atw 2018-04v6

nucmag.com

2018

4

217

Heat Transfer Systems

for Novel Nuclear

Power Plant Designs

221 ı Operation and New Build

Safety Research for GEN IV Reactors

226 ı Operation and New Build

Numerical Analysis for the MYRRHA Project

ISSN · 1431-5254

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238 ı Environment and Safety

Passive Heat Removal Systems Research

270 ı Nuclear Today

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atw Vol. 63 (2018) | Issue 4 ı April

Security of Supply ...

and the Clock is Ticking ...

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

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

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

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

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

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

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

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

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

possible further transportation length enabled through higher trans mission voltage.

207

EDITORIAL

Both current types have something in common: generation

and use need to take place simultaneously. The grid fails if

both do not fit together. Neither alternating current grids

nor direct current grids offer storage possibilities. Thus, a

stable power system also requires a stable and reliable

generation, because if a larger system “fails”, the system

restoration is, from its task and process, a large-scale

project.

Different believes e.g. from politics or other interest

groups are simply wrong, power systems are – without any

further active establishments and plants- no accumulators.

A reliable power supplying system needs at any time

reliable generation. “Surpluses”, meaning potentials for a

higher generation than demand, when so ever, cannot be

shifted or stored “electrically” in the system at a later time.

It was not an inconspicuous message, which appeared

multiple times in the press at the end of February,

beginning of March 2018. Headlines such as “Time

synchronisation per power system: Energy shortages make

watches lose time”, described a phenomenon, of which,

according to the media “one became aware of – only

( editor’s note) - after weeks”: What happened?

As an indicator for the stability of alternating power

systems stand supply voltage a well as system frequency.

For the system frequency applies that she needs to be

identical at any point of the system. If generation and

consumption do not fit, deviations occur, leaking generation

leads among others to a perceived frequency decrease

among the entire connected system. As the system

frequency is defined for our alternating electricity net with

constant 50 Hertz, it is also qualified for watches, which

use the frequency as direct clock indicator.

We can for example – due to cost reasons – renounce to

a frequency stabilising quartz oscillator. Nevertheless, this

technical simplification is bought with failures in time, if

the frequency deviates from the standard over a longer

period. Only a few hundred Hertz is enough for days and

weeks in order to, as in the current case, generate a time

deviation of minus 360 seconds, 6 minutes, and those

inside the entire affected system of 25 West, Middle- and

South European countries.

The cause for this incident was later communicated by

the European Network of Transmission System Operators for

Electricity (short ENTSO-E) and the Swiss net operator

swissgrid, that in the control zone Serbia, Macedonia,

Montenegro (the so called SMM rule block), especially in

Kosovo and Serbia less energy was fed into the system. A

deficit of 113 gigawatt hours was shown, not much, in view

of a European daily production of around 8,000 gigawatt

hours. But especially this shows how delicate our power

system is and how sensitive it reacts to the smallest

malfunctions.

Reliable measures in power generation – meaning

currently only for conventional techniques, thus need,

with all considerations on the reconstruction of electricity

supply, to be reconsidered. Additionally and almost

simultaneously another alarming “availability message”

came in: At the beginning of March 2018 European gas

storage tanks were only filled with a quantity of 26.2 per

cent, Germany even on average only with 23.8 per cent.

Thus, according to an EU-conform proceeding an early

warning level was reached, because the filling level of

storage tanks may not be lower than around 20 % due to

reasons of guaranteeing mechanical stability. On top came

the message that more natural gas was imported to Europa

than in the previous years. All first hints, that there might

not be enough natural gas in Europe for dispose filling in as

a “reserve”?

In all, these are all important references that any,

especially neither direct market- nor technically driven,

interventions – where compensation factors can con tribute

– need to be well thought in our power system. Furthermore,

does the availability of a broad basis of conventional

generation not only gain more importance, she is even

more important than it is conceded for “conventionals”

vision wise in many places in terms of an „energy

transition“. To what extend “the clock” might tick on

possible severe supply shortfalls or even large-scale loss of

off-site power… one does not know…

Christopher Weßelmann

– Editor in Chief –

Editorial

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

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atw Vol. 63 (2018) | Issue 4 ı April

EDITORIAL 208

Versorgungssicherheit und die Uhr tickt ...

Liebe Leserin, lieber Leser, vor mehr als hundert Jahren, um 1890, entbrannte eine Auseinandersetzung

zwischen den beiden bekannten Protagonisten der Elektrizitätsversorgung, Thomas Alva Edison und George Westinghouse,

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

Während Edison technisch die Gleichspannung favorisierte, setze Westinghouse die Wechselspannung dagegen.

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

der jeweiligen Unternehmen General Electric bzw. Westinghouse Electric und die dahinter stehenden Patente. Vorangegangen

waren in atemberaubendem Tempo die wichtigsten Entwicklungen für die Nutzung der Elektrizität: Im Jahr

1866 entdeckte Werner von Siemens das dynamoelektrische Prinzip, das größere Leistungen ermöglichte. Die

Entwicklung des Wechselstromtransformators im Jahr 1881 ermöglichte technisch grundsätzlich und kostengünstiger

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

zehn Kilometern. Durchgesetzt hatte sich aufgrund der durch höhere Übertragungsspannungen möglichen weiteren

Transportlängen zu jener Zeit die Wechselspannung.

Beiden Stromarten ist eines gemeinsam: Erzeugung und

Nutzung müssen exakt zeitgleich erfolgen. Sind Erzeugung

und Gebrauch nicht im Einklang, bricht das Netz

zusammen. Weder Wechsel- noch Gleichspannungsnetz

bieten „Speichermöglichkeiten“. Für ein stabiles Stromnetz

ist daher auch eine stabile und verlässliche Erzeugung

erforderlich, denn wenn einmal ein größeres Stromnetz

„zusammenbricht“, ist der Netzwiederaufbau ein von der

Aufgabe und dem zeitlichen Ablauf her aufwendiges

Vorhaben. Anderslautende Stimmen z.B. aus der Politik

oder von Interessengruppen sind schlichtweg falsch,

Strom netze sind – ohne weitere aktive Einrichtungen und

Anlagen – keine Speicher. Ein verlässliches Stromversorgungsnetz

benötigt eine jederzeit verlässliche Erzeugung.

„Überschüsse“, also Potenziale für eine höhere Erzeugung

als die vorhandene Nachfrage, wann und warum auch

immer, lassen sich „elektrisch“ im Netz nicht auf spätere

Zeiten verschieben, also speichern.

Es war eine nicht unscheinbare Nachricht, die Ende

Februar, Anfang März 2018 mehrfach durch die

Presse ging. Überschriften wie „Zeit-Synchronisation per

Stromnetz: Energieknappheit lässt Uhren nachgehen“,

beschrieben ein Phänomen, dessen man sich nach Angaben

in der Presse „nach Wochen – erst (Anm. der Red.) –

bewusst wurde“: Was war geschehen?

Für die Stabilität bzw. als Indikator für die Stabilität

von Wechselstromnetzen stehen die Netzspannung sowie

die Netzfrequenz. Für die Netzfrequenz gilt dabei, dass

diese an jedem Punkt in einem Netz identisch ist. Stimmen

Erzeugung und Verbrauch nicht überein, kommt es zu

Abweichungen, fehlende Erzeugung führt u.a. zu einer

im gesamten angebundenen Netz fühlbaren Frequenzabnahme.

Da die Netzfrequenz für unser Wechselstromnetz

mit konstant 50 Hertz vereinbart ist, eignet sich diese

auch für Uhren, die die Frequenz als direkten Taktgeber

nutzen. Auf z.B. einen frequenzstabilisierenden Quarzoszillator

kann – aus Kostengründen – verzichtet werden.

Diese technische Vereinfachung erkauft man sich allerdings

mit Fehlern in der Uhrzeit, wenn die Frequenz über

einen längeren Zeitraum vom Standard abweicht. Schon

wenige hundertstel Hertz reichen über Tage und Wochen

aus, um, wie im aktuellen Fall, eine kumulierte Zeitabweichung

von Minus 360 Sekunden, also 6 Minuten,

hervorzurufen; und dies im ganzen betroffenen Netz von

25 West-, mittel- und südosteuropäischen Ländern.

Als Ursache für dieses Ereignis wurde später vom

Verband Europäischer Übertragungsnetzbetreiber (kurz

ENTSO-E, European Network of Transmission System

Operators for Electricity) und dem Schweizer Netzbetreiber

swissgrid kommuniziert, dass in der Kontrollzone Serbien,

Mazedonien, Montenegro (dem sogenannten SMM Regelblock),

insbesondere in Kosovo und Serbien zu wenig

Energie ins Netz eingespeist wurde. Ein Fehlbetrag von

113 Gigawattstunden wurde ausgewiesen, nicht viel,

angesichts einer europaweiten Tagesproduktion von rund

8.000 Gigawattstunden. Aber gerade diese zeigt, wie

filigran unser Stromnetz ist und wie empfindlich es doch

auf kleinste Störungen reagiert.

Verlässliche Größen in der Stromerzeugung, sprich

derzeit letztendlich nur die konventionellen Techniken,

müssten von daher in allen Überlegungen zum Umbau der

Stromversorgung neu überdacht werden. Hinzu kam fast

zeitgleich eine weitere bedenkliche energiewirtschaftliche

„Verfügbarkeitsmeldung“: Europas Gasspeicher waren zu

Anfang März 2018 nur noch zu 26,2 Prozent gefüllt,

Deutschland gar im Schnitt nur zu 23,8 Prozent. Damit

war nach einem EU-einheitlichen Verfahren eine Frühwarnstufe

erreicht, denn die Speicher dürfen ihren Füllgrad

aus Gründen der Gewährleistung ihrer mechanischen

Stabilität nicht unter rund 20 % absenken. Hinzu kam die

Mitteilung, dass mehr Erdgas nach Europa importiert

wurde, als in den Vorjahren. Alles erste Hinweise darauf,

dass vielleicht in Zukunft doch nicht ausreichend Erdgas

in Europa zur Verfügung stehen wird, um als „Reserve“

einzuspringen?

In Summe sind dies alles wichtige Hinweise darauf,

dass jegliche, vor allem weder direkt markt- noch technisch

getriebenen Eingriffe – wo ausgleichende Faktoren wirken

können – in unser Stromversorgungssystem wohl überdacht

sein müssen. Zudem gewinnt die Verfügbarkeit einer

breiten Basis konventioneller Erzeugung damit nicht nur

an Bedeutung, sie ist bedeutungsvoller als vielerorts in

Visionen einer „Energiewende“ den Konventionellen

zugestanden wird.

Inwieweit „die Uhr“ möglicher schwerwiegender Versorgungsengpässe

oder gar großflächiger Netzausfälle

tickt ... man weis es nicht ...

Christopher Weßelmann

– Chefredakteur –

Editorial

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

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atw Vol. 63 (2018) | Issue 4 ı April

210

Issue 4

April

CONTENTS

217

Heat Transfer Systems

for Novel Nuclear

Power Plant Designs

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

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

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

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

Editorial

Security of Supply ... and the Clock is Ticking ... . . 207

Versorgungssicherheit und die Uhr tickt ... . . . . 208

Abstracts | English . . . . . . . . . . . . . . . . . . . 212

Abstracts | German . . . . . . . . . . . . . . . . . . . 213

Inside Nuclear with NucNet

Euratom: Industry Softens Stance

as Government Lays Out Plans for Transition . . . 214

NucNet

Calendar . . . . . . . . . . . . . . . . . . . . . . . 216

Operation and New Build

Heat Transfer Systems for Novel

Nuclear Power Plant Designs . . . . . . . . . . . . . 217

Sebastian Vlach, Christoph Fischer and Herman van Antwerpen

Experimental and Analytical Tools

for Safety Research of GEN IV Reactors . . . . . . . 221

G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak

DAtF Notes. . . . . . . . . . . . . . . . . . . . . .215

221

| | Centrum Výzkumu Řež facilities list.

217

Numerical Analysis of MYRRHA

Inter- wrapper Flow Experiment at KALLA . . . . . 226

| | Koeberg PWR steam generator and simulation model.

Abdalla Batta and Andreas G. Class

Contents

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atw Vol. 63 (2018) | Issue 4 ı April

226

CONTENTS

211

| | Velocity magnitude within bundle showing flow distribution.

Heat Balance Analysis for

Energy Conversion Systems of VHTR . . . . . . . . 230

SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon

and Soyoung Park

Spotlight on Nuclear Law

Information Requirements Versus

Confidentiality Obligations – Extension of

the In-Camera Procedure Planned . . . . . . . . . . 235

Informationsbedarf versus

Geheimhaltungspflichten – Erweiterung

des In camera-Verfahrens geplant . . . . . . . . . . 235

Tobias Leidinger

Environment and Safety

CFD Modeling and Simulation of Heat and Mass

Transfer in Passive Heat Removal Systems . . . . . 238

Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas

|241

249

| | Collecting soil samples.

Research and Innovation

Irradiation Tests of a Flat Vanadium Self- Powered

Detector with 14 MeV Neutrons . . . . . . . . . . . 246

Prasoon Raj and Axel Klix

Nanofluid Applied Thermo-hydro dynamic

Performance Analysis of Square Array

Subchannel Under PWR Condition. . . . . . . . . . 249

Jubair Ahmed Shamim and Kune Yull Suh

| Computational domain created in Star-CCM+.

KTG Inside . . . . . . . . . . . . . . . . . . . . . . 257

238

| | Liquid Volume fraction distribution.

Decommissioning and Waste Management

The Decommissioning of the ENEA RB3

Research Reactor in Montecuccolino . . . . . . . . 241

F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi,

R. Lorenzelli and A. Rizzo

News . . . . . . . . . . . . . . . . . . . . . . . . . 260

Nuclear Today

Czechs and Balances and Why ‘Ugly’

Nuclear Deserves a Political Makeover . . . . . . . 270

Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 236

AiNT. . . . . . . . . . . . . . . . . . . . . . . . . . . .Insert

AMNT 2018: Registration Form . . . . . . . . . . . Insert

Contents

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atw Vol. 63 (2018) | Issue 4 ı April

212

ABSTRACTS | ENGLISH

Euratom: Industry Softens Stance as

Government Lays Out Plans for Transition

NucNet | Page 214

The UK’s nuclear industry has welcomed a government

commitment to continuity with existing

arrangements with Euratom, Europe’s nuclear safety

and research watchdog, a softening of its earlier

stance that the UK needed to stay in the group to

protect vital nuclear research and new-build projects,

and to make sure access to nuclear fuel and

medical isotopes is not disrupted. The next phase of

discussions will focus on the UK’s future relationship

with Euratom. Specific objectives include a close

association with the Euratom Research and Training

Programme, including the Joint European Torus

(JET) and the International Thermonuclear Experimental

Reactor (ITER) projects.

Heat Transfer Systems for Novel

Nuclear Power Plant Designs

Sebastian Vlach, Christoph Fischer and

Herman van Antwerpen | Page 217

This article focuses on designing or modifying heat

exchangers found in the auxiliary systems of any

power plant. The basic premise is to show that the

software provides a one-stop solution for designing

many types of heat transfer systems, where the

interaction bet ween various loops connected by

heat exchangers can be assessed. The nuclear power

plant industry is addressed as the quality control in

the development of the software makes it most

suitable for nuclear related applications. Moreover,

the software discussed has the capability to do

contaminant tracing, which could be very useful

for nuclear contamination studies in designing

specialized ventilation systems. To highlight the

versatility of the software network approach it will

be shown how to model any setup and kind of heat

exchanger such as plate, tube-in-tube, liquid/gas,

finned tube etc. Additionally, the Koeberg pressurized

water reactor steam generator comparison and

the THTR steam generator comparison are shown

as examples.

Experimental and Analytical Tools for

Safety Research of GEN IV Reactors

G. Mazzini, M. Kyncl, Alis Musa and

M. Ruscak | Page221

Current research on nuclear safety in the world, in

addition to supporting existing nuclear power

plants is focused on the more detailed aspects of the

new reactors. The new generation reactors are

expected inter alia to use innovative types of fuel

and new types of coolants, such as e.g. Super-

Critical Water (SCW), supercritical CO 2 , liquid

metals, fluoride salts or high-temperature Helium.

The paper will describe new experimental infrastructure

build recently in Research Centre Řež

under the SUSEN (Sustainable Energy) project and

available analytical tools for supporting safety

research of GEN IV reactors. Two experimental

loops – SCWL (Supercritical Water Loop) and HTHL

(High Temperature Helium Loop) will serve as

in-pile loops in the active core of the research

reactor LVR-15. The paper provides examples of

analyses made using codes ATHLET (supercritical

water) and TRACE (high temperature He) illustrating

process of their assessment and practical use.

Numerical Analysis of MYRRHA Inter-wrapper

Flow Experiment at KALLA

Abdalla Batta and Andreas G. Class | Page 226

The MYRRHA reactor, which is developed at

SCK-SCN in Belgium, represents a multi-purpose

irradiation facility. Its prominent feature is a pool

design with the nuclear core submerged in liquid

metal lead bismuth. During transients between

normal operation and accident conditions decay

heat removal is ensured by forced and natural

convection, respectively. The flow in the gap

between the fuel assemblies plays an important role

in limiting maximum temperatures which should

not be exceeded to avoid core damage. Due to the

scarce database, within the Horizon 2020 – research

and innovation framework program of the EU, the

SESAME project was established to develop and

validate advanced numerical approaches, to

achieve a new or extended validation base and to

establish best practice guidelines including verification

& validation and uncertainty quantification.

In particular the current work supports the

inter-wrapper flow experiment at KALLA.

Heat Balance Analysis for Energy

Conversion Systems of VHTR

SangIL Lee, YeonJae Yoo, Deok Hoon Kye,

Gyunyoung Heo, Eojin Jeon

and Soyoung Park | Page 230

VHTR(Very High Temperature Gas Reactor) with

helium used as a coolant can easily produce heat

required in high-temperature thermochemical process,

and because of low heat output density, the

possibility of core melting is low. In this study, provided

that VHTR is located in the primary system,

the heat conversion system will be discussed in

which hydrogen production and power supply are

possible. In order to control the ratio between power

and hydrogen production, the helium flowing

through nuclear reactor is made to pass through

heat exchanger for hydrogen production and steam

generator or heat exchanger. This study proposes

the whole heat conversion system model, and

carries out thermodynamic feasibility calculation

according to major design variable at each point

and sensitivity analysis for efficiency optimization.

Information Requirements Versus

Confidentiality Obligations – Extension

of the In-Camera Procedure Planned

Tobias Leidinger | Page 235

The justified right of the public to detailed information

on a project requiring nuclear licensing is

opposed by the state’s interest in effective protection

of sensitive data. This conflict is manifested

in licensing procedures but also at court. The differentiated

legal provisions that regulate the balancing

of these conflicting interests are now to be supplemented

by a further facet: An expanded in-camera

trial at court. According to the coalition agreement

of 7 February 2018, the regulation is to take place in

the current 18th legislative period.

CFD Modeling and Simulation of Heat

and Mass Transfer in Passive Heat

Removal Systems

Amirhosein Moonesi, Shabestary,

Eckhard Krepper and Dirk Lucas | Page 238

The CFD-modelling and simulation of condensation

inside passive heat removal systems are presented.

Designs of future nuclear boiling water reactor concepts

are equipped with emergency cooling systems

which are passive systems for heat removal. The

emergency cooling system consists of slightly

inclined horizontal pipes which are immersed in a

tank of subcooled water. The focus of the project is

on detection of different morphologies such as

annular flow, stratified flow, slug flow and plug flow

and also modeling of the laminar film which is

occurring during the condensation near the wall.

The Decommissioning of the ENEA RB3

Research Reactor in Montecuccolino

F. Rocchi, C. M. Castellani, A. Compagno,

I. Vilardi, R. Lorenzelli and A. Rizzo | Page 241

The ENEA RB3 reactor was a 100 Wth research

installation owned and operated by ENEA, in its

center of Montecuccolino near Bologna, from 1971

to 1989. In 1989, the RB3 reactor was shut down,

and in the late 2010 ENEA received by ministerial

decree the authorization to its dismantling, with the

aim of reaching the “green field” status. This paper

presents the three main pillars of the decommissioning

of RB3, namely the strategy and methods

for the dismantling, the strategy and methods for

the radiological characterization of the building,

and finally the strategy and methods for the radiological

characterization of the site.

Irradiation Tests of a Flat Vanadium

Self-Powered Detector with

14 MeV Neutrons

Prasoon Raj and Axel Klix | Page 246

Self-powered detector (SPD) represents a class of

neutron and gamma monitoring instruments used

in the fission reactor cores worldwide. This detector

has inherent advantages of functioning without a

bias voltage, simple measurement scheme, compactness,

ease of maintenance, and high reliability.

We are studying SPD for application as flux monitors

in the European test blanket modules (TBM) of

ITER, fusion reactor under construction in southern

France.

Nanofluid Applied Thermo-hydrodynamic

Performance Analysis of Square Array

Subchannel Under PWR Condition

Jubair Ahmed Shamim and Kune Yull Suh | Page 249

Efficient engineered design of heat transfer and

fluid flow with enhanced heating or cooling requires

two pivotal aspects that must be taken into consideration

for extracting thermal energy from

nuclear fission reactions in order to save energy,

reduce process time, raise thermal rating and

increase the operating life of a reactor pressure

vessel. Hence, one of the major challenges in

designing a new nuclear power plant is the quantification

of the optimal flow of coolant and distribution

of pressure drop across the reactor core.

Recently, nanofluid has gained much renewed

attention as a promising coolant for pressurized

water reactors (PWRs) due to its enhanced thermal

capabilities with least penalty in pressure drop.

Czechs and Balances and Why ‘Ugly’

Nuclear Deserves a Political Makeover

John Shepherd | Page 270

As if Europe does not have enough on its plate

to deal with at the moment – politically and

economically just for starters – could Brussels be on

a collision course with the Czech government over

the countries plans to expand nuclear energy?

There is certainly friction over the issue between

Prague and the European Commission (EC), to put

it mildly. But why?

The veteran head of the Czech Republic’s State

Office for Nuclear Safety, Dana Drábová, last month

accused other EU member states of “pressurising”

Prague over the early closure of its oldest nuclear

reactor units.

Abstracts | English

---

atw Vol. 63 (2018) | Issue 4 ı April

Euratom: Britische Industrie zufrieden

mit Übergangsplänen der Regierung

NucNet | Seite 214

Die britische Nuklearindustrie hat die Zusage der

Regierung begrüßt, die bestehenden Vereinbarungen

mit Euratom, dem europäischen Rahmen

für nukleare Sicherheit und Forschung, aufrechtzuerhalten

und ihren früheren Standpunkt, dass

das Vereinigte Königreich im Euratom-Vertrag

verbleiben müsse, um wichtige Forschungs- und

Neubauprojekte sicherzustellen, und den Zugang

zu Kernbrennstoffen und medizinischen Isotopen

zu gewährleisten, zu relativieren. Die nächste Phase

der Gespräche im Rahmen des Brexit wird sich auf

die künftigen Beziehungen des Vereinigten Königreichs

zu Euratom konzentrieren. Zu den spezifischen

Zielen gehört eine enge Zusammenarbeit

mit den Euratom-Forschungs- und Ausbildungsprogrammen,

einschließlich der Projekte Joint

European Torus (JET) und International Thermonuclear

Experimental Reactor (ITER).

Fortgeschrittene Wärmeübertragungssysteme

für zukünftige Kernkraftwerkskonzepte

Sebastian Vlach, Christoph Fischer und

Herman van Antwerpen | Seite 217

CFD-Systemsimulation mit FlownexSE ermöglicht

es Ingenieuren, einfache und komplexe strömungstechnische

und thermische Netzwerke schnell und

effizient aufzubauen und zu analysieren. Die

Simulation ermöglicht es Ingenieuren, Systeme

aufzubauen, effizient auszulegen und bereits frühzeitig

Schwachstellen in Entwürfen zu finden

sowie geeignete Änderungen und Maßnahmen zu

entwickeln und im Netzwerkmodell zu testen.

Besondere Aufmerksamkeit wird in diesem Artikel

den vielseitigen Möglichkeiten gewidmet, einfache

und komplexe Wärmetauschersysteme der verschiedensten

Arten (Plattenwärmetauscher, Rohrbündel

etc.) für moderne Kernkraftwerke anzuwenden. Als

praktische Beispiele werden gemessene Daten von

den Kraftwerken Koeberg und Hamm-Uentrop mit

den Ergebnissen aus der Simulation verglichen.

Experimentelle und analytische

Werkzeuge für die Sicherheitsforschung

zu GEN-IV-Reaktoren

G. Mazzini, M. Kyncl, Alis Musa und

M. Ruscak | Seite 221

Die aktuelle Forschung zur Sicherheit von

Kernkraftwerken konzentriert sich neben den

Aktivitäten für bestehende Kernkraftwerke auf die

detaillierteren Aspekte neuer Reaktorkonzepte.

Hier werden u.a. innovative Brennstoffe und Kühlmittel

wie z.B. überkritisches Wasser, überkritisches

CO 2 , Flüssigmetalle, Salzschmelzen oder Helium

eingesetzt. Vorgestellt wird dazu die neue experimentelle

Infrastruktur, die im Forschungszentrum

Řež im Rahmen des SUSEN-Projekts (Sustainable

Energy) aufgebaut wurde, sowie die verfügbaren

Analyseinstrumente zur Unterstützung der Sicherheitsforschung

zu GEN IV-Reaktoren.

Numerische Analyse der Zwischenspaltströmung

im MYRRHA-Reaktor mit Ergebnissen

des Strömungsexperiment KALLA

Abdalla Batta und Andreas G. Class | Seite 226

Der am SCK-SCN in Belgien entwickelte MYRRHA-

Reaktor ist eine Mehrzweck-Bestrahlungsanlage.

Sein herausragendes Merkmal ist eine Reaktorkonstruktion

mit einer Kernkühlung aus flüssigem Blei-

Wismut. Bei Transienten zwischen Normalbetrieb

und Unfallbedingungen wird die Wärmeabfuhr

durch erzwungene bzw. natürliche Konvektion

sichergestellt. Die Strömung im Spalt zwischen den

Brennelementen spielt eine wichtige Rolle bei der

Begrenzung von Maximaltemperaturen, die zur

Vermeidung von Kernschäden nicht überschritten

werden sollten. Im Rahmenprogramm Horizon

2020 – Forschung und Innovation der EU wurde

dazu das Projekt SESAME initiiert, um fortgeschrittene

numerische Ansätze zu entwickeln und

zu validieren, die eine neue oder erweiterte

Validierungsbasis für damit verbundene Fragestellungen

zur Verfügung stellen.

Wärmebilanzanalyse für

Energieumwandlungssysteme von VHTR

SangIL Lee, YeonJae Yoo, Deok Hoon Kye,

Gyunyoung Heo, Eojin Jeon und

Soyoung Park | Seite 230

VHTR (Very High Temperature Gas Reactor) mit

Helium als Kühlmittel können Wärme bereit stellen,

die bei thermochemischen Hochtemperaturprozessen

benötigt wird. In Bezug auf die Sicherheit ist

aufgrund der geringen Wärmeleistungsdichte das

Risiko einer Kernschmelze minimiert. Diskutiert

werden Voraussetzungen für die Nutzung von

VHTR für eine Wasserstofferzeugung und Stromversorgung.

Vorgestellt wird ein Gesamtmodell des

Wärmeumwandlungssystems mit einer thermodynamischen

Machbarkeitsberechnung.

Informationsbedarf versus

Geheimhaltungspflichten – Erweiterung

des In camera-Verfahrens geplant

Tobias Leidinger | Seite 235

Dem berechtigten Anspruch der Öffentlichkeit auf

detaillierte Informationen über ein atomrechtlich

genehmigungsbedürftiges Vorhaben steht das

staatliche Interesse an einem effektiven Geheimnisschutz

sensibler Daten gegenüber. Dieser Konflikt

tritt regelmäßig im Genehmigungsverfahren aber

auch vor Gericht zu Tage. Die differenzierten

Gesetzesbestimmungen, die den Ausgleich dieser

widerstreitenden Interessen regeln, sollen nun

durch eine weitere Facette ergänzt werden: Ein

erweitertes In-camera-Verfahren bei Gericht. Nach

dem Koalitionsvertrag vom 7. Februar 2018 soll die

Regelung in der schon laufenden 18. Legislaturperiode

erfolgen.

CFD-Modellierung und Simulation

von Wärme- und Stoffaustausch

in passiven Wärmeabfuhrsystemen

Amirhosein Moonesi, Shabestary,

Eckhard Krepper und Dirk Lucas | Seite 238

Die CFD-Modellierung und Simulation der Kondensation

in passiven Wärmeabfuhrsystemen wird vorgestellt.

Zukünftige Siedewasserreaktorkonzepte

werden mit Notkühlsystemen ausgestattet, die eine

passive Wärmeabfuhr gewährleisten. Das Notkühlsystem

besteht aus leicht geneigten horizontalen

Rohren in einem Wasserbehälter. Der Schwerpunkt

des vorgestellten Projektes liegt auf der Identifikation

verschiedener Morphologien wie Ringströmung,

Schichtenströmung, Schwallströmung

und Pfropfenströmung sowie der Modellierung des

laminaren Films, der bei der Kondensation in

Wandnähe auftritt.

Die Stilllegung der ENEA RB3

Forschungsreaktor in Montecuccolino

F. Rocchi, C. M. Castellani, A. Compagno,

I. Vilardi, R. Lorenzelli und A. Rizzo | Seite 241

Der ENEA RB3-Reaktor war eine 100-Watt-Forschungsanlage,

die von 1971 bis 1989 im Zentrum

von Montecuccolino bei Bologna, Italien betrieben

wurde. 1989 wurde der RB3-Reaktor abgeschaltet

und Ende 2010 erhielt ENEA per Ministerialerlass

die Genehmigung zu seinem Rückbau mit dem Ziel,

den Status „Grünen Wiese“ zu erreichen. Vorgestellt

werden die drei wesentlichen Fragestellungen für

die Stilllegung des RB3: Strategie und Methoden

für den Rückbau, Strategie und Methoden für die

radiologische Charakterisierung des Gebäudes und

schließlich die Strategie und Methoden für die

radiologische Charakterisierung des Standortes.

Bestrahlungstests eines Vanadium-

Detektors mit 14 MeV Neutronen

Prasoon Raj und Axel Klix| | Seite 246

Self-powered Detektoren (SPD) sind eine Klasse

von Neutronen- und Gamma-Überwachungsgeräten,

die weltweit in Kernreaktoren eingesetzt

werden. Diese Detektoren besitzen die Vorteile,

dass keine Spannungsversorgung erforderlich ist,

das Messverfahren einfach und die Detektoreinheit

kompakt, wartungsfreundlich und zuverlässig ist.

SPDs werden im Rahmen des vorgestellten Projektes

für den Einsatz als Flussmonitor in den Blanketmodulen

des in Bau befindlichen Fusionsreaktors

ITER .

Einsatz von Nanofluiden und

thermohydraulische Analyse

für Druckwasserreaktoren

Jubair Ahmed Shamim und Kune Yull Suh | Seite 249

Eine effiziente Auslegung von Wärmeübertragung

und Flüssigkeitsströmung mit verbessertem

Wärme übergang, -transport oder Kühlung bedingt

zwei zentrale Aspekte, die in Kernkraftwerken

berücksichtigt werden müssen: Leistungsdichte

und technische Lebensdauer des Reaktordruckbehälters.

Eine Herausforderung für die Auslegung

neuer Kernkraftwerkskonzepte ist daher die

Quantifizierung einer optimalen Kühlmittelverteilung

und die Verteilung des Druckverlustes über

den Reaktorkern. In jüngster Zeit werden „Nanofluide“

als vielversprechendes Kühlmittel für Druckwasserreaktoren

(DWR) aufgrund verbesserter

thermischer Eigenschaften mit geringst möglichem

Druckabfall diskutiert, die auch Thema dieser

Arbeit sind..

Tschechien und Ausgewogenheit und

warum es die „hässliche“ Kernenergie verdient,

politisch neu bewertet zu werden

John Shepherd | Seite 270

Als ob Europa derzeit nicht genug zu tun hätte, mit

sich selbst – politisch und wirtschaftlich, nur um

zwei Themenbereiche zu nennen – ... könnte jetzt

Brüssel auf Kollisionskurs mit der tschechischen

Regierung zu den Plänen des Landes zum Ausbau

der Kernenergie gehen?

In der Frage zwischen Prag und der Europäischen

Kommission (EC) geht es, um es milde auszudrücken,

sicherlich um Differenzen. Aber warum?

Die langjährige Leiterin der tschechischen

Aufsichtsbehörde für nukleare Sicherheit, Dana

Drábová, warf zudem im vergangenen Monat

anderen EU-Mitgliedstaaten vor, die Regierung in

Prag unter inakzeptablem Druck zu setzen hinsichtlich

der Forderung einer vorzeitigen Stilllegung

ihrer ältesten Kernkraftwerke.

213

ABSTRACTS | GERMAN

Abstracts | German

---

atw Vol. 63 (2018) | Issue 4 ı April

214

INSIDE NUCLEAR WITH NUCNET

Euratom: Industry Softens Stance as

Government Lays Out Plans for Transition

NucNet

The UK’s nuclear industry has welcomed a government commitment to continuity with existing

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

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

sure access to nuclear fuel and medical isotopes is not disrupted.

Energy secretary Greg Clark said in a written statement to

parliament on 11 January 2018 that the government wants

to include Euratom in any implementation period agreed

as part of wider discussions on Brexit and plans to put in

place “all the necessary measures” to ensure that the UK

can operate as an independent and responsible nuclear

state from day one of Brexit and its separation from the

Euratom Treaty, which regulates the nuclear industry and

the movement of nuclear material across Europe.

According to Mr Clark’s statement, the government has

made good progress on separation issues in the last few

months as part of phase one of negotiations with the EU.

Negotiations have covered a set of legal and technical

issues related to nuclear material and waste, and safeguards

obligations and equipment.

The next phase of discussions will focus on the UK’s

future relationship with Euratom. Specific objectives

include a close association with the Euratom Research and

Training Programme, including the Joint European Torus

(JET) and the International Thermonuclear Experimental

Reactor (ITER) projects.

For the nuclear industry, rapid departure from Euratom

without a clear replacement spells disaster. Scientists have

warned that British nuclear stations may not be able to

source nuclear fuel if it cannot be legally transported

across borders. The shipment of medical isotopes used in

scans and cancer treatment might be jeopardised.

European workers on shared research projects, such as

experimental fusion reactors, face an equally uncertain

future without Euratom’s separate guarantees of freedom

of movement.

But the London-based Nuclear Industry Association

(NIA), which represents more than 260 nuclear companies,

cautiously welcomed Mr Clark’s statement, calling it a

“useful and welcome step” in setting out the government’s

approach in seeking to secure equivalent arrangements to

those the UK benefits from as a member of Euratom. The

NIA also welcomed clarity on the government’s intention

to negotiate an implementation period to ensure a smooth

transition from the current to new arrangements.

It warned, however, that there is much still to do in

equipping the UK’s regulator to take on Euratom’s safeguarding

activities. The UK needs to reach post-Euratom

agreements with the International Atomic Energy Agency,

the US, Canada, Australia, Japan and others. It needs to

agree new trading arrangements with the Euratom

community and conclude a new funding agreement for the

UK to continue its work in Euratom’s fusion R&D activities.

“It is vital government continues to prioritise these issues

in the period ahead if there is to be a successful outcome,”

the NIA said.

Unlike the dozens of other regulatory arrangements for

industries such as aviation or pharmaceuticals, Euratom

has been singled out for special treatment through the

Brexit process because it is not technically part of the EU.

Instead, the treaty that established this body to coordinate

Europe’s civil nuclear energy industry was born in parallel

with the birth of the European economic community in

1957. The UK’s participation in Euratom therefore required

a separate legal relationship with the European court of

justice to enforce it.

The nuclear industry had been hoping that because of

this separation from the “mainstream Brexit,” the UK

might decide to remain part of Euratom.

The NIA and the Brussels-based trade body Foratom

both said the UK should maintain its membership. They

argued that the nuclear industry is global, and the ease of

movement of nuclear goods, people and services enables

new build, decommissioning, R&D and other programmes

of work to continue without interruption.

The government insists that leaving Euratom is an

inevitable consequence of Brexit – a position shared by the

European negotiators. But is says it wants continuity of

open trade arrangements for nuclear goods and products

to ensure the nuclear industry is able to continue to trade

across EU borders without disruption.

Support for remaining in Euratom had come not only

from within the industry, but also from politicians.

Conservative MPs said they would for the government to

fight harder for the UK to stay in Euratom. The opposition

Labour Party said Britain should remain in Euratom,

adding it is increasingly clear that the government acted

“recklessly” by giving up on membership.

Scientists said leaving Euratom will cause widespread

confusion and have a potentially devastating impact

on the nuclear industry. They warned of potential problems

related to the transportation of nuclear materials, including

nuclear fuel; research, especially fusion research; and

overseas investment in development of British nuclear

power stations.

Mr Clark’s statement addressed another concern for the

industry – the issue of accessing a skilled pan-European

workforce for the sector once Brexit is complete.

Mr Clark said the nuclear sector needs the workforce for

decommissioning, operation of existing facilities and

new-build projects. He said proposals for the UK’s future

immigration system will be set out shortly and “we will

ensure that those businesses and communities, and

parliament have the opportunity to contribute their views

before making any decisions about the future system”.

Whatever the outcome of negotiations with the EU,

it is vital that the civil nuclear industry has a safeguards

regime that meets international standards. But this

is not dependent on the EU negotiations and the UK

government is well advanced in delivering this plan, the

statement said.

Inside Nuclear with NucNet

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

---

atw Vol. 63 (2018) | Issue 4 ı April

Advertisement

The UK is establishing a legislative and regulatory

framework for a domestic safeguards regime which will

provide legal powers to establish a domestic regime which

the Office for Nuclear Regulation will regulate. It is also

negotiating bilateral safeguards agreements with the

International Atomic Energy Agency and putting in place

bilateral nuclear cooperation agreements with key third

countries.

NIA chief executive Tom Greatrex said the UK industry

and research facilities have been consistently clear with

government about the importance of these issues since

the referendum. “Even with a suitable transition, there

remains much work for the government to do to prevent

the significant disruption that industry is concerned

about.”

Mr Clark’s statement is online:

http://bit.ly/2CQ1wwQ

Fachseminar Nuklearhaftung

Haftung und Deckung im Nuklearbereich

q Das Haftungssystem des Pariser Übereinkommens

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2 Dr. Christian Raetzke ı Rechtsanwalt

2 Achim Jansen-Tersteegen ı Geschäftsführer,

Deutsche Kernreaktor-Versicherungsgemeinschaft

DATF EDITORIAL NOTES

215

Author

NucNet

The Independent Global Nuclear News Agency

David Dalton

Editor in Chief, NucNet

Avenue des Arts 56

1000 Brussels, Belgium

www.nucnet.org

Am 5. September 2018 in Leipzig

Ï Information und Anmeldung: www.conlar.de

Rechtsanwaltskanzlei Dr. Christian Raetzke

Beethovenstraße 19 · 04107 Leipzig

Tel. 0341 – 9999 1444

christian.raetzke@conlar.de · www.conlar.de

Notes

Conlar atw 18-04 75x124.indd 1 18.03.18 15:02

Grafik des Monats

Bundesministerium für Umwelt, Naturschutz, Bau und Reaktorsicherheit (BMUB)

Fachaufsicht

Beteiligungsverwaltung

Fachaufsicht

Bundesaufsicht

Zusammenarbeit im

Länderausschuss für

Atomkernenergie (LAA)

Organigramm der Behördenstruktur

im Rahmen des

Standortauswahlverfahrens

für das Endlager für

hochradioaktive Abfälle.

Bundesamt

für kerntechnische

Entsorgungssicherheit

(BfE)

Bundesgesellschaft

für Endlagerung (BGE)

Bundesamt

für Strahlenschutz (BfS)

Landesministerien

Regulierung

von Endlagern

Planfeststellung

und Genehmigung

von Endlagern

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Private Rechtsform –

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• Standortsuche

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• Betrieb

• Stilllegung

von Endlagern

Wissenschaftliche

Bundesbehörde

für Aspekte

des Strahlenschutzes

Atomrechtliche

Vollzugsaufgaben

Bergrechtliche

Zulassungen und

wasserrechtliche

Erlaubnisse im

Benehmen mit BfE

| | Quelle: DAtF in Anlehnung

an Endlagerkommission

For further details

please contact:

Nicolas Wendler

DAtF

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

kernenergie.de

www.kernenergie.de

DAtF Notes

---

atw Vol. 63 (2018) | Issue 4 ı April

216

CALENDAR

Calendar

2018

08.04.-11.04.2018

International Congress on Advances in Nuclear

Power Plants – ICAPP 18. Charlotte, NC, USA,

American Nuclear Society (ANS), www.ans.org

08.04.-13.04.2018

11 th International Conference on Methods and

Applications of Radioanalytical Chemistry –

MARC XI. Kailua-Kona, HI, USA, American Nuclear

Society (ANS), www.ans.org

12.04.2018

Desalination Powered by Nuclear Energy. Essen,

Germany, Deutsche Meerwasser Entsalzung GmbH

in cooperation with International Atomic Energy

Agency (IAEA) and PowerTech Training Center

( Kraftwerksschule, KWS), www.dme-gmbh.de,

www.iaea.org, www.kraftwerksschule.de

16.04.-19.04.2018

Einführung in die Kerntechnik. Mannheim,

Germany, TÜV SÜD, nucleartraining@tuev-sued.de

16.04.-17.04.2018

VdTÜV Forum Kerntechnik – Sicherheit im Fokus.

Berlin, Germany, VdTÜV mit Unterstützung des

TÜV NORD, des TÜV SÜD und des TÜV Rheinland,

www.tuev-sued.de/tagungen

17.04.-19.04.2018

World Nuclear Fuel Cycle 2018. Madrid, Spain,

World Nuclear Association (WNA),

www.world-nuclear.org

18.04.-19.04.2018

9. Symposium zur Endlagerung radioaktiver

Abfälle. Vorbereitung auf KONRAD – Wege zum

G2-Gebinde. Hanover, Germany, TÜV NORD

Akademie, www.tuev-nord.de/tk-era

22.04.-26.04.2018

Reactor Physics Paving the Way Towards More

Efficient Systems – PHYSOR 2018. Cancun, Mexico,

www.physor2018.mx

24.04.-25.04.2018

Integrated Waste Management Conference.

Penrith, Cumbria, United Kingdom, The Nuclear

Institute, www.iwmeurope.com

08.05.-10.05.2018

29 th Conference of the Nuclear Societies in Israel.

Herzliya, Israel. Israel Nuclear Society and Israel

Society for Radiation Protection, ins-conference.com

13.05.-19.05.2018

BEPU-2018 – ANS International Conference on

Best-Estimate Plus Uncertainties Methods. Lucca,

Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,

ANS, IAEA, NEA, www.nineeng.com/bepu/

13.05.-18.05.2018

RadChem 2018 – 18th Radiochemical Conference.

Marianske Lazne, Czech Republic,

www.radchem.cz

14.05.-16.05.2018

ATOMEXPO 2018. Sochi, Russia,

atomexpo.ru

15.05.-17.05.2018

11 th International Conference on the Transport,

Storage, and Disposal of Radioactive Materials.

London, United Kingdom, Nuclear Institute,

www.nuclearinst.com

20.05.-23.05.2018

5 th Asian and Oceanic IRPA Regional Congress

on Radiation Protection – AOCRP5. Melbourne,

Australia, Australian Radiation Protection Society

(ARPS) and International Radiation Protection

Association (IRPA), www.aocrp-5.org

29.05.-30.05.2018

49 th Annual Meeting on Nuclear Technology

AMNT 2018 | 49. Jahrestagung Kerntechnik.

Berlin, Germany, DAtF and KTG,

www.nucleartech-meeting.com

03.06.-07.06.2018

38 th CNS Annual Conference and 42nd CNS-CNA

Student Conference. Saskotoon, SK, Canada,

Candian Nuclear Society CNS, www.cns-snc.ca

03.06.-06.06.2018

HND2018 12 th International Conference of the

Croatian Nuclear Society. Zadar, Croatia, Croatian

Nuclear Society, www.nuklearno-drustvo.hr

04.06.-05.06.2018

13 th European Nuclear Energy Forum. Bratislava,

Slova Republic, European Commission, ec.europa.eu

04.06.-07.06.2018

10 th Symposium on CBRNE Threats. Rovaniemi,

Finland, Finnish Nuclear Society, ats-fns.fi

04.06.-08.06.2018

5 th European IRPA Congress – Encouraging

Sustainability in Radiation Protection.

The Hague, The Netherlands, Dutch Society

for Radiation Protection (NVS), local organiser,

irpa2018europe.com

06.06.-08.06.2018

2 nd Workshop on Safety of Extended Dry Storage

of Spent Nuclear Fuel. Garching near Munich,

Germany, GRS, www.grs.de

25.06.-26.06.2018

index2018 – International Nuclear Digital

Experience. Paris, France, Société Française d’Energie

Nucléaire, www.sfen.org, www.sfen-index2018.org

27.06.-29.06.2018

EEM – 2018 15 th International Conference on the

European Energy Market. Lodz, Poland, Lodz

University of Technology, Institute of Electrical Power

Engineering, Association of Polish Electrical

Engineers (SEP), www.eem18.eu

24.06.-30.06.2018

ANNETTE Summer School on Nuclear Technology,

Nuclear Waste Management and Radiation

Protection. Turku, Finland, Advanced Networking

for Nuclear Education, Training and Transfer of

Expertise, annettesummerschool.org, www.enen.eu

29.07.-02.08.2018

International Nuclear Physics Conference 2019.

Glasgow, United Kingdom, www.iop.org

22.08.-31.08.2018

Frédéric Joliot/Otto Hahn (FJOH) Summer School

FJOH-2018 – Maximizing the Benefits of

Experiments for the Simulation, Design and

Analysis of Reactors. Aix-en-Provence, France,

Nuclear Energy Division of Commissariat à l’énergie

atomique et aux énergies alternatives (CEA)

and Karlsruher Institut für Technologie (KIT),

www.fjohss.eu

28.08.-31.08.2018

TINCE 2018 – Technological Innovations in

Nuclear Civil Engineering. Paris Saclay, France,

Société Française d’Energie Nucléaire, www.sfen.org,

www.sfen-tince2018.org

05.09.-07.09.2018

World Nuclear Association Symposium 2018.

London, United Kingdom, World Nuclear Association

(WNA), www.world-nuclear.org

09.09.-14.09.2018

21 st International Conference on Water

Chemistry in Nuclear Reactor Systems.

San Francisco, CA, USA, EPRI – Electric Power

Research Institute, www.epri.com

17.09.-21.09.2018

62 nd IAEA General Conference. Vienna, Austria.

International Atomic Energy Agency (IAEA),

www.iaea.org

17.09.-20.09.2018

FONTEVRAUD 9. Avignon, France,

Société Française d’Energie Nucléaire (SFEN),

www.sfen-fontevraud9.org

17.09.-19.09.2018

4 th International Conference on Physics and

Technology of Reactors and Applications –

PHYTRA4. Marrakech, Morocco, Moroccan

Association for Nuclear Engineering and Reactor

Technology (GMTR), National Center for Energy,

Sciences and Nuclear Techniques (CNESTEN) and

Moroccan Agency for Nuclear and Radiological

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

26.09.-28.09.2018

44 th Annual Meeting of the Spanish Nuclear

Society. Avila, Spain, Sociedad Nuclear Española,

www.sne.es

30.09.-04.10.2018

TopFuel 2018. Prague, Czech Republic, European

Nuclear Society (ENS), American Nuclear Society

(ANS). Atomic Energy Society of Japan, Chinese

Nuclear Society and Korean Nuclear Society,

www.euronuclear.org

02.10.-04.10.2018

7 th EU Nuclear Power Plant Simulation ENPPS

Forum. Birmingham, United Kingdom, Nuclear

Training & Simulation Group, www.enpps.tech

14.10.-18.10.2018

12 th International Topical Meeting on Nuclear

Reactor Thermal-Hydraulics, Operation and

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

www.nuthos-12.org

14.10.-18.10.2018

NuMat 2018. Seattle, United States,

www.elsevier.com

16.10.-17.10.2018

4 th GIF Symposium at the 8 th edition of Atoms

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

22.10.-24.10.2018

DEM 2018 Dismantling Challenges: Industrial

Reality, Prospects and Feedback Experience. Paris

Saclay, France, Société Française d’Energie Nucléaire,

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

22.10.-26.10.2018

NUWCEM 2018 Cement-based Materials for

Nuclear Waste. Avignon, France, French

Commission for Atomic and Alternative Energies

and Société Française d’Energie Nucléaire,

www.sfen-nuwcem2018.org

24.10.-25.10.2018

Chemistry in Power Plant. Magdeburg, Germany,

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

05.11.-08.11.2018

International Conference on Nuclear

Decom missioning – ICOND 2018. Aachen,

Eurogress, Germany, achen Institute for Nuclear

Training GmbH, www.icond.de

2019

07.05.-08.05.2019

50 th Annual Meeting on Nuclear Technology

AMNT 2019 | 50. Jahrestagung Kerntechnik.

Berlin, Germany, DAtF and KTG,

www.nucleartech-meeting.com

Calendar

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atw Vol. 63 (2018) | Issue 4 ı April

Heat Transfer Systems for Novel Nuclear

Power Plant Designs

Sebastian Vlach, Christoph Fischer and Herman van Antwerpen

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

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

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

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

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

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

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

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

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

generator comparison are shown as practical examples.

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

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

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

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

nuclear reactor during normal and abnormal conditions.” (Idaho National Laboratory).

217

OPERATION AND NEW BUILD

Based on the laws of thermodynamics 1D system

simulation is extremely robust, fast, and reliable. One

software package for 1D system simulation that gains

more and more attention recently was developed in the

early 1990ies by a South African company, namely M-Tech

Industrial. Initially, Flownex® Simulation Environment

was developed for aerospace applications and the energy

sector. Moreover, nuclear validation and verification were

supervised by the governmental ESKOM institution

through its subsidiary PBMR Ltd., who developed a

high-temperature gas-cooled (pebble-bed) reactor in

cooperation with Jülich Research Centre at that time.

Specifically for the nuclear safety analyses required by

PBMR, the software has Nuclear Quality Assurance

( NQA-1) Certification and its development process is

based on ISO 9001.

System simulation programmes provide engineers and

designers a fast and efficient way to set up simulation

models for simple as well as complex fluid dynamic

networks. Such networks commonly contain several

components such as fans, pumps, heat exchangers etc. that

can be computed almost instantly. Furthermore, dynamics

and the control of such networks can be investigated by

running different operation scenarios, such as start-up,

shut down, and various loading conditions, where steady

state and transient effects are taken into account. Thus,

weak spots within a system can be eliminated during the

design process prior to manufacturing as literally any

modification can be tested virtually.

Subsequently, the user is able to analyse the results very

quickly.

Material data that the software supports can be

gaseous, gas mixtures, as well as incompressible pure

fluids and two-phase pure fluids. The user is able to access

a vast library based on the NIST data base. Hence, complex

flows can be modelled using temperature and pressure

dependent material data as well as multiphase effects like

conden sation, evaporation, and cavitation.

The software is equipped with a vast array of components

that cover most required simulation scenarios.

Those components can be used as single components or as

building blocks of components found in thermal fluid

systems or subsystems.

Building blocks, with various levels of detail are

available to model heat transfer phenomena as shown in

Figure 1. Some of the simple heat exchanger models

utilises the Number of Transfer Units (NTU) Method while

other more complex versions employ a fully discretised

approach to heat exchanger modelling. The heat exchanger

types range from tube to plate heat exchangers that can be

modelled as parallel, counter, or cross flow types. Other

components can be vessels, reactors, tube systems, valves,

pumps, fans, compressors, seals etc. Moreover, a whole

library of com ponents for dynamics and control is available

within the software.

1D System Simulation

Flownex® Simulation Environment includes all the

necessary numerical formulations for solving all important

thermo-fluid physical phenomena and moreover, a modern

Windows-GUI that enables an intuitive and easy interaction

for the user. Therefore, the user can concentrate on

design and optimisation rather than on the complexities

usually associated with operating such calculation software.

Typical simulations are run in real time or in the

order of seconds, which makes parameter studies and

optimisation loops extremely fast and very efficient.

| | Fig. 1.

Library for heat exchangers [1].

| | Fig. 2.

Heat transfer library [1].

Operation and New Build

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

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atw Vol. 63 (2018) | Issue 4 ı April

OPERATION AND NEW BUILD 218

| | Fig. 3.

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

schematic (right) [2].

If one is looking into thermo dynamic analyses, simple

components would be used to represent radiation,

conduction, or convection as shown in Figure 2. Thus,

heat exchangers can be custom-built to answer the

question at hand.

Figure 3 shows a simple custom made plate heat

exchanger consisting of composite heat transfer components

and pipe components. The flow path is represented

with a hydraulic diameter and the flow area. The plates are

represented with heat transfer area and actual metal

thickness. User-specified correlations according to the

plate corrugation profiles are defined allowing for full

discretisation along the flow path that results in accurate

pinch-point calculation and transient response.

Another heat exchanger example is shown in Figure 4

where a finned tube air-water heat exchanger can be seen.

The air-side is modelled as a straight-through flow path

(left to right) whereas the water-side is modelled as an

up-down overall counter flow (right to left) configuration

according to the design of the header box plates. The fully

discretised flow path provides an accurate transient

response. The fin-side pressure drop and heat transfer

correlations can be specified with Chilton-Colburn J-factor

tables.

Heat exchangers are crucial for any power plant design.

Figure 5 shows the schematic of the Koeberg PWR steam

generator and the equivalent model built in the software.

For the dryer/separator a complete phase separation is

assumed. The recirculation flow rate is calculated from

buoyancy-driven flow (red circle) that is dependent on

heat transfer coefficient and flow resistance. The model

also assumes a homogeneous two-phase flow. The Chen

correlation for the shell side was implemented with

scripting. Specific material properties can be implemented

via Engineering Equation Solver (EES) coupling or

scripting if necessary.

Table 1 shows the comparison of measured and

simulated data of the Koeberg PWR (South Africa) steam

generator at 60 % and 100 % power load. The software

shows reasonably good agreement to the measured data,

especially when looking at the recirculation ratio R circ

which is a good indication of the overall calculation

accuracy.

Another power plant example is the Hamm-Uentrop

thorium high temperature reactor (THTR-300, Germany)

power plant. One challenge in modelling the THTR is that

at certain combinations of flow rate and heat input, the

flow could be oscillatory. Several types of oscillation are

possible: density wave, pressure wave, and critical heat

flux (dryout)-related oscillations. Fundamental fluiddynamic

modelling is crucial to detect this, which is

provided in the software. Furthermore, this capability is

critical to determine the minimum flow through a steam

generator because it is typically at low power levels that

the steam flow becomes oscillatory. Figure 6 shows the

schematic of the THTR-300 and an equivalent model built

in the software.

The THTR steam generator plant was modelled to

verify the steady-state performance of the assembled

steam generator model. Figure 7 shows the comparison of

measured and simulated data of the THTR-300 at 40 %

and 100 % power load. The software shows very good

agreement to the measured data. The simulation revealed

| | Fig. 4.

Model of a finned tube air-water heat exchanger with multiple water-side passes and a single air pass (top) and schematic (bottom) [2].

T pi

[C]

T po

p so

[kPa]

p si T si x so ṁ s

[kg/s]

Q boiler

[MW]

R circ

60 % power Koeberg 294 273 4889 5055 195 1.0 341 670 7.0

Simulation 294 273 4889 4919 195 1.0 341 666 6.4

100 % power Koeberg 312 279 4911 5277 220 1.0 618 1143 3.8

Simulation 312 280 4911 4951 220 1.0 618 1092 3.8

| | Tab. 1.

Koeberg PWR steam generator comparison.

Operation and New Build

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

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atw Vol. 63 (2018) | Issue 4 ı April

that the helium-side heat transfer correlation needed to

have an appropriate Reynolds-number dependence as the

error became quite large at lower power or flow levels

neglecting this.

As aforementioned, heat exchangers are crucial for

any power plant design, especially when designing new

power plants. In addition to the heat transfer modelling

capabilities and with respect to nuclear power generation

the software has recently expanded the Generic Nuclear

Reactor model to simulate the latest nuclear reactor

designs of any geometry. Novel nuclear reactor designs

include liquid fuel reactors, liquid-metal-cooled reactors,

and high temperature gas-cooled reactors (HTGR). In

more detail, there are six reactor types that have gained

researches interest all over the world:

• Very High Temperature Reactor,

• Molten Salt Reactor,

• Sodium-Cooled Fast Reactor,

• Supercritical-Water-Cooled Reactor,

• Gas-Cooled Fast Reactor, and

• Lead-Cooled Fast Reactor.

The new “generalized fuel zone” in the GNR model that is

shown in Figure 8 is capable of handling any fuel geometry

and any fluid type. It expands the geometry capability to

plate fuel, cylindrical fuel rods, spherical fuel elements,

irregular cross-section fuel (like the four-lobe cross-shape

produced by the Lightbridge Corporation), as well as

prismatic block fuel used in some HTGRs.

Appropriate pressure drop and heat transfer correlations

can be selected from the built-in library or defined by

the user. For neutronic calculations, the generalized fuel

zone can provide temperature feedback, as well as heat

generation in all solids and in the core coolant.

The default neutronics model that is supplied with the

software is the point kinetic model which requires the

following inputs:

• Temperature feedback coefficients,

• Heat distribution map, and

• Control rod worth vs. position.

This point kinetic model is provided in a user-editable C#

script, which makes it possible to replace the point kinetic

model by linking the simulation model to an external

neutronics code. The scripted neutronics model also makes

it possible for the user to define one’s own feedback

mechanisms based on the design of the specific reactor.

| | Fig. 5.

Koeberg PWR steam generator schematic (left) [2] and simulation model (right).

| | Fig. 6.

Hamm-Uentrop THTR schematic (left) [2] and simulation model (right).

OPERATION AND NEW BUILD 219

| | Fig. 7.

Hamm-Uentrop THTR-300 steam generator comparison experiment (Exp) [3] and simulation (FNX).

Operation and New Build

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

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atw Vol. 63 (2018) | Issue 4 ı April

OPERATION AND NEW BUILD 220

| | Fig. 8.

Schematic of the General Nuclear Reactor (GNR) model [1].

Being able to model all kinds of heat transfer accurately

and to include fission physics makes the software a

valuable tool for every nuclear engineer and power cycle

developer. Figure 9 shows an integrated simulation model

that includes a reactor, steam generator, heat exchange,

and some turbomachinery.

Summary

In order to size control valves or determine the control

strategy for a loop, it is necessary to have the pump

performance curve, the heat exchanger pressure drop and

heat transfer characteristics as well as reactor dynamic

behaviour in one simulation model. In this article, a fast

and efficient solution for designing many types of heat

transfer systems is presented. It was shown how to model

any setup and kind of heat exchanger such as plate, tubein-tube,

liquid/gas, finned tube etc. Flownex® Simulation

Environment offers a straight-forward workflow for

engineers who are involved in designing auxiliary systems

that usually contain one or more heat exchangers, such as

in the power plant industry. The software is a specialized

software (e.g. used by ITER, X Energy, BATAN, Hyundai

Heavy Industries) for sizing specific types of heat

exchangers or for doing basic steady-state and transient

mass-and-energy balances. The value of the software in

this area is that one can really integrate the information

from all available sources into a single representative

model, where one can size all kind of devices, test control

strategies, and do integrated system-level analysis and

design. Furthermore, examples from the nuclear power

plant industry, namely the Koeberg PWR steam generator

and the Hamm-Uentrop THTR-300 steam generator which

demonstrated the software’s usability for nuclear related

work were shown. In addition, the lately incorporated

Generic Nuclear Reactor model was introduced.

Further Reading

| | Flownex® SE: www.flownex.de

| | M-Tech Industrial: www.mtechindustrial.com

| | Idaho National Laboratory: www.inl.gov

References

[1] Flownex (2017) User Manual.

[2] Van Antwerpen, H.: Design and Optimization of Advanced

Nuclear Technologies with 1-d Simulation. 7 th Annual

International SMR and Advanced Reactor Summit 2017,

30-31 March, Atlanta, GA, USA.

[3] Esch, M., Hurtado, A., Knoche, D., and Tietsch, W.: Analysis of the

Influence of Different Heat Transfer Correlations for HTR Helical

Coil Tube Bundle Steam Generators with the System Code TRACE.

Nuclear Engineering and Design, 251, 374-380, 2012.

[4] Van Antwerpen, H., Chi, H., Brits, Y., and Botha, F.: Plant-Wide

Simulation Model for Transient Studies on the Xe-100. 2016 ANS

Winter Meeting and Nuclear Technology Expo, 6-10 November

2016, Las Vegas, NV, USA.

Authors

Sebastian Vlach

Leiter Marketing & Vertrieb

Christoph Fischer (PhD)

CFX Berlin Software GmbH

Berlin, Germany

Herman van Antwerpen (PhD)

M-Tech Industrial (Pty) Ltd

South Africa

| | Fig. 9.

Layout of a complete plant power cycle with an example reactor geometry input map (left) [4].

Operation and New Build

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

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atw Vol. 63 (2018) | Issue 4 ı April

Experimental and Analytical Tools

for Safety Research of GEN IV Reactors

G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak

Current research on nuclear safety in the world, in addition to supporting existing nuclear power plants (PLEX,

mitigation of severe accidents, the development of accident tolerant fuel, decommissioning, etc.), is focused on the

more detailed aspects of the new reactors. The new generation reactors are expected inter alia to use innovative types

of fuel and new types of coolants, such as e.g. Super-Critical Water (SCW), supercritical CO 2 , liquid metals, fluoride

salts or high-temperature Helium. The paper will describe new experimental infrastructure build recently in Research

Centre Řež under the SUSEN (Sustainable Energy) project and available analytical tools for supporting safety research

of GEN IV reactors. Two experimental loops - SCWL (Supercritical Water Loop) and HTHL (High Temperature Helium

Loop) will serve as in-pile loops in the active core of the research reactor LVR-15. The loops insertion in the reactor

LVR-15 requires performing additional safety analyses studying the mutual interference of the loops and the reactor,

especially in conditions of abnormal operation or accident conditions of the loops. The paper will provide examples of

these analyses made using codes ATHLET (supercritical water) and TRACE (high temperature He) illustrating process

of their assessment and practical use. These activities provide significant opportunity for TSO team in building its new

competencies.

Revised version

of a paper presented

at the Eurosafe,

Paris, France, 6 and

7 November 2017.

OPERATION AND NEW BUILD 221

1 Introduction

The Centrum Výzkumu Řež (CVŘ) and

its partners in the Czech Republic and

abroad are supporting the development

[1] of the Generation IV and

Fusion concepts as well as demonstrators

of these technologies such

as ALLEGRO, ALFRED, DEMO and

others. For this reason, the CVŘ has

had a large R&D program financed

from SUStainable Energy (SUSEN)

project and from its continuation

Research 4 Sustenibility (R4S) [2].

The construction and the operation of

the new SUSEN infrastructure was

supported by the grant of the Ministry

of Education, Youth and Sports as the

part of state help for the large research

infrastructure in the Czech Republic

dedicated to the period 2011–2019.

The SUSEN project consists of 4

programs:

1. Technological Experimental Circuits

(TEO)

2. Structural and System Diagnostics

(SSD)

3. Nuclear Fuel Cycle (NFC)

4. Material Research (MAT)

Within this program, several facilities

were designed and built in order to

study and to address new challenges

of such new technologies. In particular,

the paper focuses on two new

loops which are going to be inserted

inside the LVR-15 research reactor

existing in Řež. The LVR-15 is a light

water tank-type research reactor in

operation since 1957. It is placed in a

stainless steel vessel under a shielding

cover, has forced cooling, uses IRT-4M

type fuel and an has an operational

power level of 10 MWt. The reactor

operations run in campaigns that

usually last for 3 weeks, followed by

an outage lasting for 10 to 14 days

necessary for maintenance and fuel

reloading. There can be also other

campaigns which can operate for

‘short-time’ experiments. Some of the

LVR-15 applications are in the field of

material irradiation research and services,

neutron physics, development

and production of new radiopharmaceuticals

[3]. The loops in concern are

the High Temperature Helium Loop

(HTHL) and the Super Critical Water

Loop (SCWL) and their main scope

are to analyse the cladding behaviour

and structural materials under different

pressure, temperature and coolant

media conditions different from the

standard Light Water Reactors (LWR)

technology [2].

In order to get the regulatory

permit for in-pile operation of these

loops in LVR-15, CVŘ has to prepare

an amendment to the Final Safety

Analyses Report (FSAR) containing

safety analyse of the loops under

| | Fig. 1.

CVR Facilities list.

operational and accidental conditions.

Aim of this paper is to present

the methodology and the analyses

done in support of this process, starting

from code benchmarking/assessment

and the methods adopted in preparing

the safety case.

2 Facilities description

The map of experimental facilities put

into operation in 2016 and those

under preparation to be finalized in

2017 is shown in Figure 1 in the

technology – knowledge map.

In particular, the SCWL and HTHL

represent a pioneer and unique experimental

facility for Gen. IV and Fusion.

2.1 SCWL

The SCWL is going to be a part of a

research facility dedicated to GIV

technologies which will focus on

obtaining data in several areas of the

supercritical fields like: corrosion

processes of construction materials in

supercritical water, with influence of

Operation and New Build

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak

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atw Vol. 63 (2018) | Issue 4 ı April

OPERATION AND NEW BUILD 222

radiation field, supercritical water

radiolysis and its influence on materials

and water chemistry, development

and testing of sensors, mostly for

measuring of electrochemical potential

(ECP), testing and optimization of

supercritical water regimes [2]. The

specimens being tested will be placed

into the test chamber located in the

active channel where high pressure/

temperature of SCW flow parameters

will be reached.

The SCWL heart is the active

channel, where water reaches required

parameters (pressure of 25 MPa;

temperature of 600 °C; very clean

demineralised water. After successful

out-of-pile (i.e. non active, without

presence of radiation field) operation,

the active channel will be inserted into

the LVR-15 research reactor core. The

bottom part of the active channel is

then submerged between the core’s

fuel assemblies and will face a neutron

flux of up to 1.5 × 10 18 n/m 2 s (thermal

neutrons) and 3 × 10 18 n/m 2 s (fast

neutrons).

The fluid flows in the SCWL is

shown in Figure 2a while the CAD

sketches is shown in Figure 2b.

The active channel has been

modelled with the use of ATHLET

code in two different configurations

see S3.1:

• the out-of-pile configuration that

takes into consideration only pressure

and temperature conditions;

• the in-pile configuration, with the

channel placed inside the LVR-15

active core, that takes into account

also the gamma heating.

2.2 HTHL

HTHL test facility is designed for the

material testing under the simulation

of Gas-cooled Fast Reactor (GFR)

and/or Very High Temperature Reactor

(VHTR) operational conditions.

The specimens being tested will be

placed into the test chamber located

in the active channel where high

pressure/temperature helium flow

parameters will be reached. In addition

to that exposure, during the

in-pile operation, with the active

channel placed into predefined position

of LVR-15 active core rectangular

grid the irradiation effects on the

samples will be studied. The scheme

of the flows in the HTHL is shown in

Figure 3a while the CAD sketch can

be seen Figure 3b.

The active channel has been

modelled with the use of TRACE

code in two different configurations

see S3.2:

• the out-of-pile configuration that

takes into consideration only pressure

and temperature conditions;

• the in-pile configuration, with the

channel placed inside the LVR-15

active core, that takes into account

also the gamma heating.

Views of the channel and of the

coolant flow pattern can be seen in

Figure 3a and Figure 3b.

The temperature inside the channel

is reached through electrical

heater and the coolant flow circulation

is maintained using a two stages

compressor.

3 Methodology

The methodology used to select the

codes and to perform the analyses for

the amendment for the LVR-15 FSAR

consisted in 3 – independent steps:

• Searching and assessing the codes

ability to simulate helium and SCW

during steady-state and transients

conditions.

• Creating the loops model to be

used for the TH analyses and

developing it based on the steadystate

thermohydraulic parameters

• Performing analyses of the selected

scenarios in order to verify the

safety criteria and obtaining the

necessary data for the structural

analyses.

The present methodology complies

with the [4] IAEA standard in introducing

new research facilities inside

nuclear research installations such as

the LVR-15 reactor.

3.1 ATHLET 3.1A code

ATHLET 3.1 patch A code [5] is a

thermal hydraulic system code

developed by the GRS for simulating

time-dependent phenomena in the

PWRs and BWRs. Furthermore, the

code can also simulate GEN IV working

fluids like helium, liquid metals

and supercritical water.

The heat transfer behaviour in

supercritical water represents a

challenging task mainly connected

with ensuring safety and reliable

operation. Nowadays, the understanding

of the supercritical water

regimes is rather limited, specifically

regarding the close proximity of the

critical point.

For the simulation of supercritical

water, a range of properties approximation

has been extended up to a

pressure of 100 MPa. An additional

module cover the pressure range from

22.5 to 100 MPa. The transition

between subcritical and the supercritical

properties is performed by a

suitable interpolation between these

packages for pressures between 22.0

and 22.5 MPa [5].

In ATHLET 3.1A the selection of

correlations for supercritical water is

performed by switching a built in flag

found in the heat structure module.

A number of six correlations are

available which were tested against

the results obtained by IAEA-benchmark

exercise [6] and three of them

were selected for the purpose of the

certification and further use.

3.2 TRACE 5 Patch 4 codes

The TRACE code has been used as an

alternative to the RELAP5/Mod3.3

code, since US NRC decided to stop

| | Fig. 2a.

SCWL Flow.

| | Fig. 2b.

CAD Sketches.

| | Fig. 3a.

Flow in HTHL.

| | Fig. 3b.

HTHL CAD Sketches.

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| | Fig. 4.

HTHL TRACE Nodalization.

the development of RELAP starting

with next year.

TRACE has been designed to perform

best-estimate analyses of loss- ofcoolant

accidents (LOCAs), operational

transients, and other accident

scenarios in pressurized light-water

reactors (PWRs) and boiling lightwater

reactors (BWRs). It can also

model phenomena occurring in

experimental facilities designed to

simulate transients in reactor systems.

Models used include multidimensional

two-phase flow, none quilibrium

thermo-dynamics, generalized heat

transfer, reflood, level tracking, and

reactor kinetics. In addition, TRACE is

able to simulate several other coolants

such as helium and water in subcooled

condition and atmospheric pressure

(LVR-15 conditions). [7], [8]

For this reason, TRACE code was

selected and used for the simulation in

the Helium at 7 MPa with a temperature

rise from 200 °C up to 900 °C

(nominal parameters for HTHL). The

correlation adopted for simulating the

heat transfer from heat structures to

the helium coolant and vice versa

implemented in TRACE are Gnielinsky

and El Genk [7-9].

3.3 Codes assessment

The code assessment was done by

benchmarking of the codes with

available experimental results done in

different facilities around the world.

One of the most important steps

was selecting the code that can

perform the heat transfer calculation

under the high temperature He or

SCW conditions along with adequate

correlations [10], [11]. In the case of

ATHLET, the code was carefully assed

and benchmarked with experimental

results of a project coordinated by

IAEA [6] for steady state and with

Chinese SWAMUP facility [12] for

| | Fig. 5.

SCWL ATHLET Nodalization.

the transition from supercritical to

subcritical condition.

The aim of this analyses was to

simulate the deterioration phenomenon

[9] of heat transfer with fluid

transiting between subcritical and

supercritical condition. According to

Ref. [6], Mokry, Gupta and Watts-

Chou correlations show acceptable

prediction capabilities of the Heat

Transfer Coefficient (HTC). Both our

analyses and IAEA CRP program

concluded that an uncertainty for

calculating HTC is about ±25% while

the calculating wall temperature was

between ±10 to 15 %. As a result of

this exercise, the code certification

was obtained from SONS (State of

Office for Nuclear Safety) in March

2017 for using the code in simulating

the CVŘ SCWL.

The TRACE assessment was done

with the data available from the

project GoFastR [13] financed by the

EC in the Framework Program 7, in

particular with data related to the

HE-FUS3 facility [14], [15]. The

facility operational parameters are

similar to the HTHL.

The TRACE HE-FUS3 thermal hydraulic

model was developed and

compared with experimental data

from steady state loop operation and

selected transients. The comparison

showed that the TRACE T/H model

can simulate the helium temperatures

as well as the piping wall temperatures

along the different sections

of the facility accurately. After a sensitivity

analysis, the electrical heater

power has been lowered to 10.76 kW.

The certification for TRACE code was

obtained from SONS in December

2016 by CVŘ for simulating water in

PWR condition, sub-cooled water at

atmospheric pressure (such as LVR-15

operational condition) and helium

behaviour in the range of 7 MPa for a

temperature range between 200 to

900 °C. [16]

3.4 Model description for

HTHL and SCWL

The HTHL and SCWL are similar

experimental facilities characterized

by 2 steps upward and downward

flows, although some major differences

exist in the design. In particular,

the HTHL active channel contains

all necessary components for heat

transfer inside except of the compressor

and the main compensator, which

are located in the chemical control

system. The Figure 4 shows the

TRACE nodalization containing simulated

components.

The SCWL is different in such way

that it needs some extra components

larger than the HTHL to help the sub

critical water to become gas. For this

reason additional axillary facilities,

such as a recuperator, cooler, pump,

compensator and other 4 sections of

electrical heater are located in a

different building along with the

chemical control system.

The ATHET SCWL loop model

shown in Figure 5 is focused mainly

on the active channel from inlet to

outlet, although all the previous

components are also simulated as a

part of the primary and the secondary

circuits. In addition to the primary

and the secondary circuits of the

SCWL, there is the third open loop

representing the active channel position

into the LVR-15 core and providing

additional heat transfer between

active channel and reactor coolant.

3.5 Analysed Scenarios

The planned in-pile operation of both

loops requires an amendment of the

LVR-15 Final Safety Report providing

thermohydraulic and structural integrity

analyses during normal operation

OPERATION AND NEW BUILD 223

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OPERATION AND NEW BUILD 224

Normal operating

conditions

Steady State

LVR-15 Start up

LVR-15 Shutdown

Loops Start up

Loops Shutdown

and during Loss of Flow Accident

( LOFA) and Loss of Coolant Accident

(LOCA) accident conditions. In particular,

the structural integrity analyses

required the temperature profile

inside the Pressure Envelope (PE) as

boundary condition. For this reason

the normal operation and abnormal

operation conditions were calculated

using TRACE and ATHLET codes with

very narrow mesh nodal distribution

in the PE. For structural integrity

following criteria and limitation due

to the non-boiling condition in LVR-15

were used:

1. PE maximum temperature during

normal/abnormal transients is less

than 450 °C.

2. PE maximum temperature during

accident conditions is less than

500 °C.

3. Aluminium surface of the Receiver

maximum temperature in contact

with LVR-15 coolant less than 45 °C

during normal/abnormal conditions.

4. Aluminium surface of the Receiver

maximum temperature in contact

with LVR-15 coolant less than 60 °C

during accident conditions.

In the case of accident conditions,

both active channels of HTHL and

SCWL will have to be replaced. The

analysed scenarios are described in

the Table 1.

4 Illustrative results

The results described in the paper

refer to the simulations of SCWL and

Pressure

tests

(not simulated)

| | Tab. 1.

Operational and Accident Scenarios Description.

Abnormal

conditions

Switch off Loops Electrical

Heater for 1 min.

LVR-15 SCRAM and switch off

of Loops Electrical Heater

at t = 0 s + pump trip after 1 min.

Switch off Loops Electrical Heater

at t = 0 s + LVR15 SCRAM and

Pump Trip after 3 min.

Parameter Value Unit

Pressure 25 MPa

Inlet Flow

Temperature

Outlet Flow

Temperature

Max Flow

Temperature

Sample Area

Mass flow

| | Tab. 2.

SCWL main parameters calculated during

steady state.

HTHL during the steady state operation

with continuing in LOFA condition.

The results represent an extract

of the large number of calculations of

various combinations of operational

transients with the aim to demonstrate

the capabilities of the codes to

simulate behaviour the loops.

4.1 SCWL steady state and

LOFA analyses

The main parameters for the steady

state are shown in Table 2. The whole

steady state calculation was rather

long due to some inertia of the system.

The computer model simulated

behaviour during the transient of all

heat structures representing the

complete piping system. In the calculation

some numerical instability

complicated the steady state due to

Accident

conditions

385 ºC

406 ºC

600 ºC

35 %

By pass flow 65 %

Mass flow 200 kg/h

Loss of Flow Accident

(LOFA)

Loss of Coolant Accident

(LOCA)

small dimensions of the component

facing the deterioration flow phenomenon

during the heating up process.

For these reasons, the whole steady

state was completed in 25,000 s,

where 15,000 to 20,000 s were needed

to adjust the steady state and the

rest 5,000 s were used to verify the

steady behaviour of the main parameters.

After this period the model simulated

the accident scenario – LOFA

without the reactor SCRAM in order

to maximize the consequences and to

calculate the time to reach temperature

of PE (AC) 500 °C.

The scenario is described in the

following steps:

1. Pump stops in 1 s after the initialization

event (25,001 s)

2. Active channel internal electrical

heaters shut down to 0 % on the

nominal power in 7s (25,007 s)

3. The LVR-15 SCRAM starts at 40 s

when the maximum temperature

in the PE rises above the 500 °C.

(25,040 s)

4. The whole transient is completed

in 15,000 s (40,000 s), when the

SCWL and LVR-15 are in the

controlled cold state.

The Figure 6 and Figure 7 represent

the SCW maximum temperature

calculated in the sample area and the

outlet temperature from the active

channel, while the Figure 8 shows the

maximum temperature of the PE,

where there is the neutron flux peak

in the Boltzmann distribution.

4.2 HTHL steady state and

LOFA analyses

The design conditions calculated for

the active channel are described in

Table 3. And they are mainly summarized

as reported:

1. Mass flow rate of 0.0105 kg/s

2. Design pressure of 7 MPa

3. Design electrical heater power of

11.85 kW

4. Cold helium temperature of 210 °C

| | Fig. 6.

SCWL Coolant Maximum Temperature in LOFA.

| | Fig. 7.

SCWL Active Channel Outlet Temperature in LOFA.

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Location of the Thermocouples

Thermocouple

| | Fig. 8.

SCWL Maximum EP Temperature in LOFA.

Parameter Value Unit

Pressure 7 MPa

Inlet Flow

Temperature

Max Flow

Temperature

Maximum AC

Pressure Envelop (PE)

Temperature

210 ºC

900 ºC

450 ºC

Mass flow 40 kg/h

Inlet into the interpiping space of the reheater

Output from the interpiping space of the reheater

Entry into the test chamber

Inlet to the reheater piping space

Output from the reheater piping space

Output from the primary side of the heat exchanger

Maximum helium temperature

| | Tab. 4.

Thermocouples position and description.

T1

T2

T3

T4

T5

T6

Tmax

OPERATION AND NEW BUILD 225

| | Tab. 3.

HTHL main parameters calculated during

steady state.

The steady state simulation was

run in null transient mode for 5,000 s

and the stabilized conditions were

reached after 3,500 s. The LOFA

transients was characterized by an

immediate safety shutdown of the

reactor due to the loss of power. As a

result of the SCRAM, the temperature

went immediately down following the

heat generated by decay gamma flux.

Figure 9 shows the calculated

temperature for various thermocouples

positions (according to

Table 4), while Figure 10 represents

the maximum temperatures in the

HTHL PE.

| | Fig. 9.

HTHL Helium temperatures during LOFA.

5 Conclusions

The article provides a brief introduction

about the SUSEN project and the

experimental facilities built in CVŘ in

the Czech Republic for research and

development in support of the safe,

reliable and long‐term sustainable

operation of existing energy facilities

and in development of GIF IV and

fusion technologies. The SUSEN

R&D activities include four complementary

programmes, mentioned in

the introduction, which are focused

on material science, thermal hydraulics,

neutronics, radiation protection,

nuclear chemistry, waste management

and environmental studies. A

significant part of the research programme

is devoted to HTH and SCW

experimental loops, which are going

to be installed into the active core of

the research reactor LVR-15. Both of

| | Fig. 10.

HTHL PE temperature during LOFA.

these unique facilities are challenging

to model and the selection of appropriate

codes was a demanding process.

A special methodology was used for

assessing the abilities of the codes to

simulate these advanced coolants and

to obtain regulatory certificate/ permit

for their use in operational and accident

conditions and for preparation of

the amendment of the LVR-15 FSAR.

These presented activities represent

only starting steps for the further

codes validation which will be based

on benchmarking of the codes with

experimental data provided by the

SCWL and HTHL loops in their

experimental campaigns.

Aknoledgment

The authors would like to thank

Mr. Miroslav Hrehor and Dr. Vincenzo

Romanello for their kind revisions and

suggestions.

The presented work was financially

supported by the Project CZ.02.1.01/

0.0/0.0/15_008/0000293: Sustainable

energy (SUSEN) – 2 nd phase,

realized in the framework of the

Operation and New Build

Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak

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OPERATION AND NEW BUILD 226

European Structural and Investment

Funds.

This work has been supported

by the Project CZ.02.1.01/0.0/0.0/

15_008/0000293: Sustainable energy

(SUSEN) – 2 nd phase realized in the

framework of the European Structural

and Investment Funds.

References

[1] CVR Annual Report 2016.

[2] http://susen2020.cz/

[3] http://cvrez.cz/en/infrastructure/

research-reactor-lvr-15

[4] IAEA, Standards Safety in the Utilization

and Modification of Research Reactors”,

Safety Standard n° SSG-24, VIENNA,

2012.

[5] ATHLET 3.1A, 2016 User manual:

ATHLET Mod 3.1 Cycle a, G. Lerchl,

H. Austregesilo, P. Schoffel, D. von

der Cron, F. Weyermann, March 2016.

[6] Heat Transfer Behaviour and Thermohydraulics

Code Testing for Supercritical

Water Cooled Reactors (SCWRs),

IAEA. http://www-pub.iaea.org/

books/IAEABooks/10731/Heat-

Transfer-Behaviour-and-Thermo-

hydraulics-Code-Testing-for-

Supercritical-Water-Cooled-R

[7] TRACE V5.840 Theory Manual,

U.S. Nuclear Regulatory Commission,

Washington DC, March 2013.

[8] TRACE V5.840 User’s Manual, Volume 1:

Input Specification, U.S. Nuclear

Regulatory Commission, Washington

DC, February 2014.

[9] TRACE V5.840 User’s Manual, Volume 2:

Modelling Guidelines, U.S. Nuclear

Regulatory Commission, Washington

DC, February 2014.

[10] G. Mazzini et al., ATHLET 3.1A

SIMULATION CAPABILITIES FOR SUPER-

CRITICAL STATE, CVR 1581, 1.1.2017.

[11] G. Mazzini et al., ATHLET 3.1A HEAT

TRANSFER ASSESMENT FOR SUPER-

CRITICAL WATER, CVR 1582, 1.1.2017.

[12] G. Mazzini et al., ATHLET 3.1A

CAPABILITIES IN SIMULATING SWAMUP

FACILITY IN SCW CONDITIONS, CVR

1583, 1.1.2017.

[13] M. Polidori, HE-FUS3 Benchmark

Specifications, GoFastR-DEL-1.5-01,

Rev. 0, ENEA, July 2011.

[14] M. Polidori, HE-FUS3 Experimental

Campaign for the Assessment of

Thermal-Hydraulic Codes: Pre-Test

Analysis and Test Specifications,

Report RSE/2009/88.

[15] M. Polidori et al, HE-FUS3 Benchmark

Results, GoFastR-DEL-1.5-6, Rev. 0,

November 2012.

[16] Miloš Kynčl, Development and Assessment

of TRACE HTHL-2 Facility Thermal

Hydraulic Model, Internal Project Status

Report, CVŘ 1334, March 2017.

Authors

G. MazziniM. Kyncl

Alis Musa

M. Ruscak

Centrum Vyzkumu Rez (CVŘRez)

Hlavní 130

250 68 Husinec – Řež,

Czech Republic

Numerical Analysis of MYRRHA Interwrapper

Flow Experiment at KALLA

Abdalla Batta and Andreas G. Class

Introduction The MYRRHA reactor, which is developed at SCK-SCN in Belgium, represents a multi-purpose

irradiation facility. Its prominent feature is a pool design with the nuclear core submerged in liquid metal lead bismuth.

During transients between normal operation and accident conditions decay heat removal is ensured by forced and

natural convection, respectively. The flow in the gap between the fuel assemblies plays an important role in limiting

maximum temperatures which should not be exceeded to avoid core damage. The term inter-wrapper flow (IWF)

describes the convection in the small gap between the wrapper tubes of neighbouring fuel assemblies (FAs). It plays an

important role for passive decay heat removal (DHR).

Based on numerous experiments

several correlations have been proposed

for the flow within wirewrapped

rod bundles. However, for

the flow within the gap between

neighbouring bundles only few

studies are reported. Recently [1]

reviewed the existing correlations by

Rheme [2], Baxi & Dalle Donne [3]

Cheng and Tordreras [4], and Kirillov

[5] for the pressure-drop in wirewrapped

rod bundles. The existing

correlations were compared to all the

available experimental data and

showed that agreement of approximately

±20 % can be expected. For

the inter-wrapper flow within the

gap only few studies exist, see [6].

Due to the scarce database, within the

Horizon 2020 – research and innovation

framework program of the EU,

the SESAME project was established

to develop and validate advanced

numerical approaches, to achieve a

new or extended validation base and

to establish best practice guidelines

including verification & validation

and uncertainty quantification, see

[7]. In particular the current work

supports the inter-wrapper flow

experiment at KALLA. Three fuel

assemblies including the gap flow are

studied covering the full range of

thermo- hydraulic conditions expected

in the reactor application. For this

purpose, an experimental test matrix

has been established which covers

relevant scenarios. The aim of our

numerical pre-test study is to help the

design of the experiment. The current

study applied RANS-CFD methods for

design support of the experiment. In

the body of this compact the experiment,

the corresponding numerical

model, and preliminary numerical

results are provided.

1 Experimental setup

The KALLA experiment investigates

IWF between three bundles which

are thermally connected by a gap.

Figure 1 shows a cross-sectional view

of the test section which consists of

three ducts representing the fuel

assemblies. Each duct contains 7 wirewrapped

electrically-heated pins

representing the fuel rods. The gap

between the channels, i.e. assemblies,

is filled with liquid metal, so

that strong thermal coupling exists

between neighbouring assemblies.

The test matrix covers independent

variation of flow and thermal conditions

in both the gap and the bundles.

Detailed description of the experiment

is reported in [8]. The geometrical

parameters of the bundle and the

nomenclature are also shown in

Figure 1. The experimental loop

facility THESYS at KALLA and the

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| | Fig. 1.

SCWL Coolant Maximum Temperature in LOFA.

| | Fig. 2.

Left: Experimental loop facility THESYS at KALLA showing location where the inter wrapper flow

experiment (see Figure 3) will be installed; right: flow diagram for the IWF tests with four parallel

channels; the valves V2.1-V2.3 control the flow through the assemblies Q1-Q3. V.2.4 controls the

flow in the gap [8].

location where the IWF experiment

will be installed is shown in Figure 2

left. Figure 2 right shows the flow

diagram of the IWF tests with four

parallel channels representing the

three assemblies (Q1-Q3) and the gap

( illustrated by the box containing

Q1-Q3). The flow and temperature

within each assembly and the gap can

be set individually by choosing valve

openings (V2.1-V2.4) and heating

rates according to the KALLA test

matrix. Figure 3 shows the geometry

of the IWF test section.

and mesh resolution for the thermoshydraulic

investigation of the gap and

the bundle. In particular, we include

the upstream components to verify

their influence on the flow field within

the test section. We employ the k-ε

turbulence model and the commercial

CFD-code Star CCM+. Our first

studied case (i) focuses on the gap

| | Fig. 3.

Geometry of the IWF test section, dimensions are in mm, the heated part

of the bundle is marked red on the left side of the figure, 600 mm, [8].

flow and our second case (ii) on the

fuel assembly. For the study of case (i)

a computational domain including

the lower flow distributer, riser pipe

( including venture tube), upper flow

vessel, and the gap are considered (for

corresponding technical drawings of

components refer to Figure 3). For the

study of case (ii) the computational

domain includes the lower flow distributer,

riser pipe (including venture

tube), one inlet expansion and a single

7-pin bundle. Flow properties of the

liquid metal Lead-Bismuth eutectic at

200 °C are employed. Note that corresponding

upstream pipes and flow

conditioners are modelled so that

all relevant geometric details are

captured. Quantifying the effect of

the flow conditioning sections is

important for future simulations, as it

would enable the use of a simpler

computational domain, which still

provides accurate results. In the future

post-test analysis, the smallest representative

computational domain (e.g.,

potentially without flow conditioner

etc.) will be used to compose a fully

coupled thermos-hydraulic simulation

of the three bundles including

the IWF in the gap. Figures 4 left

and right show the computational

domains for the pre-test studies

OPERATION AND NEW BUILD 227

2 Numerical study

A comprehensive analysis of the

experiment requires efficient simulations.

In the pre-test analysis of the

hydraulics separate simulations of the

gap region and the fuel assembly are

performed. In a first step, we determine

suitable computational domains

| | Fig. 4.

Computational domain for IWF-gap (left) and bundle (right) including the upstream domains.

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OPERATION AND NEW BUILD 228

of cases (i) and (ii), respectively.

Obviously, a substantial effort was

undertaken to include the upstream

flow domain, so that the inflow into

the fuel assembly and the gap are

properly represented in the flow

simulations.

Since we have less experience with

the gap region, and in particular, the

applicable turbulence regime we have

considered 3 cases corresponding to

laminar flow, transitional flow, and

fully developed turbulence, respectively.

This covers the flow range 0.17

to 0.86 kg/s (Re = 1,250 to 6,250),

proposed by the test matrix. For the

investigation of the fuel assembly, we

consider the nominal flow rate, i.e. the

maximum flow rate planned in the

test matrix. This corresponds to a flow

rate of 3.58 kg/s and Re = 8,910

where Re is based on the bundle

hydraulic diameter. All cases considered

in the experimental test matrix

are within the range of transitional

flow according to Cheng and Todreas

[4] (see next section on correlations)

so that no distinction of various flow

regimes is needed for the comparison

to correlations.

2.1 Inter-wrapper flow gap

region

The objective of case study (i) is to

investigate the effects of all upstream

components on the flow distribution

entering the gap, i.e. the inter wrapper

flow region. This study employs the

computational domain shown in

Figure 4 left. For the simulation, a

mesh with approximately 0.72 million

cells has been generated. The investigated

range of flow rates results in

turbulent flow in all components

upstream of the gap, since the

Reynolds- numbers based on pipe

diameter varies between 5,200 and

26,000. However, within the gap

the Reynolds-number based on gapwidth

equates to 1,250 to 6,250 corresponding

to the transitional regime of

turbulence. The pressure drop along

the gap accounts for about 20 % of the

total pressure drop. Since we are

interested in accurately predicting the

upstream flow in the gap region the

use of a turbulent model is mandatory.

Moreover, in order to judge the uniformity

of the flow entering the gap

there is no need to use a very accurate

result within the gap. Thus, a high-

Reynolds-number turbulence model

using automatic wall functions is

used. Figure 5 shows the velocity

vectors in the gap entrance region for

the case where the Reynolds-number

is 5,200 based on pipe diameter

(Reynolds-number is 1,250 based on

the gap hydraulic diameter). We

observe that the flow within the gap

becomes near uniform after a short

length, which does not exceed 10 %

of the length of the gap region. The

heated zone starts further downstream

approximately at half the

length of the gap region. For higher

Reynolds-number a qualitative similar

result is obtained. For future simulations

aiming at accurately simulating

the temperature field, we conclude

that the effect of upstream components

is negligible. In Table 1 the

pressure drop across the simulated

region is compared to design values

for three selected cases covering

the full range of flow rates. Design

values are calculated using lumped

parameter models. Both results agree

reasonably well, indicating that

lumped parameter models well

describe the flow in the gap.

2.2 Flow within a single

wire-wrapped rod bundle

As in the previous study, we aim at

investigating whether the upstream

region that conditions the flow

entering the wire-wrapped bundle

influences the flow in the heated

section of the bundle. Here, i.e. in case

study (ii), a single Reynolds-number

of 8,900 based on the bundle hydraulic

diameter is considered. This

corresponds to the nominal flow

rate as well as the maximum flow

rate intended in the experimental

tests. The computational domain of

Figure 4 right uses approximately

1 million cells. Figure 6 shows the

velocity magnitude within the bundle.

At the entrance, we still observe pronounced

non-uniformities of the flow

distribution. These quickly equilibrate

so that a more-uniformly distributed

flow is observed well before the

heated section of the bundle is

reached (for the location of the heated

region refer to Figure 3).

This result suggests that inflow

effects are negligible for the intended

thermal analysis of the bundle. Thus

in a second simulation we remove the

flow-conditioning region to reduce

the size of the considered flow

domain. To validate our simulation

results we use higher mesh resolution

within the smaller domain. Figure 7

shows the pressure along two selected

axial lines, which are depicted in the

small inset. The influence of the wirewrap

manifests in the periodical

modulation of the pressure profile.

Obviously, development effects have

decayed at a length of approximately

100 mm. We compute the pressure

| | Fig. 5.

Velocity vectors within the gap upstream region.

Flow rate

[kg/s]

design Δp tot ,

[Pa]

CFD Δp tot , [Pa]

1. 0.86 15350 13500

2. 0.688 9964 -

3. 0.516 5667 5500

4. 0.344 2586 -

5. 0.172 663 850

| | Tab. 1.

Comparison of design values evaluated by lumped parameter model

versus computed pressure drop across the test section including the flowconditioning

components.

| | Fig. 6.

Velocity magnitude within bundle showing non-uniformities of flow distribution at leftmost plane and

more-uniformly distributed flow in subsequent planes.

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| | Fig. 7.

Pressure along two selected axial lines in the wire-wrapped rod bundle.

The inset specifies location of lines.

drop using data at corresponding

wire-wrap positions, i.e. from axial

positions 0.065 m to 1.268 m. The

mean pressure drop is 946 Pa/m.

2.3 Model validation

In this subsection, results of our

numerical study are compared to the

simplified Cheng and Todreas [1986]

correlation. The correlation was

recently recommended in (1) to

predict pressure drop (Δp) in bundles

with an accuracy of ±20 %. It applies

for a wide range of Reynolds- numbers.

The friction factor (f) is defined in

eq. 1, where d h,bdl , L, and u b

2

are

hydraulic diameter, length, and

average axial bundle velocity, respectively.

(1)

The correlation for f reads

for Re < Re L

for Re L ≤ Re ≤ Re T

for Re > Re T (2)

where

Re L = 300 x 10 1.7(P/D−1.0) (3)

Re T = 10,000 x 10 0.7(P/D−1.0) (4)

ψ = log(Re/Re L ) / log(Re T /Re L ) (5)

C fL = (-974.6 + 1612.0(P/D) −

598.5(P/D) 2 )(H/D) .06-0.085(P/D)

(6)

C fT = (0.8063 − 0.9022(log(H/D)) +

0.3526(log(H/D)) 2 ) ×

(P/D) 9.7 (H/D) 1.78-2.0(P/D) (7)

We compare the nominal flow case

of 3.580 kg/s which corresponds to a

velocity of 0.2 m/s and Re is 8910,

which is in the transient region.

According to eqns (3) and (4), Re L and

Re T are 902 and 15735, respectively.

The calculated friction factor f

equates to 0.0557. This corresponds

to a pressure drop in the bundle of

1407.2 Pa. The predicted pressure

drop resulting from the CFD study is

1,138 Pa. The difference is near 19 %,

which lays within the accuracy limits.

In future thermos-hydraulic simulations,

the current model can be

applied. For posttest analysis, additional

sensitive studies might be

necessary to further reduce the

uncertainty.

Conclusions

The flow in the gap between neighbouring

fuel assemblies plays an

important role in transients between

forced and natural convection. At

KALLA an experiment on the interwrapper

flow is currently setup and

accompanied by pre-test numerical

CFD studies. These proof that both

the flow in the gap region and the

fuel bundle are not influenced by the

upstream flow-conditioning region.

Moreover, development length are

much shorter than the unheated

length of the test section, so that

the thermal field is uninfluenced by

flow non-uniformities. Preliminary

comparison of pressure losses computed

by CFD and correlation provide

reasonable agreement for both the

gap and bundle. The result of our

study enters pre-test studies of the

thermal field within the EU-H2020

SESAME project. There complete

simulation of the test section consisting

of three bundles connected

by the gap region including conjugate

heat transfer is performed.

Acknowledgement

This project has received funding from

the Euratom research and training

programme 2014-2018 under grant

agreement No 654935 and from the

AREVA Nuclear Professional School.

References:

[1] Chen, S.; Todreas, N.; Nguyan, N.

(2014). Evaluation of existing correlations

for the prediction of pressure drop

in wire-wrapped hexagonal array pin

bundles. Nuclear Engineering and

Design 267, pp. 109 – 131

[2] Rehme, K. (1973). Pressure drop

correla tions for fuel element spacers.

Nuclear Technology 17, 15–23.

[3] Baxi, C.B., Dalle Donne, M., (1981).

Helium cooled systems, the gas cooled

fast breeder reactor. In: Fenech, H. (Ed.),

Heat Transfer and Fluid Flow in Nuclear

Systems. Pergamon Press Inc.,

pp. 410–462.

[4] Cheng, S.-K.; Todreas, N. (1986). Hydrodynamic

models and correlations for

bare and wire-wrapped hexagonal rod

bundles - Bundle friction factors,

subchannel friction factors and mixing

parameters. Nuclear Engineering and

Design 92 (2), 227 – 251.

[5] Kirillov, P.L., Bobkov, V.P., Zhukov, A.V.,

Yuriev, Y.S., (2010). Handbook on

Thermo hydraulic Calculations in

Nuclear Engineering. Thermohydraulic

Processes in Nuclear Power Facilities,

vol. 1. Energoatomizdat, Moscow.

[6] Kamide, H.; Hayashi, K.; Toda, S. (1998).

An experimental study of intersubassembly

heat transfer during

natural circulation decay heat removal

in fast breeder reactors. Nuclear

Engineering and Design 183, 97 – 106.

[7] http://sesame-h2020.eu/

[8] Pacio, J, et. al. (2016), Deliverable 2.10 –

KALLA Inter- wrapper flow setup for

SESAME (thermal hydraulics Simulations

and Experiments for the Safety

Assessment of MEtal cooled reactors)

project, activity: NFRP-01-2014

Improved safety design and operation

of fission reactors, H2020 Grant

Agreement Number: 654935.

Authors

Abdalla Batta

Andreas G. Class

AREVA Nuclear Professional School

Karlsruhe Institute of Technology

Karlsruhe, Germany

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OPERATION AND NEW BUILD 230

Heat Balance Analysis for Energy

Conversion Systems of VHTR

SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park

VHTR (Very High Temperature gas Reactor) which helium is used as a coolant can easily produce heat required in

high-temperature thermochemical process, and because of low heat output density, the possibility of core melting is

low. Helium has the advantage of safety, and the coolant can become super high temperature, thereby power production

as well as hydrogen production application is possible. In this study, provided that VHTR is located in the primary

system, the heat conversion system will be discussed in which hydrogen production and power supply are possible. In

order to control the ratio between power and hydrogen production, the helium flowing through nuclear reactor is made

to pass through heat exchanger for hydrogen production and steam generator or heat exchanger. Power production was

made to be composed of ultra-super critical steam cycle (USC) and supercritical CO 2 (SCO 2 ) cycle so that efficient

operation condition can be selected. This study proposed the whole heat conversion system model, and carried out

thermodynamic feasibility calculation according to major design variable at each point and sensitivity analysis for

efficiency optimization.

1 Introduction

Recently, an interest on hydrogen as a

clean energy source and a fossil fuel

substitute has been increasing. From

the viewpoint that hydrogen utilizes

the energy system which uses the

existing fossil fuel without the

emission of environmental pollution

material, contrary to fossil fuels,

hydrogen is emerging as a promising

future clean energy. Among hydrogen

production methods, high-temperature

pyrolysis hydrogen production

method using heat chemical process is

considered as a proper method for

mass hydrogen production. Heat is

required much for high-temperature

heat chemical process, and lightwater

reactor that uses water as coolant

does not produce heat required

for high-temperature heat chemical

process. VHTR (Very High Temperature

gas Reactor) which uses helium

as coolant can easily produce heat

required for high-temperature thermochemical

process, so recently the

study of the use of high temperature

gas for hydrogen production has been

the research trend [1, 2].

VHTR has no possibility of core

melting due to low heat output

density, and it does not use water, so

there is no risk of explosion danger

due to hydrogen generation in the

case of coolant loss accident. Besides,

it has the advantage that high-temperature

coolant can be made compared

to water-cooled reactor, so it has the

advantage of power production and

process heat supply [3]. Nuclear

reactor is in charge of heat supply, and

this can be converted variously to be

used as the production of hydrogen or

power. In this study, by borrowing

general name in the atomic power

field, VHTR is called as a primary

system, the part which hydrogen production

and power supply are possible

through heat conversion, is defined as

the secondary system. Helium flowing

in nuclear reactor delivers the heat of

the primary system to the secondary

system through HX (Heat Exchanger).

Helium flowing through the secondary

system passes first through heat

exchanger where hydrogen production

occurs, and secondly and thirdly

passes through steam generator and

heat exchanger composed of ultrasuper

critical cycle (Ultra- supercritical

steam cycle: USC) and super critical

carbon dioxide (Supercritical CO 2 :

SCO 2 ) cycle, respectively, producing

process heat and power. In this study,

the authors proposed the overall heat

conversion system model, and performed

the thermodynamic feasibility

calculation in accordance with major

design variable at each point and

sensitivity analysis for efficiency

optimization.

2 Research methodology

2.1 Concept and methodology

of hydrogen production

equipment

As a method of hydrogen production

which uses water as a raw material by

using 900 °C heat, high temperature

electrolysis using heat energy simultaneously

and the mixed method of

using thermochemistry process method

and electrolytic method. Recently,

research has been focused on Sulfur-

Iodine thermochemical cycle where

iodide and sulfuric acid were used to

break down water. This is because the

required equipment can be scaled up

and process handling material is only

composed of gas and liquid so that

continuous operation is possible.

Besides, it is advantageous to use

nuclear reactor where the safety of

load change is demanded as heat

source [2].

In the hydrogen production equipment

where high temperature heat is

used, according to the Reaction 1

below, sulfuric acid (H 2 SO 4 ) can be

broken down into water vapor

(H 2 O(g)), oxygen (O 2 (g)), and sulfur

dioxide (SO 2 (g)).

Reaction 1:

2H 2 SO 4 + Heat 2H 2 O + 2SO 2 + O 2

After decomposition, oxygen(O 2 (g))

is removed, and water vapor(H 2 O(g))

and sulfur dioxide (SO 2 (g)) are cooled

down, reacting with iodide (I).

According to Reaction 2 below,

sulfuric acid (H 2 SO 4 ) and hydrogen

iodide (HI) are formed.

Reaction 2:

4H 2 O + 2SO 2 + 2I 2 2H 2 SO 4 + 4HI

+ Heat

Finally, by using high temperature

heat, hydrogen Iodide (HI) can be

separated into hydrogen (H 2 ) and

iodide (I) according to the reaction 3

below.

Reaction 3: 4HI + heat 2I 2 + 2H 2

2.2 The concept and status

of USC and S-CO 2 cycle

USC power plant means the power

plant where vapor pressure is 254 kg/

cm 2 or higher, and main vapor’s

or reheated vapor’s temperature is

593 °C or higher. The reasons why

pressure and temperature of the

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evaporator are elevated are that the

efficiency of power plant is increased.

When the temperature of main evaporator

and reheating steam increases

by 10 °C, the efficiency increases

by 0.5 %; and pressure increases by

10 kg/cm 2 , the efficiency increases by

about 0.2 %. Domestically, in 1990’s,

500 MW-grade standard coal thermal

power plant was designed and built,

and its operation condition was pressure

246 kg/cm 2 and temperature

538 °C.

In the case of Dangjin Thermal

Power No. 9, No. 10 and Samcheok

Thermal Power No. 1, No .2 that have

been being built, the pressure of

250 kg/cm 2 , temperature of 600 °C

were accomplished [4].

SCO 2 cycle is the power generation

technology of the Gas Brayton Cycle

method where pressurized carbon

dioxide is heated by the pressure

greater than critical condition to high

temperature and turbine is driven.

Presently, CO 2 power generation cycle

can be applied to most heat sources

used, and also it can be used for large

power plant, small scale distribution

power supply, or power supply for

marine plant.

Super critical condition means the

conditions for temperature and pressure

greater than critical point in the

general material state where liquid-gas

phase change occurs, and the

temperature and pressure at the lower

pressure part is greater than 32 °C, 74

atm, and all parts of cycle are maintained

over critical condition. While

operation is carried out at high

pressure, volumetric flow decreases,

so the size of overall heat conversion

cycle can be decreased; accordingly,

construction period and production

unit price can be lowered to secure

high economic feasibility.

Besides, compared to water vapor,

the compatibility with existing material

is excellent, so it can be supplied

to turbine at the temperature higher

than evaporator cycle. From this, the

increase of additional power generation

efficiency can be possible [5].

2.3 Heat Conversion Model

Design

IHX loop of VHTR that is studied in

the present study is the system where

the high temperature heat generated

in the reactor by connecting hydrogen

generation equipment and power

generation equipment in series can be

supplied in the same manner.

IHX loop nuclear reactor shown in

Figure 1 provides 350 MWt heat output,

and the heat generated from

| | Fig. 1.

IHX Loop Modelling.

nuclear fission is supplied to helium

fluid. For heat transfer to produce

hydrogen, heat exchanger, steam generator

for the power generation via

USC cycle, and in the power generation

via SCO 2 , one heat exchanger is

provided. In order to utilize the result

of the study regarding the existing

VHTR, the major principle and

variable if heat conversion model

were set as follows. Temperature and

pressure at No. 1, 2, 3, 4, 10 were

presumed by reference literature [8].

Temperature and pressure of ultrasuper

critical cycle No. 5, 6 and SCO 2

cycle, No. 8 were assumed by using

reference literature [9]. The model to

be explained below was defined as

reference model, and then the present

authors will plan to develop a model

that considers a variety of heat

efficiency improvement method. In

the present study, in the concept

similar to general Rankine cycle’s

reheating cycle, bypass mode was

proposed.

To begin with, the reference model

is as follows. After 910 °C helium fluid

discharging from VHTR carries out

heat exchange with heat exchanger 1,

hydrogen is produced by receiving

heat from high temperature helium

fluid in the heat exchanger 1. 846 °C

helium fluid passing heat exchanger 1

enters into steam generator 2 and go

through heat exchange. The fluid of

this steam generator is ultra-super

critical state water, and produces

power. The temperature of helium

fluid that passes through steam

generator 2 is 614.8 °C, this helium

fluid enters into heat exchanger 3

where heat exchange is carried out.

The fluid of this heat exchanger is

super critical-state carbon dioxide,

and it produces power by the heat

supplied. The temperature of helium

fluid coming out of heat exchanger 3

is 450 °C. The heat output that is produced

in heat exchanger 1 producing

hydrogen is 37.37 MWt. The mass flow

of helium from IHX is m 1 , and the

mass flow of water flowing in heat

exchanger 1 is m 2 , the mass flow of

water flowing in steam generator 2 is

m 3 , and the mass flow of CO 2 flowing

in heat exchanger 3 is m 4 . In this

study, the temperatures and pressures

from No.1 to No.10 in Figure 1 were

assumed, and m 1 and m 2 were calculated

by using the Equation (1), and

m 3 and m 4 were calculated by using

the Equation (2). Besides, considering

the characteristics of general longitudinal

temperature difference of heat

exchanger, the temperature at No. 6

and No. 9 was assumed to decrease by

10 °C compared to the temperature at

No. 4 and No. 7 of the steam generator

inlet.

Major equation or relationship for

heat equilibrium analysis is as follows:

• Equation used for calculating m 1

and m 2

: W = m∆h = m(h out – h in )... (1)

Here,W : Thermal power (MWt)

m : Mass flow (kg/hr)

h : Enthalpy (kJ/kg)

in : Entrance of the equipment

out : Outlet of equipment

• Equation used for calculating m 5

and m 8

∑m in h in = ∑m out h out ... (2)

In the case of hydrogen production, it

was assumed that all heat was

converted to work required, and in

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OPERATION AND NEW BUILD 232

the case of power production, it was

assumed that only a part of the heat

delivered was converted to electricity.

Besides, it was considered that the

pumping power was consumed due to

the flow in the power generation,

therefore it was considered in the

calculation of efficiency.

The general efficiency of USC cycle

and SCO 2 cycle was 43 % and 45 %,

respectively. Using Equation (3),

efficiency was corrected, and more

realistic calculation was carried out

[6].

• efficiency correction equation

η Oper = [1.0+{(T h,oper – T h,des ) × C}]

× η Des … (3)

Here, η Des : standard efficiency

according to reference

literature

η Oper : Standard efficiency’s

correction efficiency according

to high temperature

T h,des : Standard exit temperature in

Steam Generator tube according

to reference literature

T h,oper : Exit temperature within

specified range at Steam

Generator or Heat exchanger

tube

C : Efficiency correction factor;

USC : 0.3 % / 5 °C [6],

SCO 2 : 1.0 % / 5 °C applied

(assumption)

The output of steam generator 2 and

heat exchanger 3 is as follows, and

total output W gross is the sum of all the

values.

W 1 = m 2 × (h 3 – h 2 ) kJ/hr

W 2 = η 2,Operator × m 3 ×(h 4 – h 7 ) kJ/hr

W 3 = η 3,Operator × m 4 ×(h 7 – h 10 ) kJ/hr

Here, w pump : Work used

in the pump (MWt)

η pump : Pump efficiency

v : Specific volume (m 3 /kg)

P : pressure (kPa)

In order to simulate the above model,

the flow of helium gas, water, and

carbon dioxide was calculated by

using thermodynamic system analysis

software, EES (Engineering Equation

Solver).

The following is regarding IHX

loop model to which Bypass mode was

added. Bypass mode was added to the

existing IHX loop, and the efficiency

improvement of overall heat conversion

cycle was studied. The temperature

of the entrance of evaporator 2

and heat exchanger 3 was reheated

by using high temperature helium

coming from IHX, and the output

change was studied.

In the same way as the existing IHX

loop, VHTR supplies heat generated

by nuclear fission in 350 MWt nuclear

reactor. The fluid coming from IHX is

helium, and the fluid flowing in heat

exchanger 1 is water, the fluid flowing

in steam generator 2 is ultra-supercritical-state

water, and the fluid

flowing in heat exchanger 3 is super

critical-state carbon dioxide.

The mass flow of helium from

IHX is m 1 , and m 1 is divided into m 2

and m 3 , and m 3 enters into heat

exchanger 1, and do heat exchange

with the water flowing in heat exchanger

1. At this time, the mass flow

of water flowing in heat exchanger 1

is m 5 . and m 2 is divided into m 4 and

m 9 , and m 4 enters into No. 7 in order

to reheat helium that went through

heat exchange in heat exchanger 1,

and the reheated temperature is that

of No. 8. m 9 enters No. 12 in order to

reheat helium that went through heat

exchange in the steam generator 2,

and the reheated temperature is the

temperature of No. 13. The mass flow

in steam generator 2 is m 10 , and mass

flow of CO 2 in heat exchanger 3 flowing

through heat exchanger 3 is m 14 .

The temperature and pressure at

every point except No. 12 were assumed,

and m 1 value was obtained by

using Equation (1) in the same as m 1

of the existing IHX Loop.

At this time, the temperature of

No.7 and No.12 were assumed to be

that of No. 4 and No. 7 of the existing

IHX loop. Besides, it was assumed that

the temperatures of No. 8 and No. 13

increased to 860 °C and 620 °C,

respectively due to m 4 and m 9 . By

this, the change of the output and

efficiency on the cycle of steam generator

2 and heat exchanger 3. Heat

exchanger 1 in accordance with the

addition of Bypass mode was assumed

to produce the same output, 37.37

MWt, as the existing IHX loop, and

fixed m 5 value.

Considering the characteristics of

the general longitudinal temperature

difference of the heat exchanger as

the existing IHX loop, it was assumed

that the temperatures of No. 11 and

No. 15 decrease by 10 °C compared to

that of No. 8 (at Steam Generator

entrance) and No. 13 (at Heat

Exchanger). With the obtained m 1

value, m 14 value was calculated by

using Equation (2). After that, m 5

value was fixed to the value which can

make the output as the existing IHX

loop, then m 3 value was obtained

by Equation (2). m 2 was calculated by

m 2 = m 1 – m 3 , and m 4 was obtained by

using Equation (2). m 9 was obtained

Here, W 1 : heat exchanger 1

heat output (MWt)

W 2 : Steam generator 2

heat output (MWt)

W 3 : Heat exchanger 3

heat output (MWt)

In steam generator 2 and heat

exchanger 3 in order to consider

pumping power in accordance with

mass flow, pump’s efficiency (η pump )

was assumed to be 0.9, and Equation

(4) was used.

w pump = η pump × m × v out

× (P out – P in ) kJ/hr … (4)

W net = W gross – w pump

| | Fig. 2.

Bypass mode-added IHX loop Modelling.

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by m 9 = m 2 – m 4 , and the temperature

of No. 12 can be calculated by using

Equation (2). Finally, m 10 was also

obtained by using Equation (2). Like

the existing IHX loop, in the bypass

mode-added IHX loop, correction

efficiency and pumping power in

accordance with mass flow were

considered, and pumping power used

the above Equation (4). For the

simulation for this, thermodynamic

system analysis software, EES, was

used, in the same way with the

existing IHX loop model obtained

before, and the flow of helium gas

and fluid was analyzed.

3 Result

Table 1 shows the result of physical

value at each point by simulating the

existing IHX loop EES [5]. Physical

value of each point was assumed in

accordance with reference [8], [9]

literature, and the assumed values

were colored.

The temperature of helium fluid

that leaves from the first heat exchanger

after producing the hydrogen

decreases to 846 °C from 910 °C, and

the temperature of helium fluid that

leaves from the second steam generator

is 614.8 °C, and the temperature

of helium fluid that leaves from the

last heat exchanger is 450 °C. The

temperature of helium fluid decreases

steadily, but because the fluid flowing

each steam generator and heat

exchanger is different, efficient electricity

can be produced by using each

characteristics. The existing IHX

Loop’s m 5 and m 8 are in inverse

proportion, as more mass flow moves

toward high efficiency, the amount of

overall electricity output increases.

Although the efficiency of heat

conversion cycle connected to each

steam generator may be influenced by

various causes, but in the present

study, correction factor presumed

about high temperature was used, so

the detailed design for this part would

be needed.

If heat conversion cycle connected

to each steam generator should be

operated simultaneously by a specific

objective, considering the inverse proportion

relationship between m 5 and

m 8 , the output must be distributed.

Besides, the exit temperature at the

tube part of steam generator 2 is in

inverse proportion with the exit

temperature at the shell part. This will

eventually influence on the exit

temperature of the tube part of the

heat exchanger 3. When operating

heat conversion cycle connected to

each steam generator, it is necessary

to find balanced point on the temperature

between steam generators.

In the steam generator 2 and heat

exchanger 3, exit temperature and

mass flow are in inverse proportion.

This is because if high exit enthalpy is

maintained in order to deliver the

same heat energy, less mass flow is

needed, and if a large amount of mass

flow is needed, exit enthalpy should

be maintained low. Maximum output

would be in the parabolic form as exit

enthalpy and temperature change, so

if maximum output is needed, proper

exit temperature must be selected. Or

in case there is a requirement for exit

temperature, it is possible that output

would be determined according to

that.

Table 2 shows the physical value at

each point where IHX loop added by

bypass mode is simulated with EES.

No. Fluid Temperature

(°C)

| | Tab. 1.

IHX loop Simulation Result.

In the case of IHX loop to which

bypass mode was added, the helium

fluid that passed through the first

hydrogen-producing heat exchanger

is 910 °C~ 846 °C, which is the same

as the existing IHX loop, but here by

reheating high-temperature helium

fluid, the temperature increases to

860 °C. The temperature of the helium

fluid that passed through the second

steam generator is 614.8 °C, which

is the same as that of helium fluid

that passed through the second evaporator.

However, since the temperature at

the entrance reheated, and returned,

the amount of electricity output produced

increases. When helium fluid

enters the third heat exchanger, it is

reheated from 614.8 °C to 620 °C, the

temperature of helium fluid is 450 °C,

and the amount of electricity output

Pressure

(kPa)

| | Tab. 2.

Result of IHX Loop to which Bypass Mode was added.

Enthalpy

(kJ/kg)

Mass flow

(kg/hr)

1 Helium 910.0 4000 6,161.00 527,662

2 Water 193.0 18,000 828.70 49,839

3 Water 585.0 16,500 3,528.00 49,839

4 Helium 846.0 4,000 5,829.00 527,662

5 Water 260.2 20,790 1,134.00 208,359

6 Water 836.0 16,475 4,174.00 208,359

7 Helium 614.8 4,000 4,628.00 527,662

8 CO 2 203.5 19,760 96.59 902,043

9 CO 2 604.8 19,290 597.00 902,043

10 Helium 450.0 4,000 3,773.00 527,662

No. Fluid Temperature

(°C)

Pressure

(kPa)

Enthalpy

(kJ/kg)

Mass flow

(kg/hr)

1 Helium 910.0 4,000 6,161.00 527,662

2 Helium 910.0 4,000 6,161.00 122,749

3 Helium 910.0 4,000 6,161.00 404,913

4 Helium 910.0 4,000 6,161.00 113,375

5 Water 195.0 18,000 837.50 50,000

6 Water 585.0 16,500 3,528.00 50,000

7 Helium 846.0 4,000 5,829.00 404,913

8 Helium 860.0 4,000 5,901.00 518,288

9 Helium 910.0 4,000 6,161.00 9,374

10 Water 260.2 20,790 1,134.00 214,580

11 Water 850.0 16,475 4,209.00 214,580

12 Helium 614.8 4,000 4,628.00 518,288

13 Helium 620.0 4,000 4,655.00 527,662

14 SCO 2 203.5 19,760 96.59 918,581

15 SCO 2 610.0 19,290 603.50 918,581

16 Helium 450.0 4,000 3,773.00 527,662

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OPERATION AND NEW BUILD 234

increases compared to the existing

IHX loop.

The following is the major comparison

of the result of reference

model and bypass mode model.

Mass flow IHX loop Bypass Mode Loop

Heat Exchanger 1 49839 50000

Steam Generator 2 208359 214580

Heat Exchanger 3 902043 918581

| | Tab. 3.

IHX Loop and Bypass Mode IHX Loop: Mass Flow Comparison.

Pumping power

When IHX loop and bypass modeadded

IHX loop were compared, mass

flow of m 10 and m 14 in the bypass

mode-added loop was greater compared

to the mass flow of m 5 and m 8 in

the existing loop, as shown in Table 3.

IHX loop

(MWt)

| | Tab. 4.

IHX Loop and Bypass Mode IHX loop: Pumping Power Comparison.

As shown in Table 4, depending on

the difference of mass flow value,

pumping power used in the pump also

can be high in steam generator 2 and

heat exchanger 3. However, although

pumping power is higher in the bypass

IHX loop, by reheating, efficiency of

steam generator 2 increased from

53.79 % to 54.4 %, and that of heat

exchanger 3 increased from 45.83 %

to 45.87 %; accordingly, it is seen that

the value of Power increased. As a

result, Net Power that considered

pumping power in Total Power was

178.6 MWt in the present IHX

loop, but the IHX loop to which

bypass mode was added increased

to 185.3 MWt, as shown in Table 5.

| | Tab. 5.

IHX loop and Bypass mode IHX loop: Power Comparison.

Bypass mode loop

(MWt)

Steam Generator 2 1.091 1.123

Heat Exchanger 3 9.82 10

Power

IHX loop

(MWt)

Bypass mode loop

(MWt)

Heat Exchanger 1 37.37 37.37

Steam Generator 2 94.63 99.7

Heat Exchanger 3 57.46 59.34

Net Power 178.6 185.3

Since the assumption was that

constant heat was supplied from the

primary system, it is seen that the

efficiency of the bypass model where

net power is high, and it is judged that

efficiency optimization model can be

formulated by detailed design.

4 Conclusion

In this study, VHTR system was

modelled for supplying high temperature

heat, by distribution, produced in

the high temperature gas furnace to

hydrogen producing equipment and

power generation equipment.

Provided that high temperature

gas- cooled reactor is located in

primary system, the secondary system

where hydrogen production and

power supply are possible were

explained. The helium that flows in

the nuclear reactor first passes

through the HX (heat exchanger)

whose purpose is the production of

hydrogen, and secondly and thirdly

pass through the steam generator

composed of super critical carbon

dioxide cycle, and heat exchanger,

respectively, producing the process

heat and power. In order to analyze

existing IHX loop model and bypass

mode-added IHX loop model, the

present authors studied the input &

output conditions and output change

of each steam generator and heat

exchanger, and based on this result,

by designing IHX loop in the power

production part in detail, the authors

performed the calculation of thermodynamic

physical value and efficiency

at each point. Additionally, the

authors studied the change regarding

electricity output and efficiency

according to bypass mode, when

reheating cycle is added, the possibility

on the efficiency optimization

was proposed.

References

[1] Kim. Y. W., 2015, Nuclear Hydrogen

Production Technology development

Using Very High Temperature Reactor,

Trans. Korean Soc. Mech. Eng. C, Vol. 3,

No. 4, pp. 299~305.

[2] Chang. J. H., 2006, Current Status of

Nuclear Hydrogen Development,

Journal of Energy Engineering, Vol.15,

No.2, pp. 127~137.

[3] Lee. S. I., 2015, Heat Balance Study

on Integrated Cycles for Hydrogen

and Electricity Generation in VHTR,

Transaction of the KNS Spring Meeting.

[4] Sung. H. C., 2012, Development of

Ultra-Supercritical (USC) Power Plant,

Trans. Korean Soc. Mech. Eng. B,

Vol. 36, No.2, pp.205~210.

[5] Yeom Chung-seop, Im Dong-ryeol,

Lee Jung-ik, 2014, Trend of Electricity

Generation Technology using supercritical

CO 2 , Institute for Advanced

Engineering, KIC News, Volume 17,

No.1.

[6] K.C.Cotton,1998, Evaluating and

Improving Steam Turbine Performance,

2 nd edition, Cotton Fact Inc.

[7] F-Chart Software, 2016,

Engineering Equation Solver,

http://www.fchart.com/ees/

[8] NGNP Conceptual Design Report/Steam

Cycle Modular Helium Reactor

(SC-MHR) Demonstration Plant,

Table 3-6 SC-MHR Conceptual Design

Point Design Parameter.

[9] SangIL Lee, Yeon Jae Yoo, Gyunyoung

Heo, Soyoung Park, Yeon Kwan Kang,

Heat Balance Study on Integrated

Cycles for Hydrogen and Electricity

Generation in VHTR-Part 2, Korean

Nuclear Society Autumn Meeting,

Oct 28-30, 2015.

Authors

SangIL Lee

YeonJae Yoo

Deok Hoon Kye

Department of Nuclear Team

Power & Energy Plant Division

Hyundai Engineering Company

Seoul, Korea

Gyunyoung Heo

Eojin Jeon

Soyoung Park

Department of Nuclear

Engineering

Kyung Hee University

Yongin Korea

Operation and New Build

Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park

---

atw Vol. 63 (2018) | Issue 4 ı April

Informationsbedarf versus Geheimhaltungspflichten –

Erweiterung des In-camera-Verfahrens geplant

235

Tobias Leidinger

Dem berechtigten Anspruch der Öffentlichkeit auf detaillierte Informationen über ein atomrechtlich genehmigungsbedürftiges

Vorhaben steht das staatliche Interesse an einem effektiven Geheimnisschutz sensibler Daten

gegenüber. Dieser Konflikt tritt regelmäßig im Genehmigungsverfahren aber auch vor Gericht zu Tage. Die differenzierten

Gesetzesbestimmungen, die den Ausgleich dieser widerstreitenden Interessen regeln, sollen nun durch eine

weitere Facette ergänzt werden: Ein erweitertes In-camera-Verfahren bei Gericht. Nach dem Koalitionsvertrag vom

12. März 2018 soll die Regelung in der schon laufenden 18. Legislaturperiode erfolgen.

I Grundkonflikt Informationsbedarf vs.

Geheimhaltungspflicht

In atomrechtlichen Genehmigungsverfahren zeigt sich

regelmäßig ein Grundkonflikt: Dem Interesse der Öffentlichkeit

an möglichst vertieften Informationen über

alle sicherheits- und sicherungsrelevanten Aspekte des

Vorhabens steht das Erfordernis eines effektiven Geheimnisschutzes

in Bezug auf sensible Daten gegenüber.

Genauer betrachtet lassen sich für beide Pole Grundrechtspositionen

anführen: Einerseits ist Information Voraussetzung

für Transparenz und Teilhabe der Öffentlichkeit

am Genehmigungsverfahren. Das Verfahren dient der

Gewährleistung materieller Schutzansprüche Dritter.

Ohne Information ist Kontrolle gegenüber der Verwaltung

kaum realisierbar. Information ist die Grundlage für

Partizipation und Teilhabe der Öffentlichkeit an einem

Verfahren. Das BVerfG bringt dies mit der Formel „Grundrechtsschutz

durch Verfahren und Teilhabe an Information“

auf den Punkt.

Für die andere Seite, dem Interesse an Geheimhaltung

sensibler Daten, lassen sich aber nicht minder gewichtige

Grundrechtsinteressen anführen: Die Geheimhaltung

dient ebenfalls zum Schutz der Grundrechtsträger: Ist der

Staat zum Schutz der Grundrechte („Leben, Gesundheit“)

seiner Bürger verpflichtet, bedarf es des Geheimnisschutzes

in Bezug auf sensible Daten, damit eine effektive

Terrorabwehr – gerade zum Schutz der Bürger – gewährleistet

bleibt. Die Nicht-Preisgabe sicherheits- und

sicherungsrelevanter Informationen ist mithin nicht

minder essentielle Voraussetzung für einen effektiven

Grundrechtsschutz der Bürger.

II Interessenausgleich durch differenzierte

Gesetzesregelungen

Der Gesetzgeber trägt zur Lösung dieser widerstreitenden

Interessen im atomrechtlichen Genehmigungsverfahren

bereits heute durch eine ganze Reihe differenzierter

Regelungen bei. Nach § 6 der Atomrechtlichen Verfahrensordnung

(AtVfV) sind nicht nur der Antrag, der Sicherheitsbericht

und eine Kurzbeschreibung des jeweils zu

genehmigenden Vorhabens für die Öffentlichkeit auszulegen,

sondern es besteht nach § 6a Abs. 2 Satz 1 und

Abs. 3 AtVfV die Möglichkeit, in Bezug auf das Vorhaben –

im Interesse der Sicherheit und Sicherung – geheimhaltungsbedürftige

Informationen durch eine Beschreibung

oder Inhaltsdarstellung zu ersetzen. Anstelle einer

„Schwärzung“ von Unterlagen – die letztlich eine „Verweigerung“

von Information bedeutete –, tritt so die

Möglichkeit, geheimhaltungsbedürftige Informationen zu

umschreiben, so dass der Dritte in der Lage bleibt, seine

Betroffenheit durch das Vorhaben gleichwohl erkennen

und beurteilen zu können.

Eine Einschränkung von Informationsansprüchen ist

auch jenseits dieser Regelung möglich: Während eines

atomrechtlichen Verfahrens besteht der Anspruch auf

Akteneinsicht gemäß § 6 Abs. 4 AtVfV i.V.m. § 29 Abs. 1

S. 3, Abs. 2 und 3 des Verwaltungsverfahrensgesetzes

(VwVfG) nur nach Ermessen der Behörde (also nicht

„ unbedingt“). Informationen, die sicherheits- oder

sicherungsrelevant sind, weil sie den Ansatz für die Ausschaltung

von Sicherheits- und Sicherungsmaßnahmen

oder für die Identifizierung/Lokalisierung von Schwachstellen

eröffnen könnten, können – soweit durch ihre

Preisgabe ein „Nachteil zum Wohl des Bundes oder

Landes“ zu befürchten ist – von der Offenlegung ausgeschlossen

werden. Spezialgesetzlich ist die Geheimhaltung

von sensiblen Informationen im Sicherheitsüberprüfungsgesetz

(SÜG) geregelt. Besteht danach die Gefahr

eines „Nachteils“ oder wäre die Preisgabe der Information

sogar „schädlich“ für Bund oder Land, kann sie

nach Maßgabe der Verschlusssachen-Anweisung (VS-

Anweisung) durch den Geheimschutzbeauftragten der

Behörde als „Verschlusssache – Nur für den Dienstgebrauch“

oder sogar als „Verschlusssache – Vertraulich“

eingestuft und ihre Offenlegung verweigert werden. Was

nach Maßgabe des SÜG i.V.m. VS-Anweisung geheim zu

halten ist, darf auch nicht in anderem Zusammenhang

preisgegeben werden: So bestehen – auch außerhalb eines

atomrechtlichen Verfahrens – Informationsansprüche

Dritter, z.B. auf Herausgabe von umweltrelevanten

Informationen gegen die Genehmigungsbehörde nach

Umweltinformationsgesetz (UIG) oder – soweit die

Informationen nicht umweltrelevant sind – nach Maßgabe

des Informationsfreiheitsgesetzes (IFG). Pressevertreter

können sich darüber hinaus auch auf das jeweilige

Landes-Pressegesetz stützen. In all diesen Fällen besteht

indes die Möglichkeit – mit oder ohne ausdrücklichen

Bezug auf das SÜG –, dass sicherheits- und sicherungsrelevante

Informationen im Ergebnis nicht offenbart

werden müssen, wenn die materiellen Schutzvoraussetzungen

nach SÜG i.V.m. der VS-Anweisung vorliegen.

III In-camera-Verfahren de lege lata und

de lege ferenda

Verweigert die atomrechtliche Genehmigungsbehörde die

Herausgabe sensibler Informationen unter Verweis auf

den Geheimschutz auch im Gerichtsverfahren, – in dem

z.B. über die Rechtmäßigkeit einer atomrechtlichen

Genehmigung gestritten wird – so sieht die bislang

existierende Gesetzesregelung zum sog. In-camera-

Verfahren in § 99 Verwaltungsgerichtsordnung (VwGO)

vor, dass über die Frage der Geheimhaltungsbedürftigkeit

ein speziell besetzter Fachsenat vorab entscheidet. Ihm

sind ausschließlich die geheimhaltungsbedürftigen Akten

vorzulegen („in camera“), um zu prüfen, ob die Einstufung

als „geheim“ zurecht erfolgt ist und daher die Verweigerung

der Aktenvorlage durch die Behörde Bestand hat

oder nicht. Nur wenn die Geheimhaltungsbedürftigkeit

verneint wird, ist die vorenthaltene Information dem

Verwaltungsgericht zugänglich zu machen. Nur dann

kann es darauf zugreifen und seine Entscheidung darauf

stützen.

SPOTLIGHT ON NUCLEAR LAW

Spotlight on Nuclear Law

Information Requirements Versus Confidentiality Obligations – Extension of the In-Camera Procedure Planned ı Tobias Leidinger

---

atw Vol. 63 (2018) | Issue 4 ı April

SPOTLIGHT ON NUCLEAR LAW 236

Der Koalitionsvertrag vom 12. März 2018 (vgl. Seite

141) sieht nun vor, dass die Regelungen für das In-camera-

Verfahren für das Atomrecht dahingehend erweitert

werden sollen, dass geheimhaltungsbedürftige Unter lagen

auch zum Zwecke des Nachweises der Genehmigungsvoraussetzungen

in ein verwaltungsgerichtliches Hauptsacheverfahren

– bei gleichzeitiger Wahrung des Geheimschutzes

– eingeführt werden können. Das In-camera-

Verfahren dient dann nicht (mehr allein) zur Klärung der

Frage der Geheimhaltungsbedürftigkeit einer Unterlage

(wie bisher), sondern ermöglicht darüber hinaus eine

weitergehende Prüfung in der Sache durch das Gericht.

Das Gericht prüft dann auch, ob der erforderliche Schutz

gegen Störmaßnahmen Dritter (SEWD) als gegeben unterstellt

werden darf oder nicht. Die Gewährleistung des

SEWD-Schutzes ist eine wesentliche Tatbestandsvoraussetzung,

die erfüllt sein muss, damit eine atomrechtliche

Genehmigung erteilt werden kann. Dabei ist aber auch in

einem erweiterten In-camera- Verfahren sicherzustellen,

dass die behördliche Ein schätzungsprärogative in Bezug

auf genehmigungs relevante Wertungen bei Sicherheit und

Sicherung beachtet werden. Das bedeutet, dass das Gericht

sich nicht an die Stelle der Behörde setzen darf, also eine

eigene Entscheidung anstelle der Behörde trifft, sondern

bei seiner Nachprüfung auf eine Vertretbarkeitskontrolle

beschränkt bleibt.

die Frage, ob im Ergebnis davon ausgegangen werden darf,

dass die erforderliche Schadensvorsorge und der gebotene

Schutz gegen SEWD-Ereignisse gewährleistet ist oder

nicht, könnte auf diese Weise weitergehend als bisher entschärft

werden. Idealerweise bliebe der Geheimnisschutz

auch so gewahrt, zugleich aber wäre dem Interesse

der Drittbetroffenen an einer Überprüfung essentieller

Genehmigungsvoraussetzungen unter Berücksichtigung

geheimhaltungsbedürftiger Informationen weitergehend

als bisher entsprochen. Das wäre als konstruktiver Beitrag

zur Stärkung eines effektiven Grundrechtsschutzes zu

bewerten: Ein erweitertes In-camera-Verfahren diente so

in besonderer Weise zur Gewährleistung der dem Dritten

zustehenden Schutzansprüche und wahrte dabei

gleichwohl den erforderlichen Geheimschutz, der nicht

minder einem effektiven Grundrechtsschutz der Bürger

geschuldet ist.

Allerdings bleiben die konkrete Ausgestaltung und der

Vollzug dieser Regelung in der Praxis abzuwarten: Folgt

einer guten Absicht des Gesetzgebers eine in der Praxis

tatsächlich und rechtlich brauchbare Lösung? Ziel müsste

sein, dadurch nicht neue Fragen zur Anwendung und

Reichweite eines erweiterten In-camera-Verfahrens

aufzuwerfen, sondern eine inhaltlich klare und hinreichend

bestimmte Norm zu schaffen, die das Versprechen

des Koalitionsvertrages vollzugsfähig einlöst.

IV Erweiterung des In-camera-Verfahrens:

Bedenkenswerter Schritt

Die Absicht, das In-camera-Verfahren in Bezug auf die

Prüfung materieller Genehmigungsvoraussetzungen zu

erweitern, ist ein bedenkenswerter Ansatz. Der Streit über

Autor

Prof. Dr. Tobias Leidinger

Rechtsanwalt und Fachanwalt für Verwaltungsrecht

Luther Rechtsanwaltsgesellschaft

Graf-Adolf-Platz 15

40213 Düsseldorf

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Ernst Michael Züfle

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atw Vol. 63 (2018) | Issue 4 ı April

238

ENVIRONMENT AND SAFETY

CFD Modeling and Simulation

of Heat and Mass Transfer in

Passive Heat Removal Systems

Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas

This paper is presenting the CFD-modelling and simulation of condensation inside passive heat removal systems.

Designs of future nuclear boiling water reactor concepts are equipped with emergency cooling systems which are

passive systems for heat removal. The emergency cooling system consists of slightly inclined horizontal pipes which are

immersed in a tank of subcooled water. At normal operation conditions, the pipes are filled with water and no heat

transfer to the secondary side of the condenser occurs. In the case of some accident scenarios the water level may

decrease in the core, steam enters the emergency pipes and due to the subcooled water around the pipe, this steam

condenses. The emergency condenser acts as a strong heat sink which is responsible for a quick depressurization of the

reactor core. This procedure acts passive i.e. without any additional external measures. The actual project is defined to

model the phenomena which are occurring inside the emergency condensers. The focus of the project is on detection of

different morphologies such as annular flow, stratified flow, slug flow and plug flow and also modeling of the laminar

film which is occurring during the condensation near the wall.

The condensation procedure inside the

pipe is determined by two important

phenomena. The first one is wall

condensation and the second one is the

direct contact condensation (DCC).

The Algebraic Interfacial Area Density

(AIAD) concept is used in order to

model the interface between liquid

and steam. In the next steps the Generalized

Two-Phase Flow ( GENTOP)

model will be used to model also the

dispersed phases which are occurring

inside the pipe. Finally, the results of

the simulations will be validated by

experimental data which will be available

in HZDR. In this paper the results

of the first part are presented.

1 Introduction

Condensation plays a crucial role in

the emergency condenser of passive

heat removal systems of nuclear power

plants. Passive safety systems do not

need any external power supplies and

they mostly depend on physical phenomena

such as natural circulation

and gravity driven flows. In order to

assess the performance of passive safety

systems and their efficiency mostly

one-dimensional codes are used such

as ATHLET, RELAP and TRACE. These

codes are able to calculate most of the

phe nomena in power plants; however,

they cannot reflect the 3D phenomena.

Therefore, Computational Fluid

Dynamics (CFD) methods should be

used to simulate and predict the

complex multiphase flow structure.

Despite the previous research being

done on the two-phase flow behavior,

this phenomenon needs much more

investigations. The two-phase flow

patterns and transition between vapor

and liquid are studied by Thome and

Hajal et al. [1, 2]. They introduced a

logarithmic mean void fraction (LMe)

method in order to calculate the vapor

void fractions which change from the

low pressure up to the critical pressure

point. Moreover, they proposed a new

heat transfer model based on the same

simplified flow structures that have

been used in the flow boiling model

of Kattan et al. [3]. The model can

predict the local condensation heat

transfer coefficient for different flow

regimes such as annular, intermittent,

stratified-wavy fully stratified and

wavy flow.

Many attempts have been done to

investigate the mass transfer between

liquid and gas phase in condensation.

Lee et al. [4] introduced a model for

prediction of the mass transfer. They

assumed that the interface between

liquid and steam is on saturation

temperature and introduced an

iterative technique in order to reach to

desired boundary condition inside

each cell. This model depends on a

relaxation factor which needs to be

tuned. The tuning needs many trial

and error simulations which is

time-consuming and doesn’t have any

predictive capabilities.

Moreover, there are empirical or

semi-empirical methods to calculate

the mass transfer in the interface.

Strubelj et al. [5] by using ANSYS CFX

and NEPTUNE_CFD [6] code tried to

simulate Direct Contact Condensation

(DCC) in stratified flows. In DCC the

phase change occurs due to the direct

contact interaction of subcooled water

and saturated steam. The defined

phase change mass flux depends on

thermal conductivity of the liquid and

Nusselt number of the liquid. The

Nusselt number was calculated

by Coste et al. [7] based on Surface

Renewal Theory (SRT) [8]. The SRT

theory calculates the mass transfer

according to the renewal period of

eddies and the liquid turbulent

properties. Hughes and Duffey [9]

used the surface renewal theory and

the Kolmogorov turbulent length

scale theory to define a correlation for

the heat transfer coefficient. They

considered that the heat removal from

interface occurs by smallest turbulent

scales. This model will be introduced

more detailed in the next sections.

This correlation is validated for

Pressurized Thermal Schock (PTS)

phenomenon by Egorov [10] and

Apanasevich [11]. Further to Hughes

correlation, Shen et al. [12] developed

another correlation for calculation of

heat transfer coefficient based on the

surface renewal theory. Ceuca et al.

[13] used both of these correlations

in order to simulate the direct contact

condensation for the LAOKOON

facility [14]. By comparison of Hughes

and Duffey correlation with Shen

correlation, Ceuca et al. [13] concluded

that both of the models provide

accurate results for the horizontal

stratified quasi-steady state.

Evidently, many attempts have been

done in the modeling of con densation

inside the pipes. The goal of the current

work is modeling of the transition

between different mor phologies which

are occurring during the condensation

inside the pipe ( Figure 1). In order to

do that, several CFD models such as

IMUSIG, AIAD and GENTOP which

have been developed in HZDR in cooperation

with ANSYS are available. The

Inhomo geneous MUSIG model considers

the bubble size distribution and

is used for modeling the small-scaled

dispersed gas phase [15]. The AIAD

Environment and Safety

CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems

ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas

---

atw Vol. 63 (2018) | Issue 4 ı April

| | Fig. 1.

Schematic representation of horizontal flow and different morphologies.

(Algebraic Interfacial Area Density

Model) is developed for detection of

the local morphology and corresponding

switch between them [16]. The

recently developed GENTOP- model

combines both concepts. GENTOP

(Generalized Two-Phase Flow) approach

is able to simulate co-existing

large-scaled (continuous) and smallscaled

(polydispersed) structures [17].

All these models are validated for adiabatic

cases without any phase change.

Therefore, the start point of the current

work project is using the available

models and integrating phase transition

and con densation models into

them. In the current work as initial

stages the AIAD model has been used

since in this model 2 continues phases

should be considered and it is less complicated

compare to GENTOP model

which also considers a poly- dispersed

phase. In the proceeding sections a

more detail explanation of AIAD model

will be given.

2 CFD model formulation

In the current work a multi-field twophase

CFD approach is used with

ANSYS CFX 17.2 in order to simulate

the condensation inside horizontal

pipe flows. The mass, momentum and

energy equations can be defined,

respectively, as follow:

• Mass conservation equation:

(1)

where S Mi describes user specified

mass source.

χ iβ the mass flow rate per unit volume

from phase β to phase i.

• Momentum conservation equation:

(2)

where S mi is the momentum source

caused by external body forces

and user defined momentum

sources.

M i is the interfacial forces acting

on phase i due to the presence

of other phases.

χ + iβ v β – χ + βi v i is the momentum

transfer induced

by mass transfer.

• The total energy equation:

(3)

where: h tot is the total enthalpy

related to static enthalpy by:

(4)

T i , λ i represents the temperature

and the thermal conductivity

of phase i.

S Ei describes external heat sources.

Q i is interphase heat transfer

to phase i across interfaces

with the other phase.

χ + iβ h βs – χ + βi h is denotes the interphase

mass transfer.

In ANSYS CFX in order to describe the

phase change which occurs due to the

interphase heat transfer, the Thermal

Phase Change Model has been introduced

[30]. This model is particularly

useful in simulation of the condensation

of saturated vapor. The heat

flux from the interface to phase i and

phase β is:

q i = h i (T sat – T i ) (5)

q β = h β (T sat – T β ) (6)

where h i , h β and T sat are heat transfer

coefficients of the phase i and phase

β and the saturation temperature,

respectively. ṁ iβ is the mas flux from

phase β to phase i. H is and H βs are the

interfacial enthalpy values which

come into and out of the phase due

to phase change which occurs. By

usage of the total heat balance

equation the interphase mas flux can

be determined as follow:

| | Fig. 2.

3D geometry of the pipe and mesh of the cross section.

(7)

ṁ iβ > 0 → H is = H i,sat , H βs = H β (8)

ṁ iβ < 0 → H is = H i , H βs = H β,sat (9)

In the current work, the steam

con sidered to be in saturation temperature.

Therefore, the heat flux

from the steam to the interface equals

zero since both are in saturation

temperature. As a result, the interphase

mass flux formula can be

written as:

(10)

In this work in order to model the heat

transfer coefficient the Hughes and

Duffy model has been used which is

based on the SRT model [9]. They

used the Surface Renewal Theory

(SRT) and the Kolmogorov turbulent

length scale theory to find a correlation

for heat transfer coefficient.

Therefore, the heat transfer coefficient

was derived as:

(11)

where ε is the turbulent dissipation, v l

is the kinematic viscosity and λ is the

thermal conductivity.

3 Computational grid and

boundary conditions

In Figure 2 the pipe and the boundary

conditions are shown. The pipe is

horizontal and has 1 m length

and 0.043 m diameter. In order to

define a mesh for the pipe ANSYS

ICEM software is used. Due to the

higher importance of the wall region

compare to the middle of the pipe,

the mesh near the wall needs to be

finer than the mesh in the pipe

center. The number of nodes is

1,250,000.

Mass flow rate

[Kg/s]

Temperature

(k)

Inlet 0.5 537.1

Wall - 312.18

outlet outflow -

ENVIRONMENT AND SAFETY 239

CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems

Environment and Safety

ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas

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ENVIRONMENT AND SAFETY 240

| | Fig. 3.

(a) Area averaged liquid volume fraction in different cross sections over the pipe length,

(b) Temperature distribution in the outlet of the pipe for 5 different radial lines.

4 Results

The results are obtained with the AIAD

approach for modeling the free surface

and morphologies. Moreover, the

Hughes correlation is used for the heat

transfer coefficient. Figure 3 represents

the qualitative profiles of liquid

volume fraction and tem perature. In

Figure 3 (a) the volume fraction profile

in the vertical cross section in the

middle of the pipe and in the streamwise

direction is represented. As it can

be seen at the inlet the pure steam

exists and by going further in the pipe,

due to the film condensation a liquid

film starts to generate near the wall.

The liquid film is growing and leads to

the thicker film. In a cross section

500 mm far from the inlet the liquid

film is falling down gradually and

gathering at lower part of the pipe.

The liquid film always exists near the

wall because as soon as the liquid is

falling down the steam becomes in the

direct contact with the wall and condenses

and again new film generates.

Moreover, in Figure 3 (d) the temperature

profile is shown for different

cross sections along the pipe. As

mentioned before, the steam is fixed

at the satu ration temperature, but

further along the pipe by generating

the liquid the temperature of the liquid

is decreasing because of the heat

flux to the wall. Moreover, the wall

heat flux is cooling the liquid which

causes the direct contact condensation

between liquid and steam interface. As

the steam is on the saturation temperature

there is no heat flux between

the interface which is also on the saturation

temperature and the steam.

Therefore, just the phase is changing

and the steam turns into the liquid.

| | Fig. 4.

(a) Liquid Volume fraction distribution on a cross section along the pipe, (b) temperature distribution on a cross section along the pipe,

(c) Volume fraction distribution on different cross sections, (d) temperature distribution on different cross sections.

Figure 4 (a) shows the change of

cross section averaged liquid volume

fraction along the pipe. According to

the figure the average liquid volume

fraction at the inlet is 0.0 and due to

the mass transfer it’s increasing along

the pipe and it reaches to around 0.1

at end of the pipe. Therefore, in a

horizontal pipe with one meter length

the total condensation rate is around

10 percent. In Figure 4 (b) the temperature

distribution for the five

radial lines on the outlet of the pipe

is presented. This plot shows the

temperature difference from the

center of the pipe towards the wall.

As the plot shows, in the center the

temperature is equal to the saturation

temperature. As far as getting closer to

the wall which is in subcooled

tem perature, the temperature gradient

is increasing. In other words, in

the region near the wall the temperature

difference from the saturation

temperature is higher. Moreover,

slope of the plot for L5 is higher than

L1. The reason is in lower part of the

pipe (which is showed by L5) the

amount of cooled liquid is higher

which causes higher temperature

gradient in the lower parts of the pipe.

As the pipe is symmetric and the

boundary conditions for both sides of

the pipe are same, there is no need to

plot the temperature distribution in

another half of the cross section.

5 Conclusion

The ANSYS CFX 17.2 has been used in

order to simulate the condensation

inside horizontal tubes. In order to

model the two phase flow, heat

transfer and phase change are included

in the available AIAD concept

which was developed for adiabatic

cases. Moreover, the Hughes heat

transfer coefficient correlation is

implemented for the modeling of the

direct contact condensation in the

interface. The changes of the flow

structure inside the pipe and the

volume fraction and the temperature

profiles have been studied in detail.

The liquid film which is generated

near the wall due to the wall condensation

is modeled and it can be seen in

the volume fraction profiles. By generating

the liquid film near the wall both

wall condensation and direct contact

condensation are occurring inside

the pipe at the same time. Whereas in

the actual paper only the test for

plausibility of the AIAD model was

done, in the near future the comparison

to the experiment is planned.

The next step which is an ongoing

part of the project is simulation of the

Environment and Safety

CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems

ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas

---

atw Vol. 63 (2018) | Issue 4 ı April

whole condensation phenomena and

flow morphologies by using GENTOP

concept. Further to the AIAD concept

which considers two continuous

fluids, the GENTOP approach is a

three field two fluid model and considers

also a poly dispersed phase.

Acknowledgments

This project is an ongoing project in

Helmholtz-Zentrum Dresden Rossendorf

(HZDR), which is funded by Bundesministerium

für Bildung und Forschung

(BMBF) under grant number

02NUK041B in Germany.

References

[1] Hajal, J.El.; Thome, J.; Cavallini, A.

Condensation inside horizontal tubes,

part 1: two phase flow pattern map.

International Journal of Heat and Mass

Transfer 46: 3349-3363(2003).

[2] Thome, J.; Hajal, J.El; Cavallini, A.

Condensation inside horizontal tubes,

part 2: New heat transfer model based

on flow regimes. International Journal

of Heat and Mass Transfer 46: 3365-

3387(2003).

[3] Kattan, N.; Thome, J.R.; Favrat, D. Flow

boiling in horizontal tubes:part2-New

heat transfer data for five refrigerants.

J. Heat Transfer 120: 148-155 (1998).

[4] Lee, W. H. A Pressure Iteration Scheme

for Two-Phase Flow Modeling. Multiphase

Transport Fundamentals,

Reactor Safety, Applications: 407–432,

(1980).

[5] Štrubelj, L.; Ézsöl, Gy. ; Tiselj, I. Direct

Contact Condensation Induced

Transition from Stratified to Slug Flow.

Nuclear Engineering and Design 240:

266–274 (2010).

[6] Lavieville, J.; Quemerais, E.; Boucker, M.;

Maas, L., NEPTUNE CFD V1.0 User Guide

(2005).

[7] Coste, P. ; Pouvreau, J. ; Lavieville, J.;

Boucker, M. A Two-phase CFD approach

to the PTS problem evaluated on COSI

experiment. Proceedings of the 16 th

International Conference on Nuclear

Engineering ICONE16, USA, (2008).

[8] Banerjee, S.; A surface renewal model

for interfacial heat and mass transfer in

transient two-phase flow. International

Journal of Multiphase Flow, Vol.4:

571-573 (1978).

[9] Hughes, E. D.; Duffey, R. B. Direct

Contact Condensation and Momentum

Transfer in Turbulent Separated Flows.

Internal Journal of Multiphase Flow 17:

599–619 (1991).

[10] Egorov, Y. Validation of CFD codes with

PTS relevant test cases. Technical Report

EVOL-ECORA-D07, ANSYS, Germany

(2004).

[11] Apanasevich, P. ; Lucas, D.; Beyer, M.;

Szalinski, L. CFD based approach for

modeling direct contact condensation

heat transfer in two-phase turbulent

stratified flows. International Journal of

Thermal Sciences 95: 123-135(2015).

[12] Shen, L.; Triantafyllou, G.S.; Yue. D.K.P.

Turbulent diffusion near a free surface

Journal of Fluid Mechanics 407:

145–166 (2000).

[13] Ceuca, S. C. ; Macián-Juan R. CFD

Simulation of Direct contact Condensation

with ANSYS CFX using Locally

defined Heat Transfer Coefficient.

In ICONE-20, Anaheim, California, USA,

No. 54347 (2012).

[14] Goldbrunner, M.; Karl, J. ; Hein, D.

Experimental Investigation of Heat

Transfer Phenomena During Direct

Contact Condensation in the Presence

of Noncondensable gas by means of

Linear Raman Spectroscopy. In 10 th Int.

Symp. on Laser Techniques Applied to

Fluid Mechanics, Lisbon (2000).

[15] Krepper, E.; Frank, Th.; Lucas, D.; Prasser,

H.-M.; Zwart, P.J. The Inhomogeneous

MUSIG model for the simulation of

poly-dispersed flow. Nuclear Engineering

Design 238: 1690-1702 (2008).

[16] Höhne, T.; Deendarlianto; Lucas, D.

Numerical simulations of countercurrent

two-phase flow experiments in

a PWR hot leg model using an area

density model. International Journal

of Heat and Fluid Flow 31 (5):

1047-1056 (2011).

[17] Hänsch, S.; Lucas, D.; Krepper, E.;

Höhne, T. A multi-field two-fluid

concept for transitions between

different scales of interfacial structures.

International Journal of Multiphase

Flow 47:171-182(2012).

Authors

Amirhosein Moonesi Shabestary,

Eckhard Krepper,

Dirk Lucas

Helmholtz-Zentrum

Dresden-Rossendorf

P.O.Box 510119

01314 Dresden, Germany

241

DECOMMISSIONING AND WASTE MANAGEMENT

The Decommissioning of the ENEA RB3

Research Reactor in Montecuccolino

F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo

The ENEA RB3 reactor was a 100 Wth research installation owned and operated by ENEA, in its center of Montecuccolino

near Bologna, from 1971 to 1989. It consisted of a cylindrical aluminium vessel, about 4.3 m high and 2.9 m in diameter,

which could host various types of fuel elements suspended from the top of a special adjustable rack and submerged into

moderating and cooling heavy water. Principal aim of the reactor was to provide neutronics data for the CIRENE NPP, a

SGHWR that was being designed and then partially built in Latina starting from 1979. The specific RB3 core, surrounded

by a graphite reflector and housed inside a concrete biological shielding, allowed to test easily very different fuel

element configurations by changing their pitches and by regulating the heavy water level inside the vessel. The reactor

design, similar to that of the ZED-II Canadian research facility, was originally developed by CEA for its Aquilon facility

in Saclay in 1956; in fact, through a special arrangement between ENEA and CEA, parts of the Aquilon facility were

ultimately donated to ENEA at the end of the 60s for the construction of RB3. In 1989, the RB3 reactor was shut down,

and in the late 2010 ENEA received by ministerial decree the authorization to its dismantling, with the aim of reaching

the “green field” status and with the unconditional release of its building, which is actually owned by the University of

Bologna. The dismantling activities started in May 2013 and were concluded at the end of 2014; after that, a campaign

for the radiological characterization of the building was initiated and concluded in June 2015. Now, all the necessary

site characterization activities are being conducted with the aim to present the results declaring the “green field” status

before the end of 2017. This paper will present the three main pillars of the decommissioning of RB3, namely the

strategy and methods for the dismantling, the strategy and methods for the radiological characterization of the building,

and finally the strategy and methods for the radiological characterization of the site. The radionuclide limits imposed

by the Italian Regulatory Body, together with the challenges encountered so far will be likewise shown and described.

Revised version of

a paper presented

at the Eurosafe,

Paris, France, 6 and 7

November 2017.

Decommissioning and Waste Management

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

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DECOMMISSIONING AND WASTE MANAGEMENT 242

1 Introduction

The ENEA RB3 (Reattore Bologna 3)

reactor was a 100 Wth research installation

owned and operated by ENEA

in its center of Montecuccolino, near

Bologna, from 1971 to 1989. It consisted

of a cylindrical aluminium vessel,

about 4.3 m high and 2.9 m in diameter,

which could host various types of

fuel elements suspended from the top

of a special adjustable rack, and submerged

into heavy water serving both

as moderator and coolant. Principal

aim of the reactor was to provide

neutronics data for the CIRENE NPP, a

SGHWR that was being designed, and

then partially built in Latina, starting

from 1979. The specific RB3 core, surrounded

by a graphite reflector and

housed inside a concrete biological

shielding, allowed to test easily very

different fuel element configurations

by changing their pitches and by

regulating the heavy water level inside

the vessel. The reactor design, similar

to that of the ZED-II Canadian

research facility, was originally developed

by CEA for its Aquilon facility in

Saclay in 1956; in fact, through a

special arrangement between ENEA

and CEA, parts of the Aquilon facility

were ultimately donated to ENEA at

the end of the 60s for the construction

of RB3. In 1989, after more than 18

years of operation, the RB3 reactor

was shut down, and in the late 2010,

after waiting for the entry into force of

Legislative Decree (L.D.) 230/1995

[1], which introduced new laws for

the decommissioning of NPPs, ENEA

received by ministerial decree the

authorization to its dismantling, with

the aim of reaching the “green field”

status and with the unconditional

release of its building, including the

reactor concrete biological shielding,

which is actually owned by the

University of Bologna. In fact the site

of Montecuccolino, some 3.5 km to

the South of downtown Bologna,

hosted three research reactors: RB1,

owned and operated by the University

of Bologna, RB2, owned and operated

by AGIP Nucleare, and RB3, owned

and operated by ENEA. RB1 and RB2

were decommissioned up to the green

field status well before the entry into

force of L.D. 230/1995.

Figure 1 shows an aerial view of

the Montecuccolino research center,

with the area hosting RB3 contoured

in red. Figure 2 shows a plan of the

main reactor hall, with in red the

area once occupied by the reactor

vessel, surrounded by the hectagonal

graphite reflector and encased within

a thick concrete biological shielding.

Figure 3 shows a vertical section of

the RB3 building; the lowermost floor

hosted 4 large tanks for a total of

20,000 L (in red) for the storage of the

heavy water which was daily pumped

up into the vessel to reach criticality

and then drained after the conclusion

of the experiments. Three floors are

present in the building: floor +6.0 m

corresponding to the ground level,

floor +0.0 m, corresponding to the

level of the reactor vessel, and floor

-3.0 m, with the heavy water storage

tanks, heating and cooling systems,

and other auxiliaries. The control

| | Fig. 2.

Plan of main hall of RB3.

room was located at floor +0.0 m.

While allowed to operate up to 100

Wth, operations at RB3 were always

conducted at 50 Wth.

Between 1991 and 1992, all the

fuel elements used at RB3 were either

restituted at their owner (JRC Euratom

Ispra) or sent to the ENEA Research

Center of Saluggia or to the fuel fabrication

plant of Fabbricazioni Nucleari

at Bosco Marengo. Between 1992 and

1993 all the heavy water was transferred

to the ENEA Research Center of

Borgo Sabotino, and before the end of

1997 all the sealed radioactive sources

used at the plant were disposed of.

2 Regulatory Requirements

and Classification

of Components and

Materials

In the late 2010, ENEA received, by

decree of the Italian Ministry of

Economic Development, the authorization

[2] to proceed with the dismantling

of RB3; included in the

| | Fig. 1.

Aerial view of the Montecuccolino site; the RB3 building is inside the red square.

| | Fig. 3.

Section of the RB3 building.

Decommissioning and Waste Management

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

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atw Vol. 63 (2018) | Issue 4 ı April

Radionuclide Metals Concrete Other

materials

Reused Recycle Both reuse and recycle Demolition

Surface

(Bq/cm 2 )

decree were also the requirements

imposed by the Italian Nuclear Regulatory

Body ISPRA. According to these

requirements, the components and

materials of RB3 were classified by

ENEA, taking into account the various

areas of the plant and the history of

its functioning, into four main

categories:

A) materials and components which

were both in contact with possibly

contaminated or activated process

fluids and subject to neutron flux;

B) materials and components which

were in contact with possibly

contaminated or activated process

fluids but not directly irradiated by

neutrons;

C) materials and components which

were irradiated by the neutron flux

but which never went into contact

with possibly contaminated or

activated process fluids;

D) s.c. “exempt” materials, which

were never irradiated and never

went into contact with possibly

contaminated or activated process

fluids.

The only component classified in the

A category was the aluminium vessel;

the only components in the B category

were the heavy water distribution

pipings. Exempt materials, given their

unirradiated and uncontaminated

status, were subject only to a general

screening through CANBERRA In Situ

Object Counting Systems (ISOCS) to

estimate any possible level of presence

of 60Co and 137Cs; if the measured

levels were below the decision threshold

of the measuring system in terms

of mass concentration levels, then

Surface

(Bq/cm 2 )

| | Tab. 1.

Surface or mass activity concentration levels for clearance.

Mass

(Bq/g)

these materials were automatically

discarded from the plant without any

further radiological analysis. This

demonstrates the “instrumental” zero

of this category of materials hence the

“exempt” classification. All materials

which had been classified as “exempt”

were released unconditionately, for a

total mass of about 30 tons, between

March 2013 and May 2015. For all the

other three categories, the clearance

levels imposed by the Regulatory

Authority are summarized in Table 1.

These were derived either from the

Italian L.D. n. 230/95 or from RP 89

[3] and RP 113 [4] publications. In

presence of more than one radionuclide,

the sum of the ratios of

the measured concentrations to the

respective levels must be lower than 1.

The components and materials

were further grouped by ENEA into 12

s.c. “homogeneous groups” using

material and historic criteria; homogeneous

groups are therefore constituted

by components (or parts of

them) made by the same material and

possibly with a homogeneous and

uniform activity content.

3 Radiological Characterization

of Homogeneous

Groups

Before the radiological characterization

of the batches of materials from

the various homogeneous groups

started, a preliminary, special campaign

was conducted to exclude the

presence of various isotopes among

those given in Table 1, expecially in

the most potentially activated or

contaminated materials (category A).

Surface

(Bq/cm 2 )

Mass

(Bq/g)

3 H 10,000 100,000 1 10,000 1 1

14 C 1,000 1,000 1 10,000 1 1

Mass

(Bq/g)

54 Mn 10 10 1 10 0.1 0.1

55 Fe 1,000 10,000 1 10,000 1 1

59 Ni 10,000 10,000 1 100,000 1 1

60 Co 1 10 1 1 0.1 0.1

63 Ni 1,000 10,000 1 100,000 1 1

90 Sr 10 10 1 100 1 1

125 Sb 10 100 1 10 1 1

134 Cs 1 10 0.1 10 0.1 0.1

137 Cs 10 100 1 10 1 1

152 Eu 1 10 1 10 0.1 0.1

154 Eu 1 10 1 10 0.1 0.1

Generic Alfa 0.1 0.1 0.1 0.1 0.1 0.01

241 Pu 10 10 1 100 1 1

In particular 54Mn, 59Ni, 90Sr,

125Sb, 134Cs, 137Cs, 239Pu, 240Pu

and 241Pu were excluded from

further analyses finalized to the unconditional

release of materials. Then,

for each homogeneous group, a precharacterization

measurement campaign

was con ducted with a three-fold

aim: 1) to verify if the hypothesis on

the homogeneity of activity for that

given group held; 2) to evaluate the

minimum number of samples to be

analized

for the subsequent characterization

phase; 3) to evaluate the value of

isotopic ratios of 55Fe to 60Co and

of 63Ni to 60Co, so to limit the next

analyses only to the research of 60Co

contents. After that, and using typically

13 multiple measurements for

each batch of each homogeneous

group, summations of the ratios

between measured activity concentrations

and limits (Table 1) over all

the relevant isotopes were carried out.

If these summations resulted

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atw Vol. 63 (2018) | Issue 4 ı April

DECOMMISSIONING AND WASTE MANAGEMENT 244

| | Fig. 4.

Dismantling of the lower layers of the graphite reflector.

in a 1:10 ratio with other similar

metals of warranted non-nuclear

provenance in order to be used again

for various purposes. All the homogeneous

groups were pre-characterized,

characterized and released before the

end of 2014. All the measurements

were performed by trained ENEA staff

and within qualified ENEA laboratories,

with the exception of some 14C

measurements of a small lot of rubbers

which were performed, under special

contract, by the LASE Laboratory

of CEA in Saclay. Workmen for heavy

or peculiar technological operations

were hired from the Modena Fallimenti

SaS, a private Italian company specialized

in the dismantling of special

plants. Further details about the plan

for the characterization of materials

and components can be found in

[5,6].

4 Radiological Characterization

of the Building

After the completion of all the dismantling

activities, and after the release

of all the batches of materials, a

radiological characterization of the

building of RB3 has been made. This

consisted of two main steps. The first

was the characterization of the activation

status of the baritic concrete

biological shielding of the core. This

consisted in seven core drillings, (see

Figure 5) each 16 cm long, so distributed:

1 on the floor, 1 on the

northern wall, 1 on the western wall,

1 on the eastern wall, and 3 (at

different heights) on the souther wall.

All the drilling points were at positions

where the neutron flux during

operation was maximum. From each

drilling, four aliquots, 4 cm long,

where taken, so to cover the depth

profile of any activation distribution

inside the biological shielding. Each

aliquot was subject to gamma spectrometry

to search for the presence of

60Co, 134Cs, 152Eu and 154Eu. All

the 28 samples yielded results for all

the four isotopes lower than a few

mBq/g. Then, all the samples were

subject to thorough statistical analysis,

based on several Bartlett tests, to

verify if they were all and altogether

representative of the same statistical

distribution of activity and therefore

representative of a same “homogeneous

group” constituted of the whole

biological shielding. Once this condition

has been verified, a Noether test,

using 10 randomly chosen measurements,

was put in place to verify the

minimum number of samples to be

used for the final characterization of

the biological shielding. This resulted

in 13 samples, randomly extracted

from the complete set of all the 28

available samples. However, ENEA decided

to use all the 28 samples to verify

the free release condition for the

shielding, and for all the 28 samples

the condition resulted verified, meaning

that no significative activation of

the shielding had been realized. As a

further consequence, it could be

proven that no activation of walls

outside the biological shielding was

in place, just because, due to its

screening effect, the neutron flux

outside the shielding itself was 6 to 7

orders of magnitude lower.

The second step of the characterization

consisted in the assessment of

the surface contamination of the various

areas of the building. These were

separated into three main surfaces:

1) ceiling; 2) surfaces over +6.0 m

level; 3) surfaces below +6.0 m level.

The ceiling was indeed a false ceiling

made of thin aluminium plates; these

could have been contaminated by

tritiated water vapours emerging

from the core once open for refueling

or fuel reshuffling. To investigate this,

the aluminium plates were dismantled,

taken to ground, and analyzed. It

was assumed that, if no contamination

was found, then also the real

ceiling behind it was not contaminated.

This proved indeed to be the

case. Surfaces over +6.0 m were

investigated randomly (Figure 6), by

sampling a given number of points,

quantified basing on statistical considerations.

All surfaces below +6.0 m

were completely measured, both walls

and floors. The measurement technique

consisted in using surface

contamination meters (Berthold

LB165 and LB124), properly cali brated

with large area reference sources, to

sum up count rates over 14C, 60Co,

134Cs, 152Eu and 154Eu. A similar

measurement methodology was successfully

applied for the decommissioning

of the ASTRA research reactor

in Vienna [7]. Background contributions

due to natural radionuclides in

the different materials were subtracted

after having made suitable averages

from surely clean, similar materials

to those which were to be measured

inside the building. As a further, conservative

penalization, it was decided

to attribute to each of the 5 abovementioned

nuclides the whole net

counting over each surface portion

being measured, counting time per

surface element being about 30 seconds

to reach a desired minimum

detectable activity. LB124 hand held

monitor was used for surfaces over

+6.0 m, while LB165 (wheeled monitor

as in Figure 7) was used over all

other surface portions. A special automated

vertical translational sledge

(Figure 8) was used to carry LB165

over the portions of the walls. In case

a given measurement yielded values

above the clearance limits, special

cleaning procedures were to be

adopted until subsequent measurements

proved to be below the limits

| | Fig. 5.

Core drilling of the biological shielding.

| | Fig. 6.

LB124 measurements of selected portions of walls above +6.0 m level.

Decommissioning and Waste Management

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

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atw Vol. 63 (2018) | Issue 4 ı April

| | Fig. 7.

LB165 measurement of floor portions.

themselves. No cleaning procedures

were ever necessary and all the surface

portions could be considered not

contaminated and so able to be freely

released.

5 Radiological Characterization

of the Site

The third and last pillar of the decommissioning

of RB3 is the radiological

characterization of the areas of the

site surrounding the building. In this

regard, it is important to mention that

during the operational life of the

plant, given its low power and its

construction features, no radiocativity

discharges were in place and therefore

no environmental analyses were prescribed

by the Regulatory Authority.

Another point worth of mentioning is

that no radiological status of the site

prior to the construction and exercise

of RB3 is known. However, in light of

the graded approach which is going to

be taken for this third pillar by the

Regulatory Authority, given the fact

that no activated materials have been

found and that no activation or contamination

of the building has been

detected, it is decided to base this

characterization upon the measurement

of some selected nuclides in

certain terrain samples (soil) taken

around the area of the RB3 site.

In particular, 12 measurements of

239+240Pu through alpha spectrometry

will be done, together with

25 gamma spectroscopy assessments

for 54Mn, 60Co, 125Sb, 134Cs, 137Cs,

| | Fig. 9.

Collecting soil samples from the RB3 site.

Radionuclide

152Eu and 154Eu. Each terrain

sample will be a parallelepiped of

25x20x10 cm 3 corresponding roughly

to 5 liters of humid soil (Figure 9).

The site will be sampled considering

both near-range and far-range positions

in order to find patterns of radioactivity

correlated with the distance

from the RB3 building, if any at all.

The obtained values will be confronted,

through proper summations,

with the limits for the free release

of nuclear sites prescribed by the

German national law, which correspond

to the radiological nonrelevance

value of 10 microSv/year

to the public [8,9]. The limits for

the above-mentioned isotopes are

reported in Table 2.

References

[1] D.Lgs. 17 marzo 1995, n. 230,

Attuazione delle direttive Euratom

80/836, 84/467, 84/466, 89/618,

90/64, 92/3, 96/29.

[2] D. M. 29 Novembre 2010 Ministero

dello Sviluppo Economico di

Autorizzazione alla Disattivazione

Impianto Nucleare di Ricerca Reattore

RB-3 di Montecuccolino (BO) dell’ENEA.

[3] Radiation Protection 89, Recommended

radiological protection criteria for the

recycling of metals from the

dismantling of nuclear installations,

European Commission, 1998.

| | Fig. 8.

LB165 and its translational sledge to measure wall portions.

Concentration Limit

(Bq/g)

54 Mn 0.09

60 Co 0.03

125 Sb 0.08

134 Cs 0.05

137 Cs 0.06

152 Eu 0.07

154 Eu 0.06

239 Pu 0.04

240 Pu 0.04

| | Tab. 2.

Proposed clearance limits for the free release

of the RB3 site.

[4] Radiation Protection 113,

Recommended radiological protection

criteria for the clearance of buildings

and building rubble from the

dismantling of nuclear installations,

European Commission, 2000.

[5] I. Vilardi, C. M. Castellani, D. M.

Castelluccio, F. Rocchi, Piano di

Caratterizzazione Radiologica di Materiali

provenienti dalla Disattivazione

dell’impianto Nucleare di Ricerca Rb-3

dell’enea sito in Bologna – Montecuccolino

ai Fini del loro Allontanamento,

Convegno Nazionale AIRP 2014, Aosta.

[6] M. Capone, N. Cherubini, A. Compagno,

A. Dodaro, F. Rocchi, The Dismantling of

the Montecuccolino RB3 Research

Reactor: Radiological Characterisation of

Materials for Free Release, Proceedings

of the European Reaserch Reactor

Conference RRFM 2015, Bucharest

19-23 April 2015, 528-537.

[7] F. Meyer, F. Steger, R. Steininger,

Decommissioning of the Astra Research

Reactor – Dismantling the auxiliary

Systems and Clearance and Reuse of the

Buildings, Nuclear Technology &

Radiation Protection, 1/2008, 54-62.

[8] OECD/NEA Status Report, Releasing

the Sites of Nuclear Installations,

NEA Report 6187, 2006.

[9] Bundesgesetzblatt G 5702 Teil I, Bonn

26 July 2001, Nr. 38, 2001.

Authors

F. Rocchi

ENEA FSN/SICNUC/SIN

C. M. Castellani

ENEA IRP

A. Rizzo

ENEA FSN/SICNUC/TNM

Via Martiri di Monte Sole 4

Bologna (BO), Italy

A. Compagno

ENEA FSN/FISS/CRGR

I. Vilardi

ENEA IRP/SFA

Via Anguillarese, 301

00123 S.Maria di Galeria (RM), Italy

R. Lorenzelli

ENEA FSN/SICNUC/SIN

Località Brasimone

40032 Camugnano (BO), Italy

DECOMMISSIONING AND WASTE MANAGEMENT 245

Decommissioning and Waste Management

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

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246

RESEARCH AND INNOVATION

Revised version of a

paper presented at

the Annual Meeting

of Nuclear Technology

(AMNT 2017), Berlin.

Irradiation Tests of a Flat Vanadium Self-

Powered Detector with 14 MeV Neutrons

Prasoon Raj and Axel Klix

Self-powered detector (SPD) represents a class of neutron and gamma monitoring instruments used in the fission

reactor cores worldwide. This detector has inherent advantages of functioning without a bias voltage, simple measurement

scheme, compactness, ease of maintenance, and high reliability. We are studying SPD for application as flux

monitors in the European test blanket modules (TBM) of ITER, fusion reactor under construction in southern France.

This paper presents results of experimental tests performed with 14 MeV neutrons for a flat SPD with vanadium emitter.

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

thermal and fast neutrons. Secondary electrons due to gammas from these reactions and neutron irradiation of

surrounding materials are also important contributors to the signal. Thin foils of emitter, insulator and collector

materials are used to construct the test SPD. The detector is irradiated with short and long pulses of neutrons and is

found to respond in proportion with the incident neutron flux. Further experiments with simplified and better optimized

design of detector are underway for thorough study of the signal-creation mechanism.

1 Introduction

ITER [1] is an experimental fusion

reactor based on tokamak concept,

under construction at St. Paul lez

Durance in southern France. It is an

international project aimed at proving

feasibility of fusion as a large-scale

and carbon-free source of energy. One

of the main scientific goals of this

project will be to test and prove the

concepts of tritium breeding blankets.

Tritium is an important fuel component

for devices based on D-T reaction,

which is being considered as

main reaction for fusion power plants.

Because tritium is a rare element, it is

required to breed it in the fuel cycle of

the reactor. A blanket with lithium

compounds will cover the inner wall

of the plasma vessel. Fusion neutrons

from the plasma will be absorbed by

lithium nuclei, causing reactions to

produce tritium.

There are multiple breeding

blanket designs proposed by scientists.

To determine their efficiencies in

a real fusion environment, test blanket

modules (TBM) based on different

concepts will be inserted into equatorial

ports of ITER for experimental

tests in different operational phases of

ITER. The European Union is going to

test two such concepts, namely the

Helium-Cooled Lead-Lithium (HCLL)

and Helium-Cooled Pebble Bed

(HCPB) TBMs [2]. In the neutronics

experiments, nuclear responses like

tritium production rate, material activation,

nuclear heating etc. are to be

measured and compared with the

calculations. This step will validate

the advanced computational tools

and nuclear data utilized for nuclear

analyses for fusion devices. The neutron

and gamma fluxes are important

quantities to be measured for these

experiments, for which detectors like

neutron activation system, fission

chambers and self-powered detectors

(SPD) are under study.

An SPD is a multi-layered electrical

device, which produces direct current

(DC) signal on irradiation with

neutrons and/or gammas. It can be

preferentially responsive to neutrons

(self- powered neutron detector,

SPND) or gammas (SPGD), or as

it is in most of the cases, to both.

Figure 1 shows a rough sketch of the

cross- section of a traditional detector.

Central material, called emitter produces

fast electrons on irradiation.

These fast electrons can be betas

from the decay of neutron activation

products, or secondary electrons due

to interaction of gammas in the bulk

of the material. They slow down in a

layer of insulation and stop in the

outer electrode called collector. This

electron-movement creates a potential

difference and thus, produces a

current signal proportional to the

incident particle flux. The current due

to beta electrons is “delayed” because

of the half-life of beta-emitters, e.g.

SPND based on Rh, V or Ag emitters.

Whereas that due to gamma-initiated

photoelectric or Compton electrons

is “prompt”, for example Co-based

SPND [3].

An SPD responds in a sophisticated

manner, with multiple factors

contributing to the small current

signals often totaling between 10 -12

and 10 -3 Ampere. Due to its inherent

advantages of simplicity, compactness

and high-reliability, they are highly

desirable for flux monitoring in areas

with restricted access like reactor

cores. At KIT, we are studying SPDs

for application in ITER TBM [4].

Vanadium based flat SPD is being

tested with 14 MeV neutrons, to

understand its behavior towards fast

neutrons expected in fusion environment

and ascertain the feasibility of

its application as flux monitor for

European ITER TBMs.

2 Experimental details

Vanadium is a common emitter for

fission reactor SPNDs. The response

of the detector towards thermal neutrons

is understood well. The material

is relatively inexpensive and easier to

handle. However, due to lower cross

sections the sensitivity of vanadium-

SPND towards fast neutrons reduces

(Figure 2). Commercially available

SPND cannot be directly used for

measurement of fusion neutron

fluxes, going up to approx. 14 MeV in

energy.

Characteristics of the two main

beta- emitters from 51 V (99.75 %

isotopic abundance) in case of fast

neutron irradiation, are reported in

Table 1. Cross-sections of the fast

neutron reactions in 51 V for a pure

| | Fig. 1.

Cross-sectional sketch of a cylindrical SPD showing emitter (green), insulator (dotted white) and

collector (black) layers, with connection to the lead cable and current measurement device.

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Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix

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| | Fig. 2.

Cross sections of vanadium reactions and photon production under neutron irradiation.

Reaction 51 V (n, p) 51 Ti 51 V (n, γ) 52 V

Threshold Neutron Energy 1.72 MeV 0 MeV

14 MeV Cross-section 30 mb (approx.) 0.6 mb (approx.)

Beta Emitter, Half-life 51 Ti- 5.76 m 52 V- 3.74 m

Average Beta Energy 51 Ti- 0.87 MeV 52 V- 1.07 MeV

SPND Current (14 MeV) 3.46 × 10 -12 A 6.92 × 10 -14 A

SPND Current (TBM) 7.97 × 10 -9 A 3.44 × 10 -8 A

| | Tab. 1.

Beta-emitters and corresponding currents from fast neutron reactions in vanadium based SPND.

14 MeV source are shown. Neglecting

the self-shielding of electrons in emitter

material, effect of other materials

and taking a saturation condition

(considering the short half-lives of

daughter nuclides), one can ascertain

the orders of magnitude of currents

possible with V-SPND, as reported.

For this estimation, vanadium density

of 6.1 g cm -3 , and a typical volume of

1 cm 3 are assumed. For a 14 MeV

neutron source, a flux intensity of

1 × 10 10 cm -2 s -1 is considered, which

is achievable with state of the art

14 MeV neutron generators. For TBM,

activation calculation was done [5]

with the HCLL neutron spectrum

and typical flux intensity (up to 1 ×

10 14 cm -2 s -1 ) using EASY-2007 [6].

With high-sensitivity ammeters,

currents down to the order of 1 ×

10 -14 A can be reliably measured [7].

Values in Table 1 show that a vanadium

emitter based SPND will produce

measurable signals in TBM. Due

to its high neutron threshold energy,

the (n, p) reaction can be utilized to

measure fast neutron flux exclusively.

Fast neutron reactions lead to

high-energy gamma production. This

phenomenon competes with the neutron

absorption reactions (Figure 2).

Photoelectric and Compton electron

emission from emitter causes a prompt

current which is expected to form the

major component of the signal of

V-SPND towards 14 MeV neutrons.

Secondly, vanadium being a medium-

Z nucleus can be a potential

emitter for SPGD also. With optimized

dimensions and choice of collector

material, a vanadium SPD can be

envisaged for monitoring of photon

flux in TBM.

Instead of the usual coaxial type

cylindrical geometry, we designed

our test SPD in sandwich-type flat

geometry. This provides a relatively

higher cross section area to the incident

neutrons, and ease of access for

testing various materials in the same

device. Thin foils (0.5 to 2 mm) of

emitter, insulator and collector are

arranged to form an assembly in an

aluminum case, which also serves as

an electromagnetic shield. Central

conductor of the signal cable is linked

to the emitter plates of the detector.

The collector plates, case and the

cable sheath are shorted and securely

connected to the ground. Schematic

sketch and photograph of the test

detector are shown in Figure 3 (left).

With comparable cross sections of

reactions in different materials, the

insulator and collector materials also

play an important role in SPD

response. Behaviors of different

material combinations are experimentally

tested. Alumina (Al 2 O 3 ) or

beryllia (BeO) is used as insulator and

Inconel-600 or graphite is used as

collector in our experiments. Effects

of the change of geometry and dimensions

are also studied. A Keithley 6485

Picoammeter (sensitivity range -20 fA

to 20 mA) is used as the measuring

device. A low-noise triax cable (Belden

9222) is used to reduce the interferences

in low-current measurement.

The tests are conducted at the

14 MeV neutron generator of Technical

University of Dresden (TUD-NG),

shown in Figure 3 (right). Here,

deuteron beams are impinged on a

tritiated titanium target causing D-T

reaction which leads to production of

neutrons with peak energy of approx.

14 MeV. TUD-NG offers neutron flux

intensities up to 1 × 10 10 cm -2 s -1 . The

detector is placed in front of the

tritium- target of TUD-NG and tested

under different conditions by varying

flux levels and irradiation times.

3 Results

The irradiation tests of flat sandwichtype

vanadium SPD were performed

at TUD-NG, with neutron flux intensities

around 1 × 10 9 cm -2 s -1 . DC

signals in the range of 100 fA to 100 pA

were measured. In Figure 4, a plot

shows variation of SPD signal with

change in neutron flux. The detector

was composed of 1 mm thick layers of

vanadium emitter and Inconel-600

collector. The signal was found to be

proportional to the incident flux, with

approx. 90 pA at the highest flux level.

At low fluxes and low currents,

the measurements have high uncertainties.

Interference from electromagnetic

sources of stray currents,

| | Fig. 3.

(Left) internal design of the sandwich-type flat SPD: (top)- an engineering sketch of the geometry

having sandwich of foils of emitter (green), insulator (grey) and collector (red), and (below) a photograph

of the assembly with vanadium SPD.

(Right) experimental setup showing TUD-NG beamline, tritium target, mounted SPD, and the lead cable.

RESEARCH AND INNOVATION 247

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Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix

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RESEARCH AND INNOVATION 248

| | Fig. 4.

Vanadium SPD signal (left Y-axis, red curve) variation with change in neutron flux (right Y-axis, blue

curve) plotted with respect to irradiation time.

currents generated in coaxial cables,

electrostatic effects at the contacts

and degradation of insulation layer

due to radiation, lead to background

currents in the orders of 100 fA. This

makes the measurement of low-level

currents a very challenging task.

The SPD response is often reported

in terms of sensitivity, which is SPD

current per unit of neutron (or

gamma) flux intensity, reported in

units of A cm 2 s. For the vanadium

SPND signal in Figure 4, the sensitivity

lies between 4.48 × 10 -20 A cm 2 s

± 13.4 % (at flux intensity 2.04 ×

10 9 cm -2 s -1 ) and 8.80 × 10 -19 A cm 2 s

± 51.1 % (at flux intensity 6.40 ×

10 5 cm -2 s -1 ).

In another test, a constant-flux

irradiation of around 15 minutes was

done and the TUD-NG was switched

off. This signal is shown in Figure 5. It

was found that the detector current is

dominated by a prompt component

which appeared and disappeared with

neutron flux. The delayed signal is

usually less than 10% of the total

signal. A decay of delayed current was

observed as expected.

There are parasitic beta emission

reactions in insulator, collector and

cable’s central conductor, e.g. 27 Al (n,

p) 27 Mg reaction (half-life~ 9.46 min)

in alumina insulation. Electrons

emitted due to these reactions reduce

the total delayed current. Due to this,

the analysis of decay curve becomes

very complex. After data reduction,

subtraction of background contributions,

and further analysis the major

contribution was found to be from 51 Ti

due to 51 V (n, p) 51 Ti reaction. For

reduction of aforementioned effects

materials with lower total cross

sections of beta emission reactions,

like graphite and beryllia were used as

collector and insulator, respectively.

The change in the signal characteristics

was minimal with these alterations,

leading us to conclude that the

signal was mainly due to reactions in

the vanadium emitter. A prompt current,

makes the detector suitable for

pulsed devices like ITER. However, it

is important to understand the signal

creation mechanism for calibration

and application of the SPD.

The high prompt signal is attributed

to three main reasons. First is

the interaction of photons in the

emitter volume, which release high

energy electrons producing high

positive current. Unlike thermal neutrons,

fast neutrons lead to emission

of higher-energy photons with higher

probability of secondary effects.

Moreover, the photon production

cross section is usually an order or two

higher than the fast neutron reaction

cross sections in materials of detector

and surroundings (Figure 2). Secondly,

the production of charged particles

like protons and alphas in collector

and insulator material (cross sections

of (n, xp) and (n, xα) reactions are

high for 14 MeV neutrons) lead to

further difference of charge between

electrodes and a prompt positive

contribution to the signal. Finally, the

electrical and nuclear effects in

connecting wires and cables make a

small fraction of the positive current

signal

Some of the contributing factors

will be explicitly studied in future

tests. To de-couple the effects of other

materials, a detector with simplified

geometry is under design. An air-insulated

detector with box of collector

material is being constructed. The

material thicknesses are reduced in

order to decrease the gamma interactions.

Improved ways of making

electrical contacts between cable and

emitter are studied. Other less betaactive

materials like niobium are

being considered for collector. Vanadium

detector is also planned to be

optimized for photon response. To this

end, thicker emitters and collectors

with low gamma-activity will be used

to make a test-device which will be

irradiated with high-energy bremsstrahlung

photon source.

4 Conclusions

A flat sandwich-type vanadium SPD

has been constructed, for testing the

feasibility of application of SPDs in

ITER TBMs. Irradiation tests with

14 MeV neutrons at TUD-NG resulted

in current signals in range of 100 fA to

100 pA. The signals are proportional

to the incident neutron flux. Considering

the higher flux intensities up

to 1 × 10 14 cm -2 s -1 and a wider energy

spectrum of neutrons in TBM, studies

show that vanadium SPND is expected

to produce measurable signals in ITER

| | Fig. 5.

Vanadium-SPD signal in a long constant-flux irradiation at TUD-NG showing (prompt and delayed)

currents before, during and after the irradiation.

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Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix

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TBM conditions. The high prompt

component of the SPD signal is

attributed to the interaction of high

energy photons which are produced

in the detector and surrounding

materials. Charged particles emitted

in fast neutron reactions and contributions

from wires and signal cable

contribute to the high positive signal.

Parasitic reactions in non-emitter

materials also play an important role.

These effects need to be studied

explicitly and compared for understanding

of the overall currentgeneration

mechanism. Optimization

of design, dimensions and material

combinations is underway to realize

SPD flux monitors for application in

European ITER TBMs.

Acknowledgement

The work leading to this publication

has been funded partially by Fusion

for Energy under the Specific

Grant Agreement F4E-FPA-395-1.

This publication reflects the views

only of the authors, and Fusion for

Energy cannot be held responsible for

any use which may be made of the

infor mation contained therein.

References

[1] ITER Organization – Homepage. [Online].

Available: https://www.iter.org/.

[2] P. Calderoni, Status of the HCLL and

HCPB Test Blanket System instrumentation

development, 21 st Top. Meet.

Technol. Fusion Energy (TOFE), 9-13

Nov. 2014, Anaheim, CA, USA.

[3] N. P. Goldstein and W. H. Todt, A Survey

of Self-Powered Detector - Present and

Future, IEEE Trans. Nucl. Sci., vol. 26,

no. 1, pp. 916–923, 1979.

[4] P. Raj, M. Angelone, U. Fischer, and

A. Klix, Self-powered detectors for test

blanket modules in ITER, in 2016 IEEE

Nuclear Science Symposium, Medical

Imaging Conference and Room- Tem perature

Semiconductor Detector Workshop

(NSS/MIC/RTSD), 2016, pp. 1–4.

[5] M. Angelone, A. Klix, M. Pillon, P.

Batistoni, U. Fischer, and A. Santagata,

Development of self-powered neutron

detectors for neutron flux monitoring in

HCLL and HCPB ITER-TBM, Fusion Eng.

Des., vol. 89, no. 9–10, pp. 2194–2198,

2014.

[6] R. A. Forrest, FISPACT-2007: User

manual, EASY Doc. Ser. UKAEA

FUS 534, 2007.

[7] Low Level Measurements Handbook –

7 th Edition: Precision DC Current,

Voltage, and Resistance Measurements.

Keithley- A Tektronix Company.

Authors

Prasoon Raj

Axel Klix

Institute for Neutron Physics and

Reactor Technology (INR)

Karlsruhe Institute of Technology

(KIT)

Hermann von Helmholtz Platz 1

76344 Eggenstein-Leopoldshafen

(Germany)

RESEARCH AND INNOVATION 249

Nanofluid Applied Thermo-hydrodynamic

Performance Analysis of Square

Array Subchannel Under PWR Condition

Jubair Ahmed Shamim and Kune Yull Suh

1 Introduction Efficient engineered design of heat transfer and fluid flow with enhanced heating or cooling

requires two pivotal aspects that must be taken into consideration for extracting thermal energy from nuclear fission

reactions in order to save energy, reduce process time, raise thermal rating and increase the operating life of a reactor

pressure vessel. Hence, one of the major challenges in designing a new nuclear power plant is the quantification of the

optimal flow of coolant and distribution of pressure drop across the reactor core. While higher coolant flow rates will

lead to better heat transfer and higher Departure from Nucleate Boiling (DNB) limits, it will also result in higher pressure

drop across the core, therefore additional demand of pumping powers as well as larger dynamic loads on the core

components. Thus, thermal hydraulic core analysis seeks to find proper working conditions with enhanced heat transfer

and reduced pressure drop that will assure both safe and economical operation of nuclear plants.

Recently, nanofluid has gained much

renewed attention as a promising

coolant for pressurized water reactors

(PWRs) due to its enhanced thermal

capabilities with least penalty in pressure

drop. The improved heat transfer

of nanofluids results from the fact that

the nanoparticles increase the surface

area and heat capacity of the fluid,

improve the thermal conductivity of

the fluid, cause more collisions and

interactions between the fluid, particles

and surfaces of the flow passages,

and enhance turbulence and mixing

of the fluid.

Pak & Cho [1] experimentally

observed the turbulent friction and

heat transfer of dispersed fluids in a

circular pipe using two different

metallic oxide particles, γ-alumina

(Al 2 O 3 ) and titanium dioxide (TiO 2 )

with mean diameters of 13 and 27 nm,

respectively. The results revealed

that the Nusselt number Nu for the

dispersed fluids increased with

increasing volume concentration as

well as the Reynolds number Re. But

at constant average velocity, the

convective heat transfer coefficient for

the dispersed fluid was 12% less than

that for pure water. They proposed a

new correlation for Nu under their

experimental ranges of volume concentration

(0-3%), Re (10 4 -10 5 ), and

the Prandtl number Pr (6.54-12.33)

for the dispersed fluids γ-alumina

(Al 2 O 3 ) and titanium dioxide (TiO 2 )

particles as

(1)

Xuan and Li [2] observed the flow

and convective heat transfer of the

Cu-water nanofluid flowing through

a straight brass tube of the inner

diameter of 10 mm and the length of

800 mm. They noted that suspended

nanoparticles can remarkably enhance

heat transfer given the velocities.

For instance, the heat transfer

coefficient of nanofluids containing

2.0 vol % Cu nanoparticles was increased

by as much as 40 % compared

to that of water. The conventional

Research and Innovation

Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh

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RESEARCH AND INNOVATION 250

Dittus–Boelter correlation failed to

predict this augmented heat transfer

data for nanofluids. They presented a

new correlation for turbulent flow of

nanofluids inside a tube as

(2)

Maïga et al. [3] numerically investigated

fully-developed turbulent flow

of water/Al 2 O 3 nanofluid through

circular tube using different concentrations

under the constant heat flux

boundary condition. They proposed

the following correlation for 10 4 ≤

Re ≤ 5×10 5 , 6.6 ≤ Pr ≤ 13.9 and 0 ≤

φ ≤ 10%

(3)

Asirvatham et al. [4] reviewed the

published experimental investigations

on convective heat transfer of different

nanofluids.

Despite numerous studies on both

scaled experiments and numerical

modeling on heat transfer enhancement

of nanofluids proliferate over

the past years, most of the test sections

and computational domain were

limited to round pipes. Their simulating

parameters did not reflect the

environment of a nuclear power reactor,

either. Wu and Trupp [5] demonstrated

that flow conditions inside the

fuel rod assembly are quite different

from those in typical pipes. There is

so far no appropriate correlation in

literature that can predict heat transfer

characteristics of nanofluid in a

fuel assembly under PWR operating

condition. Therefore, numerical modeling

has been performed in this study

using a commercial computational

fluid dynamic CFD tool “Star-CCM+

(ver.9.06.011)” to predict heat transfer

and pressure drop more precisely

in a square array subchannel (1.25 ≤

P/D ≤ 1.35) for different volume concentrations

of water/alumina (Al 2 O 3 )

nanofluid (0.5% ≤ φ ≤ 3.0%). Referring

to the Advanced Power Reactor

1400 MWe (APR1400).

Properties

Also, if the slip between the particles

and the continuous phase is trifling,

the flow inside the subchannel may as

well be considered as single phase and

incompressible with constant physical

properties. Both the compression

work and viscous dissipation are

neglected. Under such conditions the

general conservation equations for

mass, momentum and energy can be

written in vector notations:

∇.(ρv) = 0 (4)

∇.(ρvv) = -gradP+μΔ 2 v (5)

∇.(ρvC P T) = ∇.(k gradT) (6)

where v, P and T are fluid velocity

vector, pressure and temperature,

respectively.

2.2 Determination of physical

properties of nanofluid

Determination of physical properties

of nanofluid is key to any nanofluid

research. If the nanoparticles are

assumed to be well dispersed in the

base fluid, the particle concentration

can be considered as constant

throughout the domain and effective

physical properties of mixture can be

evaluated using some classical formulas

well known for two phase fluids

[7]. The following formulas are used

to determine such properties as density,

specific heat, dynamic viscosity

and thermal conductivity.

ρ nf = (1-ϕ)ρ bf + ϕρ P (7)

(C P ) nf = (1-ϕ)(C P ) bf + ϕ(C P ) P (8)

μ nf = (1 + 7.3ϕ + 123ϕ 2 )μ bf (9)

Base Fluid

(Pure Water)

Alumina

Nanoparticles

Density (kg/m 2 ) 734.928 3970

Thermal Conductivity (W/m.K) 0.5701 40

Specific Heat (J/kg. K) 5361.69 880

Dynamics Viscosity (Pa. s) 9.01373E-05 -

| | Tab. 1.

Physical properties of base fluid and alumina nanoparticles.

and later improved by Brinkman [10]

and another by Batchelor [11], these

formulas drastically underestimate

the viscosity of nanofluids. Therefore,

they performed a least-square curve

fitting based on some scarce experimental

data available [12, 13, 14]

which leads to Equation (9). Equation

(10) [7, 15] is introduced for the thermal

conductivity as with the dynamic

viscosity. However, the pressure and

temperature of the above investigations

sizably differ from the operating

condition of a PWR. Since no such

correlation exists for thermophysical

properties of nanofluid applicable to

the operating environment of a PWR it

is assumed that the aforementioned

correlations can also be utilized for

nuclear reactors. Different properties

of base fluid (pure water) and alumina

nanoparticles that have been used in

this study are tabulated in Table 1.

3 Numerical modelling

3.1 Computational domain

The computational domain and

boundaries considered in this study

are shown in Figure 1, which represents

a quarter of a 3-D square array

subchannel created in Star-CCM+.

The diameter of the fuel rod is taken

as 9.5 mm and pitch-to-diameter ratio

P/D of 1.25 and 1.35 are selected for

simulation. The length of the subchannel

is taken as 600 mm which

is long enough to establish a fullydeveloped

turbulent flow at the outlet

under single phase forced convection

condition up to Re = 6×10 5 according

to the following criteria [16]

2 Mathematical modelling

k nf = (1 + 2.72ϕ + 4.97ϕ 2 )k bf (10)

2.1 Governing equations

The term “nanofluid” refers to a twophase

mixture of saturated liquid and

dispersed ultrafine particles of usual

size below 40 nm. However, due to

extremely tiny size of particles, it can

be readily fluidized and thus may be

considered to behave more like a fluid

rather than heterogeneous fluid [6].

Equations (7) and (8) are general

relationships being used in literature

[1, 7, 8] to compute the density and

specific heat for a classical two phase

mixture. Regarding the dynamic

viscosity, Maïga et al. [9] showed that,

albeit several correlations exist to

calculate the dynamic viscosity of

nanofluid as proposed by Einstein

| | Fig. 1.

Computational domain created in Star-CCM+.

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P/D = 1.25

Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid

(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %

6×10 5 7.829 7.963 8.351 9.196

5.098×10 5 6.651 6.766 7.095 7.813

4×10 5 5.219 5.309 5.568 6.130

3×10 5 3.914 3.982 4.176 4.598

| | Tab. 2.

Different inlet velocities, v 0 (m/s) used in simulation.

(11)

l e = EI × D h (12)

where l e is entrance length for fullydeveloped

flow, EI is entrance length

number and Dh is the channel hydraulic

diameter.

3.2 Boundary conditions &

Physics set-up

The coolant enters the subchannel

with a uniform inlet velocity v 0 (m/s)

at the inlet temperature 569 K. Different

values of v 0 for different coolants

that have been used in the simulation

are listed in Table 2. Different properties

of base fluid (pure water)

have been calculated at temperature

569 K and at pressure 155.1375 bar.

At the outlet, a static pressure of

155.1375 bar has been imposed. On

the tube wall, the usual non-slip

conditions with the standard wall

function are considered with a constant

heat flux of 600,000 W/m 2 . The

above parameters and geometric configurations

of the computational

domain are based on the design

features of the APR1400.

The constant density model is chosen

for the material. For turbulence

modeling, the realizable k-ε model

with high y + wall treatment is selected.

Implicit coupled solver with secondorder

upwind discretization scheme in

conjunction with coupled energy

model is implemented which solves

the conservation equations for mass

and momentum simultaneously using

a pseudo time marching approach.

3.3 Turbulence modeling

By studying different literature on

numerical simulation of flow through

a rod bundle for nuclear applications,

it can be concluded that no specific

turbulence model can be regarded as

superior to others for this sort of flow

phenomena. Yadigaroglu et al. [17]

carried out an exhaustive review of

rod bundle numerical simulations

and opined that the gradient transport

models, like the standard k-ε

model, are not capable of predicting

turbulent flow in the narrow gap regions.

Hàzi [18] had demonstrated

that the Reynolds Stress Model (RSM)

could be accurately applied in simulating

the rod bundle geometry. Lee

and Choi [19] also used the RSM turbulence

model to compare the performance

of grid designs between the

small scale vortex flow (SSVF) mixing

vane and the large scale vortex flow

(LSVF) mixing vane. Liu and Ferng

[20] have also adopted RSM turbulence

model to numerically investigate

the effects of different types of

grid (standard grid and split-vane pair

one) on the turbulence mixing and

heat transfer. Palandi et al. [21] have

successfully implemented SST k-ω

model in comparing thermo-hydraulic

performance of nanofluids and

mixing vanes in VVER-440 triangular

array fuel rod bundle. However, application

of RSM turbulence model will

require 50-60% more CPU time per

iteration and 15-20% more memory

usage compared to standard k-ε and

k-ω model.

Recently Conner et al. [22] have

implemented renormalization group

(RNG) k-ε model (Yakhot et al., [23])

in simulation a 5×5 rod bundle with

mixing-vane grid using Star-CCM+.

The applicability of this model to

simulate fuel rod bundles has been

tested and validated by Westinghouse

in their extensive research (Smith et

al., [24]).

Considering the established practice

and computational time required

as discussed above, it can be concluded

that RNG k-ε model will be

sufficient in modeling turbulence for

flow through a rod bundle. However,

in this study, realizable k-ε model

(Shih et al., [25]) has been adopted

for turbulence modeling inside a

square array subchannel since it has

been statistically proved that this

model provides the best performance

among all the k-ε model versions for

separated flows and flows with complex

secondary flow features [26].

The term “realizable” means

that the model satisfies certain mathematical

constraints on the Reynolds

stresses, consistent with the physics

of turbulent flows. Neither the standard

k-ε nor the RNG k-ε model is

realizable.

The modeled transport equation

for k and ε in the realizable k-ε model

are presented by Equation (13) and

Equation (14) respectively:

(13)

and

where,

P/D = 1.35

Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid

(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %

6×10 5 5.826 5.926 6.215 6.843

5.098×10 5 4.950 5.035 5.280 5.814

4×10 5 3.884 3.951 4.143 4.562

3×10 5 2.913 2.963 3.108 3.422

(14)

(15)

(16)

In above equations, G k represents

the generation of turbulence kinetic

energy due to mean velocity gradients,

G b is the generation of turbulence

kinetic energy due to buoyancy, Y M is

the contribution of fluctuating dilatation

in compressible turbulence to

the overall dissipation rate, C 2 and C 1ε

are constants, σ k and σ ε are the

turbulent Prandtl numbers for k and ε

respectively, S k and S ε are user- defined

source terms.

3.4 Convergence of numerical

solution

Another central criteria that must be

satisfied in order to obtain proper

numerical solution is convergence.

The solver needs to be given adequate

iterations so that the problem is converged

and a solution can be treated

as converged if the following criteria

are satisfied [26]:

• The solution no longer changes

with subsequent iterations

• Overall mass, momentum, energy

and scalar balance are achieved

• All equations (momentum, energy

etc.) are obeyed in all cells to a

specific tolerance

RESEARCH AND INNOVATION 251

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RESEARCH AND INNOVATION 252

In the present study, residuals for

continuity, X & Y-momentum, Z-

momentum and turbulence kinetic

energy are decreased respectively to

an order of 10 -2 , 10 -5 , 10 -2 and 10 -4

| | Fig. 3.

Distribution of wall y + values in case of pure water

with Re=6×10 5 (P/D =1.35)

| | Fig. 2.

Convergence of mass flow averaged temperature at outlet (P/D = 1.35) for pure water at corresponding

inlet Re = 6×10 5 .

after 30,000 iterations and also a

monitor is created to check how values

for mass flow averaged temperature at

outlet is converging and it is observed

that after 30,000 iterations these

values do not change significantly

with further iterations. A typical plot

of mass flow averaged temperature at

outlet for pure water at inlet Re =

6×10 5 is shown in Figure 2.

3.5 Wall y + values

The accurate calculations of y + value

in the near-wall region, which is a

measure of non-dimensional distance

from the wall to the first mesh node

(based on local cell fluid velocity), are

of paramount importance to the success

of any simulation. In order to use

a wall function approach properly for

a particular turbulence model with

confidence, the y+ values should be

within a certain range.

In the present study, standard wall

function is used in conjunction with

realizable k-ε model and high-y + wall

treatment in which the near-wall cell

centroid are anticipated to be placed

in the log-law region with a value

30 ≤ y + ≤ 100. Results of performed

simulations demonstrate that the

wall y + values for different cases are

within this specified range. A pictorial

representation of wall y + in case

of pure water with Re = 6×10 5

(P/D = 1.35) is shown in Figure 3.

4 Code validation

4.1 Mesh convergence test

Since the accuracy of finite volume

method is directly related to the

quality of discretization used, it is

instrumental to select an optimized

mesh size that will take into account

both resolution of mesh structure and

as well as computational time and

cost.

In the present study, different

mesh settings are selected as presented

in Table 3 and values of

numerically obtained Nu are compared

against an existing correlation

for square array subchannel and for

pure water as presented by Equation

(17) through Equation (19) to check

mesh convergence for computational

domain with P/D =1.35. Results are

plotted in Figure 4 which clearly

states that a mesh setting with base

size 0.7 mm, no. of prism layer 2,

prism layer thickness 0.3mm and

prism layer stretching 3.7 will be

sufficient to produce Nu within

reasonable deviation compared to

the theoretical prediction made by

correlation.

Nu = ψ(Nu ∞ ) c.t. (17)

where,

(Nu ∞ ) c.t. = 0.023 Re 0.8 PR 0.4 (18)

for square array with 1.05 ≤ P/D ≤

1.9 and for pure water, Presser [27]

suggested:

(19)

Base Size

(mm)

No.

Prism Layers

Stretching Thickness

(mm)

Nu

(Star-CCM+)

Nu

(Presser)

Deviation

(%)

0.5 5 1.5 0.7 742.940 -35.051

0.6 4 1.5 0.5 862.627 -16.313

0.7 3 3.8 0.4 933.92 1003.35 -7.434

0.6 2 3.7 0.3 972.102 -3.214

0.7 2 3.7 0.3 1010.57 0.714

| | Fig. 4.

Mesh convergence test with different mesh settings.

| | Tab. 3.

Different mesh settings used to check mesh convergence.

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| | Fig. 5.

Validation of numerical model against correlation for P/D =1.25.

4.2 Validation of numerical

model

Since the ultimate test of any numerical

simulation is the validation of

results against well-known experimental

data, the model under consideration

in the present study has

been validated against correlation of

Presser for square array and pure

water as presented by Equation (17)

through Equation (19). Results are

plotted in Figure 5 and Figure 6

which demonstrates that there is

an excellent agreement between

numerical data and theoretical

prediction for the specified range of

inlet Re.

4.3 Validation of turbulence

model for nanofluid

Despite in the present study it is

assumed that nanofluid would behave

as a single-phase homogeneous fluid

and hence, all of the general conservation

equations of mass, momentum

and energy can directly be applied in

case of nanofluid, however, a successful

comparison of numerical Nu obtained

realizable k-ε model has been

carried out against both empirical

correlation and experimental data of

Pak & Cho [1] for turbulent flow

inside a round pipe of inside diameter

10.66 mm using alumina nanofluid

(φ=2.78%) as coolant for inlet Re

spanning from 5.03×10 4 to 1.48×10 4 .

The results are plotted in Figure 7

which clearly delineates that this

model can perform quite satisfactorily

with nanofluids.

5 Numerical results

and discussion

5.1 Temperature

Temperature profile along the centerline

of subchannel (P/D =1.25) for

different coolants at inlet Re = 6×10 5

are illustrated in Figure 8 from which

it is clear that there is a steady increase

in the coolant temperature due to absorption

of heat while flowing through

the subchannel and bulk temperature

of nanofluid is decreased with the increasing

particle volume concentration.

Numerically obtained fluid average

temperature (in case

of pure water at P/D =1.25 and

inlet Re = 6×10 5 ) at different axial

locations within the subchannel is

compared against the theoretical

predictions from energy balance

according to equation (20) [28] and

results are tabulated in Table 4.

(20)

The analogy shows that maximum

deviation between numerically obtained

axial temperature and theoretical

prediction is less than 0.6%.

5.2 Velocity

Development of axial velocity along

the centerline of subchannel (P/D

| | Fig. 6.

Validation of numerical model against correlation for P/D =1.35.

=1.25) for different coolants at inlet

Re = 6×10 5 is presented in Figure 9

which clearly states that fullydeveloped

velocity profile occurs

approximately after z=0.3 m and if

the current models are implemented

to evaluate physical properties of

nanofluid, development of velocity

| | Fig. 7.

Validation of turbulence model against Pak & Cho’s correlation.

| | Fig. 8.

Temperature along centerline of subchannel at Re = 6×10 5 .

RESEARCH AND INNOVATION 253

Axial Position

(m)

Average Bulk Fluid Temperature T m (K) %

of Deviation

Start-CCM+ Energy Balance

0 569 569 0.000

0.15 569.2431 570.6885 0.2532

0.30 570.1277 572.3771 0.3929

0.45 571.2205 574.0656 0.4956

0.60 572.4116 575.7542 0.5805

| | Tab. 4.

Comparison of numerically obtained axial temperature against theoretical predictions for pure water

(P/D =1.25 and inlet Re = 6×10 5 ).

| | Fig. 9.

Velocity along centerline of subchannel at Re = 6×10 5 .

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RESEARCH AND INNOVATION 254

| | Fig. 10.

Pressure along centerline of subchannel at Re = 6×10 5 .

profile is not affected by the inclusion

of nanoparticles. From Figure 9, it can

also be seen that there is an increase in

the velocity magnitude due to growth

of hydrodynamic boundary layer as

coolant flows from inlet towards

outlet. The inclusion of higher volume

concentration of nanoparticles also

augments the magnitude of axial

velocity as seen in Figure 9. It can be

explained from the fact that since

with the rise of volume concentration

the viscosity of the nanofluid is also

aggrandized, hence to a maintain

constant value of Reynolds number Re

at the inlet of the channel, velocity

magnitude should be increased too

according to equation (21) if the other

properties remain constant:

(21)

5.3 Pressure

A plot of static pressure along the

centerline of the subchannel (P/D

=1.25) for different coolants at inlet

Re = 6×10 5 is shown in Figure 10

which depicts that there is an increase

in axial pressure with the inclusion of

nanoparticles which is expected due

to higher viscosity and density as the

particle volume concentration is increased.

5.4 Nu and h Constant Inlet Re

Convective heat transfer is studied

with Star-CCM+ for pure water and

different concentrations of alumina

nanofluid according to Equation (22)

and Equation (23) respectively. Values

of Nu are evaluated at the outlet of the

subchannel to assure fully-developed

turbulent flow condition.

(22)

(23)

where, q '' is the constant heat flux

(W/m 2 ), k is thermal conductivity

(W/m 2 .K), D h is hydraulic diameter

(m), and T w and T m are wall and mean

bulk fluid temperature (K) respectively.

Numerical results of Nu and h for

subchannel with different pitch-todiameter

(P/D) ratio are presented

through Figure 11 to Figure 14

respectively and percentage of convective

heat transfer increment for

different nanofluid coolants are

documented in Table 5.

From the results, it is obvious that

the convective heat transfer coefficient

is remarkably increased with the

increment of nanoparticle volume

concentration and in case of 3.0 %

volume concentration, convective

heat transfer is increased above

22.0 % compared to pure water.

5.5 Comparison of Numerical

Results against Correlations

In case of nanofluid with volume

concentration, φ =3.0% numerical

results for Nu are compared against

two well cited correlations of Pak &

Cho [1] and Maïga et al. [3] as shown

in Figure 15 (a) & (b) and an attempt

has been made whether results of

present study can be represented by

either of these two correlations.

The results revealed that Pak

and Cho correlation severely underestimates

the numerical results for

Nu in subchannel and deviation lies

between 17 to 22 percent subject to

inlet Re and P/D.

Regarding correlation of Maïga

et al., it shows better approximation

compared to correlation of Pak & Cho.

Nevertheless, this correlation underestimates

the numerical results for the

| | Fig. 11.

Comparison of Nu for different coolants in subchannel (P/D 1.25).

| | Fig. 12.

Comparison of Nu for different coolants in subchannel (P/D 1.35).

| | Fig. 13.

Comparison of h for different coolants in subchannel (P/D 1.25).

| | Fig. 14.

Comparison of h for different coolants in subchannel (P/D 1.35).

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range 5×10 5 ≤ Re ≤ 6×10 5 and overestimates

for 3×10 5 ≤ Re ≤ 4×10 5

and deviations are between 0.54 and

6.66 percent depending on inlet Re

and P/D.

5.6 Pressure drop

While nanofluid enhances the convective

heat transfer, the fluid itself

P/D = 1.25

Inlet Re Increment of h (%)

φ = 0.5 % φ = 1.5 % φ = 3.0 %

6×10 5 2.75 9.62 22.46

5.098×10 5 2.75 9.58 22.37

4×10 5 2.72 9.51 22.16

3×10 5 2.74 9.42 21.89

| | Tab. 5.

Heat transfer increment (%) for different nanofluid coolants.

also gets heavier compared to pure

water. Hence, it is of utmost importance

to determine the amount of

pressure drop for the effective application

of nanofluid coolant in nuclear

reactors since it is directly related to

the pumping power required. In this

study, pressure drop along the center

line of the subchannel is evaluated for

different coolants and results are presented

in Figure 16 (a) & (b). Percentage

of pressure drop increment is

documented in Table 6.

The results shows that pressure

drop is significantly increased with

the augmentation of particle volume

concentration which in turn increases

the pumping power. For nanofluid

P/D = 1.55

Inlet Re Increment of h (%)

φ = 0.5 % φ = 1.5 % φ = 3.0 %

6×10 5 2.72 9.56 22.35

5.098×10 5 2.72 9.51 22.26

4×10 5 2.71 9.44 22.01

3×10 5 2.69 9.40 21.87

RESEARCH AND INNOVATION 255

(a) P/D = 1.25

| | Fig. 15.

Comparison of numerical Nu against different correlations.

(b) P/D = 1.35

(a) P/D = 1.25

| | Fig. 16.

Comparison of pressure drop for different coolant.

(b) P/D = 1.35

P/D = 1.25

Inlet Re Increment of ∆p (%)

φ = 0.5 % φ = 1.5 % φ = 3.0 %

6×10 5 6.22 21.53 56.60

5.098×10 5 5.82 21.17 56.62

4×10 5 5.79 21.79 56.02

3×10 5 5.24 21.65 55.83

P/D = 1.35

Inlet Re Increment of ∆p (%)

φ = 0.5 % φ = 1.5 % φ = 3.0 %

6×10 5 5.82 20.94 56.37

5.098×10 5 5.74 21.29 56.08

4×10 5 5.46 20.90 55.10

3×10 5 5.62 20.88 55.82

| | Tab. 6.

Pressure drop increment (%) for different nanofluid coolants.

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RESEARCH AND INNOVATION 256

with φ=3.0%, pressure drop increment

is about 56% higher compared

to that of pure water.

However, the typical nanoparticle

loading in PWR coolant should be

less than 1.0 vol %. At such lower

con centration, nanofluid properties

are almost similar to that of pure

water and the rise in viscosity as well

as pressure drop will be negligible too.

The present study also portrays that

pressure drop is approximately 20 %

at 1.5 vol. % of nanoparticle concentration

which can also be treated as

tolerable.

The convective heat transfer coefficient

at such low concentration of

nanofluid is yet to be improved due to

higher turbulence produced near the

grid spacers by the presence of nanoparticles

in the base fluid. Since it is

quite difficult to take into account

such effects in numerical simulation,

further experimental investigation is

required for quantification of heat

transfer increment aroused from the

presence of nanoparticles near the

spacer grids.

6 Proposed new

correction factor

Finally, a multiple regression analysis

is performed with numerical results to

propose a new correction factor, β for

the existing correlation of square

array subchannel with pure water as

suggested by Presser [27] so that Nu

for nanofluid coolant can be approximated

in such geometry. Based on

regression results, β can be expressed

as follows:

β = 1 + 0.0247ϕ 1.39 (24)

Nu for nanofluid can be calculated as

follows:

Nu nf = β*(Nu Presser ) Water (25)

The validity of above correlation is for

3×10 5 ≤ Re ≤ 6×10 5 ; 0.847 ≤ Pr ≤

1.011; 1.25 ≤ P/D ≤ 1.35 and 0.5% ≤

φ ≤ 3.0% in case of square array

subchannel.

7 Chemical and physical

stability of nanofluid

Albeit nnanofluid can readily boost

the heat transfer capability of PWR

coolant, there is still no satisfactory

explanation proposed regarding the

prevention of clustering in nanoparticle

suspensions. Agglomeration

in nanofluids containing oxide nanoparticles

can be reduced remarkably

by adjusting the pH to form electric

changes on particle surface so that

they repel each other [29]. However,

the typical values of pH should be

such that nanofluid itself becomes not

corrosive and it should be agreeable

with same allowable pH range of

nuclear reactor, since altering the

PWR coolant chemistry is not a viable

option. Besides, use of surfactants are

also not recommended since it may

undergo severe radiolysis inside the

reactor core during operation.

Hence, issues concerning chemical

and physical stabilities of nanofluid

has yet to be resolved prior to utilizing

nanofluid as a promising coolant in

PWRs to achieve both extended life

time of associated equipment and

higher thermal efficiency.

8 Conclusion

Thermo- and hydrodynamic characteristics

of water/alumina nanofluid

have been studied in a square array

subchannel featuring the pitch-todiameter

ratios of 1.25 and 1.35 under

the steady-state, incompressible,

single- phase turbulent flow condition.

Numerical results have been compared

against correlations in the

literature and the following conclusions

can be drawn.

• Convective heat transfer is increased

with increasing volume

concentration of water/alumina

nanofluid given the inlet Reynolds

number.

• The convective heat transfer increment

of nanofluid is obtained at

the expense of increased pressure

drop and hence, larger pumping

power is required. Therefore,

nano fluid as PWR coolant can be

only be implemented in reality if

the replacement of reactor coolant

pump is a feasible option compared

to higher power gained from

increased nanofluid heat transfer.

Acknowledgements

This work was supported by the

National Research Foundation of Korea

(NRF) grant funded by the Korean

Government (MSIP) under Grant No.

2008-0061900 and partly supported

by the Brain Korea 21 Plus under

Grant No. 21A20130012821.

Nomenclature

∆p Pressure Drop Pa

ρ Density kg/m 3

v Flow Velocity m/s

f Friction Factor -

L Length of Flow Channel m

le Entrance Length m

EI Entrance Length Number -

Dh Hydraulic Diameter m

μ Dynamic Viscosity N.s/m 2

Re Reynolds Number -

Nu Nusselt Number -

Pr Prandtl Number -

Pe Peclet Number -

h

Convective Heat Transfer

CoefficientW/m 2 .K

k Thermal Conductivity W/m.K

C p Specific Heat J/kg.K

T m Bulk Temperature of Fluid K

T w

Surface Temperature

of Heater Rod

P Rod Pitch m

D Rod Diameter m

Q Total Heat Input W

q” Heat Flux W/m 2

φ

ṁ Mass Flow Rate kg/sec

Subscript

nf

bf

P

Volume Concentration

of Nanoparticles %

Nanofluid

Basefluid

Particle

References

1. Pak BC, Cho YI. Hydrodynamic and heat

transfer study of dispersed fluids with

submicron metallic oxide particles.

Experimental Heat Transfer an International

Journal. 1998;11:151-70.

2. Xuan Y, Li Q. Investigation on

convective heat transfer and flow

features of nanofluids. Journal of Heat

transfer. 2003;125:151-5.

3. El Bécaye Maïga S, Tam Nguyen C,

Galanis N, Roy G, Maré T, Coqueux M.

Heat transfer enhancement in turbulent

tube flow using Al2O3 nanoparticle

suspension. International Journal of

Numerical Methods for Heat & Fluid

Flow. 2006;16:275-92.

4. Asirvatham LG, Vishal N, Gangatharan

SK, Lal DM. Experimental study on

forced convective heat transfer with low

volume fraction of CuO/water nanofluid.

Energies. 2009;2:97-119.

5. Wu X, Trupp AC. Experimental study on

the unusual turbulence intensity

distributions in rod-to-wall gap regions.

Experimental Thermal and Fluid

Science. 1993;6:360-70.

6. Xuan Y, Roetzel W. Conceptions for

heat transfer correlation of nanofluids.

International Journal of heat and Mass

transfer. 2000;43:3701-7.

7. Maïga SEB, Palm SJ, Nguyen CT, Roy G,

Galanis N. Heat transfer enhancement

by using nanofluids in forced convection

flows. International Journal of

Heat and Fluid Flow. 2005;26:530-46.

8. Bianco V, Chiacchio F, Manca O,

Nardini S. Numerical investigation of

nanofluids forced convection in circular

tubes. Applied Thermal Engineering.

2009;29:3632-42.

K

Research and Innovation

Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh

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atw Vol. 63 (2018) | Issue 4 ı April

9. Maïga S, Nguyen CT, Galanis N, Roy G,

Heat transfer enhancement in forced

convection laminar tube flow by using

nanofluids. 2004: Publisher.

10. Brinkman H. The viscosity of

concentrated suspensions and

solutions. The Journal of Chemical

Physics. 1952;20:571-.

11. Batchelor G. The effect of Brownian

motion on the bulk stress in a

suspension of spherical particles. Journal

of Fluid Mechanics. 1977;83:97-117.

12. Lee S, Choi S-S, Li S, and, Eastman J.

Measuring thermal conductivity of fluids

containing oxide nanoparticles. Journal

of Heat Transfer. 1999;121:280-9.

13. Masuda H, Ebata A, Teramae K.

Alteration of thermal conductivity and

viscosity of liquid by dispersing ultrafine

particles. Dispersion of Al 2 O 3 , SiO 2

and TiO 2 ultra-fine particles. 1993.

14. Wang X, Xu X, S. Choi SU. Thermal

conductivity of nanoparticle-fluid

mixture. Journal of thermophysics and

heat transfer. 1999;13:474-80.

15. Maïga SEB, Nguyen CT, Galanis N, Roy

G. Heat transfer behaviours of nanofluids

in a uniformly heated tube.

Superlattices and Microstructures.

2004;35:543-57.

16. Häfeli R. Fluid dynamic characterization

of single-and multiphase flow in

structured porous media: Master Thesis

ETH Zurich, 2010; 2010.

17. Yadigaroglu G, Andreani M, Dreier J,

Coddington P. Trends and needs in

experimentation and numerical simulation

for LWR safety. Nuclear Engineering

and Design. 2003;221:205-23.

18. Házi G. On turbulence models for rod

bundle flow computations. Annals of

Nuclear Energy. 2005;32:755-61.

19. Lee C, Choi Y. Comparison of thermohydraulic

performances of large scale

vortex flow (LSVF) and small scale vortex

flow (SSVF) mixing vanes in 17× 17

nuclear rod bundle. Nuclear Engineering

and Design. 2007;237:2322-31.

20. Liu CC, Ferng YM. Numerically simulating

the thermal-hydraulic characteristics

within the fuel rod bundle using CFD

methodology. Nuclear Engineering and

Design. 2010;240:3078-86.

21. Palandi SJ, Rahimi-Esbo M, Vazifeshenas

Y. Comparison of thermo- hydraulic

performance of nanofluids and mixing

vanes in a triangular fuel rod bundle.

Journal of the Brazilian Society of

Mechanical Sciences and Engineering.

2015;37:173-86.

22. Conner ME, Baglietto E, Elmahdi AM.

CFD methodology and validation for

single-phase flow in PWR fuel assemblies.

Nuclear Engineering and Design.

2010;240:2088-95.

23. Yakhot V, Orszag S, Thangam S, Gatski

T, Speziale C. Development of turbulence

models for shear flows by a

double expansion technique. Physics of

Fluids A: Fluid Dynamics (1989-1993).

1992;4:1510-20.

24. Smith III L, Conner M, Liu B, Dzodzo M,

Paramonov D, Beasley D, Langford H,

Holloway M. Benchmarking computational

fluid dynamics for application to

PWR fuel. Proceedings of ICONE.

2002;10.

25. Shih T-H, Liou W, Shabbir A, Yang Z,

Zhu J. A new k-epsilon eddy viscosity

model for high Reynolds number

turbulent flows: Model development

and validation. 1994.

26. Introduction to ANSYS FLUENT:

Customer Training Material, Release 13,

December 2010. .

27. Presser KH. Wärmeübergang und

Druckverlust an Reaktorbrennelementen

in Form längsdurchströmter

Rundstabbündel: Kernforschungsanlage,

Zentralbibliothek; 1967.

28. Azari A, Kalbasi M, Derakhshandeh M,

Rahimi M. An Experimental Study on

Nanofluids Convective Heat Transfer

Through a Straight Tube under Constant

Heat Flux. Chinese Journal of Chemical

Engineering. 2013;21:1082-8.

29. Buongiorno J, Hu L-w. Nanofluids for

Enhanced Economics and Safety of

Nuclear Reactors [Published 2007

[cited November 29].

Authors

Jubair Ahmed Shamim

Department of Nuclear

Engineering

Seoul National University

Seoul 08826, ROK

Kune Yull Suh

Seoul National University

1 Gwanak Ro, Gwanak Gu

Seoul 08826, ROK

257

KTG INSIDE

Inside

KTG-Vorstandswahl 2018

Liebe Mitglieder, gemäß unserer Satzung stehen mit Wirksamkeit zur Mitgliederversammlung am 28. Mai 2018

Wahlen zum Vorstand der KTG an. Die Wahlunterlagen mit dem vom Beirat aufgestellten funktionsbezogenen

Gesamtwahlvorschlag gehen Ihnen in den nächsten Wochen auf postalischem Wege zu. Wie gewohnt möchten sich die

Kandidaten zur KTG-Vorstandswahl 2018 Ihnen nachstehend (noch einmal) vorstellen.

Herzlichst,

Ihre KTG-Geschäftsstelle

Kandidaten

Frank Apel

Dipl.-Ing. (54), Heidelberg

Zur Person

Seit Februar 2017 Geschäftsführer bei der

Kraftanlagen Heidelberg GmbH. Vorher

leitend für den Bereich „Back-End“ bei der

AREVA (heute ORANO) und davor Leiter von

Vertrieb und Marketing der AREVA (heute

FRAMATOME) für die Region „Zentral- und Nord europa“. Darüber

hinaus verantwortlich in der Internationalen Vertriebsorganisation

für die weltweite Zusammenarbeit mit AREVAs wichtigsten Kunden

(Key Accounts). Mehr als 30 Jahre Erfahrung in der Energiewirtschaft

und in dieser Zeit in ver schiedenen Führungspositionen bei Siemens

und den Nachfolgefirmen tätig. Dabei unter anderem für alle

Service- und Wartungsarbeiten von AREVA in den deutschen

Kernkraftwerken verantwortlich. Abschluss als Diplombauingenieur

für Kernkraftwerke der Moskauer Staat lichen Universität für

Bauingenieurwesen.

Zur Wahl als Vorsitzender der KTG

In unserer letzten Vorstandsklausur haben wir erneut sehr intensiv

das Thema der Mitgliedschaft in unserem Verband diskutiert. Die

Frage, was die Mitglieder der KTG verbindet, beantworten wir

KTG Inside

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atw Vol. 63 (2018) | Issue 4 ı April

258

KTG INSIDE

mit der „Faszination Kerntechnik“. Unsere Mitglieder, das sind

aktive oder ehemalige Mitarbeiter bei den Betreibern, Herstellern,

Behörden und Gutachtern, der Lehre und der Forschung oder

Menschen in anderen Berufsgruppen, die die Kerntechnik spannend

finden und für die der Austausch in unserer (kerntechnischen)

Gesellschaft wichtig ist. Sie werden überrascht sein: Wenn Sie

nach „Faszination Kerntechnik“ googeln, ist der erste Treffer:

Faszination Kerntechnik |

Kerntechnische Gesellschaft e.V.

https://www.ktg.org/ktg/faszination-kerntechnik/

Darüber können wir uns freuen, darauf können wir stolz sein; in

puncto Kommunikation sind wir besser geworden, das zeigt auch

unsere neuer Internet-Auftritt. Wir Kerntechniker haben in Deutschland

– dem Land der Bedenkenträger – (eigentlich) häufig Grund zur

Freude:

• Die am Netz befindlichen deutschen Kernkraftwerke erzeugen im

sicheren Leistungsbetrieb umweltfreundlichen Strom, eine

Verantwortung, die wir bis zum letzten Tag des Jahres 2022 und

darüber hinaus haben.

• Der (leider größtenteils politische verordnete) Rückbau geht

voran: Im Februar dieses Jahres hat Unterweser als 5. Anlage

in der „Post-Fukushima-Ära“ die 1. Stilllegungs- und Abbaugenehmigung

erhalten.

• In vielen stillgelegten Anlagen wurden und werden die

abgebrannten Brennelemente aus den Lagerbecken in

Castoren geladen und in die standortnahen Zwischenlager

verbracht.

• Im letzten Jahr hat die Deutsche Rechtsprechung „ideologiefrei“

z. B. Urteile zur „Durchsetzung von Castor-Transporten auf dem

Neckar“ oder die „Nichtrechtmäßigkeit der Brennelement-Steuer“

verkündet.

• In unserem Nachbarland Schweiz darf das Kraftwerk Beznau den

Block 1 nach einer dreijährigen Betriebsunterbrechung wieder

in Betrieb nehmen. Der Betreiber Axpo konnte nachweisen, dass

die Einschlüsse im Stahl des Reaktordruckbehälters keinen

negativen Einfluss auf die Sicherheit haben. Die Schweizer

Gutachter und Behörden haben – ideologiefrei und gestützt auf

die Meinung internationaler Experten – die entsprechenden

technischen Nachweise geprüft und akzeptiert.

An diesen Entwicklungen haben auch Sie mit Ihren hervorragenden

kerntechnischen Kompetenzen beigetragen. Die deutschen Kerntechnikerinnen

und Kerntechniker verfügen über ein weltweit

anerkanntes und nachgefragtes Know-how, was wir in Deutschland

erhalten wollen, um unter anderem:

• den verbleibenden Leistungsbetrieb, den Nachbetrieb, die

Stilllegung und den Rückbau deutscher Anlagen sicherzustellen

und die Entsorgungsfrage nachhaltig zu lösen,

• das Exportgeschäft deutscher Anbieter und Dienstleister zu

sichern,

• nationale und internationale Sicherheitsbewertungen durchführen

zu können und

• auch in Zukunft den Beitrag deutscher Standards und Innovationen

zu internationalen Entwicklungen für neue Technologien

erhalten zu können.

Das ist – nach wie vor – die Perspektive, die ich als KTG-Vorstand

und Vorsitzender in die kerntechnische Gesellschaft einbringen

möchte. Die KTG muss als Verband der Beschäftigten der Kerntechnik

eine führende Rolle spielen, diesen Kompetenzerhalt zu

sichern. Neben wirtschaftlich erfolgreichen und weiterhin innovativen

Unternehmen brauchen wir auch eine leistungsfähige

wissenschaftliche Landschaft bei den Forschungseinrichtungen und

Universitäten. Dies alles kann bei einem absehbar schrumpfenden

Heimatmarkt durch Wachstum im Ausland gewährleistet werden,

das sowohl die wirtschaftliche Zukunft der Unternehmen als auch

die persönliche Perspektive der Kolleginnen und Kollegen unserer

Branche sichert.

Erwin Fischer

Dr.-Ing. (61), Rodenberg

Zur Person

Nach einer praktischen Ausbildung zum

Maschinenschlosser Studienabschlüsse im

zweiten Bildungsweg als Ingenieur (grad.)

für Maschinentechnik nach einem Fachhochschulstudium

und als Dipl.-Ing. für

Maschinenbau mit Vertiefungsrichtung Energie technik an der Ruhr-

Universität (RUB), Bochum. Promotion, ebenfalls an der RUB, am

Lehrstuhl für Reaktortechnik/Neue und Nukleare Energie systeme im

Fachgebiet Reaktortechnik 1991.

Zur Wahl als Schatzmeister

Seit meiner Promotion bin ich bei der PreussenElektra GmbH und

ihren Vorläuferunternehmen im Kernenergiebereich in der Nuklearen

Technik und dem Betrieb tätig. Meine Aufgaben bezogen sich auf

den Bau, Betrieb und Rückbau von Kernkraftwerken. Während

der nunmehr 27 Jahre Tätigkeit für PreussenElektra habe ich

Aufgaben in den Kernkraftwerken und der Zentralorganisation wahrgenommen.

Seit 2014 führe ich das Geschäftsführungsressort Technik

und Betrieb. Ich war 13 Jahre im KTA tätig und 5 Jahre Mitglied

der deutschen Reaktorsicherheitskommission. Seit 4 Jahren engagiere

ich mich als Governor bei der WANO – World Association for

Nuclear Operators – weltweit.

Während meines bisherigen beruflichen Werdegangs war und

ist der sichere, umweltverträgliche und wirtschaftliche Betrieb der

Kernkraftwerke mein prioritäres Anliegen. Die Kernkraft hat mich in

meinem ganzen Berufsleben fasziniert und die Faszination hält trotz

aller Rückschläge und der teilweise schwierigen Randbedingungen

für die Kernenergie in Deutschland an.

Seit 1991 bin ich Mitglied in der KTG und nunmehr schon seit

8 Jahren im Vorstand, jetzt als Schatzmeister. Ziel meiner erneuten

Kandidatur ist, die KTG als Interessengemeinschaft aller in der

Kerntechnik Tätigen und von ihr faszinierten Mitgliedern mit meinem

Wissen und meiner beruflichen Erfahrung zu unterstützen sowie die

Wissensübertragung und den Erfahrungsaustausch zu erhalten.

Wie politisch gewollt, sollten wir Kerntechniker den sicheren Betrieb

bis zum Laufzeitende und den Rückbau unserer Kernkraftwerke

in Deutschland mit Ehre abschließen. Kein Grund mit Blick auf das

Erreichte der letzten 50 Jahre nicht stolz sein zu dürfen!

Jörg Starflinger

Prof. Dr.-Ing. (51), Stuttgart

Zur Person

Nach dem Studium des Maschinenbaus an

der Ruhr-Universität Bochum (RUB) mit

Schwerpunkt Energietechnik Promotion

im Jahr 1997 am Lehrstuhl für Nukleare

und neue Energiesysteme der RUB, Prof.

Dr.-Ing. H. Unger. 1998 Wechsel als Nachwuchswissenschaftler zum

Forschungszentrum Karlsruhe, heute Karlsruhe Institut für Technologie.

Themenschwerpunkte: Wasserstofferzeugung bei schweren

Unfällen in Leichtwasserreaktoren und Kreislaufsimulation von

innovativen Reaktorkonzepten. 2006 Leiter der Gruppe „Kraftwerkstechnik“

am Institut für Kern- und Energietechnik (IKET), Prof. Dr.-Ing.

T. Schulenberg, in der innovative Kernkraftwerkskonzepte mit überkritischem

Wasser von mehreren Doktoranden untersucht wurden.

2010 Ruf an die Universität Stuttgart zum ordentlichen Professor des

Lehrstuhls für Kerntechnik und Reaktorsicherheit und Leiter des

Instituts für Kernenergetik und Energiesysteme (IKE). Neben der

Lehre im Bereich Kerntechnik Schwerpunkte in der Reaktorsicherheitsforschung,

z.B. in der Modellentwicklung zur Beschreibung der

späten Phase von Kernschmelzunfällen in Leichtwasserreaktoren

KTG Inside

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atw Vol. 63 (2018) | Issue 4 ı April

und auf dem Gebiet innovativer Sicherheitssysteme, z.B. der passiven

Lagerbeckenkühlung mit Wärmerohren (Heat pipes) und nachrüstbaren

Nachwärmeabfuhrsystemen mit überkritischem CO 2 als

Arbeitsmittel.

Zur Wahl als Vorstandsmitglied

Ich engagiere mich in der KTG auf dem Gebiet des Kompetenzerhalts

und der Kompetenzförderung. Den von Dr. Wolfgang Steinwarz ins

Leben gerufenen, sehr erfolgreichen Workshop „Kompetenzerhalt in

der Kerntechnik“ habe ich verantwortlich übernommen und möchte

ihn in den kommenden Jahren weiterführen. Dr. Steinwarz steht uns

auch als Ruheständler dankenswerterweise als Jurymitglied weiter

zur Seite. Die Umbenennung in „Young Scientists Workshop“ soll eine

Öffnung zu kerntechnisch verwandten Forschungsthemen, beispielsweise

„Kerntechnik und Gesellschaft“, symbolisieren.

Durch meine Mitarbeit im KTG-Vorstand als Vorstandsmitglied

möchte ich einen Strategieentwicklungsprozess anstoßen, der

mittelfristig eine genügende Anzahl an jungen hochqualifizierten

und motivierten Personen für die zukünftigen spannenden und

herausfordernden nationalen und internationalen kerntechnischen

Aufgaben sicherstellt. Für den Kompetenzerhalt und die Nachwuchsförderung

bieten die KTG und Ihre Mitglieder sowie unsere Tagung

„Annual Meeting on Nuclear Technology“ die ideale Plattform.

Walter Tromm

Dr.-Ing. (58), Stutensee

Zur Person

Maschinenbaustudium an der Uni (TH)

Karlsruhe mit dem Studienschwerpunkt

Kerntechnik und dort Promotion zum Thema

„Experimentelle Untersuchungen zum Nachweis

der langfristigen Kühlbarkeit von

Kernschmelzen“. Seit 1988 am damaligen Forschungszentrum

Karlsruhe, heute Karlsruher Institut für Technologie, angestellt und

schwerpunktmäßig mit Reaktorsicherheitsfragen bei auslegungsüberschreitenden

Störfällen beschäftigt. Von 1998 bis 1999 Gastwissenschaftler

am Europäischen Gemeinschaftsforschungszentrum

in Ispra (Italien) tätig.

Seit 2002 Programmbevollmächtigter in der Programmleitung

Nukleare Sicherheitsforschung des FZK bzw. heute Nukleare

Entsorgung, Sicherheit und Strahlenforschung des KIT; stellvertretender

Leiter seit 2007 wurde seit 2010 Programmleiter. 2014 im

geschäftsführenden Ausschuss des Bereichs Maschinenbau und

Elektrotechnik des KIT berufen. Seit 2015 darüber hinaus Sprecher

des vom KIT neu eingerichteten Kompetenzzentrums Rückbau und

seit 2017 Vorsitzender des Kompetenzverbundes Kerntechnik.

Tätig in nationalen und internationalen Gremien, bei der OECD/

NEA der deutsche Repräsentant des Nuclear Science Committee, bei

der IAEA in der Technical Working Group Light Water Reactors und

Mitglied im Governing Board der EU-SNETP Plattform. Weiterhin

innerhalb des VDI Vorsitzender des Fachausschusses Kraftwerkstechnik.

Seit 2016 Leiter des neu gegründeten Kompetenz-Cluster

Rückbau, der die Expertise im Rückbau mehrerer Länder zusammenführt.

Zur Wahl als stellvertretender Vorsitzender

Die Bundesregierung hat 2011 nach den Ereignissen in dem

Kernkraftwerk Fukushima Daii-chi in Japan entschieden, aus der

Stromproduktion mittels Kernkraft auszusteigen. In den nächsten

4 Jahren werden die letzten Kernkraftwerke in Deutschland

abgeschaltet. Diesen Ausstieg nach wie vor so sicher wie möglich

mitzugestalten ist eine der Aufgaben, die die in der deutschen

Kerntechnik arbeitenden Ingenieure und Naturwissenschaftler

haben. International und auf europäischer Ebene wird jedoch Kernenergie

langfristig weiterhin genutzt. Auch für den Industriestandort

Deutschland und für den Erhalt von Arbeitsplätzen ist der Export von

Komponenten für kerntechnische Anlagen nach wie vor bedeutsam.

Ebenfalls werden der Rückbau der Kernkraftwerke und die Endlagerfrage

die Gesellschaft noch über Jahrzehnte beschäftigen. Der

Ausstieg aus der Stromproduktion durch Kernenergie darf daher

nicht bedeuten, sich von den entsprechenden kerntechnischen

Kompetenzen in der Industrie, den Behörden und den Universitäten

und Forschungszentren zu verabschieden. In den Bereichen Reaktorsicherheit,

Rückbau, Endlagerung, Strahlenschutz und Krisenmanagement

sind diese Kompetenzen auch weiterhin gefragt. In

Europa stammen 27 % der Stromproduktion aus Kernkraftwerken.

Zur kompetenten Bewertung kerntechnischer Einrichtungen innerhalb

Europas und zur kritischen Begleitung internationaler Entwicklungen

sind eine enge Zusammenarbeit auf nationaler, europäischer

und internationaler Ebene unerlässlich. Deshalb sehe ich als eine

der Hauptaufgaben der KTG den Erhalt der kerntechnischen

Kompetenzen in allen genannten Bereichen.

259

KTG INSIDE

Herzlichen

Glückwunsch

April 2018

97 Jahre wird

2. Prof. Dr. Albert Ziegler, Karlsbad

87 Jahre werden

9. Dr. Klaus Penndorf, Geesthacht

11. Hubert Bairiot, Mol/B

19. Dr. Klaus Einfeld, Murnau

28. Dipl.-Ing. Rudolf Eberhart, Burgdorf

85 Jahre wird

6. Ing. Reinhard Faulhaber, Köln

84 Jahre wird

22. Dipl.-Ing. Gert Slopianka,

Gorxheimeral

83 Jahre werden

3. Dipl.-Psych. Georg Sieber,

München

5. Prof. Dr. Hans-Henning Hennies,

Karlsruhe

19. Dr. Ernst Müller, Rösrath

19. Dr. Gottfried Class,

Eggenstein-Leopoldshafen

21. Dipl.-Ing. Walter Jansing,

Bergisch Gladbach

30. Dr. Friedrich-Wilhelm Heuser,

Overath

82 Jahre werden

4. Helmut Kuhne, Neunkirchen

6. Dipl.-Ing. Hans Pirk, Rottach-Egern

10. Dipl.-Ing. Franz Stockschläder,

Bad Bentheim

11. Dipl.-Ing. Bernhard-F. Roth,

Eggenstein-Leopoldshafen

24. Dipl.-Ing. Horst Schott, Overath

81 Jahre werden

7. Dipl.-Ing. Helmut Adam, Neuenhagen

13. Dr. Martin Peehs, Bubenreuth

80 Jahre werden

4. Prof. Dr. Klaus Kühn, Clausthal- Zellerfeld

5. Dr. Hans Fuchs, Gelterkinden/CH

9. Dr. Carl Alexander Duckwitz, Alzenau

28. Prof. Dr. Georg-Friedrich Schultheiss,

Lüneburg

79 Jahre wird

8. Dr. Siegbert Storch, Aachen

78 Jahre wird

18. Dipl.-Ing. Norbert Granner,

Bergisch Gladbach

77 Jahre werden

17. Dipl.-Phys. Ernst Robinson, Gehrden

28. Dr. Ludwig Richter, Hasselroth

KTG Inside

---

atw Vol. 63 (2018) | Issue 4 ı April

260

NEWS

Wenn Sie keine

Erwähnung Ihres

Geburtstages in

der atw wünschen,

teilen Sie dies bitte

rechtzeitig der KTG-

Geschäftsstelle mit.

KTG Inside

Verantwortlich

für den Inhalt:

Die Autoren.

Lektorat:

Sibille Wingens,

Kerntechnische

Gesellschaft e. V.

(KTG)

Robert-Koch-Platz 4

10115 Berlin

T: +49 30 498555-50

F: +49 30 498555-51

E-Mail: s.wingens@

ktg.org

www.ktg.org

76 Jahre werden

9. Prof. Dr. Hans-Christoph Mehner,

Dresden

27. Dr. Dieter Sommer, Mosbach

27. Dr. Jürgen Wunschmann, Eggenstein

29. Dr. Klaus-Detlef Closs, Karlsruhe

75 Jahre werden

15. Dr. Werner Dander, Heppenheim

18. Dipl.-Betriebsw. Uwe Janßen,

Weinheim

18. Dipl.-Ing. Victor Luster, Bamberg

26. Ing. Helmut Schulz, Kürten

70 Jahre werden

6. Dr. Wolfgang Tietsch, Mannheim

9. Ing. Herbert Moryson, Essen

22. Dr. Heinz-Dietmar Maertens, Arnum

26. Dr. Rainer Heibel, Ness Neston/GB

27. Ulrich Wimmer, Erlangen

65 Jahre werden

10. Dipl.-Phys. Harold Rebohm, Berlin

24. Dipl.-Phys. Michael Beczkowiak,

Karben

60 Jahre werden

4. Dipl.-Ing. Holger Bröskamp,

Höhnhorst

4. Dipl.-Ing. (FH) Franz Xaver Pirzer,

Schwandorf

50 Jahre werden

16. Rainer Bezold, Dormitz

16. Dr. Matthias Messer, Tetbury/GB

30. Dr. Christian Raetzke, Leipzig

Mai 2018

94 Jahre wird

22. Prof. Dr. Fritz Thümmler, Karlsruhe

90 Jahre wird

10. Dr. Heinz Büchler, Sankt Augustin

89 Jahre wird

31. Dipl.-Ing. Werner-P. Kürsten,

Mannheim

88 Jahre wird

9. Dr. Hans-Jürgen Hantke, Kempten

85 Jahre werden

4. Dr. Klaus Wiendieck, Baden-Baden

25. Dr. Reinhold Mäule, Walheim

25. Georg von Klitzing, Bonn

84 Jahre werden

11. Dr. Eckhart Leischner, Rodenbach

14. Dr. Alexander Warrikoff, Frankfurt/M.

26. Dr. Günter Kußmaul, Manosque/F

83 Jahre werden

1. Dr. Willi Bermel, Jülich

8. Dipl.-Ing. Klaus Wegner, Hanau

22. Dr. Heinz Vollmer, Lampertheim

28. Dipl.-Ing. Anton Zimmermann,

Hamburg

29. Dipl.-Ing. Karlheinz Orth,

Marloffstein

82 Jahre werden

3. Ewald Jurisch, Erlangen

10. Dr. Peter Reinke, Röttenbach

18. Dipl.-Ing. Gerhard Lorenz, Bochum

81 Jahre werden

1. Prof. Dr. Dietrich Munz,

Graben-Neudorf

3. Dipl.-Ing. Harald Enderlein, Karlsruhe

6. Dr. Peter Strohbach, Mainaschaff

7. Prof. Dr. Werner Lutze,

Chevy Chase/USA

20. Dr. Norbert Krutzik, Frankfurt/M.

26. Dipl.-Ing. Rüdiger Müller, Heidelberg

27. Dr. Johannes Wolters, Düren

28. Dipl.-Ing. Heinz E. Häfner, Bruchsal

80 Jahre werden

12. Dr. Herbert Finnemann, Erlangen

13. Dipl.-Ing. Otto A. Besch, Geesthacht

13. Dr. Heinrich Werle,

Karlsdorf-Neuthard

16. Dr. Hans-Dieter Harig, Hannover

21. Dr. Hans Spenke, Bergisch Gladbach

79 Jahre werden

4. Dipl.-Ing. Norbert Albert, Ettlingen

5. Dr. Wolfgang Voigts, Linkenheim

27. Prof. Dr. Dietrich Kirsch

78 Jahre werden

11. Dr. Andreas Hölzler, Schwaig

15. Dipl.-Phys. Ludwig Aumüller,

Freigericht

18. Dr. Karl Schulte, Köln

24. Dipl.-Ing. Herbert Krinninger,

Bergisch Gladbach

77 Jahre werden

8. Prof. Dr. Helmut Alt, Aachen

12. Dipl.-Ing. Dieter Rohde, Mannheim

16. Dr. Jürgen Baier, Höchberg

76 Jahre werden

5. Hans-Bernd Maier, Aschaffenburg

9. Dr. Egbert Brandau, Alzenau

11. Dr. Erwin Lindauer, Köln

17. Dr. Heinz-Peter Holley, Forchheim

18. Dipl.-Ing. Josef Koban, Buckenhof

28. Dipl.-Ing. Wolf-Dieter Krebs,

Bubenreuth

75 Jahre werden

3. Dipl.-Ing. Hans Lettau, Effeltrich

14. Dr. Helmut-K. Hübner, Bruchsal

20. Dipl.-Ing. Dietmar Bittermann, Fürth

22. Dr. Wolfgang Schütz, Bruchsal

23. Dipl.-Ing. Max Heller, Uttenreuth

24. Dipl.-Ing. Rudolf Weh,

Stephanskirchen

27. Dr. Kurt Fischer, Erlangen

65 Jahre werden

2. Dipl.-Ing. Marc Winter, Veitshöchheim

3. Dipl.-Ing. Karl-Heinz Wiening,

Herzogenaurach

5. Michael Klein, Großenwörden

16. Ing. grad. Eckhard Raabe, Geiselbach

21. Dipl.-Ing. (FH) Reinhold Horstmann,

Erlangen

27. Dipl.-Ing. (FH) Ulrich Hudezeck,

Nürnberg

60 Jahre wird

23. Dr. Hans-Josef Zimmer, Steinfeld

50 Jahre werden

10. Dr. Astrid Petersen, Hamburg

20. Dipl.-Ing. (FH) Jürgen Bruder,

Gundremmingen

Die KTG gratuliert ihren Mitgliedern

sehr herzlich zum Geburtstag und wünscht ihnen weiterhin alles Gute!

Top

IAEA Expands International

Cooperation on Small,

Medium Sized or Modular

Nuclear Reactors

(iaea) The International Atomic

Energy Agency (IAEA) is launching an

effort to expand international cooperation

and coordination in the design,

development and deployment of

small, medium sized or modular

reactors (SMRs), among the most promising

emerging technologies in

nuclear power.

Significant advances have been

made on SMRs, some of which will use

pre-fabricated systems and components

to shorten construction schedules

and offer greater flexibility and

affordability than traditional nuclear

power plants. With some 50 SMR concepts

at various stages of development

around the world, the IAEA is forming

a Technical Working Group (TWG) to

guide its activities on SMRs and provide

a forum for Member States to

share infor mation and knowledge,

IAEA Deputy Director General Mikhail

Chudakov said.

“Innovation is crucial for nuclear

power to play a key role in de carbonising

the energy sector,” Chudakov,

who heads the IAEA Department of

Nuclear Energy, said at a conference

on SMRs in Prague on 15 February.

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atw Vol. 63 (2018) | Issue 4 ı April

“Many Member States that are

operating, expanding, introducing or

considering nuclear power are quite

keen on the development and

deployment of SMRs.”

Global interest in SMRs is growing.

SMRs have the potential to meet the

needs of a wide range of users and to

be low carbon replacements for ageing

fossil fuel fired power plants. They

also display enhanced safety features

and are suitable for non-electric applications,

such as cooling, heating and

water desalination. In addition, SMRs

offer options for remote regions with

less developed infrastructure and for

energy systems that combine nuclear

and alternative sources, including

renewables.

The first three advanced SMRs are

expected to begin commercial operation

in Argentina, China and the

Russian Federation between 2018 and

2020. SMR development is also well

advanced in about a dozen other

countries.

The TWG, comprising some 20

IAEA Member States and international

organizations, is scheduled to

meet for the first time on 23-26 April

at the IAEA’s headquarters in Vienna.

It is part of an expanding suite of

services the IAEA offers Member

States on this emerging nuclear power

technology. These include an SMR

computer simulation programme to

help educate and train nuclear professionals;

a methodology and related

IT tool for training in assessing the

reactor technology of different SMRs;

and the SMR Regulators’ Forum.

The forum, set up in 2015, enables

discussions among Member States and

other stakeholders to share SMR

regulatory knowledge and experience.

It contributes to enhancing safety by

identifying and resolving issues that

may challenge regulatory reviews of

SMRs and by facilitating robust and

thorough regulatory decisions.

Responding to requests from

Member States in Europe, the IAEA

recently launched a project to build

regional capacities for making knowledgeable

decisions on SMRs, including

technical assessments for SMRs

that are commercially available for

near term deployment. The two-year

project seeks to contribute to meeting

growing European demand for

flexible sources of electricity that do

not release greenhouse gases. Its first

meeting will be held on 13-15 March

at the IAEA in Vienna.

An expeditious deployment of

SMRs faces challenges, including the

need to develop a robust regulatory

| | IAEA Expands International Cooperation on Small, Medium Sized or Modular Nuclear Reactors.

framework, new codes and standards,

a resilient supply chain and human

resources. And although SMRs require

less upfront capital per unit, their

electricity generating cost will

probably be higher than that of large

reactors. Their competitiveness must

be weighed against alternatives and

be pursued through economies of

scale. Detailed technical information

on SMRs under construction or design

can be found at the IAEA’s Advanced

Reactor Information System.

“Realistically, we could expect the

first commercial SMR fleet to start

between 2025 and 2030,” said Hadid

Subki, Scientific Secretary of the TWG

and a Team Leader in SMR Technology

Development at the IAEA. “We

trust this new Technical Working

Group will help further the advancement

of SMR technology and guide

the Agency in its programmes and

projects in this field.”

| | (18791436), www.iaea.org

World

Poll Shows Local Residents

Support Poland’s Plans for

First Nuclear Plant

(nucnet) A poll carried out for Poland’s

PGE EJ1, the company in charge of the

country’s first nuclear power station

project, has shown that 67% of residents

in areas around the proposed

site in northern Poland support the

potential construction of a nuclear

power station in their region.

PGE said a poll was carried out

in November and December 2017

in three municipalities, Choczewo,

Gniewino, Krokowa, all close to

Poland’s Baltic coast in the northern

province of Pomerania.

According to PGE, local residents

indicated they are in favour of the

project because of the development

and job opportunities it could bring to

their regions. The poll showed 49% of

respondents expect cheaper electricity

to be one of the benefits from a

nuclear station, while 35 % expect

local infrastructure development.

In April 2017, PGE began environmental

and site selection surveys at

two locations – Lubiatowo-Kopalino in

the municipality of Choczewo and

Żarnowiec in the municipality of

Korkowa.

The studies aim to determine the

potential impact of the project on both

the environment and local residents.

An initial round of environmental

studies has already been carried out at

both locations.

The Polish government has not

made a final decision about the

country’s nuclear programme, with

the deadline being pushed back

several times. According to latest

reports, a decision is now expected in

mid-2018.

| | pgeej1.pl

SKB, Sweden: Two Statements

on the Spent Fuel Repository

(skb) The answer was a clear yes in

SSM’s statement to the Government

on SKB’s system for final disposal of

spent nuclear fuel. The Land and

Environment Court was also positive

| | Aerial photo of the planned site of the Spent Fuel Repository (centre)

at Forsmark. The picture is a photomontage. Illustration: Phosworks.

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*)

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

1)

Refueling

2)

Inspection

3)

Repair

4)

Stretch-out-operation

5)

Stretch-in-operation

6)

Hereof traction supply

7)

Incl. steam supply

8)

New nominal

capacity since

January 2016

9)

Data for the Leibstadt

(CH) NPP will

be published in a

further issue of atw

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB

in several important respects but calls

for more documentation on the

copper canisters.

The Swedish Radiation Safety

Authority (SSM) has reviewed SKB’s

applications under the Nuclear Activities

Act and recommends the Government

to grant a licence for a final

repository for spent nuclear fuel in

Forsmark and an encapsulation plant

in Oskarshamn.

The statement from the Land and

Environment Court (MMD) is also

positive in several important respects.

The court says yes to the issues

relating to the Forsmark site, the rock,

the buffer and the environmental

impact statement. The court also

approves the encapsulation plant and

increased capacity in the interim

storage facility Clab. However, the

court wants SKB to present more

documentation on the properties of

the canister and safety in the

long term. Furthermore, it wants

an investigation of the issue of responsibility

after closure, which has also

been requested by the munici pality.

We can conclude that we have

not been able to answer the court’s

questions regarding the copper

canister fully. At the same time, the

Government’s expert authority SSM

wrote in its statement that SKB has

the potential to meet the legislative

requirements on safe final disposal,

says SKB’s managing director Eva

Halldén in a comment.

SKB will provide documentation

That the two authorities have come

to such different conclusions is in

part due to the fact that they have

tried the applications under different

legislations, SSM under the Nuclear

Activities Act and MMD under the

Environmental Code. They also have

different licensing procedures. SSM

grants a licence in several steps with

continuous updates of the safety

analysis. But the court must say yes

or no based on the currently available

documentation.

The issue now lies with the

Ministry of the Environment and

Energy for further investigation and

SKB is working to develop the documentation

on the canister required

by the court.

This is material that we have

planned to produce for the preliminary

safety analysis. The difference

now is that we will prioritise the work

Operating Results November 2017

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated. gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 720 660 116 6 744 904 253 976 759 100.00 94.12 99.88 92.69 100.75 92.46

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

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

KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0

KKB 2 Beznau 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

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

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

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

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

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

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

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

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

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

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

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

Doel 3 PWR BE 1056 1006 0 0 6 732 621 251 169 221 0 78.97 0 78.79 0 79.12

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

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

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

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

Operating Results January 2018

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Time availability

[%]

Energy availability Energy utilisation

[%] *) [%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf DWR 1480 1410 744 957 521 957 521 341 149 580 100.00 100.00 93.97 93.97 86.59 86.59

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

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

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

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

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

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

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differently and complete it faster

than what was planned, says Helene

Åhsberg, SKB’s project manager for

the licensing process.

No referendum

Östhammar Municipality planned to

hold a referendum on the final repository

on March 4. But at a meeting in

the municipal council in the end of

January, it was decided to cancel the

referendum.

| | (18791534), www.skb.se

Yucca Mountain:

Can the US Finally End

the $12 Billion Impasse?

(nucnet) A US federal advisory panel

recently took a step in what could be a

lengthy process to determine if a deep

geological nuclear waste repository

should finally be built at Yucca Mountain,

a project that has been on the

drawing board since the 1970s at a

cost of around $ 12 bn (€ 9.7 bn).

The panel held a meeting to receive

input on reconstructing an electronic

library for documents needed to

decide on the US Department of

Energy’s Yucca licence application.

The meeting, at the Nuclear Regulatory

Commission’s headquarters in

Maryland, came one week after

another development: the White

House pledged $120m of funding in

its 2019 federal budget proposal to

restart licensing for the Yucca site,

north of Las Vegas in Nevada, and

to establish an interim storage programme

to address the growing

stockpile of nuclear waste produced

by nuclear plants across the nation.

After decades of wrangling, could

the US finally be on course to resolve

the question of what to do with

the high-level nuclear waste from

the nation’s 99 commercial nuclear

reactors?

| | www..energy.gov

US Nuclear Industry Calls

for Advanced Reactor Fuel

Cycle Infrastructure

(nucnet) The US Nuclear Energy

Institute has warned that preparations

should begin now to develop a

national fuel cycle infrastructure to

support the operation of advanced

reactors that are expected to begin

deployment in the 2020s and 2030s.

The Washington-based nuclear

industry lobby group said interest in

the development of advanced nuclear

reactor designs has been increasing in

recent years. Many of these designs

will require uranium fuel that is

enriched to a higher degree than

in the current worldwide fleet of lightwater

reactors. Fuel for advanced

reactors, enriched in U-235 to

between 5% and 20%, is called

high-assay low-enriched uranium

(HALEU).

Some of the advanced-performance

fuels being developed for use

with the existing reactor fleet also will

require HALEU. However, there are no

US-based facilities that manufacture

HALEU on a commercial scale. While

small quantities of HALEU materials

may be obtained on an interim basis

by “blending down” existing government

stocks of surplus high-enriched

uranium (HEU), those HEU materials

are limited in supply and not readily

available, the NEI said.

“Thus, for the long-term operation

of advanced reactors, as well as for

advanced fuels in existing reactors, a

robust new infrastructure for HALEU

fuel manufacture is needed.”

An NEI white paper says establishing

such a capability will better

position the US to advance nuclear

safety and non-proliferation policies

around the world, while helping to

ensure a robust commercial industry

domestically in the decades ahead.

On the other hand, “if the United

States and its allies have to depend on

foreign, state-owned enterprises to

meet fuel needs, it will be in a much

weaker position to influence these

policies globally”, the paper says.

| | Details online:

http://bit.ly/2FnZwOF

Reactors

IAEA Sees Safety Commitment

at Spain’s Almaraz

Nuclear Power Plant

(iaea) An International Atomic Energy

Agency (IAEA) team of experts said

the operator of Spain’s Almaraz

Nuclear Power Plant demonstrated a

commitment to the long-term safety of

the plant and noted several good practices

to share with the nuclear industry

globally. The team also identified areas

for further enhancement.

The Operational Safety Review

Team (OSART) today concluded an

18-day mission to Almaraz, whose

two 1,050-MWe pressurized-water

reactors started commercial operation

in 1983 and 1984, respectively.

Centrales Nucleares Almaraz-Trillo

(CNAT) operates the plant, located

about 200 km southwest of Madrid.

OSART missions aim to improve

operational safety by objectively

assessing safety performance using

the IAEA’s safety standards and proposing

recommendations for improvement

where appropriate. Nuclear

power generates more than 21 per

cent of electricity in Spain, whose

seven operating power reactors all

began operation in the 1980s.

“The team saw notable achievements

made by Almaraz in recent

years, such as implementing a comprehensive

management system, as

well as significant equipment renewal

plans, to establish safety as the

overriding priority at the plant,” said

Team Leader Peter Tarren, Head of the

IAEA’s Operational Safety Section.

“We found that people at every

level were willing to discuss their

work and how they might learn from

this OSART mission. They want to

keep enhancing the safety and

reliability of Almaraz.”

The 14-member team comprised

experts from Brazil, Bulgaria, France,

Germany, Mexico, the Russian Federation,

Sweden, United Arab Emirates,

the United Kingdom and the United

States of America, as well as three

IAEA officials.

The review was the 200th OSART

mission conducted by the IAEA since

the service was launched in 1982. It

covered the areas of leadership and

management for safety; training

and qualification; operations; maintenance;

technical support; operating

experience; radiation protection;

chemistry; emergency preparedness

and response; accident management;

human, technology and organizational

interactions and long-term

operation.

The team identified a number of

good practices that will be shared

with the nuclear industry globally,

including:

The use of a film-forming amine

compound to significantly reduce

the transport of potential corrosive

products to the steam generators.

The use of a cross-functional

indicator to show the cumulative

effect of equipment status and

planned activities for daily operations.

The installation of a centralized

vacuum system for cleaning, decontaminating

and discharging liquid

waste into the plant´s disposal system.

The mission made a number of

recommendations to improve operational

safety, including:

The plant should implement

further actions related to management,

staff and contractors to enforce

standards and expectations related

to industrial safety.

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The plant should take measures

to reinforce and implement standards

to enhance the performance of reactivity

manipulations in a deli berate

and carefully-controlled manner.

The plant should improve the

support, training and documented

guidance for Severe Accident Management

Guideline users in order to

mitigate complex severe accident

scenarios.

The team provided a draft report of

the mission to the plant’s management.

The plant management and the

Nuclear Safety Council (CSN), which

is responsible for nuclear safety

oversight in Spain, will have the

opportunity to make factual comments

on the draft. These will be

reviewed by the IAEA and the final report

will be submitted to the

Government of Spain within three

months.

The plant management said it

would address the areas identified

for enhancement and requested a

follow-up OSART mission in about

18 months.)

| | (18791443), www.iaea.org

Tianwan-3 Passes Commissioning

Tests at 100% Power

(nucnet) The Tianwan-3 nuclear

reactor unit in Jiangsu province,

northeastern China, has successfully

operated for 100 hours at 100% of its

design power level without interruption,

Russian state nuclear corporation

Rosatom said.

Rosatom said the 990-MW VVER

V-428M unit, which started to deliver

electrical energy to the grid on

30 December 2017, has undergone a

series of tests during the 100-hour

operation period required by regulators

before giving green light for

commercial operation.

Construction of Tianwan-3 began

in December 2012. The Tianwan

| | Swiss regulator approves safety case for restart of Beznau-1 (Photo: Axpo).

nuclear station is the largest economic

cooperation project between Russia

and China, an earlier statement had

said.

Tianwan-1 and -2, also VVER

V-428M units, began commercial

operation in 2007. The Tianwan-4

VVER V-428M unit is also under construction

by Russia while Tianwan-5

and -6 will be indigenous Generation

II+ CNP-1000 units.

| | en.cnnc.com.cn

Swiss Regulator Approves

Safety Case for Restart of

Beznau-1

(nucnet) Switzerland’s Federal

Nuclear Safety Inspectorate, ENSI,

has given the go-ahead for the restart

of the Beznau-1 nuclear unit after

approving the safety case presented

by owner Axpo following the discovery

in 2015 of flaw indications in

the reactor pressure vessel (RPV).

ENSI said in a statement that

Axpo had carried out “extensive

investigations and analyses” to

demonstrate that the RPV is safe.

Materials testing has shown

that agglomerates in the RPV do not

affect its key properties and structural

integrity analysis has shown that

the RPV does not contain any flaws

that could lead to its failure. “IRSN

is satisfied that work has been done

to all appropriate national and international

standards,” the statement

said.

Axpo said the safety case for

Beznau-1, the world’s oldest commercial

nuclear plant still in operation,

corroborates earlier assessments

and investigations, and validates the

existing safety margin for the safe

operation of the plant for 60 years.

Operator KKB will now begin the

return to service process with the

plant expected to be operating at full

load by the end of March 2018.

In December 2015 Axpo submitted

a roadmap ENSI detailing plans for

further investigations of flaw indications

in the RPV. During a scheduled

outage that began in May 2015,

inspections of the RPV registered

findings at some points in the base

material of the RPV indicating

“ minimal irregularities in the fabrication

process”, Axpo said. The company

carried out further measurements

and analyses and submitted a

report to ENSI.

In July 2015, Axpo announced

that the restart of Beznau-1 had been

postponed while the flaw indications

were investigated further. Then in

August, ENSI called for additional

investigations.

Beznau-2 was not affected by the

flaw indications and was returned to

service after its scheduled outage in

2015.

| | www.bkw.ch

Kursk II Passed

Construction Milestone

(rosatom) Kursk II began reinforcing

the foundation slab for the reactor

building of Unit 1. This operation

became the year’s key event on the

construction site of the Kursk plant.

On 21 December 2017, the first

16-ton reinforced concrete block was

installed on the rebar of the lower

foundation belt. According to the

project design, the foundation comprises

105 reinforced concrete blocks

with a total weight of 1,600 tons. This

will enable the construction team

to start concreting the foundation

slab of the reactor building in the

first half of 2018.

Prior to putting the first concrete

block, a rebar coupler engraved with

the words “The future is shaped today.

The first coupling sleeve of the innovative

VVER-TOI power unit” was

ceremonially installed into the foundation

reinforcement.

VVER-TOI (which means ‘a standard

optimized and automated power

unit based on VVER technology’)

reactors meet Russian and global

safety requirements and have a longer

service life and higher installed

capacity than existing reactors of

the Kursk Nuclear Power Plant.

Alexander Mikhailov, Governor of

the Kursk Region, noted that it was

an honor for the region to build

and commission one of the world’s

first nuclear plants with advanced

VVER-TOI reactors. “Construction of

Kursk II designed to meet the latest

global standards offers our region

development prospects for the entire

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Rosenergoatom had planned to

build two BN-1200 units at Beloyarsk

with commercial operation scheduled

by 2025. But construction depended

on the results of operating the pilot

Beloyarsk-4 BN-800 plant, which

began commercial operation in

October 2016.

There is another commercially

operational sodium-cooled FBR at

Beloyarsk, the BN-600. Both the

BN-600 and the BN-800 are smaller

versions of the BN-1200. There are

also two permanently shut-down

light-water reactors at the site.

| | www.rosatom.ru

265

NEWS

| | Kursk II passed construction milestone.

21st century. Just a few Russian

regions have such opportunities,” he

stressed.

Vyacheslav Fedyukin, Director

of Kursk NPP, noted it was symbolic

that the event happened on the

25 th anniversary of RosEnergoAtom

and 10 years after the foundation of

Rosatom, the companies that shaped

the newest history of Russia’s nuclear

industry. “Construction of Russia’s

first VVER-TOI-based power unit

proves that the national nuclear

power industry is always at the

cutting edge of science and engineering.

The new generation VVER-TOI

units are state-of-the-art facilities

made to the best of Russia’s nuclear

engineering knowledge,” he added.

At the moment, other operations

are also underway at the construction

site of Kursk II. Among them is excavation

of 1.2 million cub m of soil

to be completed in 2017, with over

800,000 cub m of sand, gravel

and aggregate already put in the

foun dation of Kursk II buildings

and structures. Construction of a

330/10 kV substation and preparation

of technical documents for its commissioning

are also drawing to a

close.

For reference:

Kursk II is designed to replace the

existing Kursk Nuclear Power Plant

that will be taken out of operation in

the years to come. Its first two units

with VVER-TOI, a new-type reactor,

will be commissioned simultaneously

with decommissioning of Units 1 and

2 of the existing nuclear station.

According to the master schedule of

Kursk II, Unit 1 will be commissioned

in late 2023 to be followed by Unit 2 in

late 2024.

| | (18791501),

ww.rosatom.ru

Russia Confirms Plans to

Revive BN-1200 Fast Breeder

Reactor Project

(nucnet) Russia plans to begin construction

of its first industrial-sized

sodium-cooled fast neutron reactor in

the 2020s after saying three years ago

that the project had been postponed,

the head of state nuclear corporation

Rosatom Alexei Likhachev told president

Vladimir Putin.

According to a transcript of a

meeting posted on the Kremlin’s

website, Mr Likhachev told Mr Putin

that fast breeder reactors (FBRs) have

significant advantages over existing

reactor types and Rosatom is proposing

that Russia goes ahead

with its plans for the Generation IV

BN-1200.

FBRs have been and are being

explored or constructed in Russia,

France, India, China, Japan and the

US. They allow a significant increase

in the amount of energy obtained

from natural, depleted and recycled

uranium. The technology also enables

plutonium and other actinides to be

used and recycled.

Russia operates the BN-600 and

BN-800 FBR units at Beloyarsk and

the BOR-60 fast breeder research

reactor at the Research Institute

of Atomic Reactors (RIAR) site in

Dimitrovgrad, southwest Russia.

BOR-60 is used to test fuel cycle,

sodium coolant technologies and a

range of design concepts for fast

breeder reactors.

In 2015, Rosatom said construction

of the planned BN-1200 at the

Beloyarsk nuclear power station in

central Russia had been postponed

until at least 2020, with state

nuclear operator Rosenergoatom

citing the need to improve fuel

for the reactor and questioning the

project’s economic feasibility.

Austria Begins Legal Action

Against EC Over Hungary’s

Paks Nuclear Project

(nucnet) Austria has filed a legal

complaint against the European Commission

with the European Court of

Justice in Luxembourg for allowing

Hungary to expand its Paks nuclear

power station.

Austrian minister of sustainability

and tourism Elisabeth Köstinger said

in a statement that nuclear power

“must have no place in Europe” and

Austria will not “not budge one

centimetre” from its anti-nuclear

stance.

The EC started an investigation

into state aid given to the Paks 2

project in November 2014. Last March

it approved the project to build two

new reactors, to be financed with the

help of Russia’s state atomic energy

corporation Rosatom, after regulators

said Hungarian authorities had

agreed to several measures to ensure

fair competition.

In January 2018, Austria announced

it planned to sue the EC over

the decision. “EU assistance is only

permissible when it is built on common

interest. For us, nuclear energy is

neither a sustainable form of energy

supply, nor is it an answer to climate

change”, a statement by the ministry

of sustainability said at the time.

The two planned units at Paks 2

nuclear power station are expected

to begin commercial operation in

2026 and 2027, Attila Aszódi,

the Hungarian government’s commissioner

for the Paks 2 project,

told a conference in Brussels late l

ast year.

An agreement signed in 2014

will see Russia supply two VVER-

1200 pressurised water reactors for

Paks 2 and a loan of up to €10bn

($12.3bn) to finance 80% of the

€12bn project.

| | www.bundeskanzleramt.gv.at

News

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atw Vol. 63 (2018) | Issue 4 ı April

266

NEWS

Company News

Framatome Completes

Purchase of Schneider

Electric’s Instrumentation and

Control Nuclear Business

(framatome) Framatome announced

that it completed its purchase of

Schneider Electric’s nuclear instrumentation

and control offering. With

this transaction, Framatome adds to

its engineering expertise and expands

its instrumentation and control (I&C)

offerings.

I&C systems are the central nervous

system of a nuclear power plant,

allowing operators to control reactor

operations. Modernizations, upgrades

and ongoing support are vital to manage

economic, long-term operation of

nuclear power plants, which provide

reliable, low-carbon electricity.

“With the integration of Schneider

Electric’s nuclear instrumentation

and control offering, we offer truly

added value to our customers with

a global technical expertise and

market know-how on I&C solutions

for the nuclear market,” said Bernard

Fontana, Chairman of the Managing

Board and Chief Executive Officer of

Framatome. “We welcome our new

colleagues to Framatome’s worldwide

team of I&C engineers and experts.”

This acquisition adds the nuclearqualified

version of Tricon and the

SPEC 200 platform to Framatome’s

nuclear safety I&C offerings, which

include the TELEPERM XS digital

platform, and non-computerized

analog solutions and instrumentation

for nuclear power plants.

This broadens the base of plants

worldwide for which Framatome

serves as the original equipment manufacturer

for safety I&C systems. It also

expands Framatome’s project and

engineering capacities for non-safety

I&C systems in the nuclear energy

market, relying on Schneider Electric’s

commercial TRICON and Foxboro

platforms.

Framatome also becomes the exclusive

service provider to the nuclear energy

market for the SPEC 200, nuclearqualified

Tricon and Foxboro systems.

| | www.framatome.com

Framatome Continues

Ramping up Production

at Its Le Creusot Site

(framatome) On January 25, 2018,

Framatome received the green light

from the French Nuclear Safety

Authority (ASN) and EDF to resume

manufacture of forgings for the

French nuclear fleet at its Le Creusot

site. This decision allows the plant to

continue ramping up its production

with a target of 80 ingots per year.

The authorization is an outcome of

the improvement plan launched at the

beginning of 2016 on the site following

a series of quality audits. With the completion

of all the actions necessary for

the resumption of production for the

French nuclear fleet and overall progress

of 90% to date, the plan will be

fully closed out in the first half of 2018.

The actions will then be incor porated

into the site’s continuous improvement

processes. Customers in France and

abroad, as well as all the safety

authorities concerned, have been kept

regularly informed of the actions

undertaken. Numerous reviews and

inspections have been conducted in order

to observe the progress of the plan

and integrate stakeholders’ feedback.

David Emond, Senior Executive

Vice President of Framatome’s Component

Manufacturing Business Unit,

comments: “The authorization to

resume manufacture of forgings for

the French nuclear fleet is a very

good news for the site that confirms

the successful execution of its improvement

plan. The 230 employees

at the Le Creusot site are engaged

in its deployment on a day to day basis

so that we can supply our customers

with equipment meeting the most

stringent safety and quality requirements

within agreed deadlines. I

want to thank them for the substantial

work they have accomplished on

the site over the last two years.”

Maintaining and developing the

skills of the Le Creusot plant teams

is a key element of the site’s improvement

plan, with a particular focus

on strengthening the nuclear safety

culture.

Framatome already invested

7.5 million euros at the site in 2017

to make the Le Creusot site a center

of excellence for the manufacture

of forgings for the nuclear industry,

and will pursue this effort in 2018.

Major milestone reached

in review of manufacturing

records

Moreover, a major milestone has

been reached in the review of legacy

manufacturing records at the Le

Creusot site. The first stage in the

inspection process which is being

applied to all records relating to

forgings produced for the nuclear

industry, a key stage consisting in

identifying findings, is now complete.

The analysis of these findings and the

processing of deviations will continue

until the end of 2018, in coordination

with customers and safety authorities.

Of the 6,000 records identified

during the initial survey, 3,854 correspond

to forgings installed on nuclear

installations.

At Framatome’s Jeumont and

Saint-Marcel sites, the audit has been

finalized since the summer of 2017 and

no deviation impacting the safety of

components has been brought to light.

| | www.framatome.com

JNFL and MHI Become

Shareholders of

Orano 2017 Revenue

(orano) The Orano Board of Directors

noted the completion of the capital

increase reserved for Japan Nuclear

Fuel Limited (JNFL) and Mitsubishi

Heavy Industries, Ltd. (MHI) for

a total of €500 million.

Pursuant to the initial agreements

signed with JNFL and MHI in

March 2017, the funds corresponding

to their total investment in Orano

had been placed in trust on July 26,

at the same time as the completion

of the capital increase reserved for

French State 2. These funds were

released and used for the subscription

of JNFL and MHI to Orano’s second

capital increase.

This transaction follows the

completion on December 31, 2017 of

the sale of the majority control of

Framatome (formerly New NP) by

AREVA SA to EDF as well as the

fulfillment of the regulatory closing

conditions related to the addition of

an equity stake in Orano of both Japanese

investors.

Orano’s capital is now held by the

French State (45.2%), the CEA

(4.8%)3, AREVA SA (40%), JNFL

(5%) and MHI (5%).

This transaction is the last major

step in the restructuring of the French

nuclear industry, undertaken in 2015,

and marks the end of the constitution

phase of the Orano group. With a

strengthened financial structure and

sound strategic partnerships, Orano

now has the means to grow and reach

its goal of being a leading player in the

production and recycling of nuclear

materials, in waste management and

dismantling.

Appointment of a new

independent director

After completion of Orano’s second

capital increase, the Orano General

Meeting, also held on February 26,

2018, appointed Patrick Pelata as

independent director.

| | www.orano.group

News

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atw Vol. 63 (2018) | Issue 4 ı April

Westinghouse Electric

Company Signs Cooperation

Agreement for Lead-cooled

Fast Reactor Development

(westinghouse) Westinghouse Electric

Company has signed a Cooperation

Agreement for lead-cooled fast

reactor (LFR) technology development

with the Italian National Agency

for New Technologies, Energy and

Sustainable Economic Development

(ENEA) and Ansaldo Nucleare. The

agreement demonstrates each party’s

commitment to collaborating toward

the development of a next-generation

nuclear plant based on LFR technology,

which is both “walk-away”

safe and economically competitive

across global energy markets.

“This agreement is an exciting

step towards the development of a

lead-cooled fast reactor for the

marketplace,” said Ken Canavan,

Westinghouse chief technology officer

and vice president, Global Technology

Office. “The LFR is game-changing

technology for clean energy industries,

and Westinghouse is pleased to

be working with such experienced

partners to bring this innovative

concept to fruition.”

Beyond baseload electricity

generation, the high-temperature

operation of the LFR will allow for

a broad range of applications such

as an effective load-following

capability enabled by an innovative

thermal energy storage system,

delivery of process heat for industrial

applications and water desalination.

ENEA is a world leader in research

and development on lead-based

systems, and currently operates

among the finest and largest experimental

facilities for LFR research in

the world.

Ansaldo Nucleare has vast experience

in nuclear power plant design,

supply, service and decommissioning,

and has played leading roles in

multiple international LFR development

programs for the past 15 years.

| | www.westinghousenuclear.com

BKW übernimmt Experten

für Strahlenschutz

(bkw) Die BKW Konzerngesellschaft

Dienstleistungen für Nukleartechnik

(DfN) übernimmt das ebenfalls

auf den kerntechnischen Bereich

spezia lisierte Unternehmen Technischer

Strahlenschutz (TSS). Dadurch

stärkt die BKW ihre Kompetenzen in

diesem Gebiet und baut sie weiter aus.

Dies vor dem Hintergrund der geplanten

Stilllegung des Kernkraftwerks

Mühleberg und zahlreicher weiterer

Kernkraftwerke in Europa.

Mit der Übernahme des Strahlenschutzunternehmens

DfN hat die BKW

bereits im letzten Jahr ihre bestehenden

und bewährten Kom petenzen im hochspezialisierten

Nukleartechnik-Bereich

erweitert. Der Eintritt der TSS in den

Unter nehmensverbund der BKW stellt

nun einen weiteren Ausbau in diesem

Gebiet dar. Die TSS ergänzt die

Strahlenschutzkompetenzen innerhalb

der BKW Gruppe und verstärkt diese

auch im Hinblick auf die Still legung

des Kernkraftwerks Mühleberg.

In Europa ist ausserdem eine Vielzahl

weiterer Stilllegungsprojekte in

Planung oder bereits im Gang. Der

Strahlenschutz spielt bereits beim

Betrieb von Kernkraftwerken eine

wichtige Rolle. Mit der Stilllegung

und den dabei ausgeführten Demontage-

und Freimessarbeiten nehmen

die Strahlenschutzarbeiten zu. Für

Strahlenschutzdienstleisterinnen wie

TSS und DfN bietet der wachsende

Stilllegungsmarkt daher ein grosses

Potenzial und die Möglichkeit, sich

weiterzuentwickeln.

Die DfN und die TSS haben bereits

verschiedentlich auf Projektbasis

zusammengearbeitet. Die erfolgreiche

Kooperation wird künftig

weiter ausgebaut, was mit einer

gegenseitigen Stärkung einhergeht.

Um eine optimale Zusammenarbeit

zu ermöglichen, wird die TSS in die

DfN integriert.

Die TSS mit Sitz in Geilenkirchen

im deutschen Bundesland Nordrhein-

Westfalen wurde 1979 gegründet

und zählt 15 Mitarbeitende. Das

Unternehmen bietet ein qualitativ

hochwertiges und breites Angebot von

Dienstleistungen im kerntechnischen

Bereich. Dazu gehören neben dem

Strahlenschutz die Dekontamination,

die Abfallentsorgung, die Dosimetrie

sowie die Abwicklung von Transporten

radioaktiver Stoffe.

| | (18791521), www.bkw.ch

Companies

China Approves $ 100 Billion

Merger of Leading

Nuclear Companies

(nucnet) China has approved the

merger of nuclear power producer

China National Nuclear Corporation

(CNNC) with nuclear plant builder

China Nuclear Engineering and

Construction Corporation (CNECC),

the state-run China Daily news agency

said.

According to the China Daily, the

combined assets of the new company

will be worth about $100bn (€80bn),

while its workforce will be about

150,000 employees.

CNNC is China’s number two

nuclear power producer and CNECC

the country’s top nuclear power plant

builder.

China Daily said the merger is in

line with efforts by China to streamline

the state-operated sector of its

economy and reduce the number of

state-owned companies administered

by central government.

Approval for the merger was

confirmed by the State-Owned Assets

Supervision and Administration

Commission (SASAC) in a one-line

statement posted on its website.

| | (18800822), en.cnnc.com.cn

267

NEWS

Research

| | BKW übernimmt Experten für Strahlenschutz © BKW.

NRG: Every Day,

30,000 Patients Benefit From

Medical Isotopes From Petten

(nrg) Medical isotopes are indispensable

for diagnosing and treating

cancer. Demand for them is set to soar

over the next 20 years, but supplies

are diminishing. To put the spotlight

News

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atw Vol. 63 (2018) | Issue 4 ı April

• Separative work: 58.00–92.00

268

NEWS

2016

January to June 2016

• Uranium: 26.50–35.25

• Conversion: 6.25–6.75

• Separative work: 58.00–62.00

July to December 2016

• Uranium: 18.75–27.80

• Conversion: 5.50–6.50

• Separative work: 47.00–62.00

| | NRG: Every day, 30000 patients benefit from medical isotopes from Petten View of the pool type reactor

core. (Courtesy: JRC)

on the world of medical isotopes, the

platform 30000perdag.nl has been

launched. The aim of the platform and

the accompanying campaign is to

boost awareness that the Netherlands

must continue leading the field in

cancer treatment.

The future

Over the next 20 years, the number of

cancer diagnoses is expected to rise by

70%. Fortunately, health care is

constantly improving, partly through

the use of medical isotopes. However,

there are only 6 reactors worldwide

which can produce medical isotopes,

one of which is closing next year.

This means that whilst demand for

medical isotopes is growing worldwide,

supplies are diminishing.

30000perdag.nl

An online information park for a

wide audience has been built on

30000perdag.nl. Visitors can learn all

about medical isotopes here: from raw

materials to the reactor in Petten to

applications in the hospital. By opening

up that world, NRG in Petten

wants to show (former) cancer

patients and their families and

acquaintances what is needed to be

able to treat cancer, and request

support for medical isotopes and good

cancer treatment in the Netherlands

and abroad.

Former cancer patients play

starring role in campaign

The campaign uses 3 video interviews

with cancer survivors. The interviews

were conducted by presenter Fien Vermeulen,

herself a former lymphoma

patient. Fien drives with former

patients Anouk (26), Alexander (42)

and Manon (34) to the research

reactor in Petten, where they talk

about their remarkable experiences in

times of uncertainty. Each of them

represents one of the 30,000 people

who benefit or have benefitted from

medical isotopes every day.

Anyone can demonstrate their

support by liking the Facebook page

30.000perdag. Another very visible

form of support is available through

the T-shirts that can be ordered via

30000perdag.nl. These enable former

patients and supporters to show their

backing for the campaign.

| | (18800822), www.nrg.eu

Market data

(All information is supplied without

guarantee.)

Nuclear Fuel Supply

Market Data

Information in current (nominal)

U.S.-$. No inflation adjustment of

prices on a base year. Separative work

data for the formerly “secondary

market”. Uranium prices [US-$/lb

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

0.385 kg U]. Conversion prices

[US-$/kg U], Separative work

[US-$/SWU (Separative work unit)].

January to December 2013

• Uranium: 34.00–43.50

• Conversion: 9.25–11.50

• Separative work: 98.00–127.00

January to December 2014

• Uranium: 28.10–42.00

• Conversion: 7.25–11.00

• Separative work: 86.00–98.00

January to December 2015

• Uranium: 35.00–39.75

• Conversion: 6.25–9.50

2017

January 2017

• Uranium: 20.25–25.50

• Conversion: 5.50–6.75

• Separative work: 47.00–50.00

February 2017

• Uranium: 23.50–26.50

• Conversion: 5.50–6.75

• Separative work: 48.00–50.00

March 2017

• Uranium: 24.00–26.00

• Conversion: 5.50–6.75

• Separative work: 47.00–50.00

April 2017

• Uranium: 22.50–23.50

• Conversion: 5.00–5.50

• Separative work: 45.50–48.50

May 2017

• Uranium: 19.25–22.75

• Conversion: 5.00–5.50

• Separative work: 42.00–45.00

June 2017

• Uranium: 19.25–20.50

• Conversion: 5.55–5.50

• Separative work: 42.00–43.00

July 2017

• Uranium: 19.75–20.50

• Conversion: 4.75–5.25

• Separative work: 42.00–43.00

August 2017

• Uranium: 19.50–21.00

• Conversion: 4.75–5.25

• Separative work: 41.00–43.00

September 2017

• Uranium: 19.75–20.75

• Conversion: 4.60–5.10

• Separative work: 40.50–42.00

October 2017

• Uranium: 19.90–20.50

• Conversion: 4.50–5.25

• Separative work: 40.00–43.00

November 2017

• Uranium: 20.00–26.00

• Conversion: 4.75–5.25

• Separative work: 40.00–43.00

December 2017

• Uranium: 23.50–25.50

• Conversion: 5.00–6.00

• Separative work: 39.00–42.00

2018

January 2018

• Uranium: 21.75–24.00

• Conversion: 6.00–7.00

• Separative work: 38.00–42.00

News

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atw Vol. 63 (2018) | Issue 4 ı April

February 2018

• Uranium: 21.25–22.50

• Conversion: 6.25–7.25

• Separative work: 37.00–40.00

| | Source: Energy Intelligence

www.energyintel.com

Cross-border Price

for Hard Coal

Cross-border price for hard coal in

[€/t TCE] and orders in [t TCE] for

use in power plants (TCE: tonnes of

coal equivalent, German border):

2012: 93.02; 27,453,635

2013: 79.12, 31,637,166

2014: 72.94, 30,591,663

2015: 67.90; 28,919,230

2016: 67.07; 29,787,178

I. quarter: 56.87; 8,627,347

II. quarter: 56.12; 5,970,240

III. quarter: 65.03, 7.257.041

IV. quarter: 88.28; 7,932,550

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

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

269

NEWS

2017:

I. quarter: 95.75; 8,385,071

II. quarter: 86.40; 5,094,233

III. quarter: 88.07; 5,504,908

| | Source: BAFA,

some data provisional

www.bafa.de

EEX Trading Results

February 2018

(eex) In February 2018, the European

Energy Exchange (EEX) achieved a

total volume of 274.3 TWh on its

power derivatives markets (February

2017: 200.8 TWh) which is a yearon-year

increase of 37 %. In doing

so, EEX was able to grow its power

derivatives volumes across all market

areas.

In total, the German and Austrian

markets (Phelix-DE, Phelix-AT and

Phelix-DE/AT) increased by 12 % to

169.7 TWh. This includes 153.4 TWh

from the benchmark product Phelix-

DE which achieved its highest volume

since launch in April 2017. Volumes

in the French market more than

doubled to 26.9 TWh (February 2017:

12.3 TWh) while Italian power

volumes grew substantially to

40.0 TWh (February 2017: 22.1 TWh).

Furthermore, on the Spanish market,

volumes increased by more than

250 % to 8.3 TWh (February 2017:

2.3 TWh).

The February volume comprised

173.5 TWh traded at EEX via Trade

Registration with subsequent clearing.

Clearing and settlement of all exchange

transactions was executed by European

Commodity Clearing (ECC).

The Settlement Price for base

load contract (Phelix Futures) with

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

)1

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

delivery in 2019 amounted to 33.85 €/

MWh. The Settlement Price for peak

load contract (Phelix Futures) with

delivery in 2019 amounted to 42.40 €/

MWh.

On the EEX markets for emission

allowances, the total trading volume

increased by 57% to 144.2 million

tonnes of CO 2 in February (February

2017: 91.7, million tonnes of CO 2 ).

Primary market auctions contributed

75.1 million tonnes of CO 2 to the total

volume. On the spot secondary

market, volumes more than doubled

to 4.5 million tonnes of CO 2 (February

2017: 2.0 million tonnes of CO 2 ). On

the EUA Futures market, EEX was able

to increase volumes by 80% to

37.3 million tonnes of CO 2 (February

2017: 20.7 million tonnes of CO 2 ).

Furthermore, 27.4 million tonnes of

CO 2 were traded in EUA Options

which is the highest monthly volume

so far in this product.

The EUA price with delivery in

December 2018 amounted to

8.80/10.15 €/ EUA (min./max.).

| | www.eex.com

MWV Crude Oil/Product Prices

January 2017

(mwv) According to information and

calculations by the Association of the

German Petroleum Industry MWV e.V.

in January 2018 the prices for

super fuel, fuel oil and heating oil

noted inconsistent compared with

the pre vious month December 2017.

The average gas station prices for Euro

super consisted of 136.84 €Cent

( December 2017: 136.84 €Cent,

approx. +-0.0 % in brackets: each

information for pre vious month or

rather previous month comparison),

for diesel fuel of 120.48 €Cent

(119.01; +1.24 %) and for heating oil

(HEL) of 62.27 €Cent (60.65 €Cent,

+2.67 %).

The tax share for super with

a consumer price of 136.84 €Cent

(136.84 €Cent) consisted of

65.45 €Cent (47.83 %, 65.45 €Cent)

for the current constant mineral oil

tax share and 21.85 €Cent (current

rate: 19.0 % = const., 22.12 €Cent)

for the value added tax. The product

price (notation Rotterdam) consisted

of 40.17 €Cent (29.36 %, 37.18 €Cent)

and the gross margin consisted of

9.37 €Cent (6.85 %; 12.36 €Cent).

Thus the overall tax share for super

results of 66.83 % (66.83 %).

Worldwide crude oil prices

(monthly average price OPEC/Brent/

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

significantly higher, approx. +8.36 %

(+2.34 %) in January 2018 compared

to December 2017.

The market showed a stable

development with higher prices; each

in US-$/bbl: OPEC basket: 66.85

(62.06); UK-Brent: 69.08 (64.37);

West Texas Inter mediate (WTI): 63.7

(57.88).

| | www.mwv.de

News

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atw Vol. 63 (2018) | Issue 4 ı April

270

Czechs and Balances and Why ‘Ugly’

Nuclear Deserves a Political Makeover

NUCLEAR TODAY

Author

John Shepherd

Shepherd

Communications

3 Brooklands

West Sussex

BN43 5FE

Links to reference

sources:

Dana Drábová

interview:

http://bit.ly/2Ik7WaJ

European Investment

Bank announcement:

http://bit.ly/2Ik7WaJ

Yonhap News

Agency report:

http://bit.ly/2FyvZkw

As if Europe does not have enough on its plate to deal with at the moment – politically and economically just for starters

– could Brussels be on a collision course with the Czech government over the country's plans to expand nuclear energy?

There is certainly friction over the issue between Prague and

the European Commission (EC), to put it mildly. But why?

The veteran head of the Czech Republic’s State Office

for Nuclear Safety, Dana Drábová, last month accused

other EU member states of “pressurising” Prague over the

early closure of its oldest nuclear reactor units.

Drábová reportedly told an energy conference in the

country: “There is immense pressure developing that the

operating life of nuclear reactors will be limited to 40 years.

That means that our political representatives, whoever they

might be, sometime around 2023 will face a battle over a

further 10-year extension for Dukovany. The current State

Energy Framework counts on the lifetime of the Dukovany

reactors ending sometime between 2030 and 2040.”

The nuclear safety chief later told Czech Radio the

pressure was coming from “the 14 countries which are not

using nuclear power and some of which regard it as

something ugly”. If the pressure continued, she predicted

there would be a concerted “willingness… to get rid of

these nuclear plants in Europe as fast as possible”.

Drábová’s comments came against a backdrop of the

Czech government saying it would appoint an expert team

to consider proposals to break up the majority state-owned

electricity firm CEZ. The move was one of several options

mooted to support financing of the construction of a new

nuclear power plant at Dukovany.

Analysts say the new nuclear plant could be built by the

traditional energy unit, which would be fully state-owned

and therefore in the best position to take on the risks of

high costs that the utility could not if it were an entity with

private owners.

Czech prime minister Andrej Babiš is backing proposals

to build the Dukovany reactor, around 50 kilometres north

of the (anti nuclear) Austrian border, to replace a Soviet-era

reactor. But this would mean persuading the EC to exempt

the project from strict EU rules on government bids.

If the Czech government fails in its quest, it could consider

doing a deal with Russia, which would undoubtedly

be very much along the lines of the nuclear construction

and financing deal Moscow signed recently with EU

member Hungary.

If, dear reader, you now have a sense of déjà vu, you

would be right. You may recall that Hungary went through

a similar nuclear battle with the EC, despite Hungary’s parliament

fully backing proposals to build two new nuclear

reactor units in that country.

Initially, the EC said in November 2015 it had started

legal action against Hungary over a contract signed with

Russia’s Rosatom to build two units at the existing Paks

plant. Brussels expressed concern about the project’s

compatibility with EU public procurement rules. However,

the EC eventually cleared the issue and a state aid investigation

into the project financing for the ‘Paks II’ project

was subsequently dropped by the EC.

There was a similar clash with the EC when the UK first

unveiled plans to invest in building the Hinkley Point C

nuclear plant.

So is the latest tussle between Prague and Brussels

really over concerns about state-aid rules or is it more a

worrying trend of interference to stop nuclear in its tracks?

And is the conflict really worth it…?

Czech PM Babiš said following an official visit to

Hungary last January, where he attended a summit of

prime ministers of the Visegrad Group countries, that he

and Hungarian counterpart Viktor Orbán discussed the

potential for “further developing” cooperation in sectors

such as the nuclear energy industry.

But far more intriguing was what Babiš claimed was the

attitude of Visegrad leaders about relations with the

institutions of the EU. According to a statement issued by

the Czech government, Babiš said the leaders agreed it was

“necessary to depoliticise Brussels and the EC”. Apparently,

the leaders believe that when it comes to EU affairs,

“ member states, prime ministers and presidents, should

have the main say”, according to Babiš.

If there is behind-the-scenes pressure to stamp out

nuclear wherever it might try to cling on or prosper in the

EU, where is that effort coming from and why? Of course,

it is no secret that Austria and Germany strongly oppose

any expansion of nuclear power in Europe. Having lived

and worked in Germany, I never understood that

politically- inspired decision – but as a guest in the country

for which I have great admiration I respect its decision.

Austria’s approach has always puzzled me more – being

willing as it is to host the International Atomic Energy

Agency (IAEA) and enjoy all the ‘fruits’ that that privilege

brings, not least in the economic benefit of having the

agency based in Vienna.

But back to the Czech project. As a possible fight with

the EC shapes up, it is not only Moscow that is set to benefit

from yet another new nuclear power order from an EU

nation.

South Korea is also reportedly circling – keen to tempt

Prague to consider its nuclear technology, according to

Seoul’s Yonhap News Agency.

Can the EU really afford such a quarrel – again – with a

member state over nuclear? And why should European

skills, knowhow and investment not be channelled into the

Czech nuclear project?

I am struck by the EC’s approach to another industrial

sector and how contrasting it is. The EC is currently working

at full tilt to develop a European battery cell industry,

with the goal of ensuring the EU is not overwhelmed by

competition from Asian battery makers for products such as

electric vehicles and energy storage devices.

The EU’s vice-president for the energy union, Maroš

Šefčovič, said in February “there are many extremely

interesting actions that we need to pursue, including (a)

simplification of approval procedures and permitting

processes in the EU”. Indeed the European Investment Bank

has already approved a loan for the construction and

operation of what it said will be a first-of-a-kind demonstration

plant in Sweden, for the manufacturing of lithium- ion

batteries.

The EC’s support for the development of such technology

across EU member states is of course admirable, but one

hears nothing of state-aid rules and complications here!

Why is it that nuclear cannot win such favourable attention

and support? Does it really have to be this way – and

hasn’t the EC learned anything from the UK’s Brexit vote

about treading carefully in issues that are seen by member

states of national importance?

Nuclear Today

Czechs and Balances and Why ‘Ugly’ Nuclear Deserves a Political Makeover ı Jubair Ahmed Shamim and Kune Yull Suh

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