release, Proceedings<br />

of the European Reaserch Reactor<br />

Conference RRFM 2015, Bucharest<br />

19-23 April 2015, 528-537.<br />

[7] F. Meyer, F. Steger, R. Steininger,<br />

Decommissioning of the Astra Research<br />

Reactor – Dismantling the auxiliary<br />

Systems and Clearance and Reuse of the<br />

Buildings, Nuclear Technology &<br />

Radiation Protection, 1/2008, 54-62.<br />

[8] OECD/NEA Status Report, Releasing<br />

the Sites of Nuclear Installations,<br />

NEA Report 6187, 2006.<br />

[9] Bundesgesetzblatt G 5702 Teil I, Bonn<br />

26 July 2001, Nr. 38, 2001.<br />

Authors<br />

F. Rocchi<br />

ENEA FSN/SICNUC/SIN<br />

C. M. Castellani<br />

ENEA IRP<br />

A. Rizzo<br />

ENEA FSN/SICNUC/TNM<br />

Via Martiri di Monte Sole 4<br />

Bologna (BO), Italy<br />

A. Compagno<br />

ENEA FSN/FISS/CRGR<br />

I. Vilardi<br />

ENEA IRP/SFA<br />

Via Anguillarese, 301<br />

00123 S.Maria di Galeria (RM), Italy<br />

R. Lorenzelli<br />

ENEA FSN/SICNUC/SIN<br />

Località Brasimone<br />

40032 Camugnano (BO), Italy<br />

DECOMMISSIONING AND WASTE MANAGEMENT 245<br />

Decommissioning and Waste Management<br />

The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

246<br />

RESEARCH AND INNOVATION<br />

Revised version of a<br />

paper presented at<br />

the Annual Meeting<br />

of Nuclear Technology<br />

(AMNT 2017), Berlin.<br />

Irradiation Tests of a Flat Vanadium Self-<br />

Powered Detector with 14 MeV Neutrons<br />

Prasoon Raj and Axel Klix<br />

Self-powered detector (SPD) represents a class of neutron and gamma monitoring instruments used in the fission<br />

reactor cores worldwide. This detector has inherent advantages of functioning without a bias voltage, simple measurement<br />

scheme, compactness, ease of maintenance, and high reliability. We are studying SPD for application as flux<br />

monitors in the European test blanket modules (TBM) of ITER, fusion reactor under construction in southern France.<br />

This paper presents results of experimental tests performed with 14 MeV neutrons for a flat SPD with vanadium emitter.<br />

Vanadium responds by beta emission from products of reactions (main routes: 51 V (n, γ) 52 V and 51 V (n, p) 51 Ti) with<br />

thermal and fast neutrons. Secondary electrons due to gammas from these reactions and neutron irradiation of<br />

surrounding materials are also important contributors to the signal. Thin foils of emitter, insulator and collector<br />

materials are used to construct the test SPD. The detector is irradiated with short and long pulses of neutrons and is<br />

found to respond in proportion with the incident neutron flux. Further experiments with simplified and better optimized<br />

design of detector are underway for thorough study of the signal-creation mechanism.<br />

1 Introduction<br />

ITER [1] is an experimental fusion<br />

reactor based on tokamak concept,<br />

under construction at St. Paul lez<br />

Durance in southern France. It is an<br />

international project aimed at proving<br />

feasibility of fusion as a large-scale<br />

and carbon-free source of energy. One<br />

of the main scientific goals of this<br />

project will be to test and prove the<br />

concepts of tritium breeding blankets.<br />

Tritium is an important fuel component<br />

for devices based on D-T reaction,<br />

which is being considered as<br />

main reaction for fusion power plants.<br />

Because tritium is a rare element, it is<br />

required to breed it in the fuel cycle of<br />

the reactor. A blanket with lithium<br />

compounds will cover the inner wall<br />

of the plasma vessel. Fusion neutrons<br />

from the plasma will be absorbed by<br />

lithium nuclei, causing reactions to<br />

produce tritium.<br />

There are multiple breeding<br />

blanket designs proposed by scientists.<br />

To determine their efficiencies in<br />

a real fusion environment, test blanket<br />

modules (TBM) based on different<br />

concepts will be inserted into equatorial<br />

ports of ITER for experimental<br />

tests in different operational phases of<br />

ITER. The European Union is going to<br />

test two such concepts, namely the<br />

Helium-Cooled Lead-Lithium (HCLL)<br />

and Helium-Cooled Pebble Bed<br />

(HCPB) TBMs [2]. In the neutronics<br />

experiments, nuclear responses like<br />

tritium production rate, material activation,<br />

nuclear heating etc. are to be<br />

measured and compared with the<br />

calculations. This step will validate<br />

the advanced computational tools<br />

and nuclear data utilized for nuclear<br />

analyses for fusion devices. The neutron<br />

and gamma fluxes are important<br />

quantities to be measured for these<br />

experiments, for which detectors like<br />

neutron activation system, fission<br />

chambers and self-powered detectors<br />

(SPD) are under study.<br />

An SPD is a multi-layered electrical<br />

device, which produces direct current<br />

(DC) signal on irradiation with<br />

neutrons and/or gammas. It can be<br />

preferentially responsive to neutrons<br />

(self- powered neutron detector,<br />

SPND) or gammas (SPGD), or as<br />

it is in most of the cases, to both.<br />

Figure 1 shows a rough sketch of the<br />

cross- section of a traditional detector.<br />

Central material, called emitter produces<br />

fast electrons on irradiation.<br />

These fast electrons can be betas<br />

from the decay of neutron activation<br />

products, or secondary electrons due<br />

to interaction of gammas in the bulk<br />

of the material. They slow down in a<br />

layer of insulation and stop in the<br />

outer electrode called collector. This<br />

electron-movement creates a potential<br />

difference and thus, produces a<br />

current signal proportional to the<br />

incident particle flux. The current due<br />

to beta electrons is “delayed” because<br />

of the half-life of beta-emitters, e.g.<br />

SPND based on Rh, V or Ag emitters.<br />

Whereas that due to gamma-initiated<br />

photoelectric or Compton electrons<br />

is “prompt”, for example Co-based<br />

SPND [3].<br />

An SPD responds in a sophisticated<br />

manner, with multiple factors<br />

contributing to the small current<br />

signals often totaling between 10 -12<br />

and 10 -3 Ampere. Due to its inherent<br />

advantages of simplicity, compactness<br />

and high-reliability, they are highly<br />

desirable for flux monitoring in areas<br />

with restricted access like reactor<br />

cores. At KIT, we are studying SPDs<br />

for application in ITER TBM [4].<br />

Vanadium based flat SPD is being<br />

tested with 14 MeV neutrons, to<br />

understand its behavior towards fast<br />

neutrons expected in fusion environment<br />

and ascertain the feasibility of<br />

its application as flux monitor for<br />

European ITER TBMs.<br />

2 Experimental details<br />

Vanadium is a common emitter for<br />

fission reactor SPNDs. The response<br />

of the detector towards thermal neutrons<br />

is understood well. The material<br />

is relatively inexpensive and easier to<br />

handle. However, due to lower cross<br />

sections the sensitivity of vanadium-<br />

SPND towards fast neutrons reduces<br />

(Figure 2). Commercially available<br />

SPND cannot be directly used for<br />

measurement of fusion neutron<br />

fluxes, going up to approx. 14 MeV in<br />

energy.<br />

Characteristics of the two main<br />

beta- emitters from 51 V (99.75 %<br />

isotopic abundance) in case of fast<br />

neutron irradiation, are reported in<br />

Table 1. Cross-sections of the fast<br />

neutron reactions in 51 V for a pure<br />

| | Fig. 1.<br />

Cross-sectional sketch of a cylindrical SPD showing emitter (green), insulator (dotted white) and<br />

collector (black) layers, with connection to the lead cable and current measurement device.<br />

Research and Innovation<br />

Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 2.<br />

Cross sections of vanadium reactions and photon production under neutron irradiation.<br />

Reaction 51 V (n, p) 51 Ti 51 V (n, γ) 52 V<br />

Threshold Neutron Energy 1.72 MeV 0 MeV<br />

14 MeV Cross-section 30 mb (approx.) 0.6 mb (approx.)<br />

Beta Emitter, Half-life 51 Ti- 5.76 m 52 V- 3.74 m<br />

Average Beta Energy 51 Ti- 0.87 MeV 52 V- 1.07 MeV<br />

SPND Current (14 MeV) 3.46 × 10 -12 A 6.92 × 10 -14 A<br />

SPND Current (TBM) 7.97 × 10 -9 A 3.44 × 10 -8 A<br />

| | Tab. 1.<br />

Beta-emitters and corresponding currents from fast neutron reactions in vanadium based SPND.<br />

14 MeV source are shown. Neglecting<br />

the self-shielding of electrons in emitter<br />

material, effect of other materials<br />

and taking a saturation condition<br />

(considering the short half-lives of<br />

daughter nuclides), one can ascertain<br />

the orders of magnitude of currents<br />

possible with V-SPND, as reported.<br />

For this estimation, vanadium density<br />

of 6.1 g cm -3 , and a typical volume of<br />

1 cm 3 are assumed. For a 14 MeV<br />

neutron source, a flux intensity of<br />

1 × 10 10 cm -2 s -1 is considered, which<br />

is achievable with state of the art<br />

14 MeV neutron generators. For TBM,<br />

activation calculation was done [5]<br />

with the HCLL neutron spectrum<br />

and typical flux intensity (up to 1 ×<br />

10 14 cm -2 s -1 ) using EASY-2007 [6].<br />

With high-sensitivity ammeters,<br />

currents down to the order of 1 ×<br />

10 -14 A can be reliably measured [7].<br />

Values in Table 1 show that a vanadium<br />

emitter based SPND will produce<br />

measurable signals in TBM. Due<br />

to its high neutron threshold energy,<br />

the (n, p) reaction can be utilized to<br />

measure fast neutron flux exclusively.<br />

Fast neutron reactions lead to<br />

high-energy gamma production. This<br />

phenomenon competes with the neutron<br />

absorption reactions (Figure 2).<br />

Photoelectric and Compton electron<br />

emission from emitter causes a prompt<br />

current which is expected to form the<br />

major component of the signal of<br />

V-SPND towards 14 MeV neutrons.<br />

Secondly, vanadium being a medium-<br />

Z nucleus can be a potential<br />

emitter for SPGD also. With optimized<br />

dimensions and choice of collector<br />

material, a vanadium SPD can be<br />

envisaged for monitoring of photon<br />

flux in TBM.<br />

Instead of the usual coaxial type<br />

cylindrical geometry, we designed<br />

our test SPD in sandwich-type flat<br />

geometry. This provides a relatively<br />

higher cross section area to the incident<br />

neutrons, and ease of access for<br />

testing various materials in the same<br />

device. Thin foils (0.5 to 2 mm) of<br />

emitter, insulator and collector are<br />

arranged to form an assembly in an<br />

aluminum case, which also serves as<br />

an electromagnetic shield. Central<br />

conductor of the signal cable is linked<br />

to the emitter plates of the detector.<br />

The collector plates, case and the<br />

cable sheath are shorted and securely<br />

connected to the ground. Schematic<br />

sketch and photograph of the test<br />

detector are shown in Figure 3 (left).<br />

With comparable cross sections of<br />

reactions in different materials, the<br />

insulator and collector materials also<br />

play an important role in SPD<br />

response. Behaviors of different<br />

material combinations are experimentally<br />

tested. Alumina (Al 2 O 3 ) or<br />

beryllia (BeO) is used as insulator and<br />

Inconel-600 or graphite is used as<br />

collector in our experiments. Effects<br />

of the change of geometry and dimensions<br />

are also studied. A Keithley 6485<br />

Picoammeter (sensitivity range -20 fA<br />

to 20 mA) is used as the measuring<br />

device. A low-noise triax cable (Belden<br />

9222) is used to reduce the interferences<br />

in low-current measurement.<br />

The tests are conducted at the<br />

14 MeV neutron generator of Technical<br />

University of Dresden (TUD-NG),<br />

shown in Figure 3 (right). Here,<br />

deuteron beams are impinged on a<br />

tritiated titanium target causing D-T<br />

reaction which leads to production of<br />

neutrons with peak energy of approx.<br />

14 MeV. TUD-NG offers neutron flux<br />

intensities up to 1 × 10 10 cm -2 s -1 . The<br />

detector is placed in front of the<br />

tritium- target of TUD-NG and tested<br />

under different conditions by varying<br />

flux levels and irradiation times.<br />

3 Results<br />

The irradiation tests of flat sandwichtype<br />

vanadium SPD were performed<br />

at TUD-NG, with neutron flux intensities<br />

around 1 × 10 9 cm -2 s -1 . DC<br />

signals in the range of 100 fA to 100 pA<br />

were measured. In Figure 4, a plot<br />

shows variation of SPD signal with<br />

change in neutron flux. The detector<br />

was composed of 1 mm thick layers of<br />

vanadium emitter and Inconel-600<br />

collector. The signal was found to be<br />

proportional to the incident flux, with<br />

approx. 90 pA at the highest flux level.<br />

At low fluxes and low currents,<br />

the measurements have high uncertainties.<br />

Interference from electromagnetic<br />

sources of stray currents,<br />

| | Fig. 3.<br />

(Left) internal design of the sandwich-type flat SPD: (top)- an engineering sketch of the geometry<br />

having sandwich of foils of emitter (green), insulator (grey) and collector (red), and (below) a photograph<br />

of the assembly with vanadium SPD.<br />

(Right) experimental setup showing TUD-NG beamline, tritium target, mounted SPD, and the lead cable.<br />

RESEARCH AND INNOVATION 247<br />

Research and Innovation<br />

Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

RESEARCH AND INNOVATION 248<br />

| | Fig. 4.<br />

Vanadium SPD signal (left Y-axis, red curve) variation with change in neutron flux (right Y-axis, blue<br />

curve) plotted with respect to irradiation time.<br />

currents generated in coaxial cables,<br />

electrostatic effects at the contacts<br />

and degradation of insulation layer<br />

due to radiation, lead to background<br />

currents in the orders of 100 fA. This<br />

makes the measurement of low-level<br />

currents a very challenging task.<br />

The SPD response is often reported<br />

in terms of sensitivity, which is SPD<br />

current per unit of neutron (or<br />

gamma) flux intensity, reported in<br />

units of A cm 2 s. For the vanadium<br />

SPND signal in Figure 4, the sensitivity<br />

lies between 4.48 × 10 -20 A cm 2 s<br />

± 13.4 % (at flux intensity 2.04 ×<br />

10 9 cm -2 s -1 ) and 8.80 × 10 -19 A cm 2 s<br />

± 51.1 % (at flux intensity 6.40 ×<br />

10 5 cm -2 s -1 ).<br />

In another test, a constant-flux<br />

irradiation of around 15 minutes was<br />

done and the TUD-NG was switched<br />

off. This signal is shown in Figure 5. It<br />

was found that the detector current is<br />

dominated by a prompt component<br />

which appeared and disappeared with<br />

neutron flux. The delayed signal is<br />

usually less than 10% of the total<br />

signal. A decay of delayed current was<br />

observed as expected.<br />

There are parasitic beta emission<br />

reactions in insulator, collector and<br />

cable’s central conductor, e.g. 27 Al (n,<br />

p) 27 Mg reaction (half-life~ 9.46 min)<br />

in alumina insulation. Electrons<br />

emitted due to these reactions reduce<br />

the total delayed current. Due to this,<br />

the analysis of decay curve becomes<br />

very complex. After data reduction,<br />

subtraction of background contributions,<br />

and further analysis the major<br />

contribution was found to be from 51 Ti<br />

due to 51 V (n, p) 51 Ti reaction. For<br />

reduction of aforementioned effects<br />

materials with lower total cross<br />

sections of beta emission reactions,<br />

like graphite and beryllia were used as<br />

collector and insulator, respectively.<br />

The change in the signal characteristics<br />

was minimal with these alterations,<br />

leading us to conclude that the<br />

signal was mainly due to reactions in<br />

the vanadium emitter. A prompt current,<br />

makes the detector suitable for<br />

pulsed devices like ITER. However, it<br />

is important to understand the signal<br />

creation mechanism for calibration<br />

and application of the SPD.<br />

The high prompt signal is attributed<br />

to three main reasons. First is<br />

the interaction of photons in the<br />

emitter volume, which release high<br />

energy electrons producing high<br />

positive current. Unlike thermal neutrons,<br />

fast neutrons lead to emission<br />

of higher-energy photons with higher<br />

probability of secondary effects.<br />

Moreover, the photon production<br />

cross section is usually an order or two<br />

higher than the fast neutron reaction<br />

cross sections in materials of detector<br />

and surroundings (Figure 2). Secondly,<br />

the production of charged particles<br />

like protons and alphas in collector<br />

and insulator material (cross sections<br />

of (n, xp) and (n, xα) reactions are<br />

high for 14 MeV neutrons) lead to<br />

further difference of charge between<br />

electrodes and a prompt positive<br />

contribution to the signal. Finally, the<br />

electrical and nuclear effects in<br />

connecting wires and cables make a<br />

small fraction of the positive current<br />

signal<br />

Some of the contributing factors<br />

will be explicitly studied in future<br />

tests. To de-couple the effects of other<br />

materials, a detector with simplified<br />

geometry is under design. An air-insulated<br />

detector with box of collector<br />

material is being constructed. The<br />

material thicknesses are reduced in<br />

order to decrease the gamma interactions.<br />

Improved ways of making<br />

electrical contacts between cable and<br />

emitter are studied. Other less betaactive<br />

materials like niobium are<br />

being considered for collector. Vanadium<br />

detector is also planned to be<br />

optimized for photon response. To this<br />

end, thicker emitters and collectors<br />

with low gamma-activity will be used<br />

to make a test-device which will be<br />

irradiated with high-energy bremsstrahlung<br />

photon source.<br />

4 Conclusions<br />

A flat sandwich-type vanadium SPD<br />

has been constructed, for testing the<br />

feasibility of application of SPDs in<br />

ITER TBMs. Irradiation tests with<br />

14 MeV neutrons at TUD-NG resulted<br />

in current signals in range of 100 fA to<br />

100 pA. The signals are proportional<br />

to the incident neutron flux. Considering<br />

the higher flux intensities up<br />

to 1 × 10 14 cm -2 s -1 and a wider energy<br />

spectrum of neutrons in TBM, studies<br />

show that vanadium SPND is expected<br />

to produce measurable signals in ITER<br />

| | Fig. 5.<br />

Vanadium-SPD signal in a long constant-flux irradiation at TUD-NG showing (prompt and delayed)<br />

currents before, during and after the irradiation.<br />

Research and Innovation<br />

Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

TBM conditions. The high prompt<br />

component of the SPD signal is<br />

attributed to the interaction of high<br />

energy photons which are produced<br />

in the detector and surrounding<br />

materials. Charged particles emitted<br />

in fast neutron reactions and contributions<br />

from wires and signal cable<br />

contribute to the high positive signal.<br />

Parasitic reactions in non-emitter<br />

materials also play an important role.<br />

These effects need to be studied<br />

explicitly and compared for understanding<br />

of the overall currentgeneration<br />

mechanism. Optimization<br />

of design, dimensions and material<br />

combinations is underway to realize<br />

SPD flux monitors for application in<br />

European ITER TBMs.<br />

Acknowledgement<br />

The work leading to this publication<br />

has been funded partially by Fusion<br />

for Energy under the Specific<br />

Grant Agreement F4E-FPA-395-1.<br />

This publication reflects the views<br />

only of the authors, and Fusion for<br />

Energy cannot be held responsible for<br />

any use which may be made of the<br />

infor mation contained therein.<br />

References<br />

[1] ITER Organization – Homepage. [Online].<br />

Available: https://www.iter.org/.<br />

[2] P. Calderoni, Status of the HCLL and<br />

HCPB Test Blanket System instrumentation<br />

development, 21 st Top. Meet.<br />

Technol. Fusion Energy (TOFE), 9-13<br />

Nov. 2014, Anaheim, CA, USA.<br />

[3] N. P. Goldstein and W. H. Todt, A Survey<br />

of Self-Powered Detector - Present and<br />

Future, IEEE Trans. Nucl. Sci., vol. 26,<br />

no. 1, pp. 916–923, 1979.<br />

[4] P. Raj, M. Angelone, U. Fischer, and<br />

A. Klix, Self-powered detectors for test<br />

blanket modules in ITER, in 2016 IEEE<br />

Nuclear Science Symposium, Medical<br />

Imaging Conference and Room- Tem perature<br />

Semiconductor Detector Workshop<br />

(NSS/MIC/RTSD), 2016, pp. 1–4.<br />

[5] M. Angelone, A. Klix, M. Pillon, P.<br />

Batistoni, U. Fischer, and A. Santagata,<br />

Development of self-powered neutron<br />

detectors for neutron flux monitoring in<br />

HCLL and HCPB ITER-TBM, Fusion Eng.<br />

Des., vol. 89, no. 9–10, pp. 2194–2198,<br />

2014.<br />

[6] R. A. Forrest, FISPACT-2007: User<br />

manual, EASY Doc. Ser. UKAEA<br />

FUS 534, 2007.<br />

[7] Low Level Measurements Handbook –<br />

7 th Edition: Precision DC Current,<br />

Voltage, and Resistance Measurements.<br />

Keithley- A Tektronix Company.<br />

Authors<br />

Prasoon Raj<br />

Axel Klix<br />

Institute for Neutron Physics and<br />

Reactor Technology (INR)<br />

Karlsruhe Institute of Technology<br />

(KIT)<br />

Hermann von Helmholtz Platz 1<br />

76344 Eggenstein-Leopoldshafen<br />

(Germany)<br />

RESEARCH AND INNOVATION 249<br />

Nanofluid Applied Thermo-hydrodynamic<br />

Performance Analysis of Square<br />

Array Subchannel Under PWR Condition<br />

Jubair Ahmed Shamim and Kune Yull Suh<br />

1 Introduction Efficient engineered design of heat transfer and fluid flow with enhanced heating or cooling<br />

requires two pivotal aspects that must be taken into consideration for extracting thermal energy from nuclear fission<br />

reactions in order to save energy, reduce process time, raise thermal rating and increase the operating life of a reactor<br />

pressure vessel. Hence, one of the major challenges in designing a new nuclear power plant is the quantification of the<br />

optimal flow of coolant and distribution of pressure drop across the reactor core. While higher coolant flow rates will<br />

lead to better heat transfer and higher Departure from Nucleate Boiling (DNB) limits, it will also result in higher pressure<br />

drop across the core, therefore additional demand of pumping powers as well as larger dynamic loads on the core<br />

components. Thus, thermal hydraulic core analysis seeks to find proper working conditions with enhanced heat transfer<br />

and reduced pressure drop that will assure both safe and economical operation of nuclear plants.<br />

Recently, nanofluid has gained much<br />

renewed attention as a promising<br />

coolant for pressurized water reactors<br />

(PWRs) due to its enhanced thermal<br />

capabilities with least penalty in pressure<br />

drop. The improved heat transfer<br />

of nanofluids results from the fact that<br />

the nanoparticles increase the surface<br />

area and heat capacity of the fluid,<br />

improve the thermal conductivity of<br />

the fluid, cause more collisions and<br />

interactions between the fluid, particles<br />

and surfaces of the flow passages,<br />

and enhance turbulence and mixing<br />

of the fluid.<br />

Pak & Cho [1] experimentally<br />

observed the turbulent friction and<br />

heat transfer of dispersed fluids in a<br />

circular pipe using two different<br />

metallic oxide particles, γ-alumina<br />

(Al 2 O 3 ) and titanium dioxide (TiO 2 )<br />

with mean diameters of 13 and 27 nm,<br />

respectively. The results revealed<br />

that the Nusselt number Nu for the<br />

dispersed fluids increased with<br />

increasing volume concentration as<br />

well as the Reynolds number Re. But<br />

at constant average velocity, the<br />

convective heat transfer coefficient for<br />

the dispersed fluid was 12% less than<br />

that for pure water. They proposed a<br />

new correlation for Nu under their<br />

experimental ranges of volume concentration<br />

(0-3%), Re (10 4 -10 5 ), and<br />

the Prandtl number Pr (6.54-12.33)<br />

for the dispersed fluids γ-alumina<br />

(Al 2 O 3 ) and titanium dioxide (TiO 2 )<br />

particles as<br />

(1)<br />

Xuan and Li [2] observed the flow<br />

and convective heat transfer of the<br />

Cu-water nanofluid flowing through<br />

a straight brass tube of the inner<br />

diameter of 10 mm and the length of<br />

800 mm. They noted that suspended<br />

nanoparticles can remarkably enhance<br />

heat transfer given the velocities.<br />

For instance, the heat transfer<br />

coefficient of nanofluids containing<br />

2.0 vol % Cu nanoparticles was increased<br />

by as much as 40 % compared<br />

to that of water. The conventional<br />

Research and Innovation<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

RESEARCH AND INNOVATION 250<br />

Dittus–Boelter correlation failed to<br />

predict this augmented heat transfer<br />

data for nanofluids. They presented a<br />

new correlation for turbulent flow of<br />

nanofluids inside a tube as<br />

(2)<br />

Maïga et al. [3] numerically investigated<br />

fully-developed turbulent flow<br />

of water/Al 2 O 3 nanofluid through<br />

circular tube using different concentrations<br />

under the constant heat flux<br />

boundary condition. They proposed<br />

the following correlation for 10 4 ≤<br />

Re ≤ 5×10 5 , 6.6 ≤ Pr ≤ 13.9 and 0 ≤<br />

φ ≤ 10%<br />

(3)<br />

Asirvatham et al. [4] reviewed the<br />

published experimental investigations<br />

on convective heat transfer of different<br />

nanofluids.<br />

Despite numerous studies on both<br />

scaled experiments and numerical<br />

modeling on heat transfer enhancement<br />

of nanofluids proliferate over<br />

the past years, most of the test sections<br />

and computational domain were<br />

limited to round pipes. Their simulating<br />

parameters did not reflect the<br />

environment of a nuclear power reactor,<br />

either. Wu and Trupp [5] demonstrated<br />

that flow conditions inside the<br />

fuel rod assembly are quite different<br />

from those in typical pipes. There is<br />

so far no appropriate correlation in<br />

literature that can predict heat transfer<br />

characteristics of nanofluid in a<br />

fuel assembly under PWR operating<br />

condition. Therefore, numerical modeling<br />

has been performed in this study<br />

using a commercial computational<br />

fluid dynamic CFD tool “Star-CCM+<br />

(ver.9.06.011)” to predict heat transfer<br />

and pressure drop more precisely<br />

in a square array subchannel (1.25 ≤<br />

P/D ≤ 1.35) for different volume concentrations<br />

of water/alumina (Al 2 O 3 )<br />

nanofluid (0.5% ≤ φ ≤ 3.0%). Referring<br />

to the Advanced Power Reactor<br />

1400 MWe (APR1400).<br />

Properties<br />

Also, if the slip between the particles<br />

and the continuous phase is trifling,<br />

the flow inside the subchannel may as<br />

well be considered as single phase and<br />

incompressible with constant physical<br />

properties. Both the compression<br />

work and viscous dissipation are<br />

neglected. Under such conditions the<br />

general conservation equations for<br />

mass, momentum and energy can be<br />

written in vector notations:<br />

∇.(ρv) = 0 (4)<br />

∇.(ρvv) = -gradP+μΔ 2 v (5)<br />

∇.(ρvC P T) = ∇.(k gradT) (6)<br />

where v, P and T are fluid velocity<br />

vector, pressure and temperature,<br />

respectively.<br />

2.2 Determination of physical<br />

properties of nanofluid<br />

Determination of physical properties<br />

of nanofluid is key to any nanofluid<br />

research. If the nanoparticles are<br />

assumed to be well dispersed in the<br />

base fluid, the particle concentration<br />

can be considered as constant<br />

throughout the domain and effective<br />

physical properties of mixture can be<br />

evaluated using some classical formulas<br />

well known for two phase fluids<br />

[7]. The following formulas are used<br />

to determine such properties as density,<br />

specific heat, dynamic viscosity<br />

and thermal conductivity.<br />

ρ nf = (1-ϕ)ρ bf + ϕρ P (7)<br />

(C P ) nf = (1-ϕ)(C P ) bf + ϕ(C P ) P (8)<br />

μ nf = (1 + 7.3ϕ + 123ϕ 2 )μ bf (9)<br />

Base Fluid<br />

(Pure Water)<br />

Alumina<br />

Nanoparticles<br />

Density (kg/m 2 ) 734.928 3970<br />

Thermal Conductivity (W/m.K) 0.5701 40<br />

Specific Heat (J/kg. K) 5361.69 880<br />

Dynamics Viscosity (Pa. s) 9.01373E-05 -<br />

| | Tab. 1.<br />

Physical properties of base fluid and alumina nanoparticles.<br />

and later improved by Brinkman [10]<br />

and another by Batchelor [11], these<br />

formulas drastically underestimate<br />

the viscosity of nanofluids. Therefore,<br />

they performed a least-square curve<br />

fitting based on some scarce experimental<br />

data available [12, 13, 14]<br />

which leads to Equation (9). Equation<br />

(10) [7, 15] is introduced for the thermal<br />

conductivity as with the dynamic<br />

viscosity. However, the pressure and<br />

temperature of the above investigations<br />

sizably differ from the operating<br />

condition of a PWR. Since no such<br />

correlation exists for thermophysical<br />

properties of nanofluid applicable to<br />

the operating environment of a PWR it<br />

is assumed that the aforementioned<br />

correlations can also be utilized for<br />

nuclear reactors. Different properties<br />

of base fluid (pure water) and alumina<br />

nanoparticles that have been used in<br />

this study are tabulated in Table 1.<br />

3 Numerical modelling<br />

3.1 Computational domain<br />

The computational domain and<br />

boundaries considered in this study<br />

are shown in Figure 1, which represents<br />

a quarter of a 3-D square array<br />

subchannel created in Star-CCM+.<br />

The diameter of the fuel rod is taken<br />

as 9.5 mm and pitch-to-diameter ratio<br />

P/D of 1.25 and 1.35 are selected for<br />

simulation. The length of the subchannel<br />

is taken as 600 mm which<br />

is long enough to establish a fullydeveloped<br />

turbulent flow at the outlet<br />

under single phase forced convection<br />

condition up to Re = 6×10 5 according<br />

to the following criteria [16]<br />

2 Mathematical modelling<br />

k nf = (1 + 2.72ϕ + 4.97ϕ 2 )k bf (10)<br />

2.1 Governing equations<br />

The term “nanofluid” refers to a twophase<br />

mixture of saturated liquid and<br />

dispersed ultrafine particles of usual<br />

size below 40 nm. However, due to<br />

extremely tiny size of particles, it can<br />

be readily fluidized and thus may be<br />

considered to behave more like a fluid<br />

rather than heterogeneous fluid [6].<br />

Equations (7) and (8) are general<br />

relationships being used in literature<br />

[1, 7, 8] to compute the density and<br />

specific heat for a classical two phase<br />

mixture. Regarding the dynamic<br />

viscosity, Maïga et al. [9] showed that,<br />

albeit several correlations exist to<br />

calculate the dynamic viscosity of<br />

nanofluid as proposed by Einstein<br />

| | Fig. 1.<br />

Computational domain created in Star-CCM+.<br />

Research and Innovation<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

P/D = 1.25<br />

Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid<br />

(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />

6×10 5 7.829 7.963 8.351 9.196<br />

5.098×10 5 6.651 6.766 7.095 7.813<br />

4×10 5 5.219 5.309 5.568 6.130<br />

3×10 5 3.914 3.982 4.176 4.598<br />

| | Tab. 2.<br />

Different inlet velocities, v 0 (m/s) used in simulation.<br />

(11)<br />

l e = EI × D h (12)<br />

where l e is entrance length for fullydeveloped<br />

flow, EI is entrance length<br />

number and Dh is the channel hydraulic<br />

diameter.<br />

3.2 Boundary conditions &<br />

Physics set-up<br />

The coolant enters the subchannel<br />

with a uniform inlet velocity v 0 (m/s)<br />

at the inlet temperature 569 K. Different<br />

values of v 0 for different coolants<br />

that have been used in the simulation<br />

are listed in Table 2. Different properties<br />

of base fluid (pure water)<br />

have been calculated at temperature<br />

569 K and at pressure 155.1375 bar.<br />

At the outlet, a static pressure of<br />

155.1375 bar has been imposed. On<br />

the tube wall, the usual non-slip<br />

conditions with the standard wall<br />

function are considered with a constant<br />

heat flux of 600,000 W/m 2 . The<br />

above parameters and geometric configurations<br />

of the computational<br />

domain are based on the design<br />

features of the APR1400.<br />

The constant density model is chosen<br />

for the material. For turbulence<br />

modeling, the realizable k-ε model<br />

with high y + wall treatment is selected.<br />

Implicit coupled solver with secondorder<br />

upwind discretization scheme in<br />

conjunction with coupled energy<br />

model is implemented which solves<br />

the conservation equations for mass<br />

and momentum simultaneously using<br />

a pseudo time marching approach.<br />

3.3 Turbulence modeling<br />

By studying different literature on<br />

numerical simulation of flow through<br />

a rod bundle for nuclear applications,<br />

it can be concluded that no specific<br />

turbulence model can be regarded as<br />

superior to others for this sort of flow<br />

phenomena. Yadigaroglu et al. [17]<br />

carried out an exhaustive review of<br />

rod bundle numerical simulations<br />

and opined that the gradient transport<br />

models, like the standard k-ε<br />

model, are not capable of predicting<br />

turbulent flow in the narrow gap regions.<br />

Hàzi [18] had demonstrated<br />

that the Reynolds Stress Model (RSM)<br />

could be accurately applied in simulating<br />

the rod bundle geometry. Lee<br />

and Choi [19] also used the RSM turbulence<br />

model to compare the performance<br />

of grid designs between the<br />

small scale vortex flow (SSVF) mixing<br />

vane and the large scale vortex flow<br />

(LSVF) mixing vane. Liu and Ferng<br />

[20] have also adopted RSM turbulence<br />

model to numerically investigate<br />

the effects of different types of<br />

grid (standard grid and split-vane pair<br />

one) on the turbulence mixing and<br />

heat transfer. Palandi et al. [21] have<br />

successfully implemented SST k-ω<br />

model in comparing thermo-hydraulic<br />

performance of nanofluids and<br />

mixing vanes in VVER-440 triangular<br />

array fuel rod bundle. However, application<br />

of RSM turbulence model will<br />

require 50-60% more CPU time per<br />

iteration and 15-20% more memory<br />

usage compared to standard k-ε and<br />

k-ω model.<br />

Recently Conner et al. [22] have<br />

implemented renormalization group<br />

(RNG) k-ε model (Yakhot et al., [23])<br />

in simulation a 5×5 rod bundle with<br />

mixing-vane grid using Star-CCM+.<br />

The applicability of this model to<br />

simulate fuel rod bundles has been<br />

tested and validated by Westinghouse<br />

in their extensive research (Smith et<br />

al., [24]).<br />

Considering the established practice<br />

and computational time required<br />

as discussed above, it can be concluded<br />

that RNG k-ε model will be<br />

sufficient in modeling turbulence for<br />

flow through a rod bundle. However,<br />

in this study, realizable k-ε model<br />

(Shih et al., [25]) has been adopted<br />

for turbulence modeling inside a<br />

square array subchannel since it has<br />

been statistically proved that this<br />

model provides the best performance<br />

among all the k-ε model versions for<br />

separated flows and flows with complex<br />

secondary flow features [26].<br />

The term “realizable” means<br />

that the model satisfies certain mathematical<br />

constraints on the Reynolds<br />

stresses, consistent with the physics<br />

of turbulent flows. Neither the standard<br />

k-ε nor the RNG k-ε model is<br />

realizable.<br />

The modeled transport equation<br />

for k and ε in the realizable k-ε model<br />

are presented by Equation (13) and<br />

Equation (14) respectively:<br />

(13)<br />

and<br />

where,<br />

P/D = 1.35<br />

Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid<br />

(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />

6×10 5 5.826 5.926 6.215 6.843<br />

5.098×10 5 4.950 5.035 5.280 5.814<br />

4×10 5 3.884 3.951 4.143 4.562<br />

3×10 5 2.913 2.963 3.108 3.422<br />

(14)<br />

(15)<br />

(16)<br />

In above equations, G k represents<br />

the generation of turbulence kinetic<br />

energy due to mean velocity gradients,<br />

G b is the generation of turbulence<br />

kinetic energy due to buoyancy, Y M is<br />

the contribution of fluctuating dilatation<br />

in compressible turbulence to<br />

the overall dissipation rate, C 2 and C 1ε<br />

are constants, σ k and σ ε are the<br />

turbulent Prandtl numbers for k and ε<br />

respectively, S k and S ε are user- defined<br />

source terms.<br />

3.4 Convergence of numerical<br />

solution<br />

Another central criteria that must be<br />

satisfied in order to obtain proper<br />

numerical solution is convergence.<br />

The solver needs to be given adequate<br />

iterations so that the problem is converged<br />

and a solution can be treated<br />

as converged if the following criteria<br />

are satisfied [26]:<br />

• The solution no longer changes<br />

with subsequent iterations<br />

• Overall mass, momentum, energy<br />

and scalar balance are achieved<br />

• All equations (momentum, energy<br />

etc.) are obeyed in all cells to a<br />

specific tolerance<br />

RESEARCH AND INNOVATION 251<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

RESEARCH AND INNOVATION 252<br />

In the present study, residuals for<br />

continuity, X & Y-momentum, Z-<br />

momentum and turbulence kinetic<br />

energy are decreased respectively to<br />

an order of 10 -2 , 10 -5 , 10 -2 and 10 -4<br />

| | Fig. 3.<br />

Distribution of wall y + values in case of pure water<br />

with Re=6×10 5 (P/D =1.35)<br />

| | Fig. 2.<br />

Convergence of mass flow averaged temperature at outlet (P/D = 1.35) for pure water at corresponding<br />

inlet Re = 6×10 5 .<br />

after 30,000 iterations and also a<br />

monitor is created to check how values<br />

for mass flow averaged temperature at<br />

outlet is converging and it is observed<br />

that after 30,000 iterations these<br />

values do not change significantly<br />

with further iterations. A typical plot<br />

of mass flow averaged temperature at<br />

outlet for pure water at inlet Re =<br />

6×10 5 is shown in Figure 2.<br />

3.5 Wall y + values<br />

The accurate calculations of y + value<br />

in the near-wall region, which is a<br />

measure of non-dimensional distance<br />

from the wall to the first mesh node<br />

(based on local cell fluid velocity), are<br />

of paramount importance to the success<br />

of any simulation. In order to use<br />

a wall function approach properly for<br />

a particular turbulence model with<br />

confidence, the y+ values should be<br />

within a certain range.<br />

In the present study, standard wall<br />

function is used in conjunction with<br />

realizable k-ε model and high-y + wall<br />

treatment in which the near-wall cell<br />

centroid are anticipated to be placed<br />

in the log-law region with a value<br />

30 ≤ y + ≤ 100. Results of performed<br />

simulations demonstrate that the<br />

wall y + values for different cases are<br />

within this specified range. A pictorial<br />

representation of wall y + in case<br />

of pure water with Re = 6×10 5<br />

(P/D = 1.35) is shown in Figure 3.<br />

4 Code validation<br />

4.1 Mesh convergence test<br />

Since the accuracy of finite volume<br />

method is directly related to the<br />

quality of discretization used, it is<br />

instrumental to select an optimized<br />

mesh size that will take into account<br />

both resolution of mesh structure and<br />

as well as computational time and<br />

cost.<br />

In the present study, different<br />

mesh settings are selected as presented<br />

in Table 3 and values of<br />

numerically obtained Nu are compared<br />

against an existing correlation<br />

for square array subchannel and for<br />

pure water as presented by Equation<br />

(17) through Equation (19) to check<br />

mesh convergence for computational<br />

domain with P/D =1.35. Results are<br />

plotted in Figure 4 which clearly<br />

states that a mesh setting with base<br />

size 0.7 mm, no. of prism layer 2,<br />

prism layer thickness 0.3mm and<br />

prism layer stretching 3.7 will be<br />

sufficient to produce Nu within<br />

reasonable deviation compared to<br />

the theoretical prediction made by<br />

correlation.<br />

Nu = ψ(Nu ∞ ) c.t. (17)<br />

where,<br />

(Nu ∞ ) c.t. = 0.023 Re 0.8 PR 0.4 (18)<br />

for square array with 1.05 ≤ P/D ≤<br />

1.9 and for pure water, Presser [27]<br />

suggested:<br />

(19)<br />

Base Size<br />

(mm)<br />

No.<br />

Prism Layers<br />

Stretching Thickness<br />

(mm)<br />

Nu<br />

(Star-CCM+)<br />

Nu<br />

(Presser)<br />

Deviation<br />

(%)<br />

0.5 5 1.5 0.7 742.940 -35.051<br />

0.6 4 1.5 0.5 862.627 -16.313<br />

0.7 3 3.8 0.4 933.92 1003.35 -7.434<br />

0.6 2 3.7 0.3 972.102 -3.214<br />

0.7 2 3.7 0.3 1010.57 0.714<br />

| | Fig. 4.<br />

Mesh convergence test with different mesh settings.<br />

| | Tab. 3.<br />

Different mesh settings used to check mesh convergence.<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

| | Fig. 5.<br />

Validation of numerical model against correlation for P/D =1.25.<br />

4.2 Validation of numerical<br />

model<br />

Since the ultimate test of any numerical<br />

simulation is the validation of<br />

results against well-known experimental<br />

data, the model under consideration<br />

in the present study has<br />

been validated against correlation of<br />

Presser for square array and pure<br />

water as presented by Equation (17)<br />

through Equation (19). Results are<br />

plotted in Figure 5 and Figure 6<br />

which demonstrates that there is<br />

an excellent agreement between<br />

numerical data and theoretical<br />

prediction for the specified range of<br />

inlet Re.<br />

4.3 Validation of turbulence<br />

model for nanofluid<br />

Despite in the present study it is<br />

assumed that nanofluid would behave<br />

as a single-phase homogeneous fluid<br />

and hence, all of the general conservation<br />

equations of mass, momentum<br />

and energy can directly be applied in<br />

case of nanofluid, however, a successful<br />

comparison of numerical Nu obtained<br />

realizable k-ε model has been<br />

carried out against both empirical<br />

correlation and experimental data of<br />

Pak & Cho [1] for turbulent flow<br />

inside a round pipe of inside diameter<br />

10.66 mm using alumina nanofluid<br />

(φ=2.78%) as coolant for inlet Re<br />

spanning from 5.03×10 4 to 1.48×10 4 .<br />

The results are plotted in Figure 7<br />

which clearly delineates that this<br />

model can perform quite satisfactorily<br />

with nanofluids.<br />

5 Numerical results<br />

and discussion<br />

5.1 Temperature<br />

Temperature profile along the centerline<br />

of subchannel (P/D =1.25) for<br />

different coolants at inlet Re = 6×10 5<br />

are illustrated in Figure 8 from which<br />

it is clear that there is a steady increase<br />

in the coolant temperature due to absorption<br />

of heat while flowing through<br />

the subchannel and bulk temperature<br />

of nanofluid is decreased with the increasing<br />

particle volume concentration.<br />

Numerically obtained fluid average<br />

temperature (in case<br />

of pure water at P/D =1.25 and<br />

inlet Re = 6×10 5 ) at different axial<br />

locations within the subchannel is<br />

compared against the theoretical<br />

predictions from energy balance<br />

according to equation (20) [28] and<br />

results are tabulated in Table 4.<br />

<br />

(20)<br />

The analogy shows that maximum<br />

deviation between numerically obtained<br />

axial temperature and theoretical<br />

prediction is less than 0.6%.<br />

5.2 Velocity<br />

Development of axial velocity along<br />

the centerline of subchannel (P/D<br />

| | Fig. 6.<br />

Validation of numerical model against correlation for P/D =1.35.<br />

=1.25) for different coolants at inlet<br />

Re = 6×10 5 is presented in Figure 9<br />

which clearly states that fullydeveloped<br />

velocity profile occurs<br />

approximately after z=0.3 m and if<br />

the current models are implemented<br />

to evaluate physical properties of<br />

nanofluid, development of velocity<br />

| | Fig. 7.<br />

Validation of turbulence model against Pak & Cho’s correlation.<br />

| | Fig. 8.<br />

Temperature along centerline of subchannel at Re = 6×10 5 .<br />

RESEARCH AND INNOVATION 253<br />

Axial Position<br />

(m)<br />

Average Bulk Fluid Temperature T m (K) %<br />

of Deviation<br />

Start-CCM+ Energy Balance<br />

0 569 569 0.000<br />

0.15 569.2431 570.6885 0.2532<br />

0.30 570.1277 572.3771 0.3929<br />

0.45 571.2205 574.0656 0.4956<br />

0.60 572.4116 575.7542 0.5805<br />

| | Tab. 4.<br />

Comparison of numerically obtained axial temperature against theoretical predictions for pure water<br />

(P/D =1.25 and inlet Re = 6×10 5 ).<br />

| | Fig. 9.<br />

Velocity along centerline of subchannel at Re = 6×10 5 .<br />

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RESEARCH AND INNOVATION 254<br />

| | Fig. 10.<br />

Pressure along centerline of subchannel at Re = 6×10 5 .<br />

profile is not affected by the inclusion<br />

of nanoparticles. From Figure 9, it can<br />

also be seen that there is an increase in<br />

the velocity magnitude due to growth<br />

of hydrodynamic boundary layer as<br />

coolant flows from inlet towards<br />

outlet. The inclusion of higher volume<br />

concentration of nanoparticles also<br />

augments the magnitude of axial<br />

velocity as seen in Figure 9. It can be<br />

explained from the fact that since<br />

with the rise of volume concentration<br />

the viscosity of the nanofluid is also<br />

aggrandized, hence to a maintain<br />

constant value of Reynolds number Re<br />

at the inlet of the channel, velocity<br />

magnitude should be increased too<br />

according to equation (21) if the other<br />

properties remain constant:<br />

<br />

(21)<br />

5.3 Pressure<br />

A plot of static pressure along the<br />

centerline of the subchannel (P/D<br />

=1.25) for different coolants at inlet<br />

Re = 6×10 5 is shown in Figure 10<br />

which depicts that there is an increase<br />

in axial pressure with the inclusion of<br />

nanoparticles which is expected due<br />

to higher viscosity and density as the<br />

particle volume concentration is increased.<br />

5.4 Nu and h Constant Inlet Re<br />

Convective heat transfer is studied<br />

with Star-CCM+ for pure water and<br />

different concentrations of alumina<br />

nanofluid according to Equation (22)<br />

and Equation (23) respectively. Values<br />

of Nu are evaluated at the outlet of the<br />

subchannel to assure fully-developed<br />

turbulent flow condition.<br />

<br />

<br />

(22)<br />

(23)<br />

where, q '' is the constant heat flux<br />

(W/m 2 ), k is thermal conductivity<br />

(W/m 2 .K), D h is hydraulic diameter<br />

(m), and T w and T m are wall and mean<br />

bulk fluid temperature (K) respectively.<br />

Numerical results of Nu and h for<br />

subchannel with different pitch-todiameter<br />

(P/D) ratio are presented<br />

through Figure 11 to Figure 14<br />

respectively and percentage of convective<br />

heat transfer increment for<br />

different nanofluid coolants are<br />

documented in Table 5.<br />

From the results, it is obvious that<br />

the convective heat transfer coefficient<br />

is remarkably increased with the<br />

increment of nanoparticle volume<br />

concentration and in case of 3.0 %<br />

volume concentration, convective<br />

heat transfer is increased above<br />

22.0 % compared to pure water.<br />

5.5 Comparison of Numerical<br />

Results against Correlations<br />

In case of nanofluid with volume<br />

concentration, φ =3.0% numerical<br />

results for Nu are compared against<br />

two well cited correlations of Pak &<br />

Cho [1] and Maïga et al. [3] as shown<br />

in Figure 15 (a) & (b) and an attempt<br />

has been made whether results of<br />

present study can be represented by<br />

either of these two correlations.<br />

The results revealed that Pak<br />

and Cho correlation severely underestimates<br />

the numerical results for<br />

Nu in subchannel and deviation lies<br />

between 17 to 22 percent subject to<br />

inlet Re and P/D.<br />

Regarding correlation of Maïga<br />

et al., it shows better approximation<br />

compared to correlation of Pak & Cho.<br />

Nevertheless, this correlation underestimates<br />

the numerical results for the<br />

| | Fig. 11.<br />

Comparison of Nu for different coolants in subchannel (P/D 1.25).<br />

| | Fig. 12.<br />

Comparison of Nu for different coolants in subchannel (P/D 1.35).<br />

| | Fig. 13.<br />

Comparison of h for different coolants in subchannel (P/D 1.25).<br />

| | Fig. 14.<br />

Comparison of h for different coolants in subchannel (P/D 1.35).<br />

Research and Innovation<br />

Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

range 5×10 5 ≤ Re ≤ 6×10 5 and overestimates<br />

for 3×10 5 ≤ Re ≤ 4×10 5<br />

and deviations are between 0.54 and<br />

6.66 percent depending on inlet Re<br />

and P/D.<br />

5.6 Pressure drop<br />

While nanofluid enhances the convective<br />

heat transfer, the fluid itself<br />

P/D = 1.25<br />

Inlet Re Increment of h (%)<br />

φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />

6×10 5 2.75 9.62 22.46<br />

5.098×10 5 2.75 9.58 22.37<br />

4×10 5 2.72 9.51 22.16<br />

3×10 5 2.74 9.42 21.89<br />

| | Tab. 5.<br />

Heat transfer increment (%) for different nanofluid coolants.<br />

also gets heavier compared to pure<br />

water. Hence, it is of utmost importance<br />

to determine the amount of<br />

pressure drop for the effective application<br />

of nanofluid coolant in nuclear<br />

reactors since it is directly related to<br />

the pumping power required. In this<br />

study, pressure drop along the center<br />

line of the subchannel is evaluated for<br />

different coolants and results are presented<br />

in Figure 16 (a) & (b). Percentage<br />

of pressure drop increment is<br />

documented in Table 6.<br />

The results shows that pressure<br />

drop is significantly increased with<br />

the augmentation of particle volume<br />

concentration which in turn increases<br />

the pumping power. For nanofluid<br />

P/D = 1.55<br />

Inlet Re Increment of h (%)<br />

φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />

6×10 5 2.72 9.56 22.35<br />

5.098×10 5 2.72 9.51 22.26<br />

4×10 5 2.71 9.44 22.01<br />

3×10 5 2.69 9.40 21.87<br />

RESEARCH AND INNOVATION 255<br />

(a) P/D = 1.25<br />

| | Fig. 15.<br />

Comparison of numerical Nu against different correlations.<br />

(b) P/D = 1.35<br />

(a) P/D = 1.25<br />

| | Fig. 16.<br />

Comparison of pressure drop for different coolant.<br />

(b) P/D = 1.35<br />

P/D = 1.25<br />

Inlet Re Increment of ∆p (%)<br />

φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />

6×10 5 6.22 21.53 56.60<br />

5.098×10 5 5.82 21.17 56.62<br />

4×10 5 5.79 21.79 56.02<br />

3×10 5 5.24 21.65 55.83<br />

P/D = 1.35<br />

Inlet Re Increment of ∆p (%)<br />

φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />

6×10 5 5.82 20.94 56.37<br />

5.098×10 5 5.74 21.29 56.08<br />

4×10 5 5.46 20.90 55.10<br />

3×10 5 5.62 20.88 55.82<br />

| | Tab. 6.<br />

Pressure drop increment (%) for different nanofluid coolants.<br />

Research and Innovation<br />

Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

RESEARCH AND INNOVATION 256<br />

with φ=3.0%, pressure drop increment<br />

is about 56% higher compared<br />

to that of pure water.<br />

However, the typical nanoparticle<br />

loading in PWR coolant should be<br />

less than 1.0 vol %. At such lower<br />

con centration, nanofluid properties<br />

are almost similar to that of pure<br />

water and the rise in viscosity as well<br />

as pressure drop will be negligible too.<br />

The present study also portrays that<br />

pressure drop is approximately 20 %<br />

at 1.5 vol. % of nanoparticle concentration<br />

which can also be treated as<br />

tolerable.<br />

The convective heat transfer coefficient<br />

at such low concentration of<br />

nanofluid is yet to be improved due to<br />

higher turbulence produced near the<br />

grid spacers by the presence of nanoparticles<br />

in the base fluid. Since it is<br />

quite difficult to take into account<br />

such effects in numerical simulation,<br />

further experimental investigation is<br />

required for quantification of heat<br />

transfer increment aroused from the<br />

presence of nanoparticles near the<br />

spacer grids.<br />

6 Proposed new<br />

correction factor<br />

Finally, a multiple regression analysis<br />

is performed with numerical results to<br />

propose a new correction factor, β for<br />

the existing correlation of square<br />

array subchannel with pure water as<br />

suggested by Presser [27] so that Nu<br />

for nanofluid coolant can be approximated<br />

in such geometry. Based on<br />

regression results, β can be expressed<br />

as follows:<br />

β = 1 + 0.0247ϕ 1.39 (24)<br />

Nu for nanofluid can be calculated as<br />

follows:<br />

Nu nf = β*(Nu Presser ) Water (25)<br />

The validity of above correlation is for<br />

3×10 5 ≤ Re ≤ 6×10 5 ; 0.847 ≤ Pr ≤<br />

1.011; 1.25 ≤ P/D ≤ 1.35 and 0.5% ≤<br />

φ ≤ 3.0% in case of square array<br />

subchannel.<br />

7 Chemical and physical<br />

stability of nanofluid<br />

Albeit nnanofluid can readily boost<br />

the heat transfer capability of PWR<br />

coolant, there is still no satisfactory<br />

explanation proposed regarding the<br />

prevention of clustering in nanoparticle<br />

suspensions. Agglomeration<br />

in nanofluids containing oxide nanoparticles<br />

can be reduced remarkably<br />

by adjusting the pH to form electric<br />

changes on particle surface so that<br />

they repel each other [29]. However,<br />

the typical values of pH should be<br />

such that nanofluid itself becomes not<br />

corrosive and it should be agreeable<br />

with same allowable pH range of<br />

nuclear reactor, since altering the<br />

PWR coolant chemistry is not a viable<br />

option. Besides, use of surfactants are<br />

also not recommended since it may<br />

undergo severe radiolysis inside the<br />

reactor core during operation.<br />

Hence, issues concerning chemical<br />

and physical stabilities of nanofluid<br />

has yet to be resolved prior to utilizing<br />

nanofluid as a promising coolant in<br />

PWRs to achieve both extended life<br />

time of associated equipment and<br />

higher thermal efficiency.<br />

8 Conclusion<br />

Thermo- and hydrodynamic characteristics<br />

of water/alumina nanofluid<br />

have been studied in a square array<br />

subchannel featuring the pitch-todiameter<br />

ratios of 1.25 and 1.35 under<br />

the steady-state, incompressible,<br />

single- phase turbulent flow condition.<br />

Numerical results have been compared<br />

against correlations in the<br />

literature and the following conclusions<br />

can be drawn.<br />

• Convective heat transfer is increased<br />

with increasing volume<br />

concentration of water/alumina<br />

nanofluid given the inlet Reynolds<br />

number.<br />

• The convective heat transfer increment<br />

of nanofluid is obtained at<br />

the expense of increased pressure<br />

drop and hence, larger pumping<br />

power is required. Therefore,<br />

nano fluid as PWR coolant can be<br />

only be implemented in reality if<br />

the replacement of reactor coolant<br />

pump is a feasible option compared<br />

to higher power gained from<br />

increased nanofluid heat transfer.<br />

Acknowledgements<br />

This work was supported by the<br />

National Research Foundation of Korea<br />

(NRF) grant funded by the Korean<br />

Government (MSIP) under Grant No.<br />

2008-0061900 and partly supported<br />

by the Brain Korea 21 Plus under<br />

Grant No. 21A20130012821.<br />

Nomenclature<br />

∆p Pressure Drop Pa<br />

ρ Density kg/m 3<br />

v Flow Velocity m/s<br />

f Friction Factor -<br />

L Length of Flow Channel m<br />

le Entrance Length m<br />

EI Entrance Length Number -<br />

Dh Hydraulic Diameter m<br />

μ Dynamic Viscosity N.s/m 2<br />

Re Reynolds Number -<br />

Nu Nusselt Number -<br />

Pr Prandtl Number -<br />

Pe Peclet Number -<br />

h<br />

Convective Heat Transfer<br />

CoefficientW/m 2 .K<br />

k Thermal Conductivity W/m.K<br />

C p Specific Heat J/kg.K<br />

T m Bulk Temperature of Fluid K<br />

T w<br />

Surface Temperature<br />

of Heater Rod<br />

P Rod Pitch m<br />

D Rod Diameter m<br />

Q Total Heat Input W<br />

q” Heat Flux W/m 2<br />

φ<br />

ṁ Mass Flow Rate kg/sec<br />

Subscript<br />

nf<br />

bf<br />

P<br />

Volume Concentration<br />

of Nanoparticles %<br />

Nanofluid<br />

Basefluid<br />

Particle<br />

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Research and Innovation<br />

Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

9. Maïga S, Nguyen CT, Galanis N, Roy G,<br />

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turbulent flows: Model development<br />

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Druckverlust an Reaktorbrennelementen<br />

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Rahimi M. An Experimental Study on<br />

Nanofluids Convective Heat Transfer<br />

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[cited November 29].<br />

Authors<br />

Jubair Ahmed Shamim<br />

Department of Nuclear<br />

Engineering<br />

Seoul National University<br />

Seoul 08826, ROK<br />

Kune Yull Suh<br />

Seoul National University<br />

1 Gwanak Ro, Gwanak Gu<br />

Seoul 08826, ROK<br />

257<br />

KTG INSIDE<br />

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Technik und dem Betrieb tätig. Meine Aufgaben bezogen sich auf<br />

den Bau, Betrieb und Rückbau von Kernkraftwerken. Während<br />

der nunmehr 27 Jahre Tätigkeit für PreussenElektra habe ich<br />

Aufgaben in den Kernkraftwerken und der Zentralorganisation wahrgenommen.<br />

Seit 2014 führe ich das Geschäftsführungsressort Technik<br />

und Betrieb. Ich war 13 Jahre im KTA tätig und 5 Jahre Mitglied<br />

der deutschen Reaktorsicherheitskommission. Seit 4 Jahren engagiere<br />

ich mich als Governor bei der WANO – World Association for<br />

Nuclear Operators – weltweit.<br />

Während meines bisherigen beruflichen Werdegangs war und<br />

ist der sichere, umweltverträgliche und wirtschaftliche Betrieb der<br />

Kernkraftwerke mein prioritäres Anliegen. Die Kernkraft hat mich in<br />

meinem ganzen Berufsleben fasziniert und die Faszination hält trotz<br />

aller Rückschläge und der teilweise schwierigen Randbedingungen<br />

für die Kernenergie in Deutschland an.<br />

Seit 1991 bin ich Mitglied in der KTG und nunmehr schon seit<br />

8 Jahren im Vorstand, jetzt als Schatzmeister. Ziel meiner erneuten<br />

Kandidatur ist, die KTG als Interessengemeinschaft aller in der<br />

Kerntechnik Tätigen und von ihr faszinierten Mitgliedern mit meinem<br />

Wissen und meiner beruflichen Erfahrung zu unterstützen sowie die<br />

Wissensübertragung und den Erfahrungsaustausch zu erhalten.<br />

Wie politisch gewollt, sollten wir Kerntechniker den sicheren Betrieb<br />

bis zum Laufzeitende und den Rückbau unserer Kernkraftwerke<br />

in Deutschland mit Ehre abschließen. Kein Grund mit Blick auf das<br />

Erreichte der letzten 50 Jahre nicht stolz sein zu dürfen!<br />

Jörg Starflinger<br />

Prof. Dr.-Ing. (51), Stuttgart<br />

Zur Person<br />

Nach dem Studium des Maschinenbaus an<br />

der Ruhr-Universität Bochum (RUB) mit<br />

Schwerpunkt Energietechnik Promotion<br />

im Jahr 1997 am Lehrstuhl für Nukleare<br />

und neue Energiesysteme der RUB, Prof.<br />

Dr.-Ing. H. Unger. 1998 Wechsel als Nachwuchswissenschaftler zum<br />

Forschungszentrum Karlsruhe, heute Karlsruhe Institut für Technologie.<br />

Themenschwerpunkte: Wasserstofferzeugung bei schweren<br />

Unfällen in Leichtwasserreaktoren und Kreislaufsimulation von<br />

innovativen Reaktorkonzepten. 2006 Leiter der Gruppe „Kraftwerkstechnik“<br />

am Institut für Kern- und Energietechnik (IKET), Prof. Dr.-Ing.<br />

T. Schulenberg, in der innovative Kernkraftwerkskonzepte mit überkritischem<br />

Wasser von mehreren Doktoranden untersucht wurden.<br />

2010 Ruf an die Universität Stuttgart zum ordentlichen Professor des<br />

Lehrstuhls für Kerntechnik und Reaktorsicherheit und Leiter des<br />

Instituts für Kernenergetik und Energiesysteme (IKE). Neben der<br />

Lehre im Bereich Kerntechnik Schwerpunkte in der Reaktorsicherheitsforschung,<br />

z.B. in der Modellentwicklung zur Beschreibung der<br />

späten Phase von Kernschmelzunfällen in Leichtwasserreaktoren<br />

KTG Inside


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

und auf dem Gebiet innovativer Sicherheitssysteme, z.B. der passiven<br />

Lagerbeckenkühlung mit Wärmerohren (Heat pipes) und nachrüstbaren<br />

Nachwärmeabfuhrsystemen mit überkritischem CO 2 als<br />

Arbeitsmittel.<br />

Zur Wahl als Vorstandsmitglied<br />

Ich engagiere mich in der KTG auf dem Gebiet des Kompetenzerhalts<br />

und der Kompetenzförderung. Den von Dr. Wolfgang Steinwarz ins<br />

Leben gerufenen, sehr erfolgreichen Workshop „Kompetenzerhalt in<br />

der Kerntechnik“ habe ich verantwortlich übernommen und möchte<br />

ihn in den kommenden Jahren weiterführen. Dr. Steinwarz steht uns<br />

auch als Ruheständler dankenswerterweise als Jurymitglied weiter<br />

zur Seite. Die Umbenennung in „Young Scientists Workshop“ soll eine<br />

Öffnung zu kerntechnisch verwandten Forschungsthemen, beispielsweise<br />

„Kerntechnik und Gesellschaft“, symbolisieren.<br />

Durch meine Mitarbeit im KTG-Vorstand als Vorstandsmitglied<br />

möchte ich einen Strategieentwicklungsprozess anstoßen, der<br />

mittelfristig eine genügende Anzahl an jungen hochqualifizierten<br />

und motivierten Personen für die zukünftigen spannenden und<br />

herausfordernden nationalen und internationalen kerntechnischen<br />

Aufgaben sicherstellt. Für den Kompetenzerhalt und die Nachwuchsförderung<br />

bieten die KTG und Ihre Mitglieder sowie unsere Tagung<br />

„Annual Meeting on Nuclear Technology“ die ideale Plattform.<br />

Walter Tromm<br />

Dr.-Ing. (58), Stutensee<br />

Zur Person<br />

Maschinenbaustudium an der Uni (TH)<br />

Karlsruhe mit dem Studienschwerpunkt<br />

Kerntechnik und dort Promotion zum Thema<br />

„Experimentelle Untersuchungen zum Nachweis<br />

der langfristigen Kühlbarkeit von<br />

Kernschmelzen“. Seit 1988 am damaligen Forschungszentrum<br />

Karlsruhe, heute Karlsruher Institut für Technologie, angestellt und<br />

schwerpunktmäßig mit Reaktorsicherheitsfragen bei auslegungsüberschreitenden<br />

Störfällen beschäftigt. Von 1998 bis 1999 Gastwissenschaftler<br />

am Europäischen Gemeinschaftsforschungszentrum<br />

in Ispra (Italien) tätig.<br />

Seit 2002 Programmbevollmächtigter in der Programmleitung<br />

Nukleare Sicherheitsforschung des FZK bzw. heute Nukleare<br />

Entsorgung, Sicherheit und Strahlenforschung des KIT; stellvertretender<br />

Leiter seit 2007 wurde seit 2010 Programmleiter. 2014 im<br />

geschäftsführenden Ausschuss des Bereichs Maschinenbau und<br />

Elektrotechnik des KIT berufen. Seit 2015 darüber hinaus Sprecher<br />

des vom KIT neu eingerichteten Kompetenzzentrums Rückbau und<br />

seit 2017 Vorsitzender des Kompetenzverbundes Kerntechnik.<br />

Tätig in nationalen und internationalen Gremien, bei der OECD/<br />

NEA der deutsche Repräsentant des Nuclear Science Committee, bei<br />

der IAEA in der Technical Working Group Light Water Reactors und<br />

Mitglied im Governing Board der EU-SNETP Plattform. Weiterhin<br />

innerhalb des VDI Vorsitzender des Fachausschusses Kraftwerkstechnik.<br />

Seit 2016 Leiter des neu gegründeten Kompetenz-Cluster<br />

Rückbau, der die Expertise im Rückbau mehrerer Länder zusammenführt.<br />

Zur Wahl als stellvertretender Vorsitzender<br />

Die Bundesregierung hat 2011 nach den Ereignissen in dem<br />

Kernkraftwerk Fukushima Daii-chi in Japan entschieden, aus der<br />

Stromproduktion mittels Kernkraft auszusteigen. In den nächsten<br />

4 Jahren werden die letzten Kernkraftwerke in Deutschland<br />

abgeschaltet. Diesen Ausstieg nach wie vor so sicher wie möglich<br />

mitzugestalten ist eine der Aufgaben, die die in der deutschen<br />

Kerntechnik arbeitenden Ingenieure und Naturwissenschaftler<br />

haben. International und auf europäischer Ebene wird jedoch Kernenergie<br />

langfristig weiterhin genutzt. Auch für den Industriestandort<br />

Deutschland und für den Erhalt von Arbeitsplätzen ist der Export von<br />

Komponenten für kerntechnische Anlagen nach wie vor bedeutsam.<br />

Ebenfalls werden der Rückbau der Kernkraftwerke und die Endlagerfrage<br />

die Gesellschaft noch über Jahrzehnte beschäftigen. Der<br />

Ausstieg aus der Stromproduktion durch Kernenergie darf daher<br />

nicht bedeuten, sich von den entsprechenden kerntechnischen<br />

Kompetenzen in der Industrie, den Behörden und den Universitäten<br />

und Forschungszentren zu verabschieden. In den Bereichen Reaktorsicherheit,<br />

Rückbau, Endlagerung, Strahlenschutz und Krisenmanagement<br />

sind diese Kompetenzen auch weiterhin gefragt. In<br />

Europa stammen 27 % der Stromproduktion aus Kernkraftwerken.<br />

Zur kompetenten Bewertung kerntechnischer Einrichtungen innerhalb<br />

Europas und zur kritischen Begleitung internationaler Entwicklungen<br />

sind eine enge Zusammenarbeit auf nationaler, europäischer<br />

und internationaler Ebene unerlässlich. Deshalb sehe ich als eine<br />

der Hauptaufgaben der KTG den Erhalt der kerntechnischen<br />

Kompetenzen in allen genannten Bereichen.<br />

259<br />

KTG INSIDE<br />

Herzlichen<br />

Glückwunsch<br />

April <strong>2018</strong><br />

97 Jahre wird<br />

2. Prof. Dr. Albert Ziegler, Karlsbad<br />

87 Jahre werden<br />

9. Dr. Klaus Penndorf, Geesthacht<br />

11. Hubert Bairiot, Mol/B<br />

19. Dr. Klaus Einfeld, Murnau<br />

28. Dipl.-Ing. Rudolf Eberhart, Burgdorf<br />

85 Jahre wird<br />

6. Ing. Reinhard Faulhaber, Köln<br />

84 Jahre wird<br />

22. Dipl.-Ing. Gert Slopianka,<br />

Gorxheimeral<br />

83 Jahre werden<br />

3. Dipl.-Psych. Georg Sieber,<br />

München<br />

5. Prof. Dr. Hans-Henning Hennies,<br />

Karlsruhe<br />

19. Dr. Ernst Müller, Rösrath<br />

19. Dr. Gottfried Class,<br />

Eggenstein-Leopoldshafen<br />

21. Dipl.-Ing. Walter Jansing,<br />

Bergisch Gladbach<br />

30. Dr. Friedrich-Wilhelm Heuser,<br />

Overath<br />

82 Jahre werden<br />

4. Helmut Kuhne, Neunkirchen<br />

6. Dipl.-Ing. Hans Pirk, Rottach-Egern<br />

10. Dipl.-Ing. Franz Stockschläder,<br />

Bad Bentheim<br />

11. Dipl.-Ing. Bernhard-F. Roth,<br />

Eggenstein-Leopoldshafen<br />

24. Dipl.-Ing. Horst Schott, Overath<br />

81 Jahre werden<br />

7. Dipl.-Ing. Helmut Adam, Neuenhagen<br />

13. Dr. Martin Peehs, Bubenreuth<br />

80 Jahre werden<br />

4. Prof. Dr. Klaus Kühn, Clausthal- Zellerfeld<br />

5. Dr. Hans Fuchs, Gelterkinden/CH<br />

9. Dr. Carl Alexander Duckwitz, Alzenau<br />

28. Prof. Dr. Georg-Friedrich Schultheiss,<br />

Lüneburg<br />

79 Jahre wird<br />

8. Dr. Siegbert Storch, Aachen<br />

78 Jahre wird<br />

18. Dipl.-Ing. Norbert Granner,<br />

Bergisch Gladbach<br />

77 Jahre werden<br />

17. Dipl.-Phys. Ernst Robinson, Gehrden<br />

28. Dr. Ludwig Richter, Hasselroth<br />

KTG Inside


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

260<br />

NEWS<br />

Wenn Sie keine<br />

Erwähnung Ihres<br />

Geburtstages in<br />

der <strong>atw</strong> wünschen,<br />

teilen Sie dies bitte<br />

rechtzeitig der KTG-<br />

Geschäftsstelle mit.<br />

KTG Inside<br />

Verantwortlich<br />

für den Inhalt:<br />

Die Autoren.<br />

Lektorat:<br />

Sibille Wingens,<br />

Kerntechnische<br />

Gesellschaft e. V.<br />

(KTG)<br />

Robert-Koch-Platz 4<br />

10115 Berlin<br />

T: +49 30 498555-50<br />

F: +49 30 498555-51<br />

E-Mail: s.wingens@<br />

ktg.org<br />

www.ktg.org<br />

76 Jahre werden<br />

9. Prof. Dr. Hans-Christoph Mehner,<br />

Dresden<br />

27. Dr. Dieter Sommer, Mosbach<br />

27. Dr. Jürgen Wunschmann, Eggenstein<br />

29. Dr. Klaus-Detlef Closs, Karlsruhe<br />

75 Jahre werden<br />

15. Dr. Werner Dander, Heppenheim<br />

18. Dipl.-Betriebsw. Uwe Janßen,<br />

Weinheim<br />

18. Dipl.-Ing. Victor Luster, Bamberg<br />

26. Ing. Helmut Schulz, Kürten<br />

70 Jahre werden<br />

6. Dr. Wolfgang Tietsch, Mannheim<br />

9. Ing. Herbert Moryson, Essen<br />

22. Dr. Heinz-Dietmar Maertens, Arnum<br />

26. Dr. Rainer Heibel, Ness Neston/GB<br />

27. Ulrich Wimmer, Erlangen<br />

65 Jahre werden<br />

10. Dipl.-Phys. Harold Rebohm, Berlin<br />

24. Dipl.-Phys. Michael Beczkowiak,<br />

Karben<br />

60 Jahre werden<br />

4. Dipl.-Ing. Holger Bröskamp,<br />

Höhnhorst<br />

4. Dipl.-Ing. (FH) Franz Xaver Pirzer,<br />

Schwandorf<br />

50 Jahre werden<br />

16. Rainer Bezold, Dormitz<br />

16. Dr. Matthias Messer, Tetbury/GB<br />

30. Dr. Christian Raetzke, Leipzig<br />

Mai <strong>2018</strong><br />

94 Jahre wird<br />

22. Prof. Dr. Fritz Thümmler, Karlsruhe<br />

90 Jahre wird<br />

10. Dr. Heinz Büchler, Sankt Augustin<br />

89 Jahre wird<br />

31. Dipl.-Ing. Werner-P. Kürsten,<br />

Mannheim<br />

88 Jahre wird<br />

9. Dr. Hans-Jürgen Hantke, Kempten<br />

85 Jahre werden<br />

4. Dr. Klaus Wiendieck, Baden-Baden<br />

25. Dr. Reinhold Mäule, Walheim<br />

25. Georg von Klitzing, Bonn<br />

84 Jahre werden<br />

11. Dr. Eckhart Leischner, Rodenbach<br />

14. Dr. Alexander Warrikoff, Frankfurt/M.<br />

26. Dr. Günter Kußmaul, Manosque/F<br />

83 Jahre werden<br />

1. Dr. Willi Bermel, Jülich<br />

8. Dipl.-Ing. Klaus Wegner, Hanau<br />

22. Dr. Heinz Vollmer, Lampertheim<br />

28. Dipl.-Ing. Anton Zimmermann,<br />

Hamburg<br />

29. Dipl.-Ing. Karlheinz Orth,<br />

Marloffstein<br />

82 Jahre werden<br />

3. Ewald Jurisch, Erlangen<br />

10. Dr. Peter Reinke, Röttenbach<br />

18. Dipl.-Ing. Gerhard Lorenz, Bochum<br />

81 Jahre werden<br />

1. Prof. Dr. Dietrich Munz,<br />

Graben-Neudorf<br />

3. Dipl.-Ing. Harald Enderlein, Karlsruhe<br />

6. Dr. Peter Strohbach, Mainaschaff<br />

7. Prof. Dr. Werner Lutze,<br />

Chevy Chase/USA<br />

20. Dr. Norbert Krutzik, Frankfurt/M.<br />

26. Dipl.-Ing. Rüdiger Müller, Heidelberg<br />

27. Dr. Johannes Wolters, Düren<br />

28. Dipl.-Ing. Heinz E. Häfner, Bruchsal<br />

80 Jahre werden<br />

12. Dr. Herbert Finnemann, Erlangen<br />

13. Dipl.-Ing. Otto A. Besch, Geesthacht<br />

13. Dr. Heinrich Werle,<br />

Karlsdorf-Neuthard<br />

16. Dr. Hans-Dieter Harig, Hannover<br />

21. Dr. Hans Spenke, Bergisch Gladbach<br />

79 Jahre werden<br />

4. Dipl.-Ing. Norbert Albert, Ettlingen<br />

5. Dr. Wolfgang Voigts, Linkenheim<br />

27. Prof. Dr. Dietrich Kirsch<br />

78 Jahre werden<br />

11. Dr. Andreas Hölzler, Schwaig<br />

15. Dipl.-Phys. Ludwig Aumüller,<br />

Freigericht<br />

18. Dr. Karl Schulte, Köln<br />

24. Dipl.-Ing. Herbert Krinninger,<br />

Bergisch Gladbach<br />

77 Jahre werden<br />

8. Prof. Dr. Helmut Alt, Aachen<br />

12. Dipl.-Ing. Dieter Rohde, Mannheim<br />

16. Dr. Jürgen Baier, Höchberg<br />

76 Jahre werden<br />

5. Hans-Bernd Maier, Aschaffenburg<br />

9. Dr. Egbert Brandau, Alzenau<br />

11. Dr. Erwin Lindauer, Köln<br />

17. Dr. Heinz-Peter Holley, Forchheim<br />

18. Dipl.-Ing. Josef Koban, Buckenhof<br />

28. Dipl.-Ing. Wolf-Dieter Krebs,<br />

Bubenreuth<br />

75 Jahre werden<br />

3. Dipl.-Ing. Hans Lettau, Effeltrich<br />

14. Dr. Helmut-K. Hübner, Bruchsal<br />

20. Dipl.-Ing. Dietmar Bittermann, Fürth<br />

22. Dr. Wolfgang Schütz, Bruchsal<br />

23. Dipl.-Ing. Max Heller, Uttenreuth<br />

24. Dipl.-Ing. Rudolf Weh,<br />

Stephanskirchen<br />

27. Dr. Kurt Fischer, Erlangen<br />

65 Jahre werden<br />

2. Dipl.-Ing. Marc Winter, Veitshöchheim<br />

3. Dipl.-Ing. Karl-Heinz Wiening,<br />

Herzogenaurach<br />

5. Michael Klein, Großenwörden<br />

16. Ing. grad. Eckhard Raabe, Geiselbach<br />

21. Dipl.-Ing. (FH) Reinhold Horstmann,<br />

Erlangen<br />

27. Dipl.-Ing. (FH) Ulrich Hudezeck,<br />

Nürnberg<br />

60 Jahre wird<br />

23. Dr. Hans-Josef Zimmer, Steinfeld<br />

50 Jahre werden<br />

10. Dr. Astrid Petersen, Hamburg<br />

20. Dipl.-Ing. (FH) Jürgen Bruder,<br />

Gundremmingen<br />

Die KTG gratuliert ihren Mitgliedern<br />

sehr herzlich zum Geburtstag und wünscht ihnen weiterhin alles Gute!<br />

Top<br />

IAEA Expands International<br />

Cooperation on Small,<br />

Medium Sized or Modular<br />

Nuclear Reactors<br />

(iaea) The International Atomic<br />

Energy Agency (IAEA) is launching an<br />

effort to expand international cooperation<br />

and coordination in the design,<br />

development and deployment of<br />

small, medium sized or modular<br />

reactors (SMRs), among the most promising<br />

emerging technologies in<br />

nuclear power.<br />

Significant advances have been<br />

made on SMRs, some of which will use<br />

pre-fabricated systems and components<br />

to shorten construction schedules<br />

and offer greater flexibility and<br />

affordability than traditional nuclear<br />

power plants. With some 50 SMR concepts<br />

at various stages of development<br />

around the world, the IAEA is forming<br />

a Technical Working Group (TWG) to<br />

guide its activities on SMRs and provide<br />

a forum for Member States to<br />

share infor mation and knowledge,<br />

IAEA Deputy Director General Mikhail<br />

Chudakov said.<br />

“Innovation is crucial for nuclear<br />

power to play a key role in de carbonising<br />

the energy sector,” Chudakov,<br />

who heads the IAEA Department of<br />

Nuclear Energy, said at a conference<br />

on SMRs in Prague on 15 February.<br />

News


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

“Many Member States that are<br />

operating, expanding, introducing or<br />

considering nuclear power are quite<br />

keen on the development and<br />

deployment of SMRs.”<br />

Global interest in SMRs is growing.<br />

SMRs have the potential to meet the<br />

needs of a wide range of users and to<br />

be low carbon replacements for ageing<br />

fossil fuel fired power plants. They<br />

also display enhanced safety features<br />

and are suitable for non-electric applications,<br />

such as cooling, heating and<br />

water desalination. In addition, SMRs<br />

offer options for remote regions with<br />

less developed infrastructure and for<br />

energy systems that combine nuclear<br />

and alternative sources, including<br />

renewables.<br />

The first three advanced SMRs are<br />

expected to begin commercial operation<br />

in Argentina, China and the<br />

Russian Federation between <strong>2018</strong> and<br />

2020. SMR development is also well<br />

advanced in about a dozen other<br />

countries.<br />

The TWG, comprising some 20<br />

IAEA Member States and international<br />

organizations, is scheduled to<br />

meet for the first time on 23-26 April<br />

at the IAEA’s headquarters in Vienna.<br />

It is part of an expanding suite of<br />

services the IAEA offers Member<br />

States on this emerging nuclear power<br />

technology. These include an SMR<br />

computer simulation programme to<br />

help educate and train nuclear professionals;<br />

a methodology and related<br />

IT tool for training in assessing the<br />

reactor technology of different SMRs;<br />

and the SMR Regulators’ Forum.<br />

The forum, set up in 2015, enables<br />

discussions among Member States and<br />

other stakeholders to share SMR<br />

regulatory knowledge and experience.<br />

It contributes to enhancing safety by<br />

identifying and resolving issues that<br />

may challenge regulatory reviews of<br />

SMRs and by facilitating robust and<br />

thorough regulatory decisions.<br />

Responding to requests from<br />

Member States in Europe, the IAEA<br />

recently launched a project to build<br />

regional capacities for making knowledgeable<br />

decisions on SMRs, including<br />

technical assessments for SMRs<br />

that are commercially available for<br />

near term deployment. The two-year<br />

project seeks to contribute to meeting<br />

growing European demand for<br />

flexible sources of electricity that do<br />

not release greenhouse gases. Its first<br />

meeting will be held on 13-15 March<br />

at the IAEA in Vienna.<br />

An expeditious deployment of<br />

SMRs faces challenges, including the<br />

need to develop a robust regulatory<br />

| | IAEA Expands International Cooperation on Small, Medium Sized or Modular Nuclear Reactors.<br />

framework, new codes and standards,<br />

a resilient supply chain and human<br />

resources. And although SMRs require<br />

less upfront capital per unit, their<br />

electricity generating cost will<br />

probably be higher than that of large<br />

reactors. Their competitiveness must<br />

be weighed against alternatives and<br />

be pursued through economies of<br />

scale. Detailed technical information<br />

on SMRs under construction or design<br />

can be found at the IAEA’s Advanced<br />

Reactor Information System.<br />

“Realistically, we could expect the<br />

first commercial SMR fleet to start<br />

between 2025 and 2030,” said Hadid<br />

Subki, Scientific Secretary of the TWG<br />

and a Team Leader in SMR Technology<br />

Development at the IAEA. “We<br />

trust this new Technical Working<br />

Group will help further the advancement<br />

of SMR technology and guide<br />

the Agency in its programmes and<br />

projects in this field.”<br />

| | (18791436), www.iaea.org<br />

World<br />

Poll Shows Local Residents<br />

Support Poland’s Plans for<br />

First Nuclear Plant<br />

(nucnet) A poll carried out for Poland’s<br />

PGE EJ1, the company in charge of the<br />

country’s first nuclear power station<br />

project, has shown that 67% of residents<br />

in areas around the proposed<br />

site in northern Poland support the<br />

potential construction of a nuclear<br />

power station in their region.<br />

PGE said a poll was carried out<br />

in November and December 2017<br />

in three municipalities, Choczewo,<br />

Gniewino, Krokowa, all close to<br />

Poland’s Baltic coast in the northern<br />

province of Pomerania.<br />

According to PGE, local residents<br />

indicated they are in favour of the<br />

project because of the development<br />

and job opportunities it could bring to<br />

their regions. The poll showed 49% of<br />

respondents expect cheaper electricity<br />

to be one of the benefits from a<br />

nuclear station, while 35 % expect<br />

local infrastructure development.<br />

In April 2017, PGE began environmental<br />

and site selection surveys at<br />

two locations – Lubiatowo-Kopalino in<br />

the municipality of Choczewo and<br />

Żarnowiec in the municipality of<br />

Korkowa.<br />

The studies aim to determine the<br />

potential impact of the project on both<br />

the environment and local residents.<br />

An initial round of environmental<br />

studies has already been carried out at<br />

both locations.<br />

The Polish government has not<br />

made a final decision about the<br />

country’s nuclear programme, with<br />

the deadline being pushed back<br />

several times. According to latest<br />

reports, a decision is now expected in<br />

mid-<strong>2018</strong>.<br />

| | pgeej1.pl<br />

SKB, Sweden: Two Statements<br />

on the Spent Fuel Repository<br />

(skb) The answer was a clear yes in<br />

SSM’s statement to the Government<br />

on SKB’s system for final disposal of<br />

spent nuclear fuel. The Land and<br />

Environment Court was also positive<br />

| | Aerial photo of the planned site of the Spent Fuel Repository (centre)<br />

at Forsmark. The picture is a photomontage. Illustration: Phosworks.<br />

261<br />

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262<br />

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*)<br />

Net-based values<br />

(Czech and Swiss<br />

nuclear power<br />

plants gross-based)<br />

1)<br />

Refueling<br />

2)<br />

Inspection<br />

3)<br />

Repair<br />

4)<br />

Stretch-out-operation<br />

5)<br />

Stretch-in-operation<br />

6)<br />

Hereof traction supply<br />

7)<br />

Incl. steam supply<br />

8)<br />

New nominal<br />

capacity since<br />

January 2016<br />

9)<br />

Data for the Leibstadt<br />

(CH) NPP will<br />

be published in a<br />

further issue of <strong>atw</strong><br />

BWR: Boiling<br />

Water Reactor<br />

PWR: Pressurised<br />

Water Reactor<br />

Source: VGB<br />

in several important respects but calls<br />

for more documentation on the<br />

copper canisters.<br />

The Swedish Radiation Safety<br />

Authority (SSM) has reviewed SKB’s<br />

applications under the Nuclear Activities<br />

Act and recommends the Government<br />

to grant a licence for a final<br />

repository for spent nuclear fuel in<br />

Forsmark and an encapsulation plant<br />

in Oskarshamn.<br />

The statement from the Land and<br />

Environment Court (MMD) is also<br />

positive in several important respects.<br />

The court says yes to the issues<br />

relating to the Forsmark site, the rock,<br />

the buffer and the environmental<br />

impact statement. The court also<br />

approves the encapsulation plant and<br />

increased capacity in the interim<br />

storage facility Clab. However, the<br />

court wants SKB to present more<br />

documentation on the properties of<br />

the canister and safety in the<br />

long term. Furthermore, it wants<br />

an investigation of the issue of responsibility<br />

after closure, which has also<br />

been requested by the munici pality.<br />

We can conclude that we have<br />

not been able to answer the court’s<br />

questions regarding the copper<br />

canister fully. At the same time, the<br />

Government’s expert authority SSM<br />

wrote in its statement that SKB has<br />

the potential to meet the legislative<br />

requirements on safe final disposal,<br />

says SKB’s managing director Eva<br />

Halldén in a comment.<br />

SKB will provide documentation<br />

That the two authorities have come<br />

to such different conclusions is in<br />

part due to the fact that they have<br />

tried the applications under different<br />

legislations, SSM under the Nuclear<br />

Activities Act and MMD under the<br />

Environmental Code. They also have<br />

different licensing procedures. SSM<br />

grants a licence in several steps with<br />

continuous updates of the safety<br />

analysis. But the court must say yes<br />

or no based on the currently available<br />

documentation.<br />

The issue now lies with the<br />

Ministry of the Environment and<br />

Energy for further investigation and<br />

SKB is working to develop the documentation<br />

on the canister required<br />

by the court.<br />

This is material that we have<br />

planned to produce for the preliminary<br />

safety analysis. The difference<br />

now is that we will prioritise the work<br />

Operating Results November 2017<br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated. gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

OL1 Olkiluoto BWR FI 910 880 720 660 116 6 744 904 253 976 759 100.00 94.12 99.88 92.69 100.75 92.46<br />

OL2 Olkiluoto BWR FI 910 880 720 662 904 5 794 145 243 611 284 100.00 79.69 99.64 78.73 101.18 79.43<br />

KCB Borssele PWR NL 512 484 720 367 102 3 021 010 157 825 451 99.78 74.15 99.78 74.55 99.75 72.16<br />

KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0<br />

KKB 2 Beznau 7) PWR CH 380 365 720 276 072 2 646 900 130 879 056 100.00 87.20 100.00 86.71 100.93 86.16<br />

KKG Gösgen 7) PWR CH 1060 1010 720 768 486 7 788 300 304 398 935 100.00 92.37 99.99 91.99 100.69 91.66<br />

KKM Mühleberg BWR CH 390 373 720 278 340 2 838 690 124 050 935 100.00 92.24 99.85 91.61 99.12 90.80<br />

CNT-I Trillo PWR ES 1066 1003 720 764 776 7 740 744 238 234 461 100.00 91.36 99.95 91.07 99.24 90.09<br />

Dukovany B1 PWR CZ 500 473 720 362 651 2 456 677 108 267 051 100.00 62.82 100.00 62.46 100.74 61.29<br />

Dukovany B2 PWR CZ 500 473 720 360 040 2 950 413 104 273 041 100.00 75.24 100.00 74.70 100.01 73.61<br />

Dukovany B3 PWR CZ 500 473 655 314 334 2 623 607 102 248 463 90.97 75.75 86.99 65.95 87.32 65.46<br />

Dukovany B4 PWR CZ 500 473 361 174 635 2 371 933 102 900 084 50.14 69.14 48.36 59.29 48.51 59.18<br />

Temelin B1 PWR CZ 1080 1030 720 781 214 8 664 341 106 292 500 100.00 100.00 99.96 99.96 100.47 100.08<br />

Temelin B2 PWR CZ 1080 1030 720 787 897 6 819 241 100 683 563 100.00 78.34 100.00 78.01 101.32 78.77<br />

Doel 1 PWR BE 454 433 720 325 983 3 277 563 133 890 536 100.00 90.76 99.47 90.23 99.32 89.84<br />

Doel 2 PWR BE 454 433 720 328 778 3 268 119 131 921 768 100.00 91.40 99.71 91.03 100.27 89.30<br />

Doel 3 PWR BE 1056 1006 0 0 6 732 621 251 169 221 0 78.97 0 78.79 0 79.12<br />

Doel 4 PWR BE 1084 1033 720 773 286 7 054 678 253 727 128 100.00 83.08 97.81 82.31 98.06 80.49<br />

Tihange 1 PWR BE 1009 962 158 124 135 2 815 111 290 078 185 21.98 36.57 17.32 35.78 17.03 34.79<br />

Tihange 2 PWR BE 1055 1008 720 766 981 6 637 622 248 156 690 100.00 82.44 100.00 78.63 101.62 78.83<br />

Tihange 3 PWR BE 1089 1038 701 759 129 8 614 260 268 094 957 97.37 99.76 96.66 99.69 96.71 98.57<br />

Operating Results January <strong>2018</strong><br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability Energy utilisation<br />

[%] *) [%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

KBR Brokdorf DWR 1480 1410 744 957 521 957 521 341 149 580 100.00 100.00 93.97 93.97 86.59 86.59<br />

KKE Emsland 4) DWR 1406 1335 744 1 010 637 1 010 637 336 333 920 100.00 100.00 100.00 100.00 96.55 96.55<br />

KWG Grohnde DWR 1430 1360 744 977 338 977 338 367 604 917 100.00 100.00 94.85 94.85 91.28 91.28<br />

KRB C Gundremmingen 4) SWR 1344 1288 744 982 159 982 159 321 562 051 100.00 100.00 100.00 100.00 97.58 97.58<br />

KKI-2 Isar DWR 1485 1410 744 1 082 908 1 082 908 342 681 231 100.00 100.00 99.98 99.98 97.73 97.72<br />

KKP-2 Philippsburg 1,2,4) DWR 1468 1402 744 1 062 603 1 062 603 356 230 119 100.00 100.00 99.92 99.92 96.06 96.06<br />

GKN-II Neckarwestheim DWR 1400 1310 744 1 006 200 1 006 200 321 129 334 100.00 100.00 99.40 99.39 96.80 96.80<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

differently and complete it faster<br />

than what was planned, says Helene<br />

Åhsberg, SKB’s project manager for<br />

the licensing process.<br />

No referendum<br />

Östhammar Municipality planned to<br />

hold a referendum on the final repository<br />

on March 4. But at a meeting in<br />

the municipal council in the end of<br />

January, it was decided to cancel the<br />

referendum.<br />

| | (18791534), www.skb.se<br />

Yucca Mountain:<br />

Can the US Finally End<br />

the $12 Billion Impasse?<br />

(nucnet) A US federal advisory panel<br />

recently took a step in what could be a<br />

lengthy process to determine if a deep<br />

geological nuclear waste repository<br />

should finally be built at Yucca Mountain,<br />

a project that has been on the<br />

drawing board since the 1970s at a<br />

cost of around $ 12 bn (€ 9.7 bn).<br />

The panel held a meeting to receive<br />

input on reconstructing an electronic<br />

library for documents needed to<br />

decide on the US Department of<br />

Energy’s Yucca licence application.<br />

The meeting, at the Nuclear Regulatory<br />

Commission’s headquarters in<br />

Maryland, came one week after<br />

another development: the White<br />

House pledged $120m of funding in<br />

its 2019 federal budget proposal to<br />

restart licensing for the Yucca site,<br />

north of Las Vegas in Nevada, and<br />

to establish an interim storage programme<br />

to address the growing<br />

stockpile of nuclear waste produced<br />

by nuclear plants across the nation.<br />

After decades of wrangling, could<br />

the US finally be on course to resolve<br />

the question of what to do with<br />

the high-level nuclear waste from<br />

the nation’s 99 commercial nuclear<br />

reactors?<br />

| | www..energy.gov<br />

US Nuclear Industry Calls<br />

for Advanced Reactor Fuel<br />

Cycle Infrastructure<br />

(nucnet) The US Nuclear Energy<br />

Institute has warned that preparations<br />

should begin now to develop a<br />

national fuel cycle infrastructure to<br />

support the operation of advanced<br />

reactors that are expected to begin<br />

deployment in the 2020s and 2030s.<br />

The Washington-based nuclear<br />

industry lobby group said interest in<br />

the development of advanced nuclear<br />

reactor designs has been increasing in<br />

recent years. Many of these designs<br />

will require uranium fuel that is<br />

enriched to a higher degree than<br />

in the current worldwide fleet of lightwater<br />

reactors. Fuel for advanced<br />

reactors, enriched in U-235 to<br />

between 5% and 20%, is called<br />

high-assay low-enriched uranium<br />

(HALEU).<br />

Some of the advanced-performance<br />

fuels being developed for use<br />

with the existing reactor fleet also will<br />

require HALEU. However, there are no<br />

US-based facilities that manufacture<br />

HALEU on a commercial scale. While<br />

small quantities of HALEU materials<br />

may be obtained on an interim basis<br />

by “blending down” existing government<br />

stocks of surplus high-enriched<br />

uranium (HEU), those HEU materials<br />

are limited in supply and not readily<br />

available, the NEI said.<br />

“Thus, for the long-term operation<br />

of advanced reactors, as well as for<br />

advanced fuels in existing reactors, a<br />

robust new infrastructure for HALEU<br />

fuel manufacture is needed.”<br />

An NEI white paper says establishing<br />

such a capability will better<br />

position the US to advance nuclear<br />

safety and non-proliferation policies<br />

around the world, while helping to<br />

ensure a robust commercial industry<br />

domestically in the decades ahead.<br />

On the other hand, “if the United<br />

States and its allies have to depend on<br />

foreign, state-owned enterprises to<br />

meet fuel needs, it will be in a much<br />

weaker position to influence these<br />

policies globally”, the paper says.<br />

| | Details online:<br />

http://bit.ly/2FnZwOF<br />

Reactors<br />

IAEA Sees Safety Commitment<br />

at Spain’s Almaraz<br />

Nuclear Power Plant<br />

(iaea) An International Atomic Energy<br />

Agency (IAEA) team of experts said<br />

the operator of Spain’s Almaraz<br />

Nuclear Power Plant demonstrated a<br />

commitment to the long-term safety of<br />

the plant and noted several good practices<br />

to share with the nuclear industry<br />

globally. The team also identified areas<br />

for further enhancement.<br />

The Operational Safety Review<br />

Team (OSART) today concluded an<br />

18-day mission to Almaraz, whose<br />

two 1,050-MWe pressurized-water<br />

reactors started commercial operation<br />

in 1983 and 1984, respectively.<br />

Centrales Nucleares Almaraz-Trillo<br />

(CNAT) operates the plant, located<br />

about 200 km southwest of Madrid.<br />

OSART missions aim to improve<br />

operational safety by objectively<br />

assessing safety performance using<br />

the IAEA’s safety standards and proposing<br />

recommendations for improvement<br />

where appropriate. Nuclear<br />

power generates more than 21 per<br />

cent of electricity in Spain, whose<br />

seven operating power reactors all<br />

began operation in the 1980s.<br />

“The team saw notable achievements<br />

made by Almaraz in recent<br />

years, such as implementing a comprehensive<br />

management system, as<br />

well as significant equipment renewal<br />

plans, to establish safety as the<br />

overriding priority at the plant,” said<br />

Team Leader Peter Tarren, Head of the<br />

IAEA’s Operational Safety Section.<br />

“We found that people at every<br />

level were willing to discuss their<br />

work and how they might learn from<br />

this OSART mission. They want to<br />

keep enhancing the safety and<br />

reliability of Almaraz.”<br />

The 14-member team comprised<br />

experts from Brazil, Bulgaria, France,<br />

Germany, Mexico, the Russian Federation,<br />

Sweden, United Arab Emirates,<br />

the United Kingdom and the United<br />

States of America, as well as three<br />

IAEA officials.<br />

The review was the 200th OSART<br />

mission conducted by the IAEA since<br />

the service was launched in 1982. It<br />

covered the areas of leadership and<br />

management for safety; training<br />

and qualification; operations; maintenance;<br />

technical support; operating<br />

experience; radiation protection;<br />

chemistry; emergency preparedness<br />

and response; accident management;<br />

human, technology and organizational<br />

interactions and long-term<br />

operation.<br />

The team identified a number of<br />

good practices that will be shared<br />

with the nuclear industry globally,<br />

including:<br />

The use of a film-forming amine<br />

compound to significantly reduce<br />

the transport of potential corrosive<br />

products to the steam generators.<br />

The use of a cross-functional<br />

indicator to show the cumulative<br />

effect of equipment status and<br />

planned activities for daily operations.<br />

The installation of a centralized<br />

vacuum system for cleaning, decontaminating<br />

and discharging liquid<br />

waste into the plant´s disposal system.<br />

The mission made a number of<br />

recommendations to improve operational<br />

safety, including:<br />

The plant should implement<br />

further actions related to management,<br />

staff and contractors to enforce<br />

standards and expectations related<br />

to industrial safety.<br />

263<br />

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264<br />

NEWS<br />

The plant should take measures<br />

to reinforce and implement standards<br />

to enhance the performance of reactivity<br />

manipulations in a deli berate<br />

and carefully-controlled manner.<br />

The plant should improve the<br />

support, training and documented<br />

guidance for Severe Accident Management<br />

Guideline users in order to<br />

mitigate complex severe accident<br />

scenarios.<br />

The team provided a draft report of<br />

the mission to the plant’s management.<br />

The plant management and the<br />

Nuclear Safety Council (CSN), which<br />

is responsible for nuclear safety<br />

oversight in Spain, will have the<br />

opportunity to make factual comments<br />

on the draft. These will be<br />

reviewed by the IAEA and the final report<br />

will be submitted to the<br />

Government of Spain within three<br />

months.<br />

The plant management said it<br />

would address the areas identified<br />

for enhancement and requested a<br />

follow-up OSART mission in about<br />

18 months.)<br />

| | (18791443), www.iaea.org<br />

Tianwan-3 Passes Commissioning<br />

Tests at 100% Power<br />

(nucnet) The Tianwan-3 nuclear<br />

reactor unit in Jiangsu province,<br />

northeastern China, has successfully<br />

operated for 100 hours at 100% of its<br />

design power level without interruption,<br />

Russian state nuclear corporation<br />

Rosatom said.<br />

Rosatom said the 990-MW VVER<br />

V-428M unit, which started to deliver<br />

electrical energy to the grid on<br />

30 December 2017, has undergone a<br />

series of tests during the 100-hour<br />

operation period required by regulators<br />

before giving green light for<br />

commercial operation.<br />

Construction of Tianwan-3 began<br />

in December 2012. The Tianwan<br />

| | Swiss regulator approves safety case for restart of Beznau-1 (Photo: Axpo).<br />

nuclear station is the largest economic<br />

cooperation project between Russia<br />

and China, an earlier statement had<br />

said.<br />

Tianwan-1 and -2, also VVER<br />

V-428M units, began commercial<br />

operation in 2007. The Tianwan-4<br />

VVER V-428M unit is also under construction<br />

by Russia while Tianwan-5<br />

and -6 will be indigenous Generation<br />

II+ CNP-1000 units.<br />

| | en.cnnc.com.cn<br />

Swiss Regulator Approves<br />

Safety Case for Restart of<br />

Beznau-1<br />

(nucnet) Switzerland’s Federal<br />

Nuclear Safety Inspectorate, ENSI,<br />

has given the go-ahead for the restart<br />

of the Beznau-1 nuclear unit after<br />

approving the safety case presented<br />

by owner Axpo following the discovery<br />

in 2015 of flaw indications in<br />

the reactor pressure vessel (RPV).<br />

ENSI said in a statement that<br />

Axpo had carried out “extensive<br />

investigations and analyses” to<br />

demonstrate that the RPV is safe.<br />

Materials testing has shown<br />

that agglomerates in the RPV do not<br />

affect its key properties and structural<br />

integrity analysis has shown that<br />

the RPV does not contain any flaws<br />

that could lead to its failure. “IRSN<br />

is satisfied that work has been done<br />

to all appropriate national and international<br />

standards,” the statement<br />

said.<br />

Axpo said the safety case for<br />

Beznau-1, the world’s oldest commercial<br />

nuclear plant still in operation,<br />

corroborates earlier assessments<br />

and investigations, and validates the<br />

existing safety margin for the safe<br />

operation of the plant for 60 years.<br />

Operator KKB will now begin the<br />

return to service process with the<br />

plant expected to be operating at full<br />

load by the end of March <strong>2018</strong>.<br />

In December 2015 Axpo submitted<br />

a roadmap ENSI detailing plans for<br />

further investigations of flaw indications<br />

in the RPV. During a scheduled<br />

outage that began in May 2015,<br />

inspections of the RPV registered<br />

findings at some points in the base<br />

material of the RPV indicating<br />

“ minimal irregularities in the fabrication<br />

process”, Axpo said. The company<br />

carried out further measurements<br />

and analyses and submitted a<br />

report to ENSI.<br />

In July 2015, Axpo announced<br />

that the restart of Beznau-1 had been<br />

postponed while the flaw indications<br />

were investigated further. Then in<br />

August, ENSI called for additional<br />

investigations.<br />

Beznau-2 was not affected by the<br />

flaw indications and was returned to<br />

service after its scheduled outage in<br />

2015.<br />

| | www.bkw.ch<br />

Kursk II Passed<br />

Construction Milestone<br />

(rosatom) Kursk II began reinforcing<br />

the foundation slab for the reactor<br />

building of Unit 1. This operation<br />

became the year’s key event on the<br />

construction site of the Kursk plant.<br />

On 21 December 2017, the first<br />

16-ton reinforced concrete block was<br />

installed on the rebar of the lower<br />

foundation belt. According to the<br />

project design, the foundation comprises<br />

105 reinforced concrete blocks<br />

with a total weight of 1,600 tons. This<br />

will enable the construction team<br />

to start concreting the foundation<br />

slab of the reactor building in the<br />

first half of <strong>2018</strong>.<br />

Prior to putting the first concrete<br />

block, a rebar coupler engraved with<br />

the words “The future is shaped today.<br />

The first coupling sleeve of the innovative<br />

VVER-TOI power unit” was<br />

ceremonially installed into the foundation<br />

reinforcement.<br />

VVER-TOI (which means ‘a standard<br />

optimized and automated power<br />

unit based on VVER technology’)<br />

reactors meet Russian and global<br />

safety requirements and have a longer<br />

service life and higher installed<br />

capacity than existing reactors of<br />

the Kursk Nuclear Power Plant.<br />

Alexander Mikhailov, Governor of<br />

the Kursk Region, noted that it was<br />

an honor for the region to build<br />

and commission one of the world’s<br />

first nuclear plants with advanced<br />

VVER-TOI reactors. “Construction of<br />

Kursk II designed to meet the latest<br />

global standards offers our region<br />

development prospects for the entire<br />

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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Rosenergoatom had planned to<br />

build two BN-1200 units at Beloyarsk<br />

with commercial operation scheduled<br />

by 2025. But construction depended<br />

on the results of operating the pilot<br />

Beloyarsk-4 BN-800 plant, which<br />

began commercial operation in<br />

October 2016.<br />

There is another commercially<br />

operational sodium-cooled FBR at<br />

Beloyarsk, the BN-600. Both the<br />

BN-600 and the BN-800 are smaller<br />

versions of the BN-1200. There are<br />

also two permanently shut-down<br />

light-water reactors at the site.<br />

| | www.rosatom.ru<br />

265<br />

NEWS<br />

| | Kursk II passed construction milestone.<br />

21st century. Just a few Russian<br />

regions have such opportunities,” he<br />

stressed.<br />

Vyacheslav Fedyukin, Director<br />

of Kursk NPP, noted it was symbolic<br />

that the event happened on the<br />

25 th anniversary of RosEnergoAtom<br />

and 10 years after the foundation of<br />

Rosatom, the companies that shaped<br />

the newest history of Russia’s nuclear<br />

industry. “Construction of Russia’s<br />

first VVER-TOI-based power unit<br />

proves that the national nuclear<br />

power industry is always at the<br />

cutting edge of science and engineering.<br />

The new generation VVER-TOI<br />

units are state-of-the-art facilities<br />

made to the best of Russia’s nuclear<br />

engineering knowledge,” he added.<br />

At the moment, other operations<br />

are also underway at the construction<br />

site of Kursk II. Among them is excavation<br />

of 1.2 million cub m of soil<br />

to be completed in 2017, with over<br />

800,000 cub m of sand, gravel<br />

and aggregate already put in the<br />

foun dation of Kursk II buildings<br />

and structures. Construction of a<br />

330/10 kV substation and preparation<br />

of technical documents for its commissioning<br />

are also drawing to a<br />

close.<br />

For reference:<br />

Kursk II is designed to replace the<br />

existing Kursk Nuclear Power Plant<br />

that will be taken out of operation in<br />

the years to come. Its first two units<br />

with VVER-TOI, a new-type reactor,<br />

will be commissioned simultaneously<br />

with decommissioning of Units 1 and<br />

2 of the existing nuclear station.<br />

According to the master schedule of<br />

Kursk II, Unit 1 will be commissioned<br />

in late 2023 to be followed by Unit 2 in<br />

late 2024.<br />

| | (18791501),<br />

ww.rosatom.ru<br />

Russia Confirms Plans to<br />

Revive BN-1200 Fast Breeder<br />

Reactor Project<br />

(nucnet) Russia plans to begin construction<br />

of its first industrial-sized<br />

sodium-cooled fast neutron reactor in<br />

the 2020s after saying three years ago<br />

that the project had been postponed,<br />

the head of state nuclear corporation<br />

Rosatom Alexei Likhachev told president<br />

Vladimir Putin.<br />

According to a transcript of a<br />

meeting posted on the Kremlin’s<br />

website, Mr Likhachev told Mr Putin<br />

that fast breeder reactors (FBRs) have<br />

significant advantages over existing<br />

reactor types and Rosatom is proposing<br />

that Russia goes ahead<br />

with its plans for the Generation IV<br />

BN-1200.<br />

FBRs have been and are being<br />

explored or constructed in Russia,<br />

France, India, China, Japan and the<br />

US. They allow a significant increase<br />

in the amount of energy obtained<br />

from natural, depleted and recycled<br />

uranium. The technology also enables<br />

plutonium and other actinides to be<br />

used and recycled.<br />

Russia operates the BN-600 and<br />

BN-800 FBR units at Beloyarsk and<br />

the BOR-60 fast breeder research<br />

reactor at the Research Institute<br />

of Atomic Reactors (RIAR) site in<br />

Dimitrovgrad, southwest Russia.<br />

BOR-60 is used to test fuel cycle,<br />

sodium coolant technologies and a<br />

range of design concepts for fast<br />

breeder reactors.<br />

In 2015, Rosatom said construction<br />

of the planned BN-1200 at the<br />

Beloyarsk nuclear power station in<br />

central Russia had been postponed<br />

until at least 2020, with state<br />

nuclear operator Rosenergoatom<br />

citing the need to improve fuel<br />

for the reactor and questioning the<br />

project’s economic feasibility.<br />

Austria Begins Legal Action<br />

Against EC Over Hungary’s<br />

Paks Nuclear Project<br />

(nucnet) Austria has filed a legal<br />

complaint against the European Commission<br />

with the European Court of<br />

Justice in Luxembourg for allowing<br />

Hungary to expand its Paks nuclear<br />

power station.<br />

Austrian minister of sustainability<br />

and tourism Elisabeth Köstinger said<br />

in a statement that nuclear power<br />

“must have no place in Europe” and<br />

Austria will not “not budge one<br />

centimetre” from its anti-nuclear<br />

stance.<br />

The EC started an investigation<br />

into state aid given to the Paks 2<br />

project in November 2014. Last March<br />

it approved the project to build two<br />

new reactors, to be financed with the<br />

help of Russia’s state atomic energy<br />

corporation Rosatom, after regulators<br />

said Hungarian authorities had<br />

agreed to several measures to ensure<br />

fair competition.<br />

In January <strong>2018</strong>, Austria announced<br />

it planned to sue the EC over<br />

the decision. “EU assistance is only<br />

permissible when it is built on common<br />

interest. For us, nuclear energy is<br />

neither a sustainable form of energy<br />

supply, nor is it an answer to climate<br />

change”, a statement by the ministry<br />

of sustainability said at the time.<br />

The two planned units at Paks 2<br />

nuclear power station are expected<br />

to begin commercial operation in<br />

2026 and 2027, Attila Aszódi,<br />

the Hungarian government’s commissioner<br />

for the Paks 2 project,<br />

told a conference in Brussels late l<br />

ast year.<br />

An agreement signed in 2014<br />

will see Russia supply two VVER-<br />

1200 pressurised water reactors for<br />

Paks 2 and a loan of up to €10bn<br />

($12.3bn) to finance 80% of the<br />

€12bn project.<br />

| | www.bundeskanzleramt.gv.at<br />

News


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

266<br />

NEWS<br />

Company News<br />

Framatome Completes<br />

Purchase of Schneider<br />

Electric’s Instrumentation and<br />

Control Nuclear Business<br />

(framatome) Framatome announced<br />

that it completed its purchase of<br />

Schneider Electric’s nuclear instrumentation<br />

and control offering. With<br />

this transaction, Framatome adds to<br />

its engineering expertise and expands<br />

its instrumentation and control (I&C)<br />

offerings.<br />

I&C systems are the central nervous<br />

system of a nuclear power plant,<br />

allowing operators to control reactor<br />

operations. Modernizations, upgrades<br />

and ongoing support are vital to manage<br />

economic, long-term operation of<br />

nuclear power plants, which provide<br />

reliable, low-carbon electricity.<br />

“With the integration of Schneider<br />

Electric’s nuclear instrumentation<br />

and control offering, we offer truly<br />

added value to our customers with<br />

a global technical expertise and<br />

market know-how on I&C solutions<br />

for the nuclear market,” said Bernard<br />

Fontana, Chairman of the Managing<br />

Board and Chief Executive Officer of<br />

Framatome. “We welcome our new<br />

colleagues to Framatome’s worldwide<br />

team of I&C engineers and experts.”<br />

This acquisition adds the nuclearqualified<br />

version of Tricon and the<br />

SPEC 200 platform to Framatome’s<br />

nuclear safety I&C offerings, which<br />

include the TELEPERM XS digital<br />

platform, and non-computerized<br />

analog solutions and instrumentation<br />

for nuclear power plants.<br />

This broadens the base of plants<br />

worldwide for which Framatome<br />

serves as the original equipment manufacturer<br />

for safety I&C systems. It also<br />

expands Framatome’s project and<br />

engineering capacities for non-safety<br />

I&C systems in the nuclear energy<br />

market, relying on Schneider Electric’s<br />

commercial TRICON and Foxboro<br />

platforms.<br />

Framatome also becomes the exclusive<br />

service provider to the nuclear energy<br />

market for the SPEC 200, nuclearqualified<br />

Tricon and Foxboro systems.<br />

| | www.framatome.com<br />

Framatome Continues<br />

Ramping up Production<br />

at Its Le Creusot Site<br />

(framatome) On January 25, <strong>2018</strong>,<br />

Framatome received the green light<br />

from the French Nuclear Safety<br />

Authority (ASN) and EDF to resume<br />

manufacture of forgings for the<br />

French nuclear fleet at its Le Creusot<br />

site. This decision allows the plant to<br />

continue ramping up its production<br />

with a target of 80 ingots per year.<br />

The authorization is an outcome of<br />

the improvement plan launched at the<br />

beginning of 2016 on the site following<br />

a series of quality audits. With the completion<br />

of all the actions necessary for<br />

the resumption of production for the<br />

French nuclear fleet and overall progress<br />

of 90% to date, the plan will be<br />

fully closed out in the first half of <strong>2018</strong>.<br />

The actions will then be incor porated<br />

into the site’s continuous improvement<br />

processes. Customers in France and<br />

abroad, as well as all the safety<br />

authorities concerned, have been kept<br />

regularly informed of the actions<br />

undertaken. Numerous reviews and<br />

inspections have been conducted in order<br />

to observe the progress of the plan<br />

and integrate stakeholders’ feedback.<br />

David Emond, Senior Executive<br />

Vice President of Framatome’s Component<br />

Manufacturing Business Unit,<br />

comments: “The authorization to<br />

resume manufacture of forgings for<br />

the French nuclear fleet is a very<br />

good news for the site that confirms<br />

the successful execution of its improvement<br />

plan. The 230 employees<br />

at the Le Creusot site are engaged<br />

in its deployment on a day to day basis<br />

so that we can supply our customers<br />

with equipment meeting the most<br />

stringent safety and quality requirements<br />

within agreed deadlines. I<br />

want to thank them for the substantial<br />

work they have accomplished on<br />

the site over the last two years.”<br />

Maintaining and developing the<br />

skills of the Le Creusot plant teams<br />

is a key element of the site’s improvement<br />

plan, with a particular focus<br />

on strengthening the nuclear safety<br />

culture.<br />

Framatome already invested<br />

7.5 million euros at the site in 2017<br />

to make the Le Creusot site a center<br />

of excellence for the manufacture<br />

of forgings for the nuclear industry,<br />

and will pursue this effort in <strong>2018</strong>.<br />

Major milestone reached<br />

in review of manufacturing<br />

records<br />

Moreover, a major milestone has<br />

been reached in the review of legacy<br />

manufacturing records at the Le<br />

Creusot site. The first stage in the<br />

inspection process which is being<br />

applied to all records relating to<br />

forgings produced for the nuclear<br />

industry, a key stage consisting in<br />

identifying findings, is now complete.<br />

The analysis of these findings and the<br />

processing of deviations will continue<br />

until the end of <strong>2018</strong>, in coordination<br />

with customers and safety authorities.<br />

Of the 6,000 records identified<br />

during the initial survey, 3,854 correspond<br />

to forgings installed on nuclear<br />

installations.<br />

At Framatome’s Jeumont and<br />

Saint-Marcel sites, the audit has been<br />

finalized since the summer of 2017 and<br />

no deviation impacting the safety of<br />

components has been brought to light.<br />

| | www.framatome.com<br />

JNFL and MHI Become<br />

Shareholders of<br />

Orano 2017 Revenue<br />

(orano) The Orano Board of Directors<br />

noted the completion of the capital<br />

increase reserved for Japan Nuclear<br />

Fuel Limited (JNFL) and Mitsubishi<br />

Heavy Industries, Ltd. (MHI) for<br />

a total of €500 million.<br />

Pursuant to the initial agreements<br />

signed with JNFL and MHI in<br />

March 2017, the funds corresponding<br />

to their total investment in Orano<br />

had been placed in trust on July 26,<br />

at the same time as the completion<br />

of the capital increase reserved for<br />

French State 2. These funds were<br />

released and used for the subscription<br />

of JNFL and MHI to Orano’s second<br />

capital increase.<br />

This transaction follows the<br />

completion on December 31, 2017 of<br />

the sale of the majority control of<br />

Framatome (formerly New NP) by<br />

AREVA SA to EDF as well as the<br />

fulfillment of the regulatory closing<br />

conditions related to the addition of<br />

an equity stake in Orano of both Japanese<br />

investors.<br />

Orano’s capital is now held by the<br />

French State (45.2%), the CEA<br />

(4.8%)3, AREVA SA (40%), JNFL<br />

(5%) and MHI (5%).<br />

This transaction is the last major<br />

step in the restructuring of the French<br />

nuclear industry, undertaken in 2015,<br />

and marks the end of the constitution<br />

phase of the Orano group. With a<br />

strengthened financial structure and<br />

sound strategic partnerships, Orano<br />

now has the means to grow and reach<br />

its goal of being a leading player in the<br />

production and recycling of nuclear<br />

materials, in waste management and<br />

dismantling.<br />

Appointment of a new<br />

independent director<br />

After completion of Orano’s second<br />

capital increase, the Orano General<br />

Meeting, also held on February 26,<br />

<strong>2018</strong>, appointed Patrick Pelata as<br />

independent director.<br />

| | www.orano.group<br />

News


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

Westinghouse Electric<br />

Company Signs Cooperation<br />

Agreement for Lead-cooled<br />

Fast Reactor Development<br />

(westinghouse) Westinghouse Electric<br />

Company has signed a Cooperation<br />

Agreement for lead-cooled fast<br />

reactor (LFR) technology development<br />

with the Italian National Agency<br />

for New Technologies, Energy and<br />

Sustainable Economic Development<br />

(ENEA) and Ansaldo Nucleare. The<br />

agreement demonstrates each party’s<br />

commitment to collaborating toward<br />

the development of a next-generation<br />

nuclear plant based on LFR technology,<br />

which is both “walk-away”<br />

safe and economically competitive<br />

across global energy markets.<br />

“This agreement is an exciting<br />

step towards the development of a<br />

lead-cooled fast reactor for the<br />

marketplace,” said Ken Canavan,<br />

Westinghouse chief technology officer<br />

and vice president, Global Technology<br />

Office. “The LFR is game-changing<br />

technology for clean energy industries,<br />

and Westinghouse is pleased to<br />

be working with such experienced<br />

partners to bring this innovative<br />

concept to fruition.”<br />

Beyond baseload electricity<br />

generation, the high-temperature<br />

operation of the LFR will allow for<br />

a broad range of applications such<br />

as an effective load-following<br />

capability enabled by an innovative<br />

thermal energy storage system,<br />

delivery of process heat for industrial<br />

applications and water desalination.<br />

ENEA is a world leader in research<br />

and development on lead-based<br />

systems, and currently operates<br />

among the finest and largest experimental<br />

facilities for LFR research in<br />

the world.<br />

Ansaldo Nucleare has vast experience<br />

in nuclear power plant design,<br />

supply, service and decommissioning,<br />

and has played leading roles in<br />

multiple international LFR development<br />

programs for the past 15 years.<br />

| | www.westinghousenuclear.com<br />

BKW übernimmt Experten<br />

für Strahlenschutz<br />

(bkw) Die BKW Konzerngesellschaft<br />

Dienstleistungen für Nukleartechnik<br />

(DfN) übernimmt das ebenfalls<br />

auf den kerntechnischen Bereich<br />

spezia lisierte Unternehmen Technischer<br />

Strahlenschutz (TSS). Dadurch<br />

stärkt die BKW ihre Kompetenzen in<br />

diesem Gebiet und baut sie weiter aus.<br />

Dies vor dem Hintergrund der geplanten<br />

Stilllegung des Kernkraftwerks<br />

Mühleberg und zahlreicher weiterer<br />

Kernkraftwerke in Europa.<br />

Mit der Übernahme des Strahlenschutzunternehmens<br />

DfN hat die BKW<br />

bereits im letzten Jahr ihre bestehenden<br />

und bewährten Kom petenzen im hochspezialisierten<br />

Nukleartechnik-Bereich<br />

erweitert. Der Eintritt der TSS in den<br />

Unter nehmensverbund der BKW stellt<br />

nun einen weiteren Ausbau in diesem<br />

Gebiet dar. Die TSS ergänzt die<br />

Strahlenschutzkompetenzen innerhalb<br />

der BKW Gruppe und verstärkt diese<br />

auch im Hinblick auf die Still legung<br />

des Kernkraftwerks Mühleberg.<br />

In Europa ist ausserdem eine Vielzahl<br />

weiterer Stilllegungsprojekte in<br />

Planung oder bereits im Gang. Der<br />

Strahlenschutz spielt bereits beim<br />

Betrieb von Kernkraftwerken eine<br />

wichtige Rolle. Mit der Stilllegung<br />

und den dabei ausgeführten Demontage-<br />

und Freimessarbeiten nehmen<br />

die Strahlenschutzarbeiten zu. Für<br />

Strahlenschutzdienstleisterinnen wie<br />

TSS und DfN bietet der wachsende<br />

Stilllegungsmarkt daher ein grosses<br />

Potenzial und die Möglichkeit, sich<br />

weiterzuentwickeln.<br />

Die DfN und die TSS haben bereits<br />

verschiedentlich auf Projektbasis<br />

zusammengearbeitet. Die erfolgreiche<br />

Kooperation wird künftig<br />

weiter ausgebaut, was mit einer<br />

gegenseitigen Stärkung einhergeht.<br />

Um eine optimale Zusammenarbeit<br />

zu ermöglichen, wird die TSS in die<br />

DfN integriert.<br />

Die TSS mit Sitz in Geilenkirchen<br />

im deutschen Bundesland Nordrhein-<br />

Westfalen wurde 1979 gegründet<br />

und zählt 15 Mitarbeitende. Das<br />

Unternehmen bietet ein qualitativ<br />

hochwertiges und breites Angebot von<br />

Dienstleistungen im kerntechnischen<br />

Bereich. Dazu gehören neben dem<br />

Strahlenschutz die Dekontamination,<br />

die Abfallentsorgung, die Dosimetrie<br />

sowie die Abwicklung von Transporten<br />

radioaktiver Stoffe.<br />

| | (18791521), www.bkw.ch<br />

Companies<br />

China Approves $ 100 Billion<br />

Merger of Leading<br />

Nuclear Companies<br />

(nucnet) China has approved the<br />

merger of nuclear power producer<br />

China National Nuclear Corporation<br />

(CNNC) with nuclear plant builder<br />

China Nuclear Engineering and<br />

Construction Corporation (CNECC),<br />

the state-run China Daily news agency<br />

said.<br />

According to the China Daily, the<br />

combined assets of the new company<br />

will be worth about $100bn (€80bn),<br />

while its workforce will be about<br />

150,000 employees.<br />

CNNC is China’s number two<br />

nuclear power producer and CNECC<br />

the country’s top nuclear power plant<br />

builder.<br />

China Daily said the merger is in<br />

line with efforts by China to streamline<br />

the state-operated sector of its<br />

economy and reduce the number of<br />

state-owned companies administered<br />

by central government.<br />

Approval for the merger was<br />

confirmed by the State-Owned Assets<br />

Supervision and Administration<br />

Commission (SASAC) in a one-line<br />

statement posted on its website.<br />

| | (18800822), en.cnnc.com.cn<br />

267<br />

NEWS<br />

Research<br />

| | BKW übernimmt Experten für Strahlenschutz © BKW.<br />

NRG: Every Day,<br />

30,000 Patients Benefit From<br />

Medical Isotopes From Petten<br />

(nrg) Medical isotopes are indispensable<br />

for diagnosing and treating<br />

cancer. Demand for them is set to soar<br />

over the next 20 years, but supplies<br />

are diminishing. To put the spotlight<br />

News


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

• Separative work: 58.00–92.00<br />

268<br />

NEWS<br />

2016<br />

January to June 2016<br />

• Uranium: 26.50–35.25<br />

• Conversion: 6.25–6.75<br />

• Separative work: 58.00–62.00<br />

July to December 2016<br />

• Uranium: 18.75–27.80<br />

• Conversion: 5.50–6.50<br />

• Separative work: 47.00–62.00<br />

| | NRG: Every day, 30000 patients benefit from medical isotopes from Petten View of the pool type reactor<br />

core. (Courtesy: JRC)<br />

on the world of medical isotopes, the<br />

platform 30000perdag.nl has been<br />

launched. The aim of the platform and<br />

the accompanying campaign is to<br />

boost awareness that the Netherlands<br />

must continue leading the field in<br />

cancer treatment.<br />

The future<br />

Over the next 20 years, the number of<br />

cancer diagnoses is expected to rise by<br />

70%. Fortunately, health care is<br />

constantly improving, partly through<br />

the use of medical isotopes. However,<br />

there are only 6 reactors worldwide<br />

which can produce medical isotopes,<br />

one of which is closing next year.<br />

This means that whilst demand for<br />

medical isotopes is growing worldwide,<br />

supplies are diminishing.<br />

30000perdag.nl<br />

An online information park for a<br />

wide audience has been built on<br />

30000perdag.nl. Visitors can learn all<br />

about medical isotopes here: from raw<br />

materials to the reactor in Petten to<br />

applications in the hospital. By opening<br />

up that world, NRG in Petten<br />

wants to show (former) cancer<br />

patients and their families and<br />

acquaintances what is needed to be<br />

able to treat cancer, and request<br />

support for medical isotopes and good<br />

cancer treatment in the Netherlands<br />

and abroad.<br />

Former cancer patients play<br />

starring role in campaign<br />

The campaign uses 3 video interviews<br />

with cancer survivors. The interviews<br />

were conducted by presenter Fien Vermeulen,<br />

herself a former lymphoma<br />

patient. Fien drives with former<br />

patients Anouk (26), Alexander (42)<br />

and Manon (34) to the research<br />

reactor in Petten, where they talk<br />

about their remarkable experiences in<br />

times of uncertainty. Each of them<br />

represents one of the 30,000 people<br />

who benefit or have benefitted from<br />

medical isotopes every day.<br />

Anyone can demonstrate their<br />

support by liking the Facebook page<br />

30.000perdag. Another very visible<br />

form of support is available through<br />

the T-shirts that can be ordered via<br />

30000perdag.nl. These enable former<br />

patients and supporters to show their<br />

backing for the campaign.<br />

| | (18800822), www.nrg.eu<br />

Market data<br />

(All information is supplied without<br />

guarantee.)<br />

Nuclear Fuel Supply<br />

Market Data<br />

Information in current (nominal)<br />

U.S.-$. No inflation adjustment of<br />

prices on a base year. Separative work<br />

data for the formerly “secondary<br />

market”. Uranium prices [US-$/lb<br />

U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =<br />

0.385 kg U]. Conversion prices<br />

[US-$/kg U], Separative work<br />

[US-$/SWU (Separative work unit)].<br />

January to December 2013<br />

• Uranium: 34.00–43.50<br />

• Conversion: 9.25–11.50<br />

• Separative work: 98.00–127.00<br />

January to December 2014<br />

• Uranium: 28.10–42.00<br />

• Conversion: 7.25–11.00<br />

• Separative work: 86.00–98.00<br />

January to December 2015<br />

• Uranium: 35.00–39.75<br />

• Conversion: 6.25–9.50<br />

2017<br />

January 2017<br />

• Uranium: 20.25–25.50<br />

• Conversion: 5.50–6.75<br />

• Separative work: 47.00–50.00<br />

February 2017<br />

• Uranium: 23.50–26.50<br />

• Conversion: 5.50–6.75<br />

• Separative work: 48.00–50.00<br />

March 2017<br />

• Uranium: 24.00–26.00<br />

• Conversion: 5.50–6.75<br />

• Separative work: 47.00–50.00<br />

April 2017<br />

• Uranium: 22.50–23.50<br />

• Conversion: 5.00–5.50<br />

• Separative work: 45.50–48.50<br />

May 2017<br />

• Uranium: 19.25–22.75<br />

• Conversion: 5.00–5.50<br />

• Separative work: 42.00–45.00<br />

June 2017<br />

• Uranium: 19.25–20.50<br />

• Conversion: 5.55–5.50<br />

• Separative work: 42.00–43.00<br />

July 2017<br />

• Uranium: 19.75–20.50<br />

• Conversion: 4.75–5.25<br />

• Separative work: 42.00–43.00<br />

August 2017<br />

• Uranium: 19.50–21.00<br />

• Conversion: 4.75–5.25<br />

• Separative work: 41.00–43.00<br />

September 2017<br />

• Uranium: 19.75–20.75<br />

• Conversion: 4.60–5.10<br />

• Separative work: 40.50–42.00<br />

October 2017<br />

• Uranium: 19.90–20.50<br />

• Conversion: 4.50–5.25<br />

• Separative work: 40.00–43.00<br />

November 2017<br />

• Uranium: 20.00–26.00<br />

• Conversion: 4.75–5.25<br />

• Separative work: 40.00–43.00<br />

December 2017<br />

• Uranium: 23.50–25.50<br />

• Conversion: 5.00–6.00<br />

• Separative work: 39.00–42.00<br />

<strong>2018</strong><br />

January <strong>2018</strong><br />

• Uranium: 21.75–24.00<br />

• Conversion: 6.00–7.00<br />

• Separative work: 38.00–42.00<br />

News


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

February <strong>2018</strong><br />

• Uranium: 21.25–22.50<br />

• Conversion: 6.25–7.25<br />

• Separative work: 37.00–40.00<br />

| | Source: Energy Intelligence<br />

www.energyintel.com<br />

Cross-border Price<br />

for Hard Coal<br />

Cross-border price for hard coal in<br />

[€/t TCE] and orders in [t TCE] for<br />

use in power plants (TCE: tonnes of<br />

coal equivalent, German border):<br />

2012: 93.02; 27,453,635<br />

2013: 79.12, 31,637,166<br />

2014: 72.94, 30,591,663<br />

2015: 67.90; 28,919,230<br />

2016: 67.07; 29,787,178<br />

I. quarter: 56.87; 8,627,347<br />

II. quarter: 56.12; 5,970,240<br />

III. quarter: 65.03, 7.257.041<br />

IV. quarter: 88.28; 7,932,550<br />

| | Uranium spot market prices from 1980 to <strong>2018</strong> and from 2008 to <strong>2018</strong>. The price range is shown.<br />

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

269<br />

NEWS<br />

2017:<br />

I. quarter: 95.75; 8,385,071<br />

II. quarter: 86.40; 5,094,233<br />

III. quarter: 88.07; 5,504,908<br />

| | Source: BAFA,<br />

some data provisional<br />

www.bafa.de<br />

EEX Trading Results<br />

February <strong>2018</strong><br />

(eex) In February <strong>2018</strong>, the European<br />

Energy Exchange (EEX) achieved a<br />

total volume of 274.3 TWh on its<br />

power derivatives markets (February<br />

2017: 200.8 TWh) which is a yearon-year<br />

increase of 37 %. In doing<br />

so, EEX was able to grow its power<br />

derivatives volumes across all market<br />

areas.<br />

In total, the German and Austrian<br />

markets (Phelix-DE, Phelix-AT and<br />

Phelix-DE/AT) increased by 12 % to<br />

169.7 TWh. This includes 153.4 TWh<br />

from the benchmark product Phelix-<br />

DE which achieved its highest volume<br />

since launch in April 2017. Volumes<br />

in the French market more than<br />

doubled to 26.9 TWh (February 2017:<br />

12.3 TWh) while Italian power<br />

volumes grew substantially to<br />

40.0 TWh (February 2017: 22.1 TWh).<br />

Furthermore, on the Spanish market,<br />

volumes increased by more than<br />

250 % to 8.3 TWh (February 2017:<br />

2.3 TWh).<br />

The February volume comprised<br />

173.5 TWh traded at EEX via Trade<br />

Registration with subsequent clearing.<br />

Clearing and settlement of all exchange<br />

transactions was executed by European<br />

Commodity Clearing (ECC).<br />

The Settlement Price for base<br />

load contract (Phelix Futures) with<br />

| | Separative work and conversion market price ranges from 2008 to <strong>2018</strong>. The price range is shown.<br />

)1<br />

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

delivery in 2019 amounted to 33.85 €/<br />

MWh. The Settlement Price for peak<br />

load contract (Phelix Futures) with<br />

delivery in 2019 amounted to 42.40 €/<br />

MWh.<br />

On the EEX markets for emission<br />

allowances, the total trading volume<br />

increased by 57% to 144.2 million<br />

tonnes of CO 2 in February (February<br />

2017: 91.7, million tonnes of CO 2 ).<br />

Primary market auctions contributed<br />

75.1 million tonnes of CO 2 to the total<br />

volume. On the spot secondary<br />

market, volumes more than doubled<br />

to 4.5 million tonnes of CO 2 (February<br />

2017: 2.0 million tonnes of CO 2 ). On<br />

the EUA Futures market, EEX was able<br />

to increase volumes by 80% to<br />

37.3 million tonnes of CO 2 (February<br />

2017: 20.7 million tonnes of CO 2 ).<br />

Furthermore, 27.4 million tonnes of<br />

CO 2 were traded in EUA Options<br />

which is the highest monthly volume<br />

so far in this product.<br />

The EUA price with delivery in<br />

December <strong>2018</strong> amounted to<br />

8.80/10.15 €/ EUA (min./max.).<br />

| | www.eex.com<br />

MWV Crude Oil/Product Prices<br />

January 2017<br />

(mwv) According to information and<br />

calculations by the Association of the<br />

German Petroleum Industry MWV e.V.<br />

in January <strong>2018</strong> the prices for<br />

super fuel, fuel oil and heating oil<br />

noted inconsistent compared with<br />

the pre vious month December 2017.<br />

The average gas station prices for Euro<br />

super consisted of 136.84 €Cent<br />

( December 2017: 136.84 €Cent,<br />

approx. +-0.0 % in brackets: each<br />

information for pre vious month or<br />

rather previous month comparison),<br />

for diesel fuel of 120.48 €Cent<br />

(119.01; +1.24 %) and for heating oil<br />

(HEL) of 62.27 €Cent (60.65 €Cent,<br />

+2.67 %).<br />

The tax share for super with<br />

a consumer price of 136.84 €Cent<br />

(136.84 €Cent) consisted of<br />

65.45 €Cent (47.83 %, 65.45 €Cent)<br />

for the current constant mineral oil<br />

tax share and 21.85 €Cent (current<br />

rate: 19.0 % = const., 22.12 €Cent)<br />

for the value added tax. The product<br />

price (notation Rotterdam) consisted<br />

of 40.17 €Cent (29.36 %, 37.18 €Cent)<br />

and the gross margin consisted of<br />

9.37 €Cent (6.85 %; 12.36 €Cent).<br />

Thus the overall tax share for super<br />

results of 66.83 % (66.83 %).<br />

Worldwide crude oil prices<br />

(monthly average price OPEC/Brent/<br />

WTI, Source: U.S. EIA) were again<br />

significantly higher, approx. +8.36 %<br />

(+2.34 %) in January <strong>2018</strong> compared<br />

to December 2017.<br />

The market showed a stable<br />

development with higher prices; each<br />

in US-$/bbl: OPEC basket: 66.85<br />

(62.06); UK-Brent: 69.08 (64.37);<br />

West Texas Inter mediate (WTI): 63.7<br />

(57.88).<br />

| | www.mwv.de<br />

News


<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />

270<br />

Czechs and Balances and Why ‘Ugly’<br />

Nuclear Deserves a Political Makeover<br />

NUCLEAR TODAY<br />

Author<br />

John Shepherd<br />

Shepherd<br />

Communications<br />

3 Brooklands<br />

West Sussex<br />

BN43 5FE<br />

Links to reference<br />

sources:<br />

Dana Drábová<br />

interview:<br />

http://bit.ly/2Ik7WaJ<br />

European Investment<br />

Bank announcement:<br />

http://bit.ly/2Ik7WaJ<br />

Yonhap News<br />

Agency report:<br />

http://bit.ly/2FyvZkw<br />

As if Europe does not have enough on its plate to deal with at the moment – politically and economically just for starters<br />

– could Brussels be on a collision course with the Czech government over the country's plans to expand nuclear energy?<br />

There is certainly friction over the issue between Prague and<br />

the European Commission (EC), to put it mildly. But why?<br />

The veteran head of the Czech Republic’s State Office<br />

for Nuclear Safety, Dana Drábová, last month accused<br />

other EU member states of “pressurising” Prague over the<br />

early closure of its oldest nuclear reactor units.<br />

Drábová reportedly told an energy conference in the<br />

country: “There is immense pressure developing that the<br />

operating life of nuclear reactors will be limited to 40 years.<br />

That means that our political representatives, whoever they<br />

might be, sometime around 2023 will face a battle over a<br />

further 10-year extension for Dukovany. The current State<br />

Energy Framework counts on the lifetime of the Dukovany<br />

reactors ending sometime between 2030 and 2040.”<br />

The nuclear safety chief later told Czech Radio the<br />

pressure was coming from “the 14 countries which are not<br />

using nuclear power and some of which regard it as<br />

something ugly”. If the pressure continued, she predicted<br />

there would be a concerted “willingness… to get rid of<br />

these nuclear plants in Europe as fast as possible”.<br />

Drábová’s comments came against a backdrop of the<br />

Czech government saying it would appoint an expert team<br />

to consider proposals to break up the majority state-owned<br />

electricity firm CEZ. The move was one of several options<br />

mooted to support financing of the construction of a new<br />

nuclear power plant at Dukovany.<br />

Analysts say the new nuclear plant could be built by the<br />

traditional energy unit, which would be fully state-owned<br />

and therefore in the best position to take on the risks of<br />

high costs that the utility could not if it were an entity with<br />

private owners.<br />

Czech prime minister Andrej Babiš is backing proposals<br />

to build the Dukovany reactor, around 50 kilometres north<br />

of the (anti nuclear) Austrian border, to replace a Soviet-era<br />

reactor. But this would mean persuading the EC to exempt<br />

the project from strict EU rules on government bids.<br />

If the Czech government fails in its quest, it could consider<br />

doing a deal with Russia, which would undoubtedly<br />

be very much along the lines of the nuclear construction<br />

and financing deal Moscow signed recently with EU<br />

member Hungary.<br />

If, dear reader, you now have a sense of déjà vu, you<br />

would be right. You may recall that Hungary went through<br />

a similar nuclear battle with the EC, despite Hungary’s parliament<br />

fully backing proposals to build two new nuclear<br />

reactor units in that country.<br />

Initially, the EC said in November 2015 it had started<br />

legal action against Hungary over a contract signed with<br />

Russia’s Rosatom to build two units at the existing Paks<br />

plant. Brussels expressed concern about the project’s<br />

compatibility with EU public procurement rules. However,<br />

the EC eventually cleared the issue and a state aid investigation<br />

into the project financing for the ‘Paks II’ project<br />

was subsequently dropped by the EC.<br />

There was a similar clash with the EC when the UK first<br />

unveiled plans to invest in building the Hinkley Point C<br />

nuclear plant.<br />

So is the latest tussle between Prague and Brussels<br />

really over concerns about state-aid rules or is it more a<br />

worrying trend of interference to stop nuclear in its tracks?<br />

And is the conflict really worth it…?<br />

Czech PM Babiš said following an official visit to<br />

Hungary last January, where he attended a summit of<br />

prime ministers of the Visegrad Group countries, that he<br />

and Hungarian counterpart Viktor Orbán discussed the<br />

potential for “further developing” cooperation in sectors<br />

such as the nuclear energy industry.<br />

But far more intriguing was what Babiš claimed was the<br />

attitude of Visegrad leaders about relations with the<br />

institutions of the EU. According to a statement issued by<br />

the Czech government, Babiš said the leaders agreed it was<br />

“necessary to depoliticise Brussels and the EC”. Apparently,<br />

the leaders believe that when it comes to EU affairs,<br />

“ member states, prime ministers and presidents, should<br />

have the main say”, according to Babiš.<br />

If there is behind-the-scenes pressure to stamp out<br />

nuclear wherever it might try to cling on or prosper in the<br />

EU, where is that effort coming from and why? Of course,<br />

it is no secret that Austria and Germany strongly oppose<br />

any expansion of nuclear power in Europe. Having lived<br />

and worked in Germany, I never understood that<br />

politically- inspired decision – but as a guest in the country<br />

for which I have great admiration I respect its decision.<br />

Austria’s approach has always puzzled me more – being<br />

willing as it is to host the International Atomic Energy<br />

Agency (IAEA) and enjoy all the ‘fruits’ that that privilege<br />

brings, not least in the economic benefit of having the<br />

agency based in Vienna.<br />

But back to the Czech project. As a possible fight with<br />

the EC shapes up, it is not only Moscow that is set to benefit<br />

from yet another new nuclear power order from an EU<br />

nation.<br />

South Korea is also reportedly circling – keen to tempt<br />

Prague to consider its nuclear technology, according to<br />

Seoul’s Yonhap News Agency.<br />

Can the EU really afford such a quarrel – again – with a<br />

member state over nuclear? And why should European<br />

skills, knowhow and investment not be channelled into the<br />

Czech nuclear project?<br />

I am struck by the EC’s approach to another industrial<br />

sector and how contrasting it is. The EC is currently working<br />

at full tilt to develop a European battery cell industry,<br />

with the goal of ensuring the EU is not overwhelmed by<br />

competition from Asian battery makers for products such as<br />

electric vehicles and energy storage devices.<br />

The EU’s vice-president for the energy union, Maroš<br />

Šefčovič, said in February “there are many extremely<br />

interesting actions that we need to pursue, including (a)<br />

simplification of approval procedures and permitting<br />

processes in the EU”. Indeed the European Investment Bank<br />

has already approved a loan for the construction and<br />

operation of what it said will be a first-of-a-kind demonstration<br />

plant in Sweden, for the manufacturing of lithium- ion<br />

batteries.<br />

The EC’s support for the development of such technology<br />

across EU member states is of course admirable, but one<br />

hears nothing of state-aid rules and complications here!<br />

Why is it that nuclear cannot win such favourable attention<br />

and support? Does it really have to be this way – and<br />

hasn’t the EC learned anything from the UK’s Brexit vote<br />

about treading carefully in issues that are seen by member<br />

states of national importance?<br />

Nuclear Today<br />

Czechs and Balances and Why ‘Ugly’ Nuclear Deserves a Political Makeover ı Jubair Ahmed Shamim and Kune Yull Suh


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