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nucmag.com<br />
<strong>2018</strong><br />
4<br />
217<br />
Heat Transfer Systems<br />
for Novel Nuclear<br />
Power Plant Designs<br />
221 ı Operation and New Build<br />
Safety Research for GEN IV Reactors<br />
226 ı Operation and New Build<br />
Numerical Analysis for the MYRRHA Project<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Security of Supply ...<br />
and the Clock is Ticking ...<br />
Deal reader, More than one hundred years ago, around 1890, a conflict flared up between the two well-known<br />
protagonists of electricity supply, Thomas Alva Edison and George Westinghouse, on the large-scale power supply and the<br />
construction of power grids in the United States of America. While Edison technically preferred D.C. voltage, Westinghouse<br />
counted instead on alternating voltage. In the end it was not a matter of the most suitable technique but of the anticipated<br />
market shares of each company General Electric or Westinghouse Electric and the patents behind. At a breath taking pace,<br />
the most important developments for the use of electricity were preceding: In the year 1866 Werner Von Siemens<br />
discovered the dynamo- electrical principle, which enabled larger performance. The development of alternating voltage<br />
in the year 1881 enabled generally technically and cost-effectively the transportation of electricity over long distances<br />
– we are talking back then about distances of some ten kilometres. Alternating voltage enforced itself at that time due to<br />
possible further transportation length enabled through higher trans mission voltage.<br />
207<br />
EDITORIAL<br />
Both current types have something in common: generation<br />
and use need to take place simultaneously. The grid fails if<br />
both do not fit together. Neither alternating current grids<br />
nor direct current grids offer storage possibilities. Thus, a<br />
stable power system also requires a stable and reliable<br />
generation, because if a larger system “fails”, the system<br />
restoration is, from its task and process, a large-scale<br />
project.<br />
Different believes e.g. from politics or other interest<br />
groups are simply wrong, power systems are – without any<br />
further active establishments and plants- no accumulators.<br />
A reliable power supplying system needs at any time<br />
reliable generation. “Surpluses”, meaning potentials for a<br />
higher generation than demand, when so ever, cannot be<br />
shifted or stored “electrically” in the system at a later time.<br />
It was not an inconspicuous message, which appeared<br />
multiple times in the press at the end of February,<br />
beginning of March <strong>2018</strong>. Headlines such as “Time<br />
synchronisation per power system: Energy shortages make<br />
watches lose time”, described a phenomenon, of which,<br />
according to the media “one became aware of – only<br />
( editor’s note) - after weeks”: What happened?<br />
As an indicator for the stability of alternating power<br />
systems stand supply voltage a well as system frequency.<br />
For the system frequency applies that she needs to be<br />
identical at any point of the system. If generation and<br />
consumption do not fit, deviations occur, leaking generation<br />
leads among others to a perceived frequency decrease<br />
among the entire connected system. As the system<br />
frequency is defined for our alternating electricity net with<br />
constant 50 Hertz, it is also qualified for watches, which<br />
use the frequency as direct clock indicator.<br />
We can for example – due to cost reasons – renounce to<br />
a frequency stabilising quartz oscillator. Nevertheless, this<br />
technical simplification is bought with failures in time, if<br />
the frequency deviates from the standard over a longer<br />
period. Only a few hundred Hertz is enough for days and<br />
weeks in order to, as in the current case, generate a time<br />
deviation of minus 360 seconds, 6 minutes, and those<br />
inside the entire affected system of 25 West, Middle- and<br />
South European countries.<br />
The cause for this incident was later communicated by<br />
the European Network of Transmission System Operators for<br />
Electricity (short ENTSO-E) and the Swiss net operator<br />
swissgrid, that in the control zone Serbia, Macedonia,<br />
Montenegro (the so called SMM rule block), especially in<br />
Kosovo and Serbia less energy was fed into the system. A<br />
deficit of 113 gigawatt hours was shown, not much, in view<br />
of a European daily production of around 8,000 gigawatt<br />
hours. But especially this shows how delicate our power<br />
system is and how sensitive it reacts to the smallest<br />
malfunctions.<br />
Reliable measures in power generation – meaning<br />
currently only for conventional techniques, thus need,<br />
with all considerations on the reconstruction of electricity<br />
supply, to be reconsidered. Additionally and almost<br />
simultaneously another alarming “availability message”<br />
came in: At the beginning of March <strong>2018</strong> European gas<br />
storage tanks were only filled with a quantity of 26.2 per<br />
cent, Germany even on average only with 23.8 per cent.<br />
Thus, according to an EU-conform proceeding an early<br />
warning level was reached, because the filling level of<br />
storage tanks may not be lower than around 20 % due to<br />
reasons of guaranteeing mechanical stability. On top came<br />
the message that more natural gas was imported to Europa<br />
than in the previous years. All first hints, that there might<br />
not be enough natural gas in Europe for dispose filling in as<br />
a “reserve”?<br />
In all, these are all important references that any,<br />
especially neither direct market- nor technically driven,<br />
interventions – where compensation factors can con tribute<br />
– need to be well thought in our power system. Furthermore,<br />
does the availability of a broad basis of conventional<br />
generation not only gain more importance, she is even<br />
more important than it is conceded for “conventionals”<br />
vision wise in many places in terms of an „energy<br />
transition“. To what extend “the clock” might tick on<br />
possible severe supply shortfalls or even large-scale loss of<br />
off-site power… one does not know…<br />
Christopher Weßelmann<br />
– Editor in Chief –<br />
Editorial<br />
Security of Supply ... and the Clock is Ticking ...
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
EDITORIAL 208<br />
Versorgungssicherheit und die Uhr tickt ...<br />
Liebe Leserin, lieber Leser, vor mehr als hundert Jahren, um 1890, entbrannte eine Auseinandersetzung<br />
zwischen den beiden bekannten Protagonisten der Elektrizitätsversorgung, Thomas Alva Edison und George Westinghouse,<br />
zur weiträumigen Versorgung der Vereinigten Staaten von Amerika mit Strom und dem Aufbau geeigneter Stromnetze.<br />
Während Edison technisch die Gleichspannung favorisierte, setze Westinghouse die Wechselspannung dagegen.<br />
Letztendlich ging es aber nicht wesentlich um die Frage der geeigneteren Technik, sondern um die avisierten Marktanteile<br />
der jeweiligen Unternehmen General Electric bzw. Westinghouse Electric und die dahinter stehenden Patente. Vorangegangen<br />
waren in atemberaubendem Tempo die wichtigsten Entwicklungen für die Nutzung der Elektrizität: Im Jahr<br />
1866 entdeckte Werner von Siemens das dynamoelektrische Prinzip, das größere Leistungen ermöglichte. Die<br />
Entwicklung des Wechselstromtransformators im Jahr 1881 ermöglichte technisch grundsätzlich und kostengünstiger<br />
den Transport von Strom über längere Strecken – wir sprechen hier zu jener Zeit über Strecken im Bereich von einigen<br />
zehn Kilometern. Durchgesetzt hatte sich aufgrund der durch höhere Übertragungsspannungen möglichen weiteren<br />
Transportlängen zu jener Zeit die Wechselspannung.<br />
Beiden Stromarten ist eines gemeinsam: Erzeugung und<br />
Nutzung müssen exakt zeitgleich erfolgen. Sind Erzeugung<br />
und Gebrauch nicht im Einklang, bricht das Netz<br />
zusammen. Weder Wechsel- noch Gleichspannungsnetz<br />
bieten „Speichermöglichkeiten“. Für ein stabiles Stromnetz<br />
ist daher auch eine stabile und verlässliche Erzeugung<br />
erforderlich, denn wenn einmal ein größeres Stromnetz<br />
„zusammenbricht“, ist der Netzwiederaufbau ein von der<br />
Aufgabe und dem zeitlichen Ablauf her aufwendiges<br />
Vorhaben. Anderslautende Stimmen z.B. aus der Politik<br />
oder von Interessengruppen sind schlichtweg falsch,<br />
Strom netze sind – ohne weitere aktive Einrichtungen und<br />
Anlagen – keine Speicher. Ein verlässliches Stromversorgungsnetz<br />
benötigt eine jederzeit verlässliche Erzeugung.<br />
„Überschüsse“, also Potenziale für eine höhere Erzeugung<br />
als die vorhandene Nachfrage, wann und warum auch<br />
immer, lassen sich „elektrisch“ im Netz nicht auf spätere<br />
Zeiten verschieben, also speichern.<br />
Es war eine nicht unscheinbare Nachricht, die Ende<br />
Februar, Anfang März <strong>2018</strong> mehrfach durch die<br />
Presse ging. Überschriften wie „Zeit-Synchronisation per<br />
Stromnetz: Energieknappheit lässt Uhren nachgehen“,<br />
beschrieben ein Phänomen, dessen man sich nach Angaben<br />
in der Presse „nach Wochen – erst (Anm. der Red.) –<br />
bewusst wurde“: Was war geschehen?<br />
Für die Stabilität bzw. als Indikator für die Stabilität<br />
von Wechselstromnetzen stehen die Netzspannung sowie<br />
die Netzfrequenz. Für die Netzfrequenz gilt dabei, dass<br />
diese an jedem Punkt in einem Netz identisch ist. Stimmen<br />
Erzeugung und Verbrauch nicht überein, kommt es zu<br />
Abweichungen, fehlende Erzeugung führt u.a. zu einer<br />
im gesamten angebundenen Netz fühlbaren Frequenzabnahme.<br />
Da die Netzfrequenz für unser Wechselstromnetz<br />
mit konstant 50 Hertz vereinbart ist, eignet sich diese<br />
auch für Uhren, die die Frequenz als direkten Taktgeber<br />
nutzen. Auf z.B. einen frequenzstabilisierenden Quarzoszillator<br />
kann – aus Kostengründen – verzichtet werden.<br />
Diese technische Vereinfachung erkauft man sich allerdings<br />
mit Fehlern in der Uhrzeit, wenn die Frequenz über<br />
einen längeren Zeitraum vom Standard abweicht. Schon<br />
wenige hundertstel Hertz reichen über Tage und Wochen<br />
aus, um, wie im aktuellen Fall, eine kumulierte Zeitabweichung<br />
von Minus 360 Sekunden, also 6 Minuten,<br />
hervorzurufen; und dies im ganzen betroffenen Netz von<br />
25 West-, mittel- und südosteuropäischen Ländern.<br />
Als Ursache für dieses Ereignis wurde später vom<br />
Verband Europäischer Übertragungsnetzbetreiber (kurz<br />
ENTSO-E, European Network of Transmission System<br />
Operators for Electricity) und dem Schweizer Netzbetreiber<br />
swissgrid kommuniziert, dass in der Kontrollzone Serbien,<br />
Mazedonien, Montenegro (dem sogenannten SMM Regelblock),<br />
insbesondere in Kosovo und Serbien zu wenig<br />
Energie ins Netz eingespeist wurde. Ein Fehlbetrag von<br />
113 Gigawattstunden wurde ausgewiesen, nicht viel,<br />
angesichts einer europaweiten Tagesproduktion von rund<br />
8.000 Gigawattstunden. Aber gerade diese zeigt, wie<br />
filigran unser Stromnetz ist und wie empfindlich es doch<br />
auf kleinste Störungen reagiert.<br />
Verlässliche Größen in der Stromerzeugung, sprich<br />
derzeit letztendlich nur die konventionellen Techniken,<br />
müssten von daher in allen Überlegungen zum Umbau der<br />
Stromversorgung neu überdacht werden. Hinzu kam fast<br />
zeitgleich eine weitere bedenkliche energiewirtschaftliche<br />
„Verfügbarkeitsmeldung“: Europas Gasspeicher waren zu<br />
Anfang März <strong>2018</strong> nur noch zu 26,2 Prozent gefüllt,<br />
Deutschland gar im Schnitt nur zu 23,8 Prozent. Damit<br />
war nach einem EU-einheitlichen Verfahren eine Frühwarnstufe<br />
erreicht, denn die Speicher dürfen ihren Füllgrad<br />
aus Gründen der Gewährleistung ihrer mechanischen<br />
Stabilität nicht unter rund 20 % absenken. Hinzu kam die<br />
Mitteilung, dass mehr Erdgas nach Europa importiert<br />
wurde, als in den Vorjahren. Alles erste Hinweise darauf,<br />
dass vielleicht in Zukunft doch nicht ausreichend Erdgas<br />
in Europa zur Verfügung stehen wird, um als „Reserve“<br />
einzuspringen?<br />
In Summe sind dies alles wichtige Hinweise darauf,<br />
dass jegliche, vor allem weder direkt markt- noch technisch<br />
getriebenen Eingriffe – wo ausgleichende Faktoren wirken<br />
können – in unser Stromversorgungssystem wohl überdacht<br />
sein müssen. Zudem gewinnt die Verfügbarkeit einer<br />
breiten Basis konventioneller Erzeugung damit nicht nur<br />
an Bedeutung, sie ist bedeutungsvoller als vielerorts in<br />
Visionen einer „Energiewende“ den Konventionellen<br />
zugestanden wird.<br />
Inwieweit „die Uhr“ möglicher schwerwiegender Versorgungsengpässe<br />
oder gar großflächiger Netzausfälle<br />
tickt ... man weis es nicht ...<br />
Christopher Weßelmann<br />
– Chefredakteur –<br />
Editorial<br />
Security of Supply... and the Clock is Ticking ...
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
210<br />
Issue 4<br />
April<br />
CONTENTS<br />
217<br />
Heat Transfer Systems<br />
for Novel Nuclear<br />
Power Plant Designs<br />
| | The Swiss nuclear power plants generate up to 40 % of the country’s electricity production. At the Beznau site, two pressurised water<br />
reactors are in operation with a gross capacity of 380 MW each and a net capacity of 365 MW. Switzerland’s Federal Nuclear Safety<br />
Inspectorate, ENSI, gave the go-ahead for the restart of Beznau-1 after approving the safety case presented by operator Axpo following<br />
the discovery in 2015 of flaw indications in the reactor pressure vessel. (Courtesy: Axpo)<br />
Editorial<br />
Security of Supply ... and the Clock is Ticking ... . . 207<br />
Versorgungssicherheit und die Uhr tickt ... . . . . 208<br />
Abstracts | English . . . . . . . . . . . . . . . . . . . 212<br />
Abstracts | German . . . . . . . . . . . . . . . . . . . 213<br />
Inside Nuclear with NucNet<br />
Euratom: Industry Softens Stance<br />
as Government Lays Out Plans for Transition . . . 214<br />
NucNet<br />
Calendar . . . . . . . . . . . . . . . . . . . . . . . 216<br />
Operation and New Build<br />
Heat Transfer Systems for Novel<br />
Nuclear Power Plant Designs . . . . . . . . . . . . . 217<br />
Sebastian Vlach, Christoph Fischer and Herman van Antwerpen<br />
Experimental and Analytical Tools<br />
for Safety Research of GEN IV Reactors . . . . . . . 221<br />
G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak<br />
DAtF Notes. . . . . . . . . . . . . . . . . . . . . .215<br />
221<br />
| | Centrum Výzkumu Řež facilities list.<br />
217<br />
Numerical Analysis of MYRRHA<br />
Inter- wrapper Flow Experiment at KALLA . . . . . 226<br />
| | Koeberg PWR steam generator and simulation model.<br />
Abdalla Batta and Andreas G. Class<br />
Contents
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
226<br />
CONTENTS<br />
211<br />
| | Velocity magnitude within bundle showing flow distribution.<br />
Heat Balance Analysis for<br />
Energy Conversion Systems of VHTR . . . . . . . . 230<br />
SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon<br />
and Soyoung Park<br />
Spotlight on Nuclear Law<br />
Information Requirements Versus<br />
Confidentiality Obligations – Extension of<br />
the In-Camera Procedure Planned . . . . . . . . . . 235<br />
Informationsbedarf versus<br />
Geheimhaltungspflichten – Erweiterung<br />
des In camera-Verfahrens geplant . . . . . . . . . . 235<br />
Tobias Leidinger<br />
Environment and Safety<br />
CFD Modeling and Simulation of Heat and Mass<br />
Transfer in Passive Heat Removal Systems . . . . . 238<br />
Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas<br />
|241<br />
249<br />
| | Collecting soil samples.<br />
Research and Innovation<br />
Irradiation Tests of a Flat Vanadium Self- Powered<br />
Detector with 14 MeV Neutrons . . . . . . . . . . . 246<br />
Prasoon Raj and Axel Klix<br />
Nanofluid Applied Thermo-hydro dynamic<br />
Performance Analysis of Square Array<br />
Subchannel Under PWR Condition. . . . . . . . . . 249<br />
Jubair Ahmed Shamim and Kune Yull Suh<br />
| Computational domain created in Star-CCM+.<br />
KTG Inside . . . . . . . . . . . . . . . . . . . . . . 257<br />
238<br />
| | Liquid Volume fraction distribution.<br />
Decommissioning and Waste Management<br />
The Decommissioning of the ENEA RB3<br />
Research Reactor in Montecuccolino . . . . . . . . 241<br />
F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi,<br />
R. Lorenzelli and A. Rizzo<br />
News . . . . . . . . . . . . . . . . . . . . . . . . . 260<br />
Nuclear Today<br />
Czechs and Balances and Why ‘Ugly’<br />
Nuclear Deserves a Political Makeover . . . . . . . 270<br />
Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 236<br />
AiNT. . . . . . . . . . . . . . . . . . . . . . . . . . . .Insert<br />
AMNT <strong>2018</strong>: Registration Form . . . . . . . . . . . Insert<br />
Contents
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
212<br />
ABSTRACTS | ENGLISH<br />
Euratom: Industry Softens Stance as<br />
Government Lays Out Plans for Transition<br />
NucNet | Page 214<br />
The UK’s nuclear industry has welcomed a government<br />
commitment to continuity with existing<br />
arrangements with Euratom, Europe’s nuclear safety<br />
and research watchdog, a softening of its earlier<br />
stance that the UK needed to stay in the group to<br />
protect vital nuclear research and new-build projects,<br />
and to make sure access to nuclear fuel and<br />
medical isotopes is not disrupted. The next phase of<br />
discussions will focus on the UK’s future relationship<br />
with Euratom. Specific objectives include a close<br />
association with the Euratom Research and Training<br />
Programme, including the Joint European Torus<br />
(JET) and the International Thermonuclear Experimental<br />
Reactor (ITER) projects.<br />
Heat Transfer Systems for Novel<br />
Nuclear Power Plant Designs<br />
Sebastian Vlach, Christoph Fischer and<br />
Herman van Antwerpen | Page 217<br />
This article focuses on designing or modifying heat<br />
exchangers found in the auxiliary systems of any<br />
power plant. The basic premise is to show that the<br />
software provides a one-stop solution for designing<br />
many types of heat transfer systems, where the<br />
interaction bet ween various loops connected by<br />
heat exchangers can be assessed. The nuclear power<br />
plant industry is addressed as the quality control in<br />
the development of the software makes it most<br />
suitable for nuclear related applications. Moreover,<br />
the software discussed has the capability to do<br />
contaminant tracing, which could be very useful<br />
for nuclear contamination studies in designing<br />
specialized ventilation systems. To highlight the<br />
versatility of the software network approach it will<br />
be shown how to model any setup and kind of heat<br />
exchanger such as plate, tube-in-tube, liquid/gas,<br />
finned tube etc. Additionally, the Koeberg pressurized<br />
water reactor steam generator comparison and<br />
the THTR steam generator comparison are shown<br />
as examples.<br />
Experimental and Analytical Tools for<br />
Safety Research of GEN IV Reactors<br />
G. Mazzini, M. Kyncl, Alis Musa and<br />
M. Ruscak | Page221<br />
Current research on nuclear safety in the world, in<br />
addition to supporting existing nuclear power<br />
plants is focused on the more detailed aspects of the<br />
new reactors. The new generation reactors are<br />
expected inter alia to use innovative types of fuel<br />
and new types of coolants, such as e.g. Super-<br />
Critical Water (SCW), supercritical CO 2 , liquid<br />
metals, fluoride salts or high-temperature Helium.<br />
The paper will describe new experimental infrastructure<br />
build recently in Research Centre Řež<br />
under the SUSEN (Sustainable Energy) project and<br />
available analytical tools for supporting safety<br />
research of GEN IV reactors. Two experimental<br />
loops – SCWL (Supercritical Water Loop) and HTHL<br />
(High Temperature Helium Loop) will serve as<br />
in-pile loops in the active core of the research<br />
reactor LVR-15. The paper provides examples of<br />
analyses made using codes ATHLET (supercritical<br />
water) and TRACE (high temperature He) illustrating<br />
process of their assessment and practical use.<br />
Numerical Analysis of MYRRHA Inter-wrapper<br />
Flow Experiment at KALLA<br />
Abdalla Batta and Andreas G. Class | Page 226<br />
The MYRRHA reactor, which is developed at<br />
SCK-SCN in Belgium, represents a multi-purpose<br />
irradiation facility. Its prominent feature is a pool<br />
design with the nuclear core submerged in liquid<br />
metal lead bismuth. During transients between<br />
normal operation and accident conditions decay<br />
heat removal is ensured by forced and natural<br />
convection, respectively. The flow in the gap<br />
between the fuel assemblies plays an important role<br />
in limiting maximum temperatures which should<br />
not be exceeded to avoid core damage. Due to the<br />
scarce database, within the Horizon 2020 – research<br />
and innovation framework program of the EU, the<br />
SESAME project was established to develop and<br />
validate advanced numerical approaches, to<br />
achieve a new or extended validation base and to<br />
establish best practice guidelines including verification<br />
& validation and uncertainty quantification.<br />
In particular the current work supports the<br />
inter-wrapper flow experiment at KALLA.<br />
Heat Balance Analysis for Energy<br />
Conversion Systems of VHTR<br />
SangIL Lee, YeonJae Yoo, Deok Hoon Kye,<br />
Gyunyoung Heo, Eojin Jeon<br />
and Soyoung Park | Page 230<br />
VHTR(Very High Temperature Gas Reactor) with<br />
helium used as a coolant can easily produce heat<br />
required in high-temperature thermochemical process,<br />
and because of low heat output density, the<br />
possibility of core melting is low. In this study, provided<br />
that VHTR is located in the primary system,<br />
the heat conversion system will be discussed in<br />
which hydrogen production and power supply are<br />
possible. In order to control the ratio between power<br />
and hydrogen production, the helium flowing<br />
through nuclear reactor is made to pass through<br />
heat exchanger for hydrogen production and steam<br />
generator or heat exchanger. This study proposes<br />
the whole heat conversion system model, and<br />
carries out thermodynamic feasibility calculation<br />
according to major design variable at each point<br />
and sensitivity analysis for efficiency optimization.<br />
Information Requirements Versus<br />
Confidentiality Obligations – Extension<br />
of the In-Camera Procedure Planned<br />
Tobias Leidinger | Page 235<br />
The justified right of the public to detailed information<br />
on a project requiring nuclear licensing is<br />
opposed by the state’s interest in effective protection<br />
of sensitive data. This conflict is manifested<br />
in licensing procedures but also at court. The differentiated<br />
legal provisions that regulate the balancing<br />
of these conflicting interests are now to be supplemented<br />
by a further facet: An expanded in-camera<br />
trial at court. According to the coalition agreement<br />
of 7 February <strong>2018</strong>, the regulation is to take place in<br />
the current 18th legislative period.<br />
CFD Modeling and Simulation of Heat<br />
and Mass Transfer in Passive Heat<br />
Removal Systems<br />
Amirhosein Moonesi, Shabestary,<br />
Eckhard Krepper and Dirk Lucas | Page 238<br />
The CFD-modelling and simulation of condensation<br />
inside passive heat removal systems are presented.<br />
Designs of future nuclear boiling water reactor concepts<br />
are equipped with emergency cooling systems<br />
which are passive systems for heat removal. The<br />
emergency cooling system consists of slightly<br />
inclined horizontal pipes which are immersed in a<br />
tank of subcooled water. The focus of the project is<br />
on detection of different morphologies such as<br />
annular flow, stratified flow, slug flow and plug flow<br />
and also modeling of the laminar film which is<br />
occurring during the condensation near the wall.<br />
The Decommissioning of the ENEA RB3<br />
Research Reactor in Montecuccolino<br />
F. Rocchi, C. M. Castellani, A. Compagno,<br />
I. Vilardi, R. Lorenzelli and A. Rizzo | Page 241<br />
The ENEA RB3 reactor was a 100 Wth research<br />
installation owned and operated by ENEA, in its<br />
center of Montecuccolino near Bologna, from 1971<br />
to 1989. In 1989, the RB3 reactor was shut down,<br />
and in the late 2010 ENEA received by ministerial<br />
decree the authorization to its dismantling, with the<br />
aim of reaching the “green field” status. This paper<br />
presents the three main pillars of the decommissioning<br />
of RB3, namely the strategy and methods<br />
for the dismantling, the strategy and methods for<br />
the radiological characterization of the building,<br />
and finally the strategy and methods for the radiological<br />
characterization of the site.<br />
Irradiation Tests of a Flat Vanadium<br />
Self-Powered Detector with<br />
14 MeV Neutrons<br />
Prasoon Raj and Axel Klix | Page 246<br />
Self-powered detector (SPD) represents a class of<br />
neutron and gamma monitoring instruments used<br />
in the fission reactor cores worldwide. This detector<br />
has inherent advantages of functioning without a<br />
bias voltage, simple measurement scheme, compactness,<br />
ease of maintenance, and high reliability.<br />
We are studying SPD for application as flux monitors<br />
in the European test blanket modules (TBM) of<br />
ITER, fusion reactor under construction in southern<br />
France.<br />
Nanofluid Applied Thermo-hydrodynamic<br />
Performance Analysis of Square Array<br />
Subchannel Under PWR Condition<br />
Jubair Ahmed Shamim and Kune Yull Suh | Page 249<br />
Efficient engineered design of heat transfer and<br />
fluid flow with enhanced heating or cooling requires<br />
two pivotal aspects that must be taken into consideration<br />
for extracting thermal energy from<br />
nuclear fission reactions in order to save energy,<br />
reduce process time, raise thermal rating and<br />
increase the operating life of a reactor pressure<br />
vessel. Hence, one of the major challenges in<br />
designing a new nuclear power plant is the quantification<br />
of the optimal flow of coolant and distribution<br />
of pressure drop across the reactor core.<br />
Recently, nanofluid has gained much renewed<br />
attention as a promising coolant for pressurized<br />
water reactors (PWRs) due to its enhanced thermal<br />
capabilities with least penalty in pressure drop.<br />
Czechs and Balances and Why ‘Ugly’<br />
Nuclear Deserves a Political Makeover<br />
John Shepherd | Page 270<br />
As if Europe does not have enough on its plate<br />
to deal with at the moment – politically and<br />
economically just for starters – could Brussels be on<br />
a collision course with the Czech government over<br />
the countries plans to expand nuclear energy?<br />
There is certainly friction over the issue between<br />
Prague and the European Commission (EC), to put<br />
it mildly. But why?<br />
The veteran head of the Czech Republic’s State<br />
Office for Nuclear Safety, Dana Drábová, last month<br />
accused other EU member states of “pressurising”<br />
Prague over the early closure of its oldest nuclear<br />
reactor units.<br />
Abstracts | English
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Euratom: Britische Industrie zufrieden<br />
mit Übergangsplänen der Regierung<br />
NucNet | Seite 214<br />
Die britische Nuklearindustrie hat die Zusage der<br />
Regierung begrüßt, die bestehenden Vereinbarungen<br />
mit Euratom, dem europäischen Rahmen<br />
für nukleare Sicherheit und Forschung, aufrechtzuerhalten<br />
und ihren früheren Standpunkt, dass<br />
das Vereinigte Königreich im Euratom-Vertrag<br />
verbleiben müsse, um wichtige Forschungs- und<br />
Neubauprojekte sicherzustellen, und den Zugang<br />
zu Kernbrennstoffen und medizinischen Isotopen<br />
zu gewährleisten, zu relativieren. Die nächste Phase<br />
der Gespräche im Rahmen des Brexit wird sich auf<br />
die künftigen Beziehungen des Vereinigten Königreichs<br />
zu Euratom konzentrieren. Zu den spezifischen<br />
Zielen gehört eine enge Zusammenarbeit<br />
mit den Euratom-Forschungs- und Ausbildungsprogrammen,<br />
einschließlich der Projekte Joint<br />
European Torus (JET) und International Thermonuclear<br />
Experimental Reactor (ITER).<br />
Fortgeschrittene Wärmeübertragungssysteme<br />
für zukünftige Kernkraftwerkskonzepte<br />
Sebastian Vlach, Christoph Fischer und<br />
Herman van Antwerpen | Seite 217<br />
CFD-Systemsimulation mit FlownexSE ermöglicht<br />
es Ingenieuren, einfache und komplexe strömungstechnische<br />
und thermische Netzwerke schnell und<br />
effizient aufzubauen und zu analysieren. Die<br />
Simulation ermöglicht es Ingenieuren, Systeme<br />
aufzubauen, effizient auszulegen und bereits frühzeitig<br />
Schwachstellen in Entwürfen zu finden<br />
sowie geeignete Änderungen und Maßnahmen zu<br />
entwickeln und im Netzwerkmodell zu testen.<br />
Besondere Aufmerksamkeit wird in diesem Artikel<br />
den vielseitigen Möglichkeiten gewidmet, einfache<br />
und komplexe Wärmetauschersysteme der verschiedensten<br />
Arten (Plattenwärmetauscher, Rohrbündel<br />
etc.) für moderne Kernkraftwerke anzuwenden. Als<br />
praktische Beispiele werden gemessene Daten von<br />
den Kraftwerken Koeberg und Hamm-Uentrop mit<br />
den Ergebnissen aus der Simulation verglichen.<br />
Experimentelle und analytische<br />
Werkzeuge für die Sicherheitsforschung<br />
zu GEN-IV-Reaktoren<br />
G. Mazzini, M. Kyncl, Alis Musa und<br />
M. Ruscak | Seite 221<br />
Die aktuelle Forschung zur Sicherheit von<br />
Kernkraftwerken konzentriert sich neben den<br />
Aktivitäten für bestehende Kernkraftwerke auf die<br />
detaillierteren Aspekte neuer Reaktorkonzepte.<br />
Hier werden u.a. innovative Brennstoffe und Kühlmittel<br />
wie z.B. überkritisches Wasser, überkritisches<br />
CO 2 , Flüssigmetalle, Salzschmelzen oder Helium<br />
eingesetzt. Vorgestellt wird dazu die neue experimentelle<br />
Infrastruktur, die im Forschungszentrum<br />
Řež im Rahmen des SUSEN-Projekts (Sustainable<br />
Energy) aufgebaut wurde, sowie die verfügbaren<br />
Analyseinstrumente zur Unterstützung der Sicherheitsforschung<br />
zu GEN IV-Reaktoren.<br />
Numerische Analyse der Zwischenspaltströmung<br />
im MYRRHA-Reaktor mit Ergebnissen<br />
des Strömungsexperiment KALLA<br />
Abdalla Batta und Andreas G. Class | Seite 226<br />
Der am SCK-SCN in Belgien entwickelte MYRRHA-<br />
Reaktor ist eine Mehrzweck-Bestrahlungsanlage.<br />
Sein herausragendes Merkmal ist eine Reaktorkonstruktion<br />
mit einer Kernkühlung aus flüssigem Blei-<br />
Wismut. Bei Transienten zwischen Normalbetrieb<br />
und Unfallbedingungen wird die Wärmeabfuhr<br />
durch erzwungene bzw. natürliche Konvektion<br />
sichergestellt. Die Strömung im Spalt zwischen den<br />
Brennelementen spielt eine wichtige Rolle bei der<br />
Begrenzung von Maximaltemperaturen, die zur<br />
Vermeidung von Kernschäden nicht überschritten<br />
werden sollten. Im Rahmenprogramm Horizon<br />
2020 – Forschung und Innovation der EU wurde<br />
dazu das Projekt SESAME initiiert, um fortgeschrittene<br />
numerische Ansätze zu entwickeln und<br />
zu validieren, die eine neue oder erweiterte<br />
Validierungsbasis für damit verbundene Fragestellungen<br />
zur Verfügung stellen.<br />
Wärmebilanzanalyse für<br />
Energieumwandlungssysteme von VHTR<br />
SangIL Lee, YeonJae Yoo, Deok Hoon Kye,<br />
Gyunyoung Heo, Eojin Jeon und<br />
Soyoung Park | Seite 230<br />
VHTR (Very High Temperature Gas Reactor) mit<br />
Helium als Kühlmittel können Wärme bereit stellen,<br />
die bei thermochemischen Hochtemperaturprozessen<br />
benötigt wird. In Bezug auf die Sicherheit ist<br />
aufgrund der geringen Wärmeleistungsdichte das<br />
Risiko einer Kernschmelze minimiert. Diskutiert<br />
werden Voraussetzungen für die Nutzung von<br />
VHTR für eine Wasserstofferzeugung und Stromversorgung.<br />
Vorgestellt wird ein Gesamtmodell des<br />
Wärmeumwandlungssystems mit einer thermodynamischen<br />
Machbarkeitsberechnung.<br />
Informationsbedarf versus<br />
Geheimhaltungspflichten – Erweiterung<br />
des In camera-Verfahrens geplant<br />
Tobias Leidinger | Seite 235<br />
Dem berechtigten Anspruch der Öffentlichkeit auf<br />
detaillierte Informationen über ein atomrechtlich<br />
genehmigungsbedürftiges Vorhaben steht das<br />
staatliche Interesse an einem effektiven Geheimnisschutz<br />
sensibler Daten gegenüber. Dieser Konflikt<br />
tritt regelmäßig im Genehmigungsverfahren aber<br />
auch vor Gericht zu Tage. Die differenzierten<br />
Gesetzesbestimmungen, die den Ausgleich dieser<br />
widerstreitenden Interessen regeln, sollen nun<br />
durch eine weitere Facette ergänzt werden: Ein<br />
erweitertes In-camera-Verfahren bei Gericht. Nach<br />
dem Koalitionsvertrag vom 7. Februar <strong>2018</strong> soll die<br />
Regelung in der schon laufenden 18. Legislaturperiode<br />
erfolgen.<br />
CFD-Modellierung und Simulation<br />
von Wärme- und Stoffaustausch<br />
in passiven Wärmeabfuhrsystemen<br />
Amirhosein Moonesi, Shabestary,<br />
Eckhard Krepper und Dirk Lucas | Seite 238<br />
Die CFD-Modellierung und Simulation der Kondensation<br />
in passiven Wärmeabfuhrsystemen wird vorgestellt.<br />
Zukünftige Siedewasserreaktorkonzepte<br />
werden mit Notkühlsystemen ausgestattet, die eine<br />
passive Wärmeabfuhr gewährleisten. Das Notkühlsystem<br />
besteht aus leicht geneigten horizontalen<br />
Rohren in einem Wasserbehälter. Der Schwerpunkt<br />
des vorgestellten Projektes liegt auf der Identifikation<br />
verschiedener Morphologien wie Ringströmung,<br />
Schichtenströmung, Schwallströmung<br />
und Pfropfenströmung sowie der Modellierung des<br />
laminaren Films, der bei der Kondensation in<br />
Wandnähe auftritt.<br />
Die Stilllegung der ENEA RB3<br />
Forschungsreaktor in Montecuccolino<br />
F. Rocchi, C. M. Castellani, A. Compagno,<br />
I. Vilardi, R. Lorenzelli und A. Rizzo | Seite 241<br />
Der ENEA RB3-Reaktor war eine 100-Watt-Forschungsanlage,<br />
die von 1971 bis 1989 im Zentrum<br />
von Montecuccolino bei Bologna, Italien betrieben<br />
wurde. 1989 wurde der RB3-Reaktor abgeschaltet<br />
und Ende 2010 erhielt ENEA per Ministerialerlass<br />
die Genehmigung zu seinem Rückbau mit dem Ziel,<br />
den Status „Grünen Wiese“ zu erreichen. Vorgestellt<br />
werden die drei wesentlichen Fragestellungen für<br />
die Stilllegung des RB3: Strategie und Methoden<br />
für den Rückbau, Strategie und Methoden für die<br />
radiologische Charakterisierung des Gebäudes und<br />
schließlich die Strategie und Methoden für die<br />
radiologische Charakterisierung des Standortes.<br />
Bestrahlungstests eines Vanadium-<br />
Detektors mit 14 MeV Neutronen<br />
Prasoon Raj und Axel Klix| | Seite 246<br />
Self-powered Detektoren (SPD) sind eine Klasse<br />
von Neutronen- und Gamma-Überwachungsgeräten,<br />
die weltweit in Kernreaktoren eingesetzt<br />
werden. Diese Detektoren besitzen die Vorteile,<br />
dass keine Spannungsversorgung erforderlich ist,<br />
das Messverfahren einfach und die Detektoreinheit<br />
kompakt, wartungsfreundlich und zuverlässig ist.<br />
SPDs werden im Rahmen des vorgestellten Projektes<br />
für den Einsatz als Flussmonitor in den Blanketmodulen<br />
des in Bau befindlichen Fusionsreaktors<br />
ITER .<br />
Einsatz von Nanofluiden und<br />
thermohydraulische Analyse<br />
für Druckwasserreaktoren<br />
Jubair Ahmed Shamim und Kune Yull Suh | Seite 249<br />
Eine effiziente Auslegung von Wärmeübertragung<br />
und Flüssigkeitsströmung mit verbessertem<br />
Wärme übergang, -transport oder Kühlung bedingt<br />
zwei zentrale Aspekte, die in Kernkraftwerken<br />
berücksichtigt werden müssen: Leistungsdichte<br />
und technische Lebensdauer des Reaktordruckbehälters.<br />
Eine Herausforderung für die Auslegung<br />
neuer Kernkraftwerkskonzepte ist daher die<br />
Quantifizierung einer optimalen Kühlmittelverteilung<br />
und die Verteilung des Druckverlustes über<br />
den Reaktorkern. In jüngster Zeit werden „Nanofluide“<br />
als vielversprechendes Kühlmittel für Druckwasserreaktoren<br />
(DWR) aufgrund verbesserter<br />
thermischer Eigenschaften mit geringst möglichem<br />
Druckabfall diskutiert, die auch Thema dieser<br />
Arbeit sind..<br />
Tschechien und Ausgewogenheit und<br />
warum es die „hässliche“ Kernenergie verdient,<br />
politisch neu bewertet zu werden<br />
John Shepherd | Seite 270<br />
Als ob Europa derzeit nicht genug zu tun hätte, mit<br />
sich selbst – politisch und wirtschaftlich, nur um<br />
zwei Themenbereiche zu nennen – ... könnte jetzt<br />
Brüssel auf Kollisionskurs mit der tschechischen<br />
Regierung zu den Plänen des Landes zum Ausbau<br />
der Kernenergie gehen?<br />
In der Frage zwischen Prag und der Europäischen<br />
Kommission (EC) geht es, um es milde auszudrücken,<br />
sicherlich um Differenzen. Aber warum?<br />
Die langjährige Leiterin der tschechischen<br />
Aufsichtsbehörde für nukleare Sicherheit, Dana<br />
Drábová, warf zudem im vergangenen Monat<br />
anderen EU-Mitgliedstaaten vor, die Regierung in<br />
Prag unter inakzeptablem Druck zu setzen hinsichtlich<br />
der Forderung einer vorzeitigen Stilllegung<br />
ihrer ältesten Kernkraftwerke.<br />
213<br />
ABSTRACTS | GERMAN<br />
Abstracts | German
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
214<br />
INSIDE NUCLEAR WITH NUCNET<br />
Euratom: Industry Softens Stance as<br />
Government Lays Out Plans for Transition<br />
NucNet<br />
The UK’s nuclear industry has welcomed a government commitment to continuity with existing<br />
arrangements with Euratom, Europe’s nuclear safety and research watchdog, a softening of its earlier stance<br />
that the UK needed to stay in the group to protect vital nuclear research and new-build projects, and to make<br />
sure access to nuclear fuel and medical isotopes is not disrupted.<br />
Energy secretary Greg Clark said in a written statement to<br />
parliament on 11 January <strong>2018</strong> that the government wants<br />
to include Euratom in any implementation period agreed<br />
as part of wider discussions on Brexit and plans to put in<br />
place “all the necessary measures” to ensure that the UK<br />
can operate as an independent and responsible nuclear<br />
state from day one of Brexit and its separation from the<br />
Euratom Treaty, which regulates the nuclear industry and<br />
the movement of nuclear material across Europe.<br />
According to Mr Clark’s statement, the government has<br />
made good progress on separation issues in the last few<br />
months as part of phase one of negotiations with the EU.<br />
Negotiations have covered a set of legal and technical<br />
issues related to nuclear material and waste, and safeguards<br />
obligations and equipment.<br />
The next phase of discussions will focus on the UK’s<br />
future relationship with Euratom. Specific objectives<br />
include a close association with the Euratom Research and<br />
Training Programme, including the Joint European Torus<br />
(JET) and the International Thermonuclear Experimental<br />
Reactor (ITER) projects.<br />
For the nuclear industry, rapid departure from Euratom<br />
without a clear replacement spells disaster. Scientists have<br />
warned that British nuclear stations may not be able to<br />
source nuclear fuel if it cannot be legally transported<br />
across borders. The shipment of medical isotopes used in<br />
scans and cancer treatment might be jeopardised.<br />
European workers on shared research projects, such as<br />
experimental fusion reactors, face an equally uncertain<br />
future without Euratom’s separate guarantees of freedom<br />
of movement.<br />
But the London-based Nuclear Industry Association<br />
(NIA), which represents more than 260 nuclear companies,<br />
cautiously welcomed Mr Clark’s statement, calling it a<br />
“useful and welcome step” in setting out the government’s<br />
approach in seeking to secure equivalent arrangements to<br />
those the UK benefits from as a member of Euratom. The<br />
NIA also welcomed clarity on the government’s intention<br />
to negotiate an implementation period to ensure a smooth<br />
transition from the current to new arrangements.<br />
It warned, however, that there is much still to do in<br />
equipping the UK’s regulator to take on Euratom’s safeguarding<br />
activities. The UK needs to reach post-Euratom<br />
agreements with the International Atomic Energy Agency,<br />
the US, Canada, Australia, Japan and others. It needs to<br />
agree new trading arrangements with the Euratom<br />
community and conclude a new funding agreement for the<br />
UK to continue its work in Euratom’s fusion R&D activities.<br />
“It is vital government continues to prioritise these issues<br />
in the period ahead if there is to be a successful outcome,”<br />
the NIA said.<br />
Unlike the dozens of other regulatory arrangements for<br />
industries such as aviation or pharmaceuticals, Euratom<br />
has been singled out for special treatment through the<br />
Brexit process because it is not technically part of the EU.<br />
Instead, the treaty that established this body to coordinate<br />
Europe’s civil nuclear energy industry was born in parallel<br />
with the birth of the European economic community in<br />
1957. The UK’s participation in Euratom therefore required<br />
a separate legal relationship with the European court of<br />
justice to enforce it.<br />
The nuclear industry had been hoping that because of<br />
this separation from the “mainstream Brexit,” the UK<br />
might decide to remain part of Euratom.<br />
The NIA and the Brussels-based trade body Foratom<br />
both said the UK should maintain its membership. They<br />
argued that the nuclear industry is global, and the ease of<br />
movement of nuclear goods, people and services enables<br />
new build, decommissioning, R&D and other programmes<br />
of work to continue without interruption.<br />
The government insists that leaving Euratom is an<br />
inevitable consequence of Brexit – a position shared by the<br />
European negotiators. But is says it wants continuity of<br />
open trade arrangements for nuclear goods and products<br />
to ensure the nuclear industry is able to continue to trade<br />
across EU borders without disruption.<br />
Support for remaining in Euratom had come not only<br />
from within the industry, but also from politicians.<br />
Conservative MPs said they would for the government to<br />
fight harder for the UK to stay in Euratom. The opposition<br />
Labour Party said Britain should remain in Euratom,<br />
adding it is increasingly clear that the government acted<br />
“recklessly” by giving up on membership.<br />
Scientists said leaving Euratom will cause widespread<br />
confusion and have a potentially devastating impact<br />
on the nuclear industry. They warned of potential problems<br />
related to the transportation of nuclear materials, including<br />
nuclear fuel; research, especially fusion research; and<br />
overseas investment in development of British nuclear<br />
power stations.<br />
Mr Clark’s statement addressed another concern for the<br />
industry – the issue of accessing a skilled pan-European<br />
workforce for the sector once Brexit is complete.<br />
Mr Clark said the nuclear sector needs the workforce for<br />
decommissioning, operation of existing facilities and<br />
new-build projects. He said proposals for the UK’s future<br />
immigration system will be set out shortly and “we will<br />
ensure that those businesses and communities, and<br />
parliament have the opportunity to contribute their views<br />
before making any decisions about the future system”.<br />
Whatever the outcome of negotiations with the EU,<br />
it is vital that the civil nuclear industry has a safeguards<br />
regime that meets international standards. But this<br />
is not dependent on the EU negotiations and the UK<br />
government is well advanced in delivering this plan, the<br />
statement said.<br />
Inside Nuclear with NucNet<br />
Euratom: Industry Softens Stance as Government Lays Out Plans for Transition ı NucNet
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Advertisement<br />
The UK is establishing a legislative and regulatory<br />
framework for a domestic safeguards regime which will<br />
provide legal powers to establish a domestic regime which<br />
the Office for Nuclear Regulation will regulate. It is also<br />
negotiating bilateral safeguards agreements with the<br />
International Atomic Energy Agency and putting in place<br />
bilateral nuclear cooperation agreements with key third<br />
countries.<br />
NIA chief executive Tom Greatrex said the UK industry<br />
and research facilities have been consistently clear with<br />
government about the importance of these issues since<br />
the referendum. “Even with a suitable transition, there<br />
remains much work for the government to do to prevent<br />
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Mr Clark’s statement is online:<br />
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Fachseminar Nuklearhaftung<br />
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2 Achim Jansen-Tersteegen ı Geschäftsführer,<br />
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DATF EDITORIAL NOTES<br />
215<br />
Author<br />
NucNet<br />
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David Dalton<br />
Editor in Chief, NucNet<br />
Avenue des Arts 56<br />
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www.nucnet.org<br />
Am 5. September <strong>2018</strong> in Leipzig<br />
Ï Information und Anmeldung: www.conlar.de<br />
Rechtsanwaltskanzlei Dr. Christian Raetzke<br />
Beethovenstraße 19 · 04107 Leipzig<br />
Tel. 0341 – 9999 1444<br />
christian.raetzke@conlar.de · www.conlar.de<br />
Notes<br />
Conlar <strong>atw</strong> 18-04 75x124.indd 1 18.03.18 15:02<br />
Grafik des Monats<br />
Bundesministerium für Umwelt, Naturschutz, Bau und Reaktorsicherheit (BMUB)<br />
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Beteiligungsverwaltung<br />
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für Strahlenschutz (BfS)<br />
Landesministerien<br />
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www.kernenergie.de<br />
DAtF Notes
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
216<br />
CALENDAR<br />
Calendar<br />
<strong>2018</strong><br />
08.04.-11.04.<strong>2018</strong><br />
International Congress on Advances in Nuclear<br />
Power Plants – ICAPP 18. Charlotte, NC, USA,<br />
American Nuclear Society (ANS), www.ans.org<br />
08.04.-13.04.<strong>2018</strong><br />
11 th International Conference on Methods and<br />
Applications of Radioanalytical Chemistry –<br />
MARC XI. Kailua-Kona, HI, USA, American Nuclear<br />
Society (ANS), www.ans.org<br />
12.04.<strong>2018</strong><br />
Desalination Powered by Nuclear Energy. Essen,<br />
Germany, Deutsche Meerwasser Entsalzung GmbH<br />
in cooperation with International Atomic Energy<br />
Agency (IAEA) and PowerTech Training Center<br />
( Kraftwerksschule, KWS), www.dme-gmbh.de,<br />
www.iaea.org, www.kraftwerksschule.de<br />
16.04.-19.04.<strong>2018</strong><br />
Einführung in die Kerntechnik. Mannheim,<br />
Germany, TÜV SÜD, nucleartraining@tuev-sued.de<br />
16.04.-17.04.<strong>2018</strong><br />
VdTÜV Forum Kerntechnik – Sicherheit im Fokus.<br />
Berlin, Germany, VdTÜV mit Unterstützung des<br />
TÜV NORD, des TÜV SÜD und des TÜV Rheinland,<br />
www.tuev-sued.de/tagungen<br />
17.04.-19.04.<strong>2018</strong><br />
World Nuclear Fuel Cycle <strong>2018</strong>. Madrid, Spain,<br />
World Nuclear Association (WNA),<br />
www.world-nuclear.org<br />
18.04.-19.04.<strong>2018</strong><br />
9. Symposium zur Endlagerung radioaktiver<br />
Abfälle. Vorbereitung auf KONRAD – Wege zum<br />
G2-Gebinde. Hanover, Germany, TÜV NORD<br />
Akademie, www.tuev-nord.de/tk-era<br />
22.04.-26.04.<strong>2018</strong><br />
Reactor Physics Paving the Way Towards More<br />
Efficient Systems – PHYSOR <strong>2018</strong>. Cancun, Mexico,<br />
www.physor<strong>2018</strong>.mx<br />
24.04.-25.04.<strong>2018</strong><br />
Integrated Waste Management Conference.<br />
Penrith, Cumbria, United Kingdom, The Nuclear<br />
Institute, www.iwmeurope.com<br />
08.05.-10.05.<strong>2018</strong><br />
29 th Conference of the Nuclear Societies in Israel.<br />
Herzliya, Israel. Israel Nuclear Society and Israel<br />
Society for Radiation Protection, ins-conference.com<br />
13.05.-19.05.<strong>2018</strong><br />
BEPU-<strong>2018</strong> – ANS International Conference on<br />
Best-Estimate Plus Uncertainties Methods. Lucca,<br />
Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,<br />
ANS, IAEA, NEA, www.nineeng.com/bepu/<br />
13.05.-18.05.<strong>2018</strong><br />
RadChem <strong>2018</strong> – 18th Radiochemical Conference.<br />
Marianske Lazne, Czech Republic,<br />
www.radchem.cz<br />
14.05.-16.05.<strong>2018</strong><br />
ATOMEXPO <strong>2018</strong>. Sochi, Russia,<br />
atomexpo.ru<br />
15.05.-17.05.<strong>2018</strong><br />
11 th International Conference on the Transport,<br />
Storage, and Disposal of Radioactive Materials.<br />
London, United Kingdom, Nuclear Institute,<br />
www.nuclearinst.com<br />
20.05.-23.05.<strong>2018</strong><br />
5 th Asian and Oceanic IRPA Regional Congress<br />
on Radiation Protection – AOCRP5. Melbourne,<br />
Australia, Australian Radiation Protection Society<br />
(ARPS) and International Radiation Protection<br />
Association (IRPA), www.aocrp-5.org<br />
29.05.-30.05.<strong>2018</strong><br />
49 th Annual Meeting on Nuclear Technology<br />
AMNT <strong>2018</strong> | 49. Jahrestagung Kerntechnik.<br />
Berlin, Germany, DAtF and KTG,<br />
www.nucleartech-meeting.com<br />
03.06.-07.06.<strong>2018</strong><br />
38 th CNS Annual Conference and 42nd CNS-CNA<br />
Student Conference. Saskotoon, SK, Canada,<br />
Candian Nuclear Society CNS, www.cns-snc.ca<br />
03.06.-06.06.<strong>2018</strong><br />
HND<strong>2018</strong> 12 th International Conference of the<br />
Croatian Nuclear Society. Zadar, Croatia, Croatian<br />
Nuclear Society, www.nuklearno-drustvo.hr<br />
04.06.-05.06.<strong>2018</strong><br />
13 th European Nuclear Energy Forum. Bratislava,<br />
Slova Republic, European Commission, ec.europa.eu<br />
04.06.-07.06.<strong>2018</strong><br />
10 th Symposium on CBRNE Threats. Rovaniemi,<br />
Finland, Finnish Nuclear Society, ats-fns.fi<br />
04.06.-08.06.<strong>2018</strong><br />
5 th European IRPA Congress – Encouraging<br />
Sustainability in Radiation Protection.<br />
The Hague, The Netherlands, Dutch Society<br />
for Radiation Protection (NVS), local organiser,<br />
irpa<strong>2018</strong>europe.com<br />
06.06.-08.06.<strong>2018</strong><br />
2 nd Workshop on Safety of Extended Dry Storage<br />
of Spent Nuclear Fuel. Garching near Munich,<br />
Germany, GRS, www.grs.de<br />
25.06.-26.06.<strong>2018</strong><br />
index<strong>2018</strong> – International Nuclear Digital<br />
Experience. Paris, France, Société Française d’Energie<br />
Nucléaire, www.sfen.org, www.sfen-index<strong>2018</strong>.org<br />
27.06.-29.06.<strong>2018</strong><br />
EEM – <strong>2018</strong> 15 th International Conference on the<br />
European Energy Market. Lodz, Poland, Lodz<br />
University of Technology, Institute of Electrical Power<br />
Engineering, Association of Polish Electrical<br />
Engineers (SEP), www.eem18.eu<br />
24.06.-30.06.<strong>2018</strong><br />
ANNETTE Summer School on Nuclear Technology,<br />
Nuclear Waste Management and Radiation<br />
Protection. Turku, Finland, Advanced Networking<br />
for Nuclear Education, Training and Transfer of<br />
Expertise, annettesummerschool.org, www.enen.eu<br />
29.07.-02.08.<strong>2018</strong><br />
International Nuclear Physics Conference 2019.<br />
Glasgow, United Kingdom, www.iop.org<br />
22.08.-31.08.<strong>2018</strong><br />
Frédéric Joliot/Otto Hahn (FJOH) Summer School<br />
FJOH-<strong>2018</strong> – Maximizing the Benefits of<br />
Experiments for the Simulation, Design and<br />
Analysis of Reactors. Aix-en-Provence, France,<br />
Nuclear Energy Division of Commissariat à l’énergie<br />
atomique et aux énergies alternatives (CEA)<br />
and Karlsruher Institut für Technologie (KIT),<br />
www.fjohss.eu<br />
28.08.-31.08.<strong>2018</strong><br />
TINCE <strong>2018</strong> – Technological Innovations in<br />
Nuclear Civil Engineering. Paris Saclay, France,<br />
Société Française d’Energie Nucléaire, www.sfen.org,<br />
www.sfen-tince<strong>2018</strong>.org<br />
05.09.-07.09.<strong>2018</strong><br />
World Nuclear Association Symposium <strong>2018</strong>.<br />
London, United Kingdom, World Nuclear Association<br />
(WNA), www.world-nuclear.org<br />
09.09.-14.09.<strong>2018</strong><br />
21 st International Conference on Water<br />
Chemistry in Nuclear Reactor Systems.<br />
San Francisco, CA, USA, EPRI – Electric Power<br />
Research Institute, www.epri.com<br />
17.09.-21.09.<strong>2018</strong><br />
62 nd IAEA General Conference. Vienna, Austria.<br />
International Atomic Energy Agency (IAEA),<br />
www.iaea.org<br />
17.09.-20.09.<strong>2018</strong><br />
FONTEVRAUD 9. Avignon, France,<br />
Société Française d’Energie Nucléaire (SFEN),<br />
www.sfen-fontevraud9.org<br />
17.09.-19.09.<strong>2018</strong><br />
4 th International Conference on Physics and<br />
Technology of Reactors and Applications –<br />
PHYTRA4. Marrakech, Morocco, Moroccan<br />
Association for Nuclear Engineering and Reactor<br />
Technology (GMTR), National Center for Energy,<br />
Sciences and Nuclear Techniques (CNESTEN) and<br />
Moroccan Agency for Nuclear and Radiological<br />
Safety and Security (AMSSNuR), phytra4.gmtr.ma<br />
26.09.-28.09.<strong>2018</strong><br />
44 th Annual Meeting of the Spanish Nuclear<br />
Society. Avila, Spain, Sociedad Nuclear Española,<br />
www.sne.es<br />
30.09.-04.10.<strong>2018</strong><br />
TopFuel <strong>2018</strong>. Prague, Czech Republic, European<br />
Nuclear Society (ENS), American Nuclear Society<br />
(ANS). Atomic Energy Society of Japan, Chinese<br />
Nuclear Society and Korean Nuclear Society,<br />
www.euronuclear.org<br />
02.10.-04.10.<strong>2018</strong><br />
7 th EU Nuclear Power Plant Simulation ENPPS<br />
Forum. Birmingham, United Kingdom, Nuclear<br />
Training & Simulation Group, www.enpps.tech<br />
14.10.-18.10.<strong>2018</strong><br />
12 th International Topical Meeting on Nuclear<br />
Reactor Thermal-Hydraulics, Operation and<br />
Safety – NUTHOS-12. Qingdao, China, Elsevier,<br />
www.nuthos-12.org<br />
14.10.-18.10.<strong>2018</strong><br />
NuMat <strong>2018</strong>. Seattle, United States,<br />
www.elsevier.com<br />
16.10.-17.10.<strong>2018</strong><br />
4 th GIF Symposium at the 8 th edition of Atoms<br />
for the Future. Paris, France, www.gen-4.org<br />
22.10.-24.10.<strong>2018</strong><br />
DEM <strong>2018</strong> Dismantling Challenges: Industrial<br />
Reality, Prospects and Feedback Experience. Paris<br />
Saclay, France, Société Française d’Energie Nucléaire,<br />
www.sfen.org, www.sfen-dem<strong>2018</strong>.org<br />
22.10.-26.10.<strong>2018</strong><br />
NUWCEM <strong>2018</strong> Cement-based Materials for<br />
Nuclear Waste. Avignon, France, French<br />
Commission for Atomic and Alternative Energies<br />
and Société Française d’Energie Nucléaire,<br />
www.sfen-nuwcem<strong>2018</strong>.org<br />
24.10.-25.10.<strong>2018</strong><br />
Chemistry in Power Plant. Magdeburg, Germany,<br />
VGB PowerTech e.V., www.vgb.org<br />
05.11.-08.11.<strong>2018</strong><br />
International Conference on Nuclear<br />
Decom missioning – ICOND <strong>2018</strong>. Aachen,<br />
Eurogress, Germany, achen Institute for Nuclear<br />
Training GmbH, www.icond.de<br />
2019<br />
07.05.-08.05.2019<br />
50 th Annual Meeting on Nuclear Technology<br />
AMNT 2019 | 50. Jahrestagung Kerntechnik.<br />
Berlin, Germany, DAtF and KTG,<br />
www.nucleartech-meeting.com<br />
Calendar
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Heat Transfer Systems for Novel Nuclear<br />
Power Plant Designs<br />
Sebastian Vlach, Christoph Fischer and Herman van Antwerpen<br />
This article focuses on work that involves designing or modifying heat exchangers that usually can be found in the auxiliary<br />
systems of any power plant. The basic premise of the article is to show that the software provides a one-stop solution for<br />
designing many types of heat transfer systems, where the interaction between various loops connected by heat exchangers can<br />
be assessed. This article especially addresses the audience among nuclear power plants as the quality control in the development<br />
of the software makes it most suitable for nuclear related work. Moreover, the software discussed in this article has the<br />
capability to do contaminant tracing, which could be very useful for nuclear contamination studies in designing specialized<br />
ventilation systems. To highlight the versatility of the software network approach it will be shown how to model any setup and<br />
kind of heat exchanger such as plate, tube-in-tube, liquid/gas, finned tube etc. Additionally, the Koeberg pressurized water<br />
reactor (PWR) steam generator comparison and the Hamm-Uentrop thorium high temperature reactor (THTR) steam<br />
generator comparison are shown as practical examples.<br />
Introduction “Every type of technology benefits from advances inspired by new knowledge and understanding.<br />
Although nuclear energy has operated mostly safely in the past, nuclear engineers do continue to devise new ideas for<br />
making nuclear energy even safer and more secure. The future of reliable nuclear energy requires scientific research to<br />
verify that new types of advanced nuclear fuels and materials are robust enough to withstand the conditions inside a<br />
nuclear reactor during normal and abnormal conditions.” (Idaho National Laboratory).<br />
217<br />
OPERATION AND NEW BUILD<br />
Based on the laws of thermodynamics 1D system<br />
simulation is extremely robust, fast, and reliable. One<br />
software package for 1D system simulation that gains<br />
more and more attention recently was developed in the<br />
early 1990ies by a South African company, namely M-Tech<br />
Industrial. Initially, Flownex® Simulation Environment<br />
was developed for aerospace applications and the energy<br />
sector. Moreover, nuclear validation and verification were<br />
supervised by the governmental ESKOM institution<br />
through its subsidiary PBMR Ltd., who developed a<br />
high-temperature gas-cooled (pebble-bed) reactor in<br />
cooperation with Jülich Research Centre at that time.<br />
Specifically for the nuclear safety analyses required by<br />
PBMR, the software has Nuclear Quality Assurance<br />
( NQA-1) Certification and its development process is<br />
based on ISO 9001.<br />
System simulation programmes provide engineers and<br />
designers a fast and efficient way to set up simulation<br />
models for simple as well as complex fluid dynamic<br />
networks. Such networks commonly contain several<br />
components such as fans, pumps, heat exchangers etc. that<br />
can be computed almost instantly. Furthermore, dynamics<br />
and the control of such networks can be investigated by<br />
running different operation scenarios, such as start-up,<br />
shut down, and various loading conditions, where steady<br />
state and transient effects are taken into account. Thus,<br />
weak spots within a system can be eliminated during the<br />
design process prior to manufacturing as literally any<br />
modification can be tested virtually.<br />
Subsequently, the user is able to analyse the results very<br />
quickly.<br />
Material data that the software supports can be<br />
gaseous, gas mixtures, as well as incompressible pure<br />
fluids and two-phase pure fluids. The user is able to access<br />
a vast library based on the NIST data base. Hence, complex<br />
flows can be modelled using temperature and pressure<br />
dependent material data as well as multiphase effects like<br />
conden sation, evaporation, and cavitation.<br />
The software is equipped with a vast array of components<br />
that cover most required simulation scenarios.<br />
Those components can be used as single components or as<br />
building blocks of components found in thermal fluid<br />
systems or subsystems.<br />
Building blocks, with various levels of detail are<br />
available to model heat transfer phenomena as shown in<br />
Figure 1. Some of the simple heat exchanger models<br />
utilises the Number of Transfer Units (NTU) Method while<br />
other more complex versions employ a fully discretised<br />
approach to heat exchanger modelling. The heat exchanger<br />
types range from tube to plate heat exchangers that can be<br />
modelled as parallel, counter, or cross flow types. Other<br />
components can be vessels, reactors, tube systems, valves,<br />
pumps, fans, compressors, seals etc. Moreover, a whole<br />
library of com ponents for dynamics and control is available<br />
within the software.<br />
1D System Simulation<br />
Flownex® Simulation Environment includes all the<br />
necessary numerical formulations for solving all important<br />
thermo-fluid physical phenomena and moreover, a modern<br />
Windows-GUI that enables an intuitive and easy interaction<br />
for the user. Therefore, the user can concentrate on<br />
design and optimisation rather than on the complexities<br />
usually associated with operating such calculation software.<br />
Typical simulations are run in real time or in the<br />
order of seconds, which makes parameter studies and<br />
optimisation loops extremely fast and very efficient.<br />
| | Fig. 1.<br />
Library for heat exchangers [1].<br />
| | Fig. 2.<br />
Heat transfer library [1].<br />
Operation and New Build<br />
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 218<br />
| | Fig. 3.<br />
Plate heat exchanger model with a two-pass hot side and a single-pass cold side (left) and<br />
schematic (right) [2].<br />
If one is looking into thermo dynamic analyses, simple<br />
components would be used to represent radiation,<br />
conduction, or convection as shown in Figure 2. Thus,<br />
heat exchangers can be custom-built to answer the<br />
question at hand.<br />
Figure 3 shows a simple custom made plate heat<br />
exchanger consisting of composite heat transfer components<br />
and pipe components. The flow path is represented<br />
with a hydraulic diameter and the flow area. The plates are<br />
represented with heat transfer area and actual metal<br />
thickness. User-specified correlations according to the<br />
plate corrugation profiles are defined allowing for full<br />
discretisation along the flow path that results in accurate<br />
pinch-point calculation and transient response.<br />
Another heat exchanger example is shown in Figure 4<br />
where a finned tube air-water heat exchanger can be seen.<br />
The air-side is modelled as a straight-through flow path<br />
(left to right) whereas the water-side is modelled as an<br />
up-down overall counter flow (right to left) configuration<br />
according to the design of the header box plates. The fully<br />
discretised flow path provides an accurate transient<br />
response. The fin-side pressure drop and heat transfer<br />
correlations can be specified with Chilton-Colburn J-factor<br />
tables.<br />
Heat exchangers are crucial for any power plant design.<br />
Figure 5 shows the schematic of the Koeberg PWR steam<br />
generator and the equivalent model built in the software.<br />
For the dryer/separator a complete phase separation is<br />
assumed. The recirculation flow rate is calculated from<br />
buoyancy-driven flow (red circle) that is dependent on<br />
heat transfer coefficient and flow resistance. The model<br />
also assumes a homogeneous two-phase flow. The Chen<br />
correlation for the shell side was implemented with<br />
scripting. Specific material properties can be implemented<br />
via Engineering Equation Solver (EES) coupling or<br />
scripting if necessary.<br />
Table 1 shows the comparison of measured and<br />
simulated data of the Koeberg PWR (South Africa) steam<br />
generator at 60 % and 100 % power load. The software<br />
shows reasonably good agreement to the measured data,<br />
especially when looking at the recirculation ratio R circ<br />
which is a good indication of the overall calculation<br />
accuracy.<br />
Another power plant example is the Hamm-Uentrop<br />
thorium high temperature reactor (THTR-300, Germany)<br />
power plant. One challenge in modelling the THTR is that<br />
at certain combinations of flow rate and heat input, the<br />
flow could be oscillatory. Several types of oscillation are<br />
possible: density wave, pressure wave, and critical heat<br />
flux (dryout)-related oscillations. Fundamental fluiddynamic<br />
modelling is crucial to detect this, which is<br />
provided in the software. Furthermore, this capability is<br />
critical to determine the minimum flow through a steam<br />
generator because it is typically at low power levels that<br />
the steam flow becomes oscillatory. Figure 6 shows the<br />
schematic of the THTR-300 and an equivalent model built<br />
in the software.<br />
The THTR steam generator plant was modelled to<br />
verify the steady-state performance of the assembled<br />
steam generator model. Figure 7 shows the comparison of<br />
measured and simulated data of the THTR-300 at 40 %<br />
and 100 % power load. The software shows very good<br />
agreement to the measured data. The simulation revealed<br />
| | Fig. 4.<br />
Model of a finned tube air-water heat exchanger with multiple water-side passes and a single air pass (top) and schematic (bottom) [2].<br />
T pi<br />
[C]<br />
T po<br />
p so<br />
[kPa]<br />
p si T si x so ṁ s<br />
[kg/s]<br />
Q boiler<br />
[MW]<br />
R circ<br />
60 % power Koeberg 294 273 4889 5055 195 1.0 341 670 7.0<br />
Simulation 294 273 4889 4919 195 1.0 341 666 6.4<br />
100 % power Koeberg 312 279 4911 5277 220 1.0 618 1143 3.8<br />
Simulation 312 280 4911 4951 220 1.0 618 1092 3.8<br />
| | Tab. 1.<br />
Koeberg PWR steam generator comparison.<br />
Operation and New Build<br />
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
that the helium-side heat transfer correlation needed to<br />
have an appropriate Reynolds-number dependence as the<br />
error became quite large at lower power or flow levels<br />
neglecting this.<br />
As aforementioned, heat exchangers are crucial for<br />
any power plant design, especially when designing new<br />
power plants. In addition to the heat transfer modelling<br />
capabilities and with respect to nuclear power generation<br />
the software has recently expanded the Generic Nuclear<br />
Reactor model to simulate the latest nuclear reactor<br />
designs of any geometry. Novel nuclear reactor designs<br />
include liquid fuel reactors, liquid-metal-cooled reactors,<br />
and high temperature gas-cooled reactors (HTGR). In<br />
more detail, there are six reactor types that have gained<br />
researches interest all over the world:<br />
• Very High Temperature Reactor,<br />
• Molten Salt Reactor,<br />
• Sodium-Cooled Fast Reactor,<br />
• Supercritical-Water-Cooled Reactor,<br />
• Gas-Cooled Fast Reactor, and<br />
• Lead-Cooled Fast Reactor.<br />
The new “generalized fuel zone” in the GNR model that is<br />
shown in Figure 8 is capable of handling any fuel geometry<br />
and any fluid type. It expands the geometry capability to<br />
plate fuel, cylindrical fuel rods, spherical fuel elements,<br />
irregular cross-section fuel (like the four-lobe cross-shape<br />
produced by the Lightbridge Corporation), as well as<br />
prismatic block fuel used in some HTGRs.<br />
Appropriate pressure drop and heat transfer correlations<br />
can be selected from the built-in library or defined by<br />
the user. For neutronic calculations, the generalized fuel<br />
zone can provide temperature feedback, as well as heat<br />
generation in all solids and in the core coolant.<br />
The default neutronics model that is supplied with the<br />
software is the point kinetic model which requires the<br />
following inputs:<br />
• Temperature feedback coefficients,<br />
• Heat distribution map, and<br />
• Control rod worth vs. position.<br />
This point kinetic model is provided in a user-editable C#<br />
script, which makes it possible to replace the point kinetic<br />
model by linking the simulation model to an external<br />
neutronics code. The scripted neutronics model also makes<br />
it possible for the user to define one’s own feedback<br />
mechanisms based on the design of the specific reactor.<br />
| | Fig. 5.<br />
Koeberg PWR steam generator schematic (left) [2] and simulation model (right).<br />
| | Fig. 6.<br />
Hamm-Uentrop THTR schematic (left) [2] and simulation model (right).<br />
OPERATION AND NEW BUILD 219<br />
| | Fig. 7.<br />
Hamm-Uentrop THTR-300 steam generator comparison experiment (Exp) [3] and simulation (FNX).<br />
Operation and New Build<br />
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 220<br />
| | Fig. 8.<br />
Schematic of the General Nuclear Reactor (GNR) model [1].<br />
Being able to model all kinds of heat transfer accurately<br />
and to include fission physics makes the software a<br />
valuable tool for every nuclear engineer and power cycle<br />
developer. Figure 9 shows an integrated simulation model<br />
that includes a reactor, steam generator, heat exchange,<br />
and some turbomachinery.<br />
Summary<br />
In order to size control valves or determine the control<br />
strategy for a loop, it is necessary to have the pump<br />
performance curve, the heat exchanger pressure drop and<br />
heat transfer characteristics as well as reactor dynamic<br />
behaviour in one simulation model. In this article, a fast<br />
and efficient solution for designing many types of heat<br />
transfer systems is presented. It was shown how to model<br />
any setup and kind of heat exchanger such as plate, tubein-tube,<br />
liquid/gas, finned tube etc. Flownex® Simulation<br />
Environment offers a straight-forward workflow for<br />
engineers who are involved in designing auxiliary systems<br />
that usually contain one or more heat exchangers, such as<br />
in the power plant industry. The software is a specialized<br />
software (e.g. used by ITER, X Energy, BATAN, Hyundai<br />
Heavy Industries) for sizing specific types of heat<br />
exchangers or for doing basic steady-state and transient<br />
mass-and-energy balances. The value of the software in<br />
this area is that one can really integrate the information<br />
from all available sources into a single representative<br />
model, where one can size all kind of devices, test control<br />
strategies, and do integrated system-level analysis and<br />
design. Furthermore, examples from the nuclear power<br />
plant industry, namely the Koeberg PWR steam generator<br />
and the Hamm-Uentrop THTR-300 steam generator which<br />
demonstrated the software’s usability for nuclear related<br />
work were shown. In addition, the lately incorporated<br />
Generic Nuclear Reactor model was introduced.<br />
Further Reading<br />
| | Flownex® SE: www.flownex.de<br />
| | M-Tech Industrial: www.mtechindustrial.com<br />
| | Idaho National Laboratory: www.inl.gov<br />
References<br />
[1] Flownex (2017) User Manual.<br />
[2] Van Antwerpen, H.: Design and Optimization of Advanced<br />
Nuclear Technologies with 1-d Simulation. 7 th Annual<br />
International SMR and Advanced Reactor Summit 2017,<br />
30-31 March, Atlanta, GA, USA.<br />
[3] Esch, M., Hurtado, A., Knoche, D., and Tietsch, W.: Analysis of the<br />
Influence of Different Heat Transfer Correlations for HTR Helical<br />
Coil Tube Bundle Steam Generators with the System Code TRACE.<br />
Nuclear Engineering and Design, 251, 374-380, 2012.<br />
[4] Van Antwerpen, H., Chi, H., Brits, Y., and Botha, F.: Plant-Wide<br />
Simulation Model for Transient Studies on the Xe-100. 2016 ANS<br />
Winter Meeting and Nuclear Technology Expo, 6-10 November<br />
2016, Las Vegas, NV, USA.<br />
Authors<br />
Sebastian Vlach<br />
Leiter Marketing & Vertrieb<br />
Christoph Fischer (PhD)<br />
CFX Berlin Software GmbH<br />
Berlin, Germany<br />
Herman van Antwerpen (PhD)<br />
M-Tech Industrial (Pty) Ltd<br />
South Africa<br />
| | Fig. 9.<br />
Layout of a complete plant power cycle with an example reactor geometry input map (left) [4].<br />
Operation and New Build<br />
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Experimental and Analytical Tools<br />
for Safety Research of GEN IV Reactors<br />
G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak<br />
Current research on nuclear safety in the world, in addition to supporting existing nuclear power plants (PLEX,<br />
mitigation of severe accidents, the development of accident tolerant fuel, decommissioning, etc.), is focused on the<br />
more detailed aspects of the new reactors. The new generation reactors are expected inter alia to use innovative types<br />
of fuel and new types of coolants, such as e.g. Super-Critical Water (SCW), supercritical CO 2 , liquid metals, fluoride<br />
salts or high-temperature Helium. The paper will describe new experimental infrastructure build recently in Research<br />
Centre Řež under the SUSEN (Sustainable Energy) project and available analytical tools for supporting safety research<br />
of GEN IV reactors. Two experimental loops - SCWL (Supercritical Water Loop) and HTHL (High Temperature Helium<br />
Loop) will serve as in-pile loops in the active core of the research reactor LVR-15. The loops insertion in the reactor<br />
LVR-15 requires performing additional safety analyses studying the mutual interference of the loops and the reactor,<br />
especially in conditions of abnormal operation or accident conditions of the loops. The paper will provide examples of<br />
these analyses made using codes ATHLET (supercritical water) and TRACE (high temperature He) illustrating process<br />
of their assessment and practical use. These activities provide significant opportunity for TSO team in building its new<br />
competencies.<br />
Revised version<br />
of a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and<br />
7 November 2017.<br />
OPERATION AND NEW BUILD 221<br />
1 Introduction<br />
The Centrum Výzkumu Řež (CVŘ) and<br />
its partners in the Czech Republic and<br />
abroad are supporting the development<br />
[1] of the Generation IV and<br />
Fusion concepts as well as demonstrators<br />
of these technologies such<br />
as ALLEGRO, ALFRED, DEMO and<br />
others. For this reason, the CVŘ has<br />
had a large R&D program financed<br />
from SUStainable Energy (SUSEN)<br />
project and from its continuation<br />
Research 4 Sustenibility (R4S) [2].<br />
The construction and the operation of<br />
the new SUSEN infrastructure was<br />
supported by the grant of the Ministry<br />
of Education, Youth and Sports as the<br />
part of state help for the large research<br />
infrastructure in the Czech Republic<br />
dedicated to the period 2011–2019.<br />
The SUSEN project consists of 4<br />
programs:<br />
1. Technological Experimental Circuits<br />
(TEO)<br />
2. Structural and System Diagnostics<br />
(SSD)<br />
3. Nuclear Fuel Cycle (NFC)<br />
4. Material Research (MAT)<br />
Within this program, several facilities<br />
were designed and built in order to<br />
study and to address new challenges<br />
of such new technologies. In particular,<br />
the paper focuses on two new<br />
loops which are going to be inserted<br />
inside the LVR-15 research reactor<br />
existing in Řež. The LVR-15 is a light<br />
water tank-type research reactor in<br />
operation since 1957. It is placed in a<br />
stainless steel vessel under a shielding<br />
cover, has forced cooling, uses IRT-4M<br />
type fuel and an has an operational<br />
power level of 10 MWt. The reactor<br />
operations run in campaigns that<br />
usually last for 3 weeks, followed by<br />
an outage lasting for 10 to 14 days<br />
necessary for maintenance and fuel<br />
reloading. There can be also other<br />
campaigns which can operate for<br />
‘short-time’ experiments. Some of the<br />
LVR-15 applications are in the field of<br />
material irradiation research and services,<br />
neutron physics, development<br />
and production of new radiopharmaceuticals<br />
[3]. The loops in concern are<br />
the High Temperature Helium Loop<br />
(HTHL) and the Super Critical Water<br />
Loop (SCWL) and their main scope<br />
are to analyse the cladding behaviour<br />
and structural materials under different<br />
pressure, temperature and coolant<br />
media conditions different from the<br />
standard Light Water Reactors (LWR)<br />
technology [2].<br />
In order to get the regulatory<br />
permit for in-pile operation of these<br />
loops in LVR-15, CVŘ has to prepare<br />
an amendment to the Final Safety<br />
Analyses Report (FSAR) containing<br />
safety analyse of the loops under<br />
| | Fig. 1.<br />
CVR Facilities list.<br />
operational and accidental conditions.<br />
Aim of this paper is to present<br />
the methodology and the analyses<br />
done in support of this process, starting<br />
from code benchmarking/assessment<br />
and the methods adopted in preparing<br />
the safety case.<br />
2 Facilities description<br />
The map of experimental facilities put<br />
into operation in 2016 and those<br />
under preparation to be finalized in<br />
2017 is shown in Figure 1 in the<br />
technology – knowledge map.<br />
In particular, the SCWL and HTHL<br />
represent a pioneer and unique experimental<br />
facility for Gen. IV and Fusion.<br />
2.1 SCWL<br />
The SCWL is going to be a part of a<br />
research facility dedicated to GIV<br />
technologies which will focus on<br />
obtaining data in several areas of the<br />
supercritical fields like: corrosion<br />
processes of construction materials in<br />
supercritical water, with influence of<br />
Operation and New Build<br />
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 222<br />
radiation field, supercritical water<br />
radiolysis and its influence on materials<br />
and water chemistry, development<br />
and testing of sensors, mostly for<br />
measuring of electrochemical potential<br />
(ECP), testing and optimization of<br />
supercritical water regimes [2]. The<br />
specimens being tested will be placed<br />
into the test chamber located in the<br />
active channel where high pressure/<br />
temperature of SCW flow parameters<br />
will be reached.<br />
The SCWL heart is the active<br />
channel, where water reaches required<br />
parameters (pressure of 25 MPa;<br />
temperature of 600 °C; very clean<br />
demineralised water. After successful<br />
out-of-pile (i.e. non active, without<br />
presence of radiation field) operation,<br />
the active channel will be inserted into<br />
the LVR-15 research reactor core. The<br />
bottom part of the active channel is<br />
then submerged between the core’s<br />
fuel assemblies and will face a neutron<br />
flux of up to 1.5 × 10 18 n/m 2 s (thermal<br />
neutrons) and 3 × 10 18 n/m 2 s (fast<br />
neutrons).<br />
The fluid flows in the SCWL is<br />
shown in Figure 2a while the CAD<br />
sketches is shown in Figure 2b.<br />
The active channel has been<br />
modelled with the use of ATHLET<br />
code in two different configurations<br />
see S3.1:<br />
• the out-of-pile configuration that<br />
takes into consideration only pressure<br />
and temperature conditions;<br />
• the in-pile configuration, with the<br />
channel placed inside the LVR-15<br />
active core, that takes into account<br />
also the gamma heating.<br />
2.2 HTHL<br />
HTHL test facility is designed for the<br />
material testing under the simulation<br />
of Gas-cooled Fast Reactor (GFR)<br />
and/or Very High Temperature Reactor<br />
(VHTR) operational conditions.<br />
The specimens being tested will be<br />
placed into the test chamber located<br />
in the active channel where high<br />
pressure/temperature helium flow<br />
parameters will be reached. In addition<br />
to that exposure, during the<br />
in-pile operation, with the active<br />
channel placed into predefined position<br />
of LVR-15 active core rectangular<br />
grid the irradiation effects on the<br />
samples will be studied. The scheme<br />
of the flows in the HTHL is shown in<br />
Figure 3a while the CAD sketch can<br />
be seen Figure 3b.<br />
The active channel has been<br />
modelled with the use of TRACE<br />
code in two different configurations<br />
see S3.2:<br />
• the out-of-pile configuration that<br />
takes into consideration only pressure<br />
and temperature conditions;<br />
• the in-pile configuration, with the<br />
channel placed inside the LVR-15<br />
active core, that takes into account<br />
also the gamma heating.<br />
Views of the channel and of the<br />
coolant flow pattern can be seen in<br />
Figure 3a and Figure 3b.<br />
The temperature inside the channel<br />
is reached through electrical<br />
heater and the coolant flow circulation<br />
is maintained using a two stages<br />
compressor.<br />
3 Methodology<br />
The methodology used to select the<br />
codes and to perform the analyses for<br />
the amendment for the LVR-15 FSAR<br />
consisted in 3 – independent steps:<br />
• Searching and assessing the codes<br />
ability to simulate helium and SCW<br />
during steady-state and transients<br />
conditions.<br />
• Creating the loops model to be<br />
used for the TH analyses and<br />
developing it based on the steadystate<br />
thermohydraulic parameters<br />
• Performing analyses of the selected<br />
scenarios in order to verify the<br />
safety criteria and obtaining the<br />
necessary data for the structural<br />
analyses.<br />
The present methodology complies<br />
with the [4] IAEA standard in introducing<br />
new research facilities inside<br />
nuclear research installations such as<br />
the LVR-15 reactor.<br />
3.1 ATHLET 3.1A code<br />
ATHLET 3.1 patch A code [5] is a<br />
thermal hydraulic system code<br />
developed by the GRS for simulating<br />
time-dependent phenomena in the<br />
PWRs and BWRs. Furthermore, the<br />
code can also simulate GEN IV working<br />
fluids like helium, liquid metals<br />
and supercritical water.<br />
The heat transfer behaviour in<br />
supercritical water represents a<br />
challenging task mainly connected<br />
with ensuring safety and reliable<br />
operation. Nowadays, the understanding<br />
of the supercritical water<br />
regimes is rather limited, specifically<br />
regarding the close proximity of the<br />
critical point.<br />
For the simulation of supercritical<br />
water, a range of properties approximation<br />
has been extended up to a<br />
pressure of 100 MPa. An additional<br />
module cover the pressure range from<br />
22.5 to 100 MPa. The transition<br />
between subcritical and the supercritical<br />
properties is performed by a<br />
suitable interpolation between these<br />
packages for pressures between 22.0<br />
and 22.5 MPa [5].<br />
In ATHLET 3.1A the selection of<br />
correlations for supercritical water is<br />
performed by switching a built in flag<br />
found in the heat structure module.<br />
A number of six correlations are<br />
available which were tested against<br />
the results obtained by IAEA-benchmark<br />
exercise [6] and three of them<br />
were selected for the purpose of the<br />
certification and further use.<br />
3.2 TRACE 5 Patch 4 codes<br />
The TRACE code has been used as an<br />
alternative to the RELAP5/Mod3.3<br />
code, since US NRC decided to stop<br />
| | Fig. 2a.<br />
SCWL Flow.<br />
| | Fig. 2b.<br />
CAD Sketches.<br />
| | Fig. 3a.<br />
Flow in HTHL.<br />
| | Fig. 3b.<br />
HTHL CAD Sketches.<br />
Operation and New Build<br />
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
| | Fig. 4.<br />
HTHL TRACE Nodalization.<br />
the development of RELAP starting<br />
with next year.<br />
TRACE has been designed to perform<br />
best-estimate analyses of loss- ofcoolant<br />
accidents (LOCAs), operational<br />
transients, and other accident<br />
scenarios in pressurized light-water<br />
reactors (PWRs) and boiling lightwater<br />
reactors (BWRs). It can also<br />
model phenomena occurring in<br />
experimental facilities designed to<br />
simulate transients in reactor systems.<br />
Models used include multidimensional<br />
two-phase flow, none quilibrium<br />
thermo-dynamics, generalized heat<br />
transfer, reflood, level tracking, and<br />
reactor kinetics. In addition, TRACE is<br />
able to simulate several other coolants<br />
such as helium and water in subcooled<br />
condition and atmospheric pressure<br />
(LVR-15 conditions). [7], [8]<br />
For this reason, TRACE code was<br />
selected and used for the simulation in<br />
the Helium at 7 MPa with a temperature<br />
rise from 200 °C up to 900 °C<br />
(nominal parameters for HTHL). The<br />
correlation adopted for simulating the<br />
heat transfer from heat structures to<br />
the helium coolant and vice versa<br />
implemented in TRACE are Gnielinsky<br />
and El Genk [7-9].<br />
3.3 Codes assessment<br />
The code assessment was done by<br />
benchmarking of the codes with<br />
available experimental results done in<br />
different facilities around the world.<br />
One of the most important steps<br />
was selecting the code that can<br />
perform the heat transfer calculation<br />
under the high temperature He or<br />
SCW conditions along with adequate<br />
correlations [10], [11]. In the case of<br />
ATHLET, the code was carefully assed<br />
and benchmarked with experimental<br />
results of a project coordinated by<br />
IAEA [6] for steady state and with<br />
Chinese SWAMUP facility [12] for<br />
| | Fig. 5.<br />
SCWL ATHLET Nodalization.<br />
the transition from supercritical to<br />
subcritical condition.<br />
The aim of this analyses was to<br />
simulate the deterioration phenomenon<br />
[9] of heat transfer with fluid<br />
transiting between subcritical and<br />
supercritical condition. According to<br />
Ref. [6], Mokry, Gupta and Watts-<br />
Chou correlations show acceptable<br />
prediction capabilities of the Heat<br />
Transfer Coefficient (HTC). Both our<br />
analyses and IAEA CRP program<br />
concluded that an uncertainty for<br />
calculating HTC is about ±25% while<br />
the calculating wall temperature was<br />
between ±10 to 15 %. As a result of<br />
this exercise, the code certification<br />
was obtained from SONS (State of<br />
Office for Nuclear Safety) in March<br />
2017 for using the code in simulating<br />
the CVŘ SCWL.<br />
The TRACE assessment was done<br />
with the data available from the<br />
project GoFastR [13] financed by the<br />
EC in the Framework Program 7, in<br />
particular with data related to the<br />
HE-FUS3 facility [14], [15]. The<br />
facility operational parameters are<br />
similar to the HTHL.<br />
The TRACE HE-FUS3 thermal hydraulic<br />
model was developed and<br />
compared with experimental data<br />
from steady state loop operation and<br />
selected transients. The comparison<br />
showed that the TRACE T/H model<br />
can simulate the helium temperatures<br />
as well as the piping wall temperatures<br />
along the different sections<br />
of the facility accurately. After a sensitivity<br />
analysis, the electrical heater<br />
power has been lowered to 10.76 kW.<br />
The certification for TRACE code was<br />
obtained from SONS in December<br />
2016 by CVŘ for simulating water in<br />
PWR condition, sub-cooled water at<br />
atmospheric pressure (such as LVR-15<br />
operational condition) and helium<br />
behaviour in the range of 7 MPa for a<br />
temperature range between 200 to<br />
900 °C. [16]<br />
3.4 Model description for<br />
HTHL and SCWL<br />
The HTHL and SCWL are similar<br />
experimental facilities characterized<br />
by 2 steps upward and downward<br />
flows, although some major differences<br />
exist in the design. In particular,<br />
the HTHL active channel contains<br />
all necessary components for heat<br />
transfer inside except of the compressor<br />
and the main compensator, which<br />
are located in the chemical control<br />
system. The Figure 4 shows the<br />
TRACE nodalization containing simulated<br />
components.<br />
The SCWL is different in such way<br />
that it needs some extra components<br />
larger than the HTHL to help the sub<br />
critical water to become gas. For this<br />
reason additional axillary facilities,<br />
such as a recuperator, cooler, pump,<br />
compensator and other 4 sections of<br />
electrical heater are located in a<br />
different building along with the<br />
chemical control system.<br />
The ATHET SCWL loop model<br />
shown in Figure 5 is focused mainly<br />
on the active channel from inlet to<br />
outlet, although all the previous<br />
components are also simulated as a<br />
part of the primary and the secondary<br />
circuits. In addition to the primary<br />
and the secondary circuits of the<br />
SCWL, there is the third open loop<br />
representing the active channel position<br />
into the LVR-15 core and providing<br />
additional heat transfer between<br />
active channel and reactor coolant.<br />
3.5 Analysed Scenarios<br />
The planned in-pile operation of both<br />
loops requires an amendment of the<br />
LVR-15 Final Safety Report providing<br />
thermohydraulic and structural integrity<br />
analyses during normal operation<br />
OPERATION AND NEW BUILD 223<br />
Operation and New Build<br />
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 224<br />
Normal operating<br />
conditions<br />
Steady State<br />
LVR-15 Start up<br />
LVR-15 Shutdown<br />
Loops Start up<br />
Loops Shutdown<br />
and during Loss of Flow Accident<br />
( LOFA) and Loss of Coolant Accident<br />
(LOCA) accident conditions. In particular,<br />
the structural integrity analyses<br />
required the temperature profile<br />
inside the Pressure Envelope (PE) as<br />
boundary condition. For this reason<br />
the normal operation and abnormal<br />
operation conditions were calculated<br />
using TRACE and ATHLET codes with<br />
very narrow mesh nodal distribution<br />
in the PE. For structural integrity<br />
following criteria and limitation due<br />
to the non-boiling condition in LVR-15<br />
were used:<br />
1. PE maximum temperature during<br />
normal/abnormal transients is less<br />
than 450 °C.<br />
2. PE maximum temperature during<br />
accident conditions is less than<br />
500 °C.<br />
3. Aluminium surface of the Receiver<br />
maximum temperature in contact<br />
with LVR-15 coolant less than 45 °C<br />
during normal/abnormal conditions.<br />
4. Aluminium surface of the Receiver<br />
maximum temperature in contact<br />
with LVR-15 coolant less than 60 °C<br />
during accident conditions.<br />
In the case of accident conditions,<br />
both active channels of HTHL and<br />
SCWL will have to be replaced. The<br />
analysed scenarios are described in<br />
the Table 1.<br />
4 Illustrative results<br />
The results described in the paper<br />
refer to the simulations of SCWL and<br />
Pressure<br />
tests<br />
(not simulated)<br />
| | Tab. 1.<br />
Operational and Accident Scenarios Description.<br />
Abnormal<br />
conditions<br />
Switch off Loops Electrical<br />
Heater for 1 min.<br />
LVR-15 SCRAM and switch off<br />
of Loops Electrical Heater<br />
at t = 0 s + pump trip after 1 min.<br />
Switch off Loops Electrical Heater<br />
at t = 0 s + LVR15 SCRAM and<br />
Pump Trip after 3 min.<br />
Parameter Value Unit<br />
Pressure 25 MPa<br />
Inlet Flow<br />
Temperature<br />
Outlet Flow<br />
Temperature<br />
Max Flow<br />
Temperature<br />
Sample Area<br />
Mass flow<br />
| | Tab. 2.<br />
SCWL main parameters calculated during<br />
steady state.<br />
HTHL during the steady state operation<br />
with continuing in LOFA condition.<br />
The results represent an extract<br />
of the large number of calculations of<br />
various combinations of operational<br />
transients with the aim to demonstrate<br />
the capabilities of the codes to<br />
simulate behaviour the loops.<br />
4.1 SCWL steady state and<br />
LOFA analyses<br />
The main parameters for the steady<br />
state are shown in Table 2. The whole<br />
steady state calculation was rather<br />
long due to some inertia of the system.<br />
The computer model simulated<br />
behaviour during the transient of all<br />
heat structures representing the<br />
complete piping system. In the calculation<br />
some numerical instability<br />
complicated the steady state due to<br />
Accident<br />
conditions<br />
385 ºC<br />
406 ºC<br />
600 ºC<br />
35 %<br />
By pass flow 65 %<br />
Mass flow 200 kg/h<br />
Loss of Flow Accident<br />
(LOFA)<br />
Loss of Coolant Accident<br />
(LOCA)<br />
small dimensions of the component<br />
facing the deterioration flow phenomenon<br />
during the heating up process.<br />
For these reasons, the whole steady<br />
state was completed in 25,000 s,<br />
where 15,000 to 20,000 s were needed<br />
to adjust the steady state and the<br />
rest 5,000 s were used to verify the<br />
steady behaviour of the main parameters.<br />
After this period the model simulated<br />
the accident scenario – LOFA<br />
without the reactor SCRAM in order<br />
to maximize the consequences and to<br />
calculate the time to reach temperature<br />
of PE (AC) 500 °C.<br />
The scenario is described in the<br />
following steps:<br />
1. Pump stops in 1 s after the initialization<br />
event (25,001 s)<br />
2. Active channel internal electrical<br />
heaters shut down to 0 % on the<br />
nominal power in 7s (25,007 s)<br />
3. The LVR-15 SCRAM starts at 40 s<br />
when the maximum temperature<br />
in the PE rises above the 500 °C.<br />
(25,040 s)<br />
4. The whole transient is completed<br />
in 15,000 s (40,000 s), when the<br />
SCWL and LVR-15 are in the<br />
controlled cold state.<br />
The Figure 6 and Figure 7 represent<br />
the SCW maximum temperature<br />
calculated in the sample area and the<br />
outlet temperature from the active<br />
channel, while the Figure 8 shows the<br />
maximum temperature of the PE,<br />
where there is the neutron flux peak<br />
in the Boltzmann distribution.<br />
4.2 HTHL steady state and<br />
LOFA analyses<br />
The design conditions calculated for<br />
the active channel are described in<br />
Table 3. And they are mainly summarized<br />
as reported:<br />
1. Mass flow rate of 0.0105 kg/s<br />
2. Design pressure of 7 MPa<br />
3. Design electrical heater power of<br />
11.85 kW<br />
4. Cold helium temperature of 210 °C<br />
| | Fig. 6.<br />
SCWL Coolant Maximum Temperature in LOFA.<br />
| | Fig. 7.<br />
SCWL Active Channel Outlet Temperature in LOFA.<br />
Operation and New Build<br />
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Location of the Thermocouples<br />
Thermocouple<br />
| | Fig. 8.<br />
SCWL Maximum EP Temperature in LOFA.<br />
Parameter Value Unit<br />
Pressure 7 MPa<br />
Inlet Flow<br />
Temperature<br />
Max Flow<br />
Temperature<br />
Maximum AC<br />
Pressure Envelop (PE)<br />
Temperature<br />
210 ºC<br />
900 ºC<br />
450 ºC<br />
Mass flow 40 kg/h<br />
Inlet into the interpiping space of the reheater<br />
Output from the interpiping space of the reheater<br />
Entry into the test chamber<br />
Inlet to the reheater piping space<br />
Output from the reheater piping space<br />
Output from the primary side of the heat exchanger<br />
Maximum helium temperature<br />
| | Tab. 4.<br />
Thermocouples position and description.<br />
T1<br />
T2<br />
T3<br />
T4<br />
T5<br />
T6<br />
Tmax<br />
OPERATION AND NEW BUILD 225<br />
| | Tab. 3.<br />
HTHL main parameters calculated during<br />
steady state.<br />
The steady state simulation was<br />
run in null transient mode for 5,000 s<br />
and the stabilized conditions were<br />
reached after 3,500 s. The LOFA<br />
transients was characterized by an<br />
immediate safety shutdown of the<br />
reactor due to the loss of power. As a<br />
result of the SCRAM, the temperature<br />
went immediately down following the<br />
heat generated by decay gamma flux.<br />
Figure 9 shows the calculated<br />
temperature for various thermocouples<br />
positions (according to<br />
Table 4), while Figure 10 represents<br />
the maximum temperatures in the<br />
HTHL PE.<br />
| | Fig. 9.<br />
HTHL Helium temperatures during LOFA.<br />
5 Conclusions<br />
The article provides a brief introduction<br />
about the SUSEN project and the<br />
experimental facilities built in CVŘ in<br />
the Czech Republic for research and<br />
development in support of the safe,<br />
reliable and long‐term sustainable<br />
operation of existing energy facilities<br />
and in development of GIF IV and<br />
fusion technologies. The SUSEN<br />
R&D activities include four complementary<br />
programmes, mentioned in<br />
the introduction, which are focused<br />
on material science, thermal hydraulics,<br />
neutronics, radiation protection,<br />
nuclear chemistry, waste management<br />
and environmental studies. A<br />
significant part of the research programme<br />
is devoted to HTH and SCW<br />
experimental loops, which are going<br />
to be installed into the active core of<br />
the research reactor LVR-15. Both of<br />
| | Fig. 10.<br />
HTHL PE temperature during LOFA.<br />
these unique facilities are challenging<br />
to model and the selection of appropriate<br />
codes was a demanding process.<br />
A special methodology was used for<br />
assessing the abilities of the codes to<br />
simulate these advanced coolants and<br />
to obtain regulatory certificate/ permit<br />
for their use in operational and accident<br />
conditions and for preparation of<br />
the amendment of the LVR-15 FSAR.<br />
These presented activities represent<br />
only starting steps for the further<br />
codes validation which will be based<br />
on benchmarking of the codes with<br />
experimental data provided by the<br />
SCWL and HTHL loops in their<br />
experimental campaigns.<br />
Aknoledgment<br />
The authors would like to thank<br />
Mr. Miroslav Hrehor and Dr. Vincenzo<br />
Romanello for their kind revisions and<br />
suggestions.<br />
The presented work was financially<br />
supported by the Project CZ.02.1.01/<br />
0.0/0.0/15_008/0000293: Sustainable<br />
energy (SUSEN) – 2 nd phase,<br />
realized in the framework of the<br />
Operation and New Build<br />
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 226<br />
European Structural and Investment<br />
Funds.<br />
This work has been supported<br />
by the Project CZ.02.1.01/0.0/0.0/<br />
15_008/0000293: Sustainable energy<br />
(SUSEN) – 2 nd phase realized in the<br />
framework of the European Structural<br />
and Investment Funds.<br />
References<br />
[1] CVR Annual Report 2016.<br />
[2] http://susen2020.cz/<br />
[3] http://cvrez.cz/en/infrastructure/<br />
research-reactor-lvr-15<br />
[4] IAEA, Standards Safety in the Utilization<br />
and Modification of Research Reactors”,<br />
Safety Standard n° SSG-24, VIENNA,<br />
2012.<br />
[5] ATHLET 3.1A, 2016 User manual:<br />
ATHLET Mod 3.1 Cycle a, G. Lerchl,<br />
H. Austregesilo, P. Schoffel, D. von<br />
der Cron, F. Weyermann, March 2016.<br />
[6] Heat Transfer Behaviour and Thermohydraulics<br />
Code Testing for Supercritical<br />
Water Cooled Reactors (SCWRs),<br />
IAEA. http://www-pub.iaea.org/<br />
books/IAEABooks/10731/Heat-<br />
Transfer-Behaviour-and-Thermo-<br />
hydraulics-Code-Testing-for-<br />
Supercritical-Water-Cooled-R<br />
[7] TRACE V5.840 Theory Manual,<br />
U.S. Nuclear Regulatory Commission,<br />
Washington DC, March 2013.<br />
[8] TRACE V5.840 User’s Manual, Volume 1:<br />
Input Specification, U.S. Nuclear<br />
Regulatory Commission, Washington<br />
DC, February 2014.<br />
[9] TRACE V5.840 User’s Manual, Volume 2:<br />
Modelling Guidelines, U.S. Nuclear<br />
Regulatory Commission, Washington<br />
DC, February 2014.<br />
[10] G. Mazzini et al., ATHLET 3.1A<br />
SIMULATION CAPABILITIES FOR SUPER-<br />
CRITICAL STATE, CVR 1581, 1.1.2017.<br />
[11] G. Mazzini et al., ATHLET 3.1A HEAT<br />
TRANSFER ASSESMENT FOR SUPER-<br />
CRITICAL WATER, CVR 1582, 1.1.2017.<br />
[12] G. Mazzini et al., ATHLET 3.1A<br />
CAPABILITIES IN SIMULATING SWAMUP<br />
FACILITY IN SCW CONDITIONS, CVR<br />
1583, 1.1.2017.<br />
[13] M. Polidori, HE-FUS3 Benchmark<br />
Specifications, GoFastR-DEL-1.5-01,<br />
Rev. 0, ENEA, July 2011.<br />
[14] M. Polidori, HE-FUS3 Experimental<br />
Campaign for the Assessment of<br />
Thermal-Hydraulic Codes: Pre-Test<br />
Analysis and Test Specifications,<br />
Report RSE/2009/88.<br />
[15] M. Polidori et al, HE-FUS3 Benchmark<br />
Results, GoFastR-DEL-1.5-6, Rev. 0,<br />
November 2012.<br />
[16] Miloš Kynčl, Development and Assessment<br />
of TRACE HTHL-2 Facility Thermal<br />
Hydraulic Model, Internal Project Status<br />
Report, CVŘ 1334, March 2017.<br />
Authors<br />
G. MazziniM. Kyncl<br />
Alis Musa<br />
M. Ruscak<br />
Centrum Vyzkumu Rez (CVŘRez)<br />
Hlavní 130<br />
250 68 Husinec – Řež,<br />
Czech Republic<br />
Numerical Analysis of MYRRHA Interwrapper<br />
Flow Experiment at KALLA<br />
Abdalla Batta and Andreas G. Class<br />
Introduction The MYRRHA reactor, which is developed at SCK-SCN in Belgium, represents a multi-purpose<br />
irradiation facility. Its prominent feature is a pool design with the nuclear core submerged in liquid metal lead bismuth.<br />
During transients between normal operation and accident conditions decay heat removal is ensured by forced and<br />
natural convection, respectively. The flow in the gap between the fuel assemblies plays an important role in limiting<br />
maximum temperatures which should not be exceeded to avoid core damage. The term inter-wrapper flow (IWF)<br />
describes the convection in the small gap between the wrapper tubes of neighbouring fuel assemblies (FAs). It plays an<br />
important role for passive decay heat removal (DHR).<br />
Based on numerous experiments<br />
several correlations have been proposed<br />
for the flow within wirewrapped<br />
rod bundles. However, for<br />
the flow within the gap between<br />
neighbouring bundles only few<br />
studies are reported. Recently [1]<br />
reviewed the existing correlations by<br />
Rheme [2], Baxi & Dalle Donne [3]<br />
Cheng and Tordreras [4], and Kirillov<br />
[5] for the pressure-drop in wirewrapped<br />
rod bundles. The existing<br />
correlations were compared to all the<br />
available experimental data and<br />
showed that agreement of approximately<br />
±20 % can be expected. For<br />
the inter-wrapper flow within the<br />
gap only few studies exist, see [6].<br />
Due to the scarce database, within the<br />
Horizon 2020 – research and innovation<br />
framework program of the EU,<br />
the SESAME project was established<br />
to develop and validate advanced<br />
numerical approaches, to achieve a<br />
new or extended validation base and<br />
to establish best practice guidelines<br />
including verification & validation<br />
and uncertainty quantification, see<br />
[7]. In particular the current work<br />
supports the inter-wrapper flow<br />
experiment at KALLA. Three fuel<br />
assemblies including the gap flow are<br />
studied covering the full range of<br />
thermo- hydraulic conditions expected<br />
in the reactor application. For this<br />
purpose, an experimental test matrix<br />
has been established which covers<br />
relevant scenarios. The aim of our<br />
numerical pre-test study is to help the<br />
design of the experiment. The current<br />
study applied RANS-CFD methods for<br />
design support of the experiment. In<br />
the body of this compact the experiment,<br />
the corresponding numerical<br />
model, and preliminary numerical<br />
results are provided.<br />
1 Experimental setup<br />
The KALLA experiment investigates<br />
IWF between three bundles which<br />
are thermally connected by a gap.<br />
Figure 1 shows a cross-sectional view<br />
of the test section which consists of<br />
three ducts representing the fuel<br />
assemblies. Each duct contains 7 wirewrapped<br />
electrically-heated pins<br />
representing the fuel rods. The gap<br />
between the channels, i.e. assemblies,<br />
is filled with liquid metal, so<br />
that strong thermal coupling exists<br />
between neighbouring assemblies.<br />
The test matrix covers independent<br />
variation of flow and thermal conditions<br />
in both the gap and the bundles.<br />
Detailed description of the experiment<br />
is reported in [8]. The geometrical<br />
parameters of the bundle and the<br />
nomenclature are also shown in<br />
Figure 1. The experimental loop<br />
facility THESYS at KALLA and the<br />
Operation and New Build<br />
Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
| | Fig. 1.<br />
SCWL Coolant Maximum Temperature in LOFA.<br />
| | Fig. 2.<br />
Left: Experimental loop facility THESYS at KALLA showing location where the inter wrapper flow<br />
experiment (see Figure 3) will be installed; right: flow diagram for the IWF tests with four parallel<br />
channels; the valves V2.1-V2.3 control the flow through the assemblies Q1-Q3. V.2.4 controls the<br />
flow in the gap [8].<br />
location where the IWF experiment<br />
will be installed is shown in Figure 2<br />
left. Figure 2 right shows the flow<br />
diagram of the IWF tests with four<br />
parallel channels representing the<br />
three assemblies (Q1-Q3) and the gap<br />
( illustrated by the box containing<br />
Q1-Q3). The flow and temperature<br />
within each assembly and the gap can<br />
be set individually by choosing valve<br />
openings (V2.1-V2.4) and heating<br />
rates according to the KALLA test<br />
matrix. Figure 3 shows the geometry<br />
of the IWF test section.<br />
and mesh resolution for the thermoshydraulic<br />
investigation of the gap and<br />
the bundle. In particular, we include<br />
the upstream components to verify<br />
their influence on the flow field within<br />
the test section. We employ the k-ε<br />
turbulence model and the commercial<br />
CFD-code Star CCM+. Our first<br />
studied case (i) focuses on the gap<br />
| | Fig. 3.<br />
Geometry of the IWF test section, dimensions are in mm, the heated part<br />
of the bundle is marked red on the left side of the figure, 600 mm, [8].<br />
flow and our second case (ii) on the<br />
fuel assembly. For the study of case (i)<br />
a computational domain including<br />
the lower flow distributer, riser pipe<br />
( including venture tube), upper flow<br />
vessel, and the gap are considered (for<br />
corresponding technical drawings of<br />
components refer to Figure 3). For the<br />
study of case (ii) the computational<br />
domain includes the lower flow distributer,<br />
riser pipe (including venture<br />
tube), one inlet expansion and a single<br />
7-pin bundle. Flow properties of the<br />
liquid metal Lead-Bismuth eutectic at<br />
200 °C are employed. Note that corresponding<br />
upstream pipes and flow<br />
conditioners are modelled so that<br />
all relevant geometric details are<br />
captured. Quantifying the effect of<br />
the flow conditioning sections is<br />
important for future simulations, as it<br />
would enable the use of a simpler<br />
computational domain, which still<br />
provides accurate results. In the future<br />
post-test analysis, the smallest representative<br />
computational domain (e.g.,<br />
potentially without flow conditioner<br />
etc.) will be used to compose a fully<br />
coupled thermos-hydraulic simulation<br />
of the three bundles including<br />
the IWF in the gap. Figures 4 left<br />
and right show the computational<br />
domains for the pre-test studies<br />
OPERATION AND NEW BUILD 227<br />
2 Numerical study<br />
A comprehensive analysis of the<br />
experiment requires efficient simulations.<br />
In the pre-test analysis of the<br />
hydraulics separate simulations of the<br />
gap region and the fuel assembly are<br />
performed. In a first step, we determine<br />
suitable computational domains<br />
| | Fig. 4.<br />
Computational domain for IWF-gap (left) and bundle (right) including the upstream domains.<br />
Operation and New Build<br />
Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 228<br />
of cases (i) and (ii), respectively.<br />
Obviously, a substantial effort was<br />
undertaken to include the upstream<br />
flow domain, so that the inflow into<br />
the fuel assembly and the gap are<br />
properly represented in the flow<br />
simulations.<br />
Since we have less experience with<br />
the gap region, and in particular, the<br />
applicable turbulence regime we have<br />
considered 3 cases corresponding to<br />
laminar flow, transitional flow, and<br />
fully developed turbulence, respectively.<br />
This covers the flow range 0.17<br />
to 0.86 kg/s (Re = 1,250 to 6,250),<br />
proposed by the test matrix. For the<br />
investigation of the fuel assembly, we<br />
consider the nominal flow rate, i.e. the<br />
maximum flow rate planned in the<br />
test matrix. This corresponds to a flow<br />
rate of 3.58 kg/s and Re = 8,910<br />
where Re is based on the bundle<br />
hydraulic diameter. All cases considered<br />
in the experimental test matrix<br />
are within the range of transitional<br />
flow according to Cheng and Todreas<br />
[4] (see next section on correlations)<br />
so that no distinction of various flow<br />
regimes is needed for the comparison<br />
to correlations.<br />
2.1 Inter-wrapper flow gap<br />
region<br />
The objective of case study (i) is to<br />
investigate the effects of all upstream<br />
components on the flow distribution<br />
entering the gap, i.e. the inter wrapper<br />
flow region. This study employs the<br />
computational domain shown in<br />
Figure 4 left. For the simulation, a<br />
mesh with approximately 0.72 million<br />
cells has been generated. The investigated<br />
range of flow rates results in<br />
turbulent flow in all components<br />
upstream of the gap, since the<br />
Reynolds- numbers based on pipe<br />
diameter varies between 5,200 and<br />
26,000. However, within the gap<br />
the Reynolds-number based on gapwidth<br />
equates to 1,250 to 6,250 corresponding<br />
to the transitional regime of<br />
turbulence. The pressure drop along<br />
the gap accounts for about 20 % of the<br />
total pressure drop. Since we are<br />
interested in accurately predicting the<br />
upstream flow in the gap region the<br />
use of a turbulent model is mandatory.<br />
Moreover, in order to judge the uniformity<br />
of the flow entering the gap<br />
there is no need to use a very accurate<br />
result within the gap. Thus, a high-<br />
Reynolds-number turbulence model<br />
using automatic wall functions is<br />
used. Figure 5 shows the velocity<br />
vectors in the gap entrance region for<br />
the case where the Reynolds-number<br />
is 5,200 based on pipe diameter<br />
(Reynolds-number is 1,250 based on<br />
the gap hydraulic diameter). We<br />
observe that the flow within the gap<br />
becomes near uniform after a short<br />
length, which does not exceed 10 %<br />
of the length of the gap region. The<br />
heated zone starts further downstream<br />
approximately at half the<br />
length of the gap region. For higher<br />
Reynolds-number a qualitative similar<br />
result is obtained. For future simulations<br />
aiming at accurately simulating<br />
the temperature field, we conclude<br />
that the effect of upstream components<br />
is negligible. In Table 1 the<br />
pressure drop across the simulated<br />
region is compared to design values<br />
for three selected cases covering<br />
the full range of flow rates. Design<br />
values are calculated using lumped<br />
parameter models. Both results agree<br />
reasonably well, indicating that<br />
lumped parameter models well<br />
describe the flow in the gap.<br />
2.2 Flow within a single<br />
wire-wrapped rod bundle<br />
As in the previous study, we aim at<br />
investigating whether the upstream<br />
region that conditions the flow<br />
entering the wire-wrapped bundle<br />
influences the flow in the heated<br />
section of the bundle. Here, i.e. in case<br />
study (ii), a single Reynolds-number<br />
of 8,900 based on the bundle hydraulic<br />
diameter is considered. This<br />
corresponds to the nominal flow<br />
rate as well as the maximum flow<br />
rate intended in the experimental<br />
tests. The computational domain of<br />
Figure 4 right uses approximately<br />
1 million cells. Figure 6 shows the<br />
velocity magnitude within the bundle.<br />
At the entrance, we still observe pronounced<br />
non-uniformities of the flow<br />
distribution. These quickly equilibrate<br />
so that a more-uniformly distributed<br />
flow is observed well before the<br />
heated section of the bundle is<br />
reached (for the location of the heated<br />
region refer to Figure 3).<br />
This result suggests that inflow<br />
effects are negligible for the intended<br />
thermal analysis of the bundle. Thus<br />
in a second simulation we remove the<br />
flow-conditioning region to reduce<br />
the size of the considered flow<br />
domain. To validate our simulation<br />
results we use higher mesh resolution<br />
within the smaller domain. Figure 7<br />
shows the pressure along two selected<br />
axial lines, which are depicted in the<br />
small inset. The influence of the wirewrap<br />
manifests in the periodical<br />
modulation of the pressure profile.<br />
Obviously, development effects have<br />
decayed at a length of approximately<br />
100 mm. We compute the pressure<br />
| | Fig. 5.<br />
Velocity vectors within the gap upstream region.<br />
Flow rate<br />
[kg/s]<br />
design Δp tot ,<br />
[Pa]<br />
CFD Δp tot , [Pa]<br />
1. 0.86 15350 13500<br />
2. 0.688 9964 -<br />
3. 0.516 5667 5500<br />
4. 0.344 2586 -<br />
5. 0.172 663 850<br />
| | Tab. 1.<br />
Comparison of design values evaluated by lumped parameter model<br />
versus computed pressure drop across the test section including the flowconditioning<br />
components.<br />
| | Fig. 6.<br />
Velocity magnitude within bundle showing non-uniformities of flow distribution at leftmost plane and<br />
more-uniformly distributed flow in subsequent planes.<br />
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Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
| | Fig. 7.<br />
Pressure along two selected axial lines in the wire-wrapped rod bundle.<br />
The inset specifies location of lines.<br />
drop using data at corresponding<br />
wire-wrap positions, i.e. from axial<br />
positions 0.065 m to 1.268 m. The<br />
mean pressure drop is 946 Pa/m.<br />
2.3 Model validation<br />
In this subsection, results of our<br />
numerical study are compared to the<br />
simplified Cheng and Todreas [1986]<br />
correlation. The correlation was<br />
recently recommended in (1) to<br />
predict pressure drop (Δp) in bundles<br />
with an accuracy of ±20 %. It applies<br />
for a wide range of Reynolds- numbers.<br />
The friction factor (f) is defined in<br />
eq. 1, where d h,bdl , L, and u b<br />
2<br />
are<br />
hydraulic diameter, length, and<br />
average axial bundle velocity, respectively.<br />
(1)<br />
The correlation for f reads<br />
for Re < Re L<br />
for Re L ≤ Re ≤ Re T<br />
for Re > Re T (2)<br />
where<br />
Re L = 300 x 10 1.7(P/D−1.0) (3)<br />
Re T = 10,000 x 10 0.7(P/D−1.0) (4)<br />
ψ = log(Re/Re L ) / log(Re T /Re L ) (5)<br />
C fL = (-974.6 + 1612.0(P/D) −<br />
598.5(P/D) 2 )(H/D) .06-0.085(P/D)<br />
(6)<br />
C fT = (0.8063 − 0.9022(log(H/D)) +<br />
0.3526(log(H/D)) 2 ) ×<br />
(P/D) 9.7 (H/D) 1.78-2.0(P/D) (7)<br />
We compare the nominal flow case<br />
of 3.580 kg/s which corresponds to a<br />
velocity of 0.2 m/s and Re is 8910,<br />
which is in the transient region.<br />
According to eqns (3) and (4), Re L and<br />
Re T are 902 and 15735, respectively.<br />
The calculated friction factor f<br />
equates to 0.0557. This corresponds<br />
to a pressure drop in the bundle of<br />
1407.2 Pa. The predicted pressure<br />
drop resulting from the CFD study is<br />
1,138 Pa. The difference is near 19 %,<br />
which lays within the accuracy limits.<br />
In future thermos-hydraulic simulations,<br />
the current model can be<br />
applied. For posttest analysis, additional<br />
sensitive studies might be<br />
necessary to further reduce the<br />
uncertainty.<br />
Conclusions<br />
The flow in the gap between neighbouring<br />
fuel assemblies plays an<br />
important role in transients between<br />
forced and natural convection. At<br />
KALLA an experiment on the interwrapper<br />
flow is currently setup and<br />
accompanied by pre-test numerical<br />
CFD studies. These proof that both<br />
the flow in the gap region and the<br />
fuel bundle are not influenced by the<br />
upstream flow-conditioning region.<br />
Moreover, development length are<br />
much shorter than the unheated<br />
length of the test section, so that<br />
the thermal field is uninfluenced by<br />
flow non-uniformities. Preliminary<br />
comparison of pressure losses computed<br />
by CFD and correlation provide<br />
reasonable agreement for both the<br />
gap and bundle. The result of our<br />
study enters pre-test studies of the<br />
thermal field within the EU-H2020<br />
SESAME project. There complete<br />
simulation of the test section consisting<br />
of three bundles connected<br />
by the gap region including conjugate<br />
heat transfer is performed.<br />
Acknowledgement<br />
This project has received funding from<br />
the Euratom research and training<br />
programme 2014-<strong>2018</strong> under grant<br />
agreement No 654935 and from the<br />
AREVA Nuclear Professional School.<br />
References:<br />
[1] Chen, S.; Todreas, N.; Nguyan, N.<br />
(2014). Evaluation of existing correlations<br />
for the prediction of pressure drop<br />
in wire-wrapped hexagonal array pin<br />
bundles. Nuclear Engineering and<br />
Design 267, pp. 109 – 131<br />
[2] Rehme, K. (1973). Pressure drop<br />
correla tions for fuel element spacers.<br />
Nuclear Technology 17, 15–23.<br />
[3] Baxi, C.B., Dalle Donne, M., (1981).<br />
Helium cooled systems, the gas cooled<br />
fast breeder reactor. In: Fenech, H. (Ed.),<br />
Heat Transfer and Fluid Flow in Nuclear<br />
Systems. Pergamon Press Inc.,<br />
pp. 410–462.<br />
[4] Cheng, S.-K.; Todreas, N. (1986). Hydrodynamic<br />
models and correlations for<br />
bare and wire-wrapped hexagonal rod<br />
bundles - Bundle friction factors,<br />
subchannel friction factors and mixing<br />
parameters. Nuclear Engineering and<br />
Design 92 (2), 227 – 251.<br />
[5] Kirillov, P.L., Bobkov, V.P., Zhukov, A.V.,<br />
Yuriev, Y.S., (2010). Handbook on<br />
Thermo hydraulic Calculations in<br />
Nuclear Engineering. Thermohydraulic<br />
Processes in Nuclear Power Facilities,<br />
vol. 1. Energoatomizdat, Moscow.<br />
[6] Kamide, H.; Hayashi, K.; Toda, S. (1998).<br />
An experimental study of intersubassembly<br />
heat transfer during<br />
natural circulation decay heat removal<br />
in fast breeder reactors. Nuclear<br />
Engineering and Design 183, 97 – 106.<br />
[7] http://sesame-h2020.eu/<br />
[8] Pacio, J, et. al. (2016), Deliverable 2.10 –<br />
KALLA Inter- wrapper flow setup for<br />
SESAME (thermal hydraulics Simulations<br />
and Experiments for the Safety<br />
Assessment of MEtal cooled reactors)<br />
project, activity: NFRP-01-2014<br />
Improved safety design and operation<br />
of fission reactors, H2020 Grant<br />
Agreement Number: 654935.<br />
Authors<br />
Abdalla Batta<br />
Andreas G. Class<br />
AREVA Nuclear Professional School<br />
Karlsruhe Institute of Technology<br />
Karlsruhe, Germany<br />
OPERATION AND NEW BUILD 229<br />
Operation and New Build<br />
Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 230<br />
Heat Balance Analysis for Energy<br />
Conversion Systems of VHTR<br />
SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park<br />
VHTR (Very High Temperature gas Reactor) which helium is used as a coolant can easily produce heat required in<br />
high-temperature thermochemical process, and because of low heat output density, the possibility of core melting is<br />
low. Helium has the advantage of safety, and the coolant can become super high temperature, thereby power production<br />
as well as hydrogen production application is possible. In this study, provided that VHTR is located in the primary<br />
system, the heat conversion system will be discussed in which hydrogen production and power supply are possible. In<br />
order to control the ratio between power and hydrogen production, the helium flowing through nuclear reactor is made<br />
to pass through heat exchanger for hydrogen production and steam generator or heat exchanger. Power production was<br />
made to be composed of ultra-super critical steam cycle (USC) and supercritical CO 2 (SCO 2 ) cycle so that efficient<br />
operation condition can be selected. This study proposed the whole heat conversion system model, and carried out<br />
thermodynamic feasibility calculation according to major design variable at each point and sensitivity analysis for<br />
efficiency optimization.<br />
1 Introduction<br />
Recently, an interest on hydrogen as a<br />
clean energy source and a fossil fuel<br />
substitute has been increasing. From<br />
the viewpoint that hydrogen utilizes<br />
the energy system which uses the<br />
existing fossil fuel without the<br />
emission of environmental pollution<br />
material, contrary to fossil fuels,<br />
hydrogen is emerging as a promising<br />
future clean energy. Among hydrogen<br />
production methods, high-temperature<br />
pyrolysis hydrogen production<br />
method using heat chemical process is<br />
considered as a proper method for<br />
mass hydrogen production. Heat is<br />
required much for high-temperature<br />
heat chemical process, and lightwater<br />
reactor that uses water as coolant<br />
does not produce heat required<br />
for high-temperature heat chemical<br />
process. VHTR (Very High Temperature<br />
gas Reactor) which uses helium<br />
as coolant can easily produce heat<br />
required for high-temperature thermochemical<br />
process, so recently the<br />
study of the use of high temperature<br />
gas for hydrogen production has been<br />
the research trend [1, 2].<br />
VHTR has no possibility of core<br />
melting due to low heat output<br />
density, and it does not use water, so<br />
there is no risk of explosion danger<br />
due to hydrogen generation in the<br />
case of coolant loss accident. Besides,<br />
it has the advantage that high-temperature<br />
coolant can be made compared<br />
to water-cooled reactor, so it has the<br />
advantage of power production and<br />
process heat supply [3]. Nuclear<br />
reactor is in charge of heat supply, and<br />
this can be converted variously to be<br />
used as the production of hydrogen or<br />
power. In this study, by borrowing<br />
general name in the atomic power<br />
field, VHTR is called as a primary<br />
system, the part which hydrogen production<br />
and power supply are possible<br />
through heat conversion, is defined as<br />
the secondary system. Helium flowing<br />
in nuclear reactor delivers the heat of<br />
the primary system to the secondary<br />
system through HX (Heat Exchanger).<br />
Helium flowing through the secondary<br />
system passes first through heat<br />
exchanger where hydrogen production<br />
occurs, and secondly and thirdly<br />
passes through steam generator and<br />
heat exchanger composed of ultrasuper<br />
critical cycle (Ultra- supercritical<br />
steam cycle: USC) and super critical<br />
carbon dioxide (Supercritical CO 2 :<br />
SCO 2 ) cycle, respectively, producing<br />
process heat and power. In this study,<br />
the authors proposed the overall heat<br />
conversion system model, and performed<br />
the thermodynamic feasibility<br />
calculation in accordance with major<br />
design variable at each point and<br />
sensitivity analysis for efficiency<br />
optimization.<br />
2 Research methodology<br />
2.1 Concept and methodology<br />
of hydrogen production<br />
equipment<br />
As a method of hydrogen production<br />
which uses water as a raw material by<br />
using 900 °C heat, high temperature<br />
electrolysis using heat energy simultaneously<br />
and the mixed method of<br />
using thermochemistry process method<br />
and electrolytic method. Recently,<br />
research has been focused on Sulfur-<br />
Iodine thermochemical cycle where<br />
iodide and sulfuric acid were used to<br />
break down water. This is because the<br />
required equipment can be scaled up<br />
and process handling material is only<br />
composed of gas and liquid so that<br />
continuous operation is possible.<br />
Besides, it is advantageous to use<br />
nuclear reactor where the safety of<br />
load change is demanded as heat<br />
source [2].<br />
In the hydrogen production equipment<br />
where high temperature heat is<br />
used, according to the Reaction 1<br />
below, sulfuric acid (H 2 SO 4 ) can be<br />
broken down into water vapor<br />
(H 2 O(g)), oxygen (O 2 (g)), and sulfur<br />
dioxide (SO 2 (g)).<br />
Reaction 1:<br />
2H 2 SO 4 + Heat 2H 2 O + 2SO 2 + O 2<br />
After decomposition, oxygen(O 2 (g))<br />
is removed, and water vapor(H 2 O(g))<br />
and sulfur dioxide (SO 2 (g)) are cooled<br />
down, reacting with iodide (I).<br />
According to Reaction 2 below,<br />
sulfuric acid (H 2 SO 4 ) and hydrogen<br />
iodide (HI) are formed.<br />
Reaction 2:<br />
4H 2 O + 2SO 2 + 2I 2 2H 2 SO 4 + 4HI<br />
+ Heat<br />
Finally, by using high temperature<br />
heat, hydrogen Iodide (HI) can be<br />
separated into hydrogen (H 2 ) and<br />
iodide (I) according to the reaction 3<br />
below.<br />
Reaction 3: 4HI + heat 2I 2 + 2H 2<br />
2.2 The concept and status<br />
of USC and S-CO 2 cycle<br />
USC power plant means the power<br />
plant where vapor pressure is 254 kg/<br />
cm 2 or higher, and main vapor’s<br />
or reheated vapor’s temperature is<br />
593 °C or higher. The reasons why<br />
pressure and temperature of the<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
evaporator are elevated are that the<br />
efficiency of power plant is increased.<br />
When the temperature of main evaporator<br />
and reheating steam increases<br />
by 10 °C, the efficiency increases<br />
by 0.5 %; and pressure increases by<br />
10 kg/cm 2 , the efficiency increases by<br />
about 0.2 %. Domestically, in 1990’s,<br />
500 MW-grade standard coal thermal<br />
power plant was designed and built,<br />
and its operation condition was pressure<br />
246 kg/cm 2 and temperature<br />
538 °C.<br />
In the case of Dangjin Thermal<br />
Power No. 9, No. 10 and Samcheok<br />
Thermal Power No. 1, No .2 that have<br />
been being built, the pressure of<br />
250 kg/cm 2 , temperature of 600 °C<br />
were accomplished [4].<br />
SCO 2 cycle is the power generation<br />
technology of the Gas Brayton Cycle<br />
method where pressurized carbon<br />
dioxide is heated by the pressure<br />
greater than critical condition to high<br />
temperature and turbine is driven.<br />
Presently, CO 2 power generation cycle<br />
can be applied to most heat sources<br />
used, and also it can be used for large<br />
power plant, small scale distribution<br />
power supply, or power supply for<br />
marine plant.<br />
Super critical condition means the<br />
conditions for temperature and pressure<br />
greater than critical point in the<br />
general material state where liquid-gas<br />
phase change occurs, and the<br />
temperature and pressure at the lower<br />
pressure part is greater than 32 °C, 74<br />
atm, and all parts of cycle are maintained<br />
over critical condition. While<br />
operation is carried out at high<br />
pressure, volumetric flow decreases,<br />
so the size of overall heat conversion<br />
cycle can be decreased; accordingly,<br />
construction period and production<br />
unit price can be lowered to secure<br />
high economic feasibility.<br />
Besides, compared to water vapor,<br />
the compatibility with existing material<br />
is excellent, so it can be supplied<br />
to turbine at the temperature higher<br />
than evaporator cycle. From this, the<br />
increase of additional power generation<br />
efficiency can be possible [5].<br />
2.3 Heat Conversion Model<br />
Design<br />
IHX loop of VHTR that is studied in<br />
the present study is the system where<br />
the high temperature heat generated<br />
in the reactor by connecting hydrogen<br />
generation equipment and power<br />
generation equipment in series can be<br />
supplied in the same manner.<br />
IHX loop nuclear reactor shown in<br />
Figure 1 provides 350 MWt heat output,<br />
and the heat generated from<br />
| | Fig. 1.<br />
IHX Loop Modelling.<br />
nuclear fission is supplied to helium<br />
fluid. For heat transfer to produce<br />
hydrogen, heat exchanger, steam generator<br />
for the power generation via<br />
USC cycle, and in the power generation<br />
via SCO 2 , one heat exchanger is<br />
provided. In order to utilize the result<br />
of the study regarding the existing<br />
VHTR, the major principle and<br />
variable if heat conversion model<br />
were set as follows. Temperature and<br />
pressure at No. 1, 2, 3, 4, 10 were<br />
presumed by reference literature [8].<br />
Temperature and pressure of ultrasuper<br />
critical cycle No. 5, 6 and SCO 2<br />
cycle, No. 8 were assumed by using<br />
reference literature [9]. The model to<br />
be explained below was defined as<br />
reference model, and then the present<br />
authors will plan to develop a model<br />
that considers a variety of heat<br />
efficiency improvement method. In<br />
the present study, in the concept<br />
similar to general Rankine cycle’s<br />
reheating cycle, bypass mode was<br />
proposed.<br />
To begin with, the reference model<br />
is as follows. After 910 °C helium fluid<br />
discharging from VHTR carries out<br />
heat exchange with heat exchanger 1,<br />
hydrogen is produced by receiving<br />
heat from high temperature helium<br />
fluid in the heat exchanger 1. 846 °C<br />
helium fluid passing heat exchanger 1<br />
enters into steam generator 2 and go<br />
through heat exchange. The fluid of<br />
this steam generator is ultra-super<br />
critical state water, and produces<br />
power. The temperature of helium<br />
fluid that passes through steam<br />
generator 2 is 614.8 °C, this helium<br />
fluid enters into heat exchanger 3<br />
where heat exchange is carried out.<br />
The fluid of this heat exchanger is<br />
super critical-state carbon dioxide,<br />
and it produces power by the heat<br />
supplied. The temperature of helium<br />
fluid coming out of heat exchanger 3<br />
is 450 °C. The heat output that is produced<br />
in heat exchanger 1 producing<br />
hydrogen is 37.37 MWt. The mass flow<br />
of helium from IHX is m 1 , and the<br />
mass flow of water flowing in heat<br />
exchanger 1 is m 2 , the mass flow of<br />
water flowing in steam generator 2 is<br />
m 3 , and the mass flow of CO 2 flowing<br />
in heat exchanger 3 is m 4 . In this<br />
study, the temperatures and pressures<br />
from No.1 to No.10 in Figure 1 were<br />
assumed, and m 1 and m 2 were calculated<br />
by using the Equation (1), and<br />
m 3 and m 4 were calculated by using<br />
the Equation (2). Besides, considering<br />
the characteristics of general longitudinal<br />
temperature difference of heat<br />
exchanger, the temperature at No. 6<br />
and No. 9 was assumed to decrease by<br />
10 °C compared to the temperature at<br />
No. 4 and No. 7 of the steam generator<br />
inlet.<br />
Major equation or relationship for<br />
heat equilibrium analysis is as follows:<br />
• Equation used for calculating m 1<br />
and m 2<br />
: W = m∆h = m(h out – h in )... (1)<br />
Here,W : Thermal power (MWt)<br />
m : Mass flow (kg/hr)<br />
h : Enthalpy (kJ/kg)<br />
in : Entrance of the equipment<br />
out : Outlet of equipment<br />
• Equation used for calculating m 5<br />
and m 8<br />
∑m in h in = ∑m out h out ... (2)<br />
In the case of hydrogen production, it<br />
was assumed that all heat was<br />
converted to work required, and in<br />
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OPERATION AND NEW BUILD 232<br />
the case of power production, it was<br />
assumed that only a part of the heat<br />
delivered was converted to electricity.<br />
Besides, it was considered that the<br />
pumping power was consumed due to<br />
the flow in the power generation,<br />
therefore it was considered in the<br />
calculation of efficiency.<br />
The general efficiency of USC cycle<br />
and SCO 2 cycle was 43 % and 45 %,<br />
respectively. Using Equation (3),<br />
efficiency was corrected, and more<br />
realistic calculation was carried out<br />
[6].<br />
• efficiency correction equation<br />
η Oper = [1.0+{(T h,oper – T h,des ) × C}]<br />
× η Des … (3)<br />
Here, η Des : standard efficiency<br />
according to reference<br />
literature<br />
η Oper : Standard efficiency’s<br />
correction efficiency according<br />
to high temperature<br />
T h,des : Standard exit temperature in<br />
Steam Generator tube according<br />
to reference literature<br />
T h,oper : Exit temperature within<br />
specified range at Steam<br />
Generator or Heat exchanger<br />
tube<br />
C : Efficiency correction factor;<br />
USC : 0.3 % / 5 °C [6],<br />
SCO 2 : 1.0 % / 5 °C applied<br />
(assumption)<br />
The output of steam generator 2 and<br />
heat exchanger 3 is as follows, and<br />
total output W gross is the sum of all the<br />
values.<br />
W 1 = m 2 × (h 3 – h 2 ) kJ/hr<br />
W 2 = η 2,Operator × m 3 ×(h 4 – h 7 ) kJ/hr<br />
W 3 = η 3,Operator × m 4 ×(h 7 – h 10 ) kJ/hr<br />
Here, w pump : Work used<br />
in the pump (MWt)<br />
η pump : Pump efficiency<br />
v : Specific volume (m 3 /kg)<br />
P : pressure (kPa)<br />
In order to simulate the above model,<br />
the flow of helium gas, water, and<br />
carbon dioxide was calculated by<br />
using thermodynamic system analysis<br />
software, EES (Engineering Equation<br />
Solver).<br />
The following is regarding IHX<br />
loop model to which Bypass mode was<br />
added. Bypass mode was added to the<br />
existing IHX loop, and the efficiency<br />
improvement of overall heat conversion<br />
cycle was studied. The temperature<br />
of the entrance of evaporator 2<br />
and heat exchanger 3 was reheated<br />
by using high temperature helium<br />
coming from IHX, and the output<br />
change was studied.<br />
In the same way as the existing IHX<br />
loop, VHTR supplies heat generated<br />
by nuclear fission in 350 MWt nuclear<br />
reactor. The fluid coming from IHX is<br />
helium, and the fluid flowing in heat<br />
exchanger 1 is water, the fluid flowing<br />
in steam generator 2 is ultra-supercritical-state<br />
water, and the fluid<br />
flowing in heat exchanger 3 is super<br />
critical-state carbon dioxide.<br />
The mass flow of helium from<br />
IHX is m 1 , and m 1 is divided into m 2<br />
and m 3 , and m 3 enters into heat<br />
exchanger 1, and do heat exchange<br />
with the water flowing in heat exchanger<br />
1. At this time, the mass flow<br />
of water flowing in heat exchanger 1<br />
is m 5 . and m 2 is divided into m 4 and<br />
m 9 , and m 4 enters into No. 7 in order<br />
to reheat helium that went through<br />
heat exchange in heat exchanger 1,<br />
and the reheated temperature is that<br />
of No. 8. m 9 enters No. 12 in order to<br />
reheat helium that went through heat<br />
exchange in the steam generator 2,<br />
and the reheated temperature is the<br />
temperature of No. 13. The mass flow<br />
in steam generator 2 is m 10 , and mass<br />
flow of CO 2 in heat exchanger 3 flowing<br />
through heat exchanger 3 is m 14 .<br />
The temperature and pressure at<br />
every point except No. 12 were assumed,<br />
and m 1 value was obtained by<br />
using Equation (1) in the same as m 1<br />
of the existing IHX Loop.<br />
At this time, the temperature of<br />
No.7 and No.12 were assumed to be<br />
that of No. 4 and No. 7 of the existing<br />
IHX loop. Besides, it was assumed that<br />
the temperatures of No. 8 and No. 13<br />
increased to 860 °C and 620 °C,<br />
respectively due to m 4 and m 9 . By<br />
this, the change of the output and<br />
efficiency on the cycle of steam generator<br />
2 and heat exchanger 3. Heat<br />
exchanger 1 in accordance with the<br />
addition of Bypass mode was assumed<br />
to produce the same output, 37.37<br />
MWt, as the existing IHX loop, and<br />
fixed m 5 value.<br />
Considering the characteristics of<br />
the general longitudinal temperature<br />
difference of the heat exchanger as<br />
the existing IHX loop, it was assumed<br />
that the temperatures of No. 11 and<br />
No. 15 decrease by 10 °C compared to<br />
that of No. 8 (at Steam Generator<br />
entrance) and No. 13 (at Heat<br />
Exchanger). With the obtained m 1<br />
value, m 14 value was calculated by<br />
using Equation (2). After that, m 5<br />
value was fixed to the value which can<br />
make the output as the existing IHX<br />
loop, then m 3 value was obtained<br />
by Equation (2). m 2 was calculated by<br />
m 2 = m 1 – m 3 , and m 4 was obtained by<br />
using Equation (2). m 9 was obtained<br />
Here, W 1 : heat exchanger 1<br />
heat output (MWt)<br />
W 2 : Steam generator 2<br />
heat output (MWt)<br />
W 3 : Heat exchanger 3<br />
heat output (MWt)<br />
In steam generator 2 and heat<br />
exchanger 3 in order to consider<br />
pumping power in accordance with<br />
mass flow, pump’s efficiency (η pump )<br />
was assumed to be 0.9, and Equation<br />
(4) was used.<br />
w pump = η pump × m × v out<br />
× (P out – P in ) kJ/hr … (4)<br />
W net = W gross – w pump<br />
| | Fig. 2.<br />
Bypass mode-added IHX loop Modelling.<br />
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by m 9 = m 2 – m 4 , and the temperature<br />
of No. 12 can be calculated by using<br />
Equation (2). Finally, m 10 was also<br />
obtained by using Equation (2). Like<br />
the existing IHX loop, in the bypass<br />
mode-added IHX loop, correction<br />
efficiency and pumping power in<br />
accordance with mass flow were<br />
considered, and pumping power used<br />
the above Equation (4). For the<br />
simulation for this, thermodynamic<br />
system analysis software, EES, was<br />
used, in the same way with the<br />
existing IHX loop model obtained<br />
before, and the flow of helium gas<br />
and fluid was analyzed.<br />
3 Result<br />
Table 1 shows the result of physical<br />
value at each point by simulating the<br />
existing IHX loop EES [5]. Physical<br />
value of each point was assumed in<br />
accordance with reference [8], [9]<br />
literature, and the assumed values<br />
were colored.<br />
The temperature of helium fluid<br />
that leaves from the first heat exchanger<br />
after producing the hydrogen<br />
decreases to 846 °C from 910 °C, and<br />
the temperature of helium fluid that<br />
leaves from the second steam generator<br />
is 614.8 °C, and the temperature<br />
of helium fluid that leaves from the<br />
last heat exchanger is 450 °C. The<br />
temperature of helium fluid decreases<br />
steadily, but because the fluid flowing<br />
each steam generator and heat<br />
exchanger is different, efficient electricity<br />
can be produced by using each<br />
characteristics. The existing IHX<br />
Loop’s m 5 and m 8 are in inverse<br />
proportion, as more mass flow moves<br />
toward high efficiency, the amount of<br />
overall electricity output increases.<br />
Although the efficiency of heat<br />
conversion cycle connected to each<br />
steam generator may be influenced by<br />
various causes, but in the present<br />
study, correction factor presumed<br />
about high temperature was used, so<br />
the detailed design for this part would<br />
be needed.<br />
If heat conversion cycle connected<br />
to each steam generator should be<br />
operated simultaneously by a specific<br />
objective, considering the inverse proportion<br />
relationship between m 5 and<br />
m 8 , the output must be distributed.<br />
Besides, the exit temperature at the<br />
tube part of steam generator 2 is in<br />
inverse proportion with the exit<br />
temperature at the shell part. This will<br />
eventually influence on the exit<br />
temperature of the tube part of the<br />
heat exchanger 3. When operating<br />
heat conversion cycle connected to<br />
each steam generator, it is necessary<br />
to find balanced point on the temperature<br />
between steam generators.<br />
In the steam generator 2 and heat<br />
exchanger 3, exit temperature and<br />
mass flow are in inverse proportion.<br />
This is because if high exit enthalpy is<br />
maintained in order to deliver the<br />
same heat energy, less mass flow is<br />
needed, and if a large amount of mass<br />
flow is needed, exit enthalpy should<br />
be maintained low. Maximum output<br />
would be in the parabolic form as exit<br />
enthalpy and temperature change, so<br />
if maximum output is needed, proper<br />
exit temperature must be selected. Or<br />
in case there is a requirement for exit<br />
temperature, it is possible that output<br />
would be determined according to<br />
that.<br />
Table 2 shows the physical value at<br />
each point where IHX loop added by<br />
bypass mode is simulated with EES.<br />
No. Fluid Temperature<br />
(°C)<br />
| | Tab. 1.<br />
IHX loop Simulation Result.<br />
In the case of IHX loop to which<br />
bypass mode was added, the helium<br />
fluid that passed through the first<br />
hydrogen-producing heat exchanger<br />
is 910 °C~ 846 °C, which is the same<br />
as the existing IHX loop, but here by<br />
reheating high-temperature helium<br />
fluid, the temperature increases to<br />
860 °C. The temperature of the helium<br />
fluid that passed through the second<br />
steam generator is 614.8 °C, which<br />
is the same as that of helium fluid<br />
that passed through the second evaporator.<br />
However, since the temperature at<br />
the entrance reheated, and returned,<br />
the amount of electricity output produced<br />
increases. When helium fluid<br />
enters the third heat exchanger, it is<br />
reheated from 614.8 °C to 620 °C, the<br />
temperature of helium fluid is 450 °C,<br />
and the amount of electricity output<br />
Pressure<br />
(kPa)<br />
| | Tab. 2.<br />
Result of IHX Loop to which Bypass Mode was added.<br />
Enthalpy<br />
(kJ/kg)<br />
Mass flow<br />
(kg/hr)<br />
1 Helium 910.0 4000 6,161.00 527,662<br />
2 Water 193.0 18,000 828.70 49,839<br />
3 Water 585.0 16,500 3,528.00 49,839<br />
4 Helium 846.0 4,000 5,829.00 527,662<br />
5 Water 260.2 20,790 1,134.00 208,359<br />
6 Water 836.0 16,475 4,174.00 208,359<br />
7 Helium 614.8 4,000 4,628.00 527,662<br />
8 CO 2 203.5 19,760 96.59 902,043<br />
9 CO 2 604.8 19,290 597.00 902,043<br />
10 Helium 450.0 4,000 3,773.00 527,662<br />
No. Fluid Temperature<br />
(°C)<br />
Pressure<br />
(kPa)<br />
Enthalpy<br />
(kJ/kg)<br />
Mass flow<br />
(kg/hr)<br />
1 Helium 910.0 4,000 6,161.00 527,662<br />
2 Helium 910.0 4,000 6,161.00 122,749<br />
3 Helium 910.0 4,000 6,161.00 404,913<br />
4 Helium 910.0 4,000 6,161.00 113,375<br />
5 Water 195.0 18,000 837.50 50,000<br />
6 Water 585.0 16,500 3,528.00 50,000<br />
7 Helium 846.0 4,000 5,829.00 404,913<br />
8 Helium 860.0 4,000 5,901.00 518,288<br />
9 Helium 910.0 4,000 6,161.00 9,374<br />
10 Water 260.2 20,790 1,134.00 214,580<br />
11 Water 850.0 16,475 4,209.00 214,580<br />
12 Helium 614.8 4,000 4,628.00 518,288<br />
13 Helium 620.0 4,000 4,655.00 527,662<br />
14 SCO 2 203.5 19,760 96.59 918,581<br />
15 SCO 2 610.0 19,290 603.50 918,581<br />
16 Helium 450.0 4,000 3,773.00 527,662<br />
OPERATION AND NEW BUILD 233<br />
Operation and New Build<br />
Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
OPERATION AND NEW BUILD 234<br />
increases compared to the existing<br />
IHX loop.<br />
The following is the major comparison<br />
of the result of reference<br />
model and bypass mode model.<br />
Mass flow IHX loop Bypass Mode Loop<br />
Heat Exchanger 1 49839 50000<br />
Steam Generator 2 208359 214580<br />
Heat Exchanger 3 902043 918581<br />
| | Tab. 3.<br />
IHX Loop and Bypass Mode IHX Loop: Mass Flow Comparison.<br />
Pumping power<br />
When IHX loop and bypass modeadded<br />
IHX loop were compared, mass<br />
flow of m 10 and m 14 in the bypass<br />
mode-added loop was greater compared<br />
to the mass flow of m 5 and m 8 in<br />
the existing loop, as shown in Table 3.<br />
IHX loop<br />
(MWt)<br />
| | Tab. 4.<br />
IHX Loop and Bypass Mode IHX loop: Pumping Power Comparison.<br />
As shown in Table 4, depending on<br />
the difference of mass flow value,<br />
pumping power used in the pump also<br />
can be high in steam generator 2 and<br />
heat exchanger 3. However, although<br />
pumping power is higher in the bypass<br />
IHX loop, by reheating, efficiency of<br />
steam generator 2 increased from<br />
53.79 % to 54.4 %, and that of heat<br />
exchanger 3 increased from 45.83 %<br />
to 45.87 %; accordingly, it is seen that<br />
the value of Power increased. As a<br />
result, Net Power that considered<br />
pumping power in Total Power was<br />
178.6 MWt in the present IHX<br />
loop, but the IHX loop to which<br />
bypass mode was added increased<br />
to 185.3 MWt, as shown in Table 5.<br />
| | Tab. 5.<br />
IHX loop and Bypass mode IHX loop: Power Comparison.<br />
Bypass mode loop<br />
(MWt)<br />
Steam Generator 2 1.091 1.123<br />
Heat Exchanger 3 9.82 10<br />
Power<br />
IHX loop<br />
(MWt)<br />
Bypass mode loop<br />
(MWt)<br />
Heat Exchanger 1 37.37 37.37<br />
Steam Generator 2 94.63 99.7<br />
Heat Exchanger 3 57.46 59.34<br />
Net Power 178.6 185.3<br />
Since the assumption was that<br />
constant heat was supplied from the<br />
primary system, it is seen that the<br />
efficiency of the bypass model where<br />
net power is high, and it is judged that<br />
efficiency optimization model can be<br />
formulated by detailed design.<br />
4 Conclusion<br />
In this study, VHTR system was<br />
modelled for supplying high temperature<br />
heat, by distribution, produced in<br />
the high temperature gas furnace to<br />
hydrogen producing equipment and<br />
power generation equipment.<br />
Provided that high temperature<br />
gas- cooled reactor is located in<br />
primary system, the secondary system<br />
where hydrogen production and<br />
power supply are possible were<br />
explained. The helium that flows in<br />
the nuclear reactor first passes<br />
through the HX (heat exchanger)<br />
whose purpose is the production of<br />
hydrogen, and secondly and thirdly<br />
pass through the steam generator<br />
composed of super critical carbon<br />
dioxide cycle, and heat exchanger,<br />
respectively, producing the process<br />
heat and power. In order to analyze<br />
existing IHX loop model and bypass<br />
mode-added IHX loop model, the<br />
present authors studied the input &<br />
output conditions and output change<br />
of each steam generator and heat<br />
exchanger, and based on this result,<br />
by designing IHX loop in the power<br />
production part in detail, the authors<br />
performed the calculation of thermodynamic<br />
physical value and efficiency<br />
at each point. Additionally, the<br />
authors studied the change regarding<br />
electricity output and efficiency<br />
according to bypass mode, when<br />
reheating cycle is added, the possibility<br />
on the efficiency optimization<br />
was proposed.<br />
References<br />
[1] Kim. Y. W., 2015, Nuclear Hydrogen<br />
Production Technology development<br />
Using Very High Temperature Reactor,<br />
Trans. Korean Soc. Mech. Eng. C, Vol. 3,<br />
No. 4, pp. 299~305.<br />
[2] Chang. J. H., 2006, Current Status of<br />
Nuclear Hydrogen Development,<br />
Journal of Energy Engineering, Vol.15,<br />
No.2, pp. 127~137.<br />
[3] Lee. S. I., 2015, Heat Balance Study<br />
on Integrated Cycles for Hydrogen<br />
and Electricity Generation in VHTR,<br />
Transaction of the KNS Spring Meeting.<br />
[4] Sung. H. C., 2012, Development of<br />
Ultra-Supercritical (USC) Power Plant,<br />
Trans. Korean Soc. Mech. Eng. B,<br />
Vol. 36, No.2, pp.205~210.<br />
[5] Yeom Chung-seop, Im Dong-ryeol,<br />
Lee Jung-ik, 2014, Trend of Electricity<br />
Generation Technology using supercritical<br />
CO 2 , Institute for Advanced<br />
Engineering, KIC News, Volume 17,<br />
No.1.<br />
[6] K.C.Cotton,1998, Evaluating and<br />
Improving Steam Turbine Performance,<br />
2 nd edition, Cotton Fact Inc.<br />
[7] F-Chart Software, 2016,<br />
Engineering Equation Solver,<br />
http://www.fchart.com/ees/<br />
[8] NGNP Conceptual Design Report/Steam<br />
Cycle Modular Helium Reactor<br />
(SC-MHR) Demonstration Plant,<br />
Table 3-6 SC-MHR Conceptual Design<br />
Point Design Parameter.<br />
[9] SangIL Lee, Yeon Jae Yoo, Gyunyoung<br />
Heo, Soyoung Park, Yeon Kwan Kang,<br />
Heat Balance Study on Integrated<br />
Cycles for Hydrogen and Electricity<br />
Generation in VHTR-Part 2, Korean<br />
Nuclear Society Autumn Meeting,<br />
Oct 28-30, 2015.<br />
Authors<br />
SangIL Lee<br />
YeonJae Yoo<br />
Deok Hoon Kye<br />
Department of Nuclear Team<br />
Power & Energy Plant Division<br />
Hyundai Engineering Company<br />
Seoul, Korea<br />
Gyunyoung Heo<br />
Eojin Jeon<br />
Soyoung Park<br />
Department of Nuclear<br />
Engineering<br />
Kyung Hee University<br />
Yongin Korea<br />
Operation and New Build<br />
Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Informationsbedarf versus Geheimhaltungspflichten –<br />
Erweiterung des In-camera-Verfahrens geplant<br />
235<br />
Tobias Leidinger<br />
Dem berechtigten Anspruch der Öffentlichkeit auf detaillierte Informationen über ein atomrechtlich genehmigungsbedürftiges<br />
Vorhaben steht das staatliche Interesse an einem effektiven Geheimnisschutz sensibler Daten<br />
gegenüber. Dieser Konflikt tritt regelmäßig im Genehmigungsverfahren aber auch vor Gericht zu Tage. Die differenzierten<br />
Gesetzesbestimmungen, die den Ausgleich dieser widerstreitenden Interessen regeln, sollen nun durch eine<br />
weitere Facette ergänzt werden: Ein erweitertes In-camera-Verfahren bei Gericht. Nach dem Koalitionsvertrag vom<br />
12. März <strong>2018</strong> soll die Regelung in der schon laufenden 18. Legislaturperiode erfolgen.<br />
I Grundkonflikt Informationsbedarf vs.<br />
Geheimhaltungspflicht<br />
In atomrechtlichen Genehmigungsverfahren zeigt sich<br />
regelmäßig ein Grundkonflikt: Dem Interesse der Öffentlichkeit<br />
an möglichst vertieften Informationen über<br />
alle sicherheits- und sicherungsrelevanten Aspekte des<br />
Vorhabens steht das Erfordernis eines effektiven Geheimnisschutzes<br />
in Bezug auf sensible Daten gegenüber.<br />
Genauer betrachtet lassen sich für beide Pole Grundrechtspositionen<br />
anführen: Einerseits ist Information Voraussetzung<br />
für Transparenz und Teilhabe der Öffentlichkeit<br />
am Genehmigungsverfahren. Das Verfahren dient der<br />
Gewährleistung materieller Schutzansprüche Dritter.<br />
Ohne Information ist Kontrolle gegenüber der Verwaltung<br />
kaum realisierbar. Information ist die Grundlage für<br />
Partizipation und Teilhabe der Öffentlichkeit an einem<br />
Verfahren. Das BVerfG bringt dies mit der Formel „Grundrechtsschutz<br />
durch Verfahren und Teilhabe an Information“<br />
auf den Punkt.<br />
Für die andere Seite, dem Interesse an Geheimhaltung<br />
sensibler Daten, lassen sich aber nicht minder gewichtige<br />
Grundrechtsinteressen anführen: Die Geheimhaltung<br />
dient ebenfalls zum Schutz der Grundrechtsträger: Ist der<br />
Staat zum Schutz der Grundrechte („Leben, Gesundheit“)<br />
seiner Bürger verpflichtet, bedarf es des Geheimnisschutzes<br />
in Bezug auf sensible Daten, damit eine effektive<br />
Terrorabwehr – gerade zum Schutz der Bürger – gewährleistet<br />
bleibt. Die Nicht-Preisgabe sicherheits- und<br />
sicherungsrelevanter Informationen ist mithin nicht<br />
minder essentielle Voraussetzung für einen effektiven<br />
Grundrechtsschutz der Bürger.<br />
II Interessenausgleich durch differenzierte<br />
Gesetzesregelungen<br />
Der Gesetzgeber trägt zur Lösung dieser widerstreitenden<br />
Interessen im atomrechtlichen Genehmigungsverfahren<br />
bereits heute durch eine ganze Reihe differenzierter<br />
Regelungen bei. Nach § 6 der Atomrechtlichen Verfahrensordnung<br />
(AtVfV) sind nicht nur der Antrag, der Sicherheitsbericht<br />
und eine Kurzbeschreibung des jeweils zu<br />
genehmigenden Vorhabens für die Öffentlichkeit auszulegen,<br />
sondern es besteht nach § 6a Abs. 2 Satz 1 und<br />
Abs. 3 AtVfV die Möglichkeit, in Bezug auf das Vorhaben –<br />
im Interesse der Sicherheit und Sicherung – geheimhaltungsbedürftige<br />
Informationen durch eine Beschreibung<br />
oder Inhaltsdarstellung zu ersetzen. Anstelle einer<br />
„Schwärzung“ von Unterlagen – die letztlich eine „Verweigerung“<br />
von Information bedeutete –, tritt so die<br />
Möglichkeit, geheimhaltungsbedürftige Informationen zu<br />
umschreiben, so dass der Dritte in der Lage bleibt, seine<br />
Betroffenheit durch das Vorhaben gleichwohl erkennen<br />
und beurteilen zu können.<br />
Eine Einschränkung von Informationsansprüchen ist<br />
auch jenseits dieser Regelung möglich: Während eines<br />
atomrechtlichen Verfahrens besteht der Anspruch auf<br />
Akteneinsicht gemäß § 6 Abs. 4 AtVfV i.V.m. § 29 Abs. 1<br />
S. 3, Abs. 2 und 3 des Verwaltungsverfahrensgesetzes<br />
(VwVfG) nur nach Ermessen der Behörde (also nicht<br />
„ unbedingt“). Informationen, die sicherheits- oder<br />
sicherungsrelevant sind, weil sie den Ansatz für die Ausschaltung<br />
von Sicherheits- und Sicherungsmaßnahmen<br />
oder für die Identifizierung/Lokalisierung von Schwachstellen<br />
eröffnen könnten, können – soweit durch ihre<br />
Preisgabe ein „Nachteil zum Wohl des Bundes oder<br />
Landes“ zu befürchten ist – von der Offenlegung ausgeschlossen<br />
werden. Spezialgesetzlich ist die Geheimhaltung<br />
von sensiblen Informationen im Sicherheitsüberprüfungsgesetz<br />
(SÜG) geregelt. Besteht danach die Gefahr<br />
eines „Nachteils“ oder wäre die Preisgabe der Information<br />
sogar „schädlich“ für Bund oder Land, kann sie<br />
nach Maßgabe der Verschlusssachen-Anweisung (VS-<br />
Anweisung) durch den Geheimschutzbeauftragten der<br />
Behörde als „Verschlusssache – Nur für den Dienstgebrauch“<br />
oder sogar als „Verschlusssache – Vertraulich“<br />
eingestuft und ihre Offenlegung verweigert werden. Was<br />
nach Maßgabe des SÜG i.V.m. VS-Anweisung geheim zu<br />
halten ist, darf auch nicht in anderem Zusammenhang<br />
preisgegeben werden: So bestehen – auch außerhalb eines<br />
atomrechtlichen Verfahrens – Informationsansprüche<br />
Dritter, z.B. auf Herausgabe von umweltrelevanten<br />
Informationen gegen die Genehmigungsbehörde nach<br />
Umweltinformationsgesetz (UIG) oder – soweit die<br />
Informationen nicht umweltrelevant sind – nach Maßgabe<br />
des Informationsfreiheitsgesetzes (IFG). Pressevertreter<br />
können sich darüber hinaus auch auf das jeweilige<br />
Landes-Pressegesetz stützen. In all diesen Fällen besteht<br />
indes die Möglichkeit – mit oder ohne ausdrücklichen<br />
Bezug auf das SÜG –, dass sicherheits- und sicherungsrelevante<br />
Informationen im Ergebnis nicht offenbart<br />
werden müssen, wenn die materiellen Schutzvoraussetzungen<br />
nach SÜG i.V.m. der VS-Anweisung vorliegen.<br />
III In-camera-Verfahren de lege lata und<br />
de lege ferenda<br />
Verweigert die atomrechtliche Genehmigungsbehörde die<br />
Herausgabe sensibler Informationen unter Verweis auf<br />
den Geheimschutz auch im Gerichtsverfahren, – in dem<br />
z.B. über die Rechtmäßigkeit einer atomrechtlichen<br />
Genehmigung gestritten wird – so sieht die bislang<br />
existierende Gesetzesregelung zum sog. In-camera-<br />
Verfahren in § 99 Verwaltungsgerichtsordnung (VwGO)<br />
vor, dass über die Frage der Geheimhaltungsbedürftigkeit<br />
ein speziell besetzter Fachsenat vorab entscheidet. Ihm<br />
sind ausschließlich die geheimhaltungsbedürftigen Akten<br />
vorzulegen („in camera“), um zu prüfen, ob die Einstufung<br />
als „geheim“ zurecht erfolgt ist und daher die Verweigerung<br />
der Aktenvorlage durch die Behörde Bestand hat<br />
oder nicht. Nur wenn die Geheimhaltungsbedürftigkeit<br />
verneint wird, ist die vorenthaltene Information dem<br />
Verwaltungsgericht zugänglich zu machen. Nur dann<br />
kann es darauf zugreifen und seine Entscheidung darauf<br />
stützen.<br />
SPOTLIGHT ON NUCLEAR LAW<br />
Spotlight on Nuclear Law<br />
Information Requirements Versus Confidentiality Obligations – Extension of the In-Camera Procedure Planned ı Tobias Leidinger
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
SPOTLIGHT ON NUCLEAR LAW 236<br />
Der Koalitionsvertrag vom 12. März <strong>2018</strong> (vgl. Seite<br />
141) sieht nun vor, dass die Regelungen für das In-camera-<br />
Verfahren für das Atomrecht dahingehend erweitert<br />
werden sollen, dass geheimhaltungsbedürftige Unter lagen<br />
auch zum Zwecke des Nachweises der Genehmigungsvoraussetzungen<br />
in ein verwaltungsgerichtliches Hauptsacheverfahren<br />
– bei gleichzeitiger Wahrung des Geheimschutzes<br />
– eingeführt werden können. Das In-camera-<br />
Verfahren dient dann nicht (mehr allein) zur Klärung der<br />
Frage der Geheimhaltungsbedürftigkeit einer Unterlage<br />
(wie bisher), sondern ermöglicht darüber hinaus eine<br />
weitergehende Prüfung in der Sache durch das Gericht.<br />
Das Gericht prüft dann auch, ob der erforderliche Schutz<br />
gegen Störmaßnahmen Dritter (SEWD) als gegeben unterstellt<br />
werden darf oder nicht. Die Gewährleistung des<br />
SEWD-Schutzes ist eine wesentliche Tatbestandsvoraussetzung,<br />
die erfüllt sein muss, damit eine atomrechtliche<br />
Genehmigung erteilt werden kann. Dabei ist aber auch in<br />
einem erweiterten In-camera- Verfahren sicherzustellen,<br />
dass die behördliche Ein schätzungsprärogative in Bezug<br />
auf genehmigungs relevante Wertungen bei Sicherheit und<br />
Sicherung beachtet werden. Das bedeutet, dass das Gericht<br />
sich nicht an die Stelle der Behörde setzen darf, also eine<br />
eigene Entscheidung anstelle der Behörde trifft, sondern<br />
bei seiner Nachprüfung auf eine Vertretbarkeitskontrolle<br />
beschränkt bleibt.<br />
die Frage, ob im Ergebnis davon ausgegangen werden darf,<br />
dass die erforderliche Schadensvorsorge und der gebotene<br />
Schutz gegen SEWD-Ereignisse gewährleistet ist oder<br />
nicht, könnte auf diese Weise weitergehend als bisher entschärft<br />
werden. Idealerweise bliebe der Geheimnisschutz<br />
auch so gewahrt, zugleich aber wäre dem Interesse<br />
der Drittbetroffenen an einer Überprüfung essentieller<br />
Genehmigungsvoraussetzungen unter Berücksichtigung<br />
geheimhaltungsbedürftiger Informationen weitergehend<br />
als bisher entsprochen. Das wäre als konstruktiver Beitrag<br />
zur Stärkung eines effektiven Grundrechtsschutzes zu<br />
bewerten: Ein erweitertes In-camera-Verfahren diente so<br />
in besonderer Weise zur Gewährleistung der dem Dritten<br />
zustehenden Schutzansprüche und wahrte dabei<br />
gleichwohl den erforderlichen Geheimschutz, der nicht<br />
minder einem effektiven Grundrechtsschutz der Bürger<br />
geschuldet ist.<br />
Allerdings bleiben die konkrete Ausgestaltung und der<br />
Vollzug dieser Regelung in der Praxis abzuwarten: Folgt<br />
einer guten Absicht des Gesetzgebers eine in der Praxis<br />
tatsächlich und rechtlich brauchbare Lösung? Ziel müsste<br />
sein, dadurch nicht neue Fragen zur Anwendung und<br />
Reichweite eines erweiterten In-camera-Verfahrens<br />
aufzuwerfen, sondern eine inhaltlich klare und hinreichend<br />
bestimmte Norm zu schaffen, die das Versprechen<br />
des Koalitionsvertrages vollzugsfähig einlöst.<br />
IV Erweiterung des In-camera-Verfahrens:<br />
Bedenkenswerter Schritt<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
238<br />
ENVIRONMENT AND SAFETY<br />
CFD Modeling and Simulation<br />
of Heat and Mass Transfer in<br />
Passive Heat Removal Systems<br />
Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas<br />
This paper is presenting the CFD-modelling and simulation of condensation inside passive heat removal systems.<br />
Designs of future nuclear boiling water reactor concepts are equipped with emergency cooling systems which are<br />
passive systems for heat removal. The emergency cooling system consists of slightly inclined horizontal pipes which are<br />
immersed in a tank of subcooled water. At normal operation conditions, the pipes are filled with water and no heat<br />
transfer to the secondary side of the condenser occurs. In the case of some accident scenarios the water level may<br />
decrease in the core, steam enters the emergency pipes and due to the subcooled water around the pipe, this steam<br />
condenses. The emergency condenser acts as a strong heat sink which is responsible for a quick depressurization of the<br />
reactor core. This procedure acts passive i.e. without any additional external measures. The actual project is defined to<br />
model the phenomena which are occurring inside the emergency condensers. The focus of the project is on detection of<br />
different morphologies such as annular flow, stratified flow, slug flow and plug flow and also modeling of the laminar<br />
film which is occurring during the condensation near the wall.<br />
The condensation procedure inside the<br />
pipe is determined by two important<br />
phenomena. The first one is wall<br />
condensation and the second one is the<br />
direct contact condensation (DCC).<br />
The Algebraic Interfacial Area Density<br />
(AIAD) concept is used in order to<br />
model the interface between liquid<br />
and steam. In the next steps the Generalized<br />
Two-Phase Flow ( GENTOP)<br />
model will be used to model also the<br />
dispersed phases which are occurring<br />
inside the pipe. Finally, the results of<br />
the simulations will be validated by<br />
experimental data which will be available<br />
in HZDR. In this paper the results<br />
of the first part are presented.<br />
1 Introduction<br />
Condensation plays a crucial role in<br />
the emergency condenser of passive<br />
heat removal systems of nuclear power<br />
plants. Passive safety systems do not<br />
need any external power supplies and<br />
they mostly depend on physical phenomena<br />
such as natural circulation<br />
and gravity driven flows. In order to<br />
assess the performance of passive safety<br />
systems and their efficiency mostly<br />
one-dimensional codes are used such<br />
as ATHLET, RELAP and TRACE. These<br />
codes are able to calculate most of the<br />
phe nomena in power plants; however,<br />
they cannot reflect the 3D phenomena.<br />
Therefore, Computational Fluid<br />
Dynamics (CFD) methods should be<br />
used to simulate and predict the<br />
complex multiphase flow structure.<br />
Despite the previous research being<br />
done on the two-phase flow behavior,<br />
this phenomenon needs much more<br />
investigations. The two-phase flow<br />
patterns and transition between vapor<br />
and liquid are studied by Thome and<br />
Hajal et al. [1, 2]. They introduced a<br />
logarithmic mean void fraction (LMe)<br />
method in order to calculate the vapor<br />
void fractions which change from the<br />
low pressure up to the critical pressure<br />
point. Moreover, they proposed a new<br />
heat transfer model based on the same<br />
simplified flow structures that have<br />
been used in the flow boiling model<br />
of Kattan et al. [3]. The model can<br />
predict the local condensation heat<br />
transfer coefficient for different flow<br />
regimes such as annular, intermittent,<br />
stratified-wavy fully stratified and<br />
wavy flow.<br />
Many attempts have been done to<br />
investigate the mass transfer between<br />
liquid and gas phase in condensation.<br />
Lee et al. [4] introduced a model for<br />
prediction of the mass transfer. They<br />
assumed that the interface between<br />
liquid and steam is on saturation<br />
temperature and introduced an<br />
iterative technique in order to reach to<br />
desired boundary condition inside<br />
each cell. This model depends on a<br />
relaxation factor which needs to be<br />
tuned. The tuning needs many trial<br />
and error simulations which is<br />
time-consuming and doesn’t have any<br />
predictive capabilities.<br />
Moreover, there are empirical or<br />
semi-empirical methods to calculate<br />
the mass transfer in the interface.<br />
Strubelj et al. [5] by using ANSYS CFX<br />
and NEPTUNE_CFD [6] code tried to<br />
simulate Direct Contact Condensation<br />
(DCC) in stratified flows. In DCC the<br />
phase change occurs due to the direct<br />
contact interaction of subcooled water<br />
and saturated steam. The defined<br />
phase change mass flux depends on<br />
thermal conductivity of the liquid and<br />
Nusselt number of the liquid. The<br />
Nusselt number was calculated<br />
by Coste et al. [7] based on Surface<br />
Renewal Theory (SRT) [8]. The SRT<br />
theory calculates the mass transfer<br />
according to the renewal period of<br />
eddies and the liquid turbulent<br />
properties. Hughes and Duffey [9]<br />
used the surface renewal theory and<br />
the Kolmogorov turbulent length<br />
scale theory to define a correlation for<br />
the heat transfer coefficient. They<br />
considered that the heat removal from<br />
interface occurs by smallest turbulent<br />
scales. This model will be introduced<br />
more detailed in the next sections.<br />
This correlation is validated for<br />
Pressurized Thermal Schock (PTS)<br />
phenomenon by Egorov [10] and<br />
Apanasevich [11]. Further to Hughes<br />
correlation, Shen et al. [12] developed<br />
another correlation for calculation of<br />
heat transfer coefficient based on the<br />
surface renewal theory. Ceuca et al.<br />
[13] used both of these correlations<br />
in order to simulate the direct contact<br />
condensation for the LAOKOON<br />
facility [14]. By comparison of Hughes<br />
and Duffey correlation with Shen<br />
correlation, Ceuca et al. [13] concluded<br />
that both of the models provide<br />
accurate results for the horizontal<br />
stratified quasi-steady state.<br />
Evidently, many attempts have been<br />
done in the modeling of con densation<br />
inside the pipes. The goal of the current<br />
work is modeling of the transition<br />
between different mor phologies which<br />
are occurring during the condensation<br />
inside the pipe ( Figure 1). In order to<br />
do that, several CFD models such as<br />
IMUSIG, AIAD and GENTOP which<br />
have been developed in HZDR in cooperation<br />
with ANSYS are available. The<br />
Inhomo geneous MUSIG model considers<br />
the bubble size distribution and<br />
is used for modeling the small-scaled<br />
dispersed gas phase [15]. The AIAD<br />
Environment and Safety<br />
CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems<br />
ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
| | Fig. 1.<br />
Schematic representation of horizontal flow and different morphologies.<br />
(Algebraic Interfacial Area Density<br />
Model) is developed for detection of<br />
the local morphology and corresponding<br />
switch between them [16]. The<br />
recently developed GENTOP- model<br />
combines both concepts. GENTOP<br />
(Generalized Two-Phase Flow) approach<br />
is able to simulate co-existing<br />
large-scaled (continuous) and smallscaled<br />
(polydispersed) structures [17].<br />
All these models are validated for adiabatic<br />
cases without any phase change.<br />
Therefore, the start point of the current<br />
work project is using the available<br />
models and integrating phase transition<br />
and con densation models into<br />
them. In the current work as initial<br />
stages the AIAD model has been used<br />
since in this model 2 continues phases<br />
should be considered and it is less complicated<br />
compare to GENTOP model<br />
which also considers a poly- dispersed<br />
phase. In the proceeding sections a<br />
more detail explanation of AIAD model<br />
will be given.<br />
2 CFD model formulation<br />
In the current work a multi-field twophase<br />
CFD approach is used with<br />
ANSYS CFX 17.2 in order to simulate<br />
the condensation inside horizontal<br />
pipe flows. The mass, momentum and<br />
energy equations can be defined,<br />
respectively, as follow:<br />
• Mass conservation equation:<br />
(1)<br />
where S Mi describes user specified<br />
mass source.<br />
χ iβ the mass flow rate per unit volume<br />
from phase β to phase i.<br />
• Momentum conservation equation:<br />
(2)<br />
where S mi is the momentum source<br />
caused by external body forces<br />
and user defined momentum<br />
sources.<br />
M i is the interfacial forces acting<br />
on phase i due to the presence<br />
of other phases.<br />
χ + iβ v β – χ + βi v i is the momentum<br />
transfer induced<br />
by mass transfer.<br />
• The total energy equation:<br />
(3)<br />
where: h tot is the total enthalpy<br />
related to static enthalpy by:<br />
(4)<br />
<br />
T i , λ i represents the temperature<br />
and the thermal conductivity<br />
of phase i.<br />
S Ei describes external heat sources.<br />
Q i is interphase heat transfer<br />
to phase i across interfaces<br />
with the other phase.<br />
χ + iβ h βs – χ + βi h is denotes the interphase<br />
mass transfer.<br />
In ANSYS CFX in order to describe the<br />
phase change which occurs due to the<br />
interphase heat transfer, the Thermal<br />
Phase Change Model has been introduced<br />
[30]. This model is particularly<br />
useful in simulation of the condensation<br />
of saturated vapor. The heat<br />
flux from the interface to phase i and<br />
phase β is:<br />
q i = h i (T sat – T i ) (5)<br />
q β = h β (T sat – T β ) (6)<br />
where h i , h β and T sat are heat transfer<br />
coefficients of the phase i and phase<br />
β and the saturation temperature,<br />
respectively. ṁ iβ is the mas flux from<br />
phase β to phase i. H is and H βs are the<br />
interfacial enthalpy values which<br />
come into and out of the phase due<br />
to phase change which occurs. By<br />
usage of the total heat balance<br />
equation the interphase mas flux can<br />
be determined as follow:<br />
<br />
| | Fig. 2.<br />
3D geometry of the pipe and mesh of the cross section.<br />
(7)<br />
ṁ iβ > 0 → H is = H i,sat , H βs = H β (8)<br />
ṁ iβ < 0 → H is = H i , H βs = H β,sat (9)<br />
In the current work, the steam<br />
con sidered to be in saturation temperature.<br />
Therefore, the heat flux<br />
from the steam to the interface equals<br />
zero since both are in saturation<br />
temperature. As a result, the interphase<br />
mass flux formula can be<br />
written as:<br />
<br />
(10)<br />
In this work in order to model the heat<br />
transfer coefficient the Hughes and<br />
Duffy model has been used which is<br />
based on the SRT model [9]. They<br />
used the Surface Renewal Theory<br />
(SRT) and the Kolmogorov turbulent<br />
length scale theory to find a correlation<br />
for heat transfer coefficient.<br />
Therefore, the heat transfer coefficient<br />
was derived as:<br />
<br />
(11)<br />
where ε is the turbulent dissipation, v l<br />
is the kinematic viscosity and λ is the<br />
thermal conductivity.<br />
3 Computational grid and<br />
boundary conditions<br />
In Figure 2 the pipe and the boundary<br />
conditions are shown. The pipe is<br />
horizontal and has 1 m length<br />
and 0.043 m diameter. In order to<br />
define a mesh for the pipe ANSYS<br />
ICEM software is used. Due to the<br />
higher importance of the wall region<br />
compare to the middle of the pipe,<br />
the mesh near the wall needs to be<br />
finer than the mesh in the pipe<br />
center. The number of nodes is<br />
1,250,000.<br />
Mass flow rate<br />
[Kg/s]<br />
Temperature<br />
(k)<br />
Inlet 0.5 537.1<br />
Wall - 312.18<br />
outlet outflow -<br />
ENVIRONMENT AND SAFETY 239<br />
CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems<br />
Environment and Safety<br />
ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
ENVIRONMENT AND SAFETY 240<br />
| | Fig. 3.<br />
(a) Area averaged liquid volume fraction in different cross sections over the pipe length,<br />
(b) Temperature distribution in the outlet of the pipe for 5 different radial lines.<br />
4 Results<br />
The results are obtained with the AIAD<br />
approach for modeling the free surface<br />
and morphologies. Moreover, the<br />
Hughes correlation is used for the heat<br />
transfer coefficient. Figure 3 represents<br />
the qualitative profiles of liquid<br />
volume fraction and tem perature. In<br />
Figure 3 (a) the volume fraction profile<br />
in the vertical cross section in the<br />
middle of the pipe and in the streamwise<br />
direction is represented. As it can<br />
be seen at the inlet the pure steam<br />
exists and by going further in the pipe,<br />
due to the film condensation a liquid<br />
film starts to generate near the wall.<br />
The liquid film is growing and leads to<br />
the thicker film. In a cross section<br />
500 mm far from the inlet the liquid<br />
film is falling down gradually and<br />
gathering at lower part of the pipe.<br />
The liquid film always exists near the<br />
wall because as soon as the liquid is<br />
falling down the steam becomes in the<br />
direct contact with the wall and condenses<br />
and again new film generates.<br />
Moreover, in Figure 3 (d) the temperature<br />
profile is shown for different<br />
cross sections along the pipe. As<br />
mentioned before, the steam is fixed<br />
at the satu ration temperature, but<br />
further along the pipe by generating<br />
the liquid the temperature of the liquid<br />
is decreasing because of the heat<br />
flux to the wall. Moreover, the wall<br />
heat flux is cooling the liquid which<br />
causes the direct contact condensation<br />
between liquid and steam interface. As<br />
the steam is on the saturation temperature<br />
there is no heat flux between<br />
the interface which is also on the saturation<br />
temperature and the steam.<br />
Therefore, just the phase is changing<br />
and the steam turns into the liquid.<br />
| | Fig. 4.<br />
(a) Liquid Volume fraction distribution on a cross section along the pipe, (b) temperature distribution on a cross section along the pipe,<br />
(c) Volume fraction distribution on different cross sections, (d) temperature distribution on different cross sections.<br />
Figure 4 (a) shows the change of<br />
cross section averaged liquid volume<br />
fraction along the pipe. According to<br />
the figure the average liquid volume<br />
fraction at the inlet is 0.0 and due to<br />
the mass transfer it’s increasing along<br />
the pipe and it reaches to around 0.1<br />
at end of the pipe. Therefore, in a<br />
horizontal pipe with one meter length<br />
the total condensation rate is around<br />
10 percent. In Figure 4 (b) the temperature<br />
distribution for the five<br />
radial lines on the outlet of the pipe<br />
is presented. This plot shows the<br />
temperature difference from the<br />
center of the pipe towards the wall.<br />
As the plot shows, in the center the<br />
temperature is equal to the saturation<br />
temperature. As far as getting closer to<br />
the wall which is in subcooled<br />
tem perature, the temperature gradient<br />
is increasing. In other words, in<br />
the region near the wall the temperature<br />
difference from the saturation<br />
temperature is higher. Moreover,<br />
slope of the plot for L5 is higher than<br />
L1. The reason is in lower part of the<br />
pipe (which is showed by L5) the<br />
amount of cooled liquid is higher<br />
which causes higher temperature<br />
gradient in the lower parts of the pipe.<br />
As the pipe is symmetric and the<br />
boundary conditions for both sides of<br />
the pipe are same, there is no need to<br />
plot the temperature distribution in<br />
another half of the cross section.<br />
5 Conclusion<br />
The ANSYS CFX 17.2 has been used in<br />
order to simulate the condensation<br />
inside horizontal tubes. In order to<br />
model the two phase flow, heat<br />
transfer and phase change are included<br />
in the available AIAD concept<br />
which was developed for adiabatic<br />
cases. Moreover, the Hughes heat<br />
transfer coefficient correlation is<br />
implemented for the modeling of the<br />
direct contact condensation in the<br />
interface. The changes of the flow<br />
structure inside the pipe and the<br />
volume fraction and the temperature<br />
profiles have been studied in detail.<br />
The liquid film which is generated<br />
near the wall due to the wall condensation<br />
is modeled and it can be seen in<br />
the volume fraction profiles. By generating<br />
the liquid film near the wall both<br />
wall condensation and direct contact<br />
condensation are occurring inside<br />
the pipe at the same time. Whereas in<br />
the actual paper only the test for<br />
plausibility of the AIAD model was<br />
done, in the near future the comparison<br />
to the experiment is planned.<br />
The next step which is an ongoing<br />
part of the project is simulation of the<br />
Environment and Safety<br />
CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems<br />
ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
whole condensation phenomena and<br />
flow morphologies by using GENTOP<br />
concept. Further to the AIAD concept<br />
which considers two continuous<br />
fluids, the GENTOP approach is a<br />
three field two fluid model and considers<br />
also a poly dispersed phase.<br />
Acknowledgments<br />
This project is an ongoing project in<br />
Helmholtz-Zentrum Dresden Rossendorf<br />
(HZDR), which is funded by Bundesministerium<br />
für Bildung und Forschung<br />
(BMBF) under grant number<br />
02NUK041B in Germany.<br />
References<br />
[1] Hajal, J.El.; Thome, J.; Cavallini, A.<br />
Condensation inside horizontal tubes,<br />
part 1: two phase flow pattern map.<br />
International Journal of Heat and Mass<br />
Transfer 46: 3349-3363(2003).<br />
[2] Thome, J.; Hajal, J.El; Cavallini, A.<br />
Condensation inside horizontal tubes,<br />
part 2: New heat transfer model based<br />
on flow regimes. International Journal<br />
of Heat and Mass Transfer 46: 3365-<br />
3387(2003).<br />
[3] Kattan, N.; Thome, J.R.; Favrat, D. Flow<br />
boiling in horizontal tubes:part2-New<br />
heat transfer data for five refrigerants.<br />
J. Heat Transfer 120: 148-155 (1998).<br />
[4] Lee, W. H. A Pressure Iteration Scheme<br />
for Two-Phase Flow Modeling. Multiphase<br />
Transport Fundamentals,<br />
Reactor Safety, Applications: 407–432,<br />
(1980).<br />
[5] Štrubelj, L.; Ézsöl, Gy. ; Tiselj, I. Direct<br />
Contact Condensation Induced<br />
Transition from Stratified to Slug Flow.<br />
Nuclear Engineering and Design 240:<br />
266–274 (2010).<br />
[6] Lavieville, J.; Quemerais, E.; Boucker, M.;<br />
Maas, L., NEPTUNE CFD V1.0 User Guide<br />
(2005).<br />
[7] Coste, P. ; Pouvreau, J. ; Lavieville, J.;<br />
Boucker, M. A Two-phase CFD approach<br />
to the PTS problem evaluated on COSI<br />
experiment. Proceedings of the 16 th<br />
International Conference on Nuclear<br />
Engineering ICONE16, USA, (2008).<br />
[8] Banerjee, S.; A surface renewal model<br />
for interfacial heat and mass transfer in<br />
transient two-phase flow. International<br />
Journal of Multiphase Flow, Vol.4:<br />
571-573 (1978).<br />
[9] Hughes, E. D.; Duffey, R. B. Direct<br />
Contact Condensation and Momentum<br />
Transfer in Turbulent Separated Flows.<br />
Internal Journal of Multiphase Flow 17:<br />
599–619 (1991).<br />
[10] Egorov, Y. Validation of CFD codes with<br />
PTS relevant test cases. Technical Report<br />
EVOL-ECORA-D07, ANSYS, Germany<br />
(2004).<br />
[11] Apanasevich, P. ; Lucas, D.; Beyer, M.;<br />
Szalinski, L. CFD based approach for<br />
modeling direct contact condensation<br />
heat transfer in two-phase turbulent<br />
stratified flows. International Journal of<br />
Thermal Sciences 95: 123-135(2015).<br />
[12] Shen, L.; Triantafyllou, G.S.; Yue. D.K.P.<br />
Turbulent diffusion near a free surface<br />
Journal of Fluid Mechanics 407:<br />
145–166 (2000).<br />
[13] Ceuca, S. C. ; Macián-Juan R. CFD<br />
Simulation of Direct contact Condensation<br />
with ANSYS CFX using Locally<br />
defined Heat Transfer Coefficient.<br />
In ICONE-20, Anaheim, California, USA,<br />
No. 54347 (2012).<br />
[14] Goldbrunner, M.; Karl, J. ; Hein, D.<br />
Experimental Investigation of Heat<br />
Transfer Phenomena During Direct<br />
Contact Condensation in the Presence<br />
of Noncondensable gas by means of<br />
Linear Raman Spectroscopy. In 10 th Int.<br />
Symp. on Laser Techniques Applied to<br />
Fluid Mechanics, Lisbon (2000).<br />
[15] Krepper, E.; Frank, Th.; Lucas, D.; Prasser,<br />
H.-M.; Zwart, P.J. The Inhomogeneous<br />
MUSIG model for the simulation of<br />
poly-dispersed flow. Nuclear Engineering<br />
Design 238: 1690-1702 (2008).<br />
[16] Höhne, T.; Deendarlianto; Lucas, D.<br />
Numerical simulations of countercurrent<br />
two-phase flow experiments in<br />
a PWR hot leg model using an area<br />
density model. International Journal<br />
of Heat and Fluid Flow 31 (5):<br />
1047-1056 (2011).<br />
[17] Hänsch, S.; Lucas, D.; Krepper, E.;<br />
Höhne, T. A multi-field two-fluid<br />
concept for transitions between<br />
different scales of interfacial structures.<br />
International Journal of Multiphase<br />
Flow 47:171-182(2012).<br />
Authors<br />
Amirhosein Moonesi Shabestary,<br />
Eckhard Krepper,<br />
Dirk Lucas<br />
Helmholtz-Zentrum<br />
Dresden-Rossendorf<br />
P.O.Box 510119<br />
01314 Dresden, Germany<br />
241<br />
DECOMMISSIONING AND WASTE MANAGEMENT<br />
The Decommissioning of the ENEA RB3<br />
Research Reactor in Montecuccolino<br />
F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo<br />
The ENEA RB3 reactor was a 100 Wth research installation owned and operated by ENEA, in its center of Montecuccolino<br />
near Bologna, from 1971 to 1989. It consisted of a cylindrical aluminium vessel, about 4.3 m high and 2.9 m in diameter,<br />
which could host various types of fuel elements suspended from the top of a special adjustable rack and submerged into<br />
moderating and cooling heavy water. Principal aim of the reactor was to provide neutronics data for the CIRENE NPP, a<br />
SGHWR that was being designed and then partially built in Latina starting from 1979. The specific RB3 core, surrounded<br />
by a graphite reflector and housed inside a concrete biological shielding, allowed to test easily very different fuel<br />
element configurations by changing their pitches and by regulating the heavy water level inside the vessel. The reactor<br />
design, similar to that of the ZED-II Canadian research facility, was originally developed by CEA for its Aquilon facility<br />
in Saclay in 1956; in fact, through a special arrangement between ENEA and CEA, parts of the Aquilon facility were<br />
ultimately donated to ENEA at the end of the 60s for the construction of RB3. In 1989, the RB3 reactor was shut down,<br />
and in the late 2010 ENEA received by ministerial decree the authorization to its dismantling, with the aim of reaching<br />
the “green field” status and with the unconditional release of its building, which is actually owned by the University of<br />
Bologna. The dismantling activities started in May 2013 and were concluded at the end of 2014; after that, a campaign<br />
for the radiological characterization of the building was initiated and concluded in June 2015. Now, all the necessary<br />
site characterization activities are being conducted with the aim to present the results declaring the “green field” status<br />
before the end of 2017. This paper will present the three main pillars of the decommissioning of RB3, namely the<br />
strategy and methods for the dismantling, the strategy and methods for the radiological characterization of the building,<br />
and finally the strategy and methods for the radiological characterization of the site. The radionuclide limits imposed<br />
by the Italian Regulatory Body, together with the challenges encountered so far will be likewise shown and described.<br />
Revised version of<br />
a paper presented<br />
at the Eurosafe,<br />
Paris, France, 6 and 7<br />
November 2017.<br />
Decommissioning and Waste Management<br />
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
DECOMMISSIONING AND WASTE MANAGEMENT 242<br />
1 Introduction<br />
The ENEA RB3 (Reattore Bologna 3)<br />
reactor was a 100 Wth research installation<br />
owned and operated by ENEA<br />
in its center of Montecuccolino, near<br />
Bologna, from 1971 to 1989. It consisted<br />
of a cylindrical aluminium vessel,<br />
about 4.3 m high and 2.9 m in diameter,<br />
which could host various types of<br />
fuel elements suspended from the top<br />
of a special adjustable rack, and submerged<br />
into heavy water serving both<br />
as moderator and coolant. Principal<br />
aim of the reactor was to provide<br />
neutronics data for the CIRENE NPP, a<br />
SGHWR that was being designed, and<br />
then partially built in Latina, starting<br />
from 1979. The specific RB3 core, surrounded<br />
by a graphite reflector and<br />
housed inside a concrete biological<br />
shielding, allowed to test easily very<br />
different fuel element configurations<br />
by changing their pitches and by<br />
regulating the heavy water level inside<br />
the vessel. The reactor design, similar<br />
to that of the ZED-II Canadian<br />
research facility, was originally developed<br />
by CEA for its Aquilon facility in<br />
Saclay in 1956; in fact, through a<br />
special arrangement between ENEA<br />
and CEA, parts of the Aquilon facility<br />
were ultimately donated to ENEA at<br />
the end of the 60s for the construction<br />
of RB3. In 1989, after more than 18<br />
years of operation, the RB3 reactor<br />
was shut down, and in the late 2010,<br />
after waiting for the entry into force of<br />
Legislative Decree (L.D.) 230/1995<br />
[1], which introduced new laws for<br />
the decommissioning of NPPs, ENEA<br />
received by ministerial decree the<br />
authorization to its dismantling, with<br />
the aim of reaching the “green field”<br />
status and with the unconditional<br />
release of its building, including the<br />
reactor concrete biological shielding,<br />
which is actually owned by the<br />
University of Bologna. In fact the site<br />
of Montecuccolino, some 3.5 km to<br />
the South of downtown Bologna,<br />
hosted three research reactors: RB1,<br />
owned and operated by the University<br />
of Bologna, RB2, owned and operated<br />
by AGIP Nucleare, and RB3, owned<br />
and operated by ENEA. RB1 and RB2<br />
were decommissioned up to the green<br />
field status well before the entry into<br />
force of L.D. 230/1995.<br />
Figure 1 shows an aerial view of<br />
the Montecuccolino research center,<br />
with the area hosting RB3 contoured<br />
in red. Figure 2 shows a plan of the<br />
main reactor hall, with in red the<br />
area once occupied by the reactor<br />
vessel, surrounded by the hectagonal<br />
graphite reflector and encased within<br />
a thick concrete biological shielding.<br />
Figure 3 shows a vertical section of<br />
the RB3 building; the lowermost floor<br />
hosted 4 large tanks for a total of<br />
20,000 L (in red) for the storage of the<br />
heavy water which was daily pumped<br />
up into the vessel to reach criticality<br />
and then drained after the conclusion<br />
of the experiments. Three floors are<br />
present in the building: floor +6.0 m<br />
corresponding to the ground level,<br />
floor +0.0 m, corresponding to the<br />
level of the reactor vessel, and floor<br />
-3.0 m, with the heavy water storage<br />
tanks, heating and cooling systems,<br />
and other auxiliaries. The control<br />
| | Fig. 2.<br />
Plan of main hall of RB3.<br />
room was located at floor +0.0 m.<br />
While allowed to operate up to 100<br />
Wth, operations at RB3 were always<br />
conducted at 50 Wth.<br />
Between 1991 and 1992, all the<br />
fuel elements used at RB3 were either<br />
restituted at their owner (JRC Euratom<br />
Ispra) or sent to the ENEA Research<br />
Center of Saluggia or to the fuel fabrication<br />
plant of Fabbricazioni Nucleari<br />
at Bosco Marengo. Between 1992 and<br />
1993 all the heavy water was transferred<br />
to the ENEA Research Center of<br />
Borgo Sabotino, and before the end of<br />
1997 all the sealed radioactive sources<br />
used at the plant were disposed of.<br />
2 Regulatory Requirements<br />
and Classification<br />
of Components and<br />
Materials<br />
In the late 2010, ENEA received, by<br />
decree of the Italian Ministry of<br />
Economic Development, the authorization<br />
[2] to proceed with the dismantling<br />
of RB3; included in the<br />
| | Fig. 1.<br />
Aerial view of the Montecuccolino site; the RB3 building is inside the red square.<br />
| | Fig. 3.<br />
Section of the RB3 building.<br />
Decommissioning and Waste Management<br />
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Radionuclide Metals Concrete Other<br />
materials<br />
Reused Recycle Both reuse and recycle Demolition<br />
Surface<br />
(Bq/cm 2 )<br />
decree were also the requirements<br />
imposed by the Italian Nuclear Regulatory<br />
Body ISPRA. According to these<br />
requirements, the components and<br />
materials of RB3 were classified by<br />
ENEA, taking into account the various<br />
areas of the plant and the history of<br />
its functioning, into four main<br />
categories:<br />
A) materials and components which<br />
were both in contact with possibly<br />
contaminated or activated process<br />
fluids and subject to neutron flux;<br />
B) materials and components which<br />
were in contact with possibly<br />
contaminated or activated process<br />
fluids but not directly irradiated by<br />
neutrons;<br />
C) materials and components which<br />
were irradiated by the neutron flux<br />
but which never went into contact<br />
with possibly contaminated or<br />
activated process fluids;<br />
D) s.c. “exempt” materials, which<br />
were never irradiated and never<br />
went into contact with possibly<br />
contaminated or activated process<br />
fluids.<br />
The only component classified in the<br />
A category was the aluminium vessel;<br />
the only components in the B category<br />
were the heavy water distribution<br />
pipings. Exempt materials, given their<br />
unirradiated and uncontaminated<br />
status, were subject only to a general<br />
screening through CANBERRA In Situ<br />
Object Counting Systems (ISOCS) to<br />
estimate any possible level of presence<br />
of 60Co and 137Cs; if the measured<br />
levels were below the decision threshold<br />
of the measuring system in terms<br />
of mass concentration levels, then<br />
Surface<br />
(Bq/cm 2 )<br />
| | Tab. 1.<br />
Surface or mass activity concentration levels for clearance.<br />
Mass<br />
(Bq/g)<br />
these materials were automatically<br />
discarded from the plant without any<br />
further radiological analysis. This<br />
demonstrates the “instrumental” zero<br />
of this category of materials hence the<br />
“exempt” classification. All materials<br />
which had been classified as “exempt”<br />
were released unconditionately, for a<br />
total mass of about 30 tons, between<br />
March 2013 and May 2015. For all the<br />
other three categories, the clearance<br />
levels imposed by the Regulatory<br />
Authority are summarized in Table 1.<br />
These were derived either from the<br />
Italian L.D. n. 230/95 or from RP 89<br />
[3] and RP 113 [4] publications. In<br />
presence of more than one radionuclide,<br />
the sum of the ratios of<br />
the measured concentrations to the<br />
respective levels must be lower than 1.<br />
The components and materials<br />
were further grouped by ENEA into 12<br />
s.c. “homogeneous groups” using<br />
material and historic criteria; homogeneous<br />
groups are therefore constituted<br />
by components (or parts of<br />
them) made by the same material and<br />
possibly with a homogeneous and<br />
uniform activity content.<br />
3 Radiological Characterization<br />
of Homogeneous<br />
Groups<br />
Before the radiological characterization<br />
of the batches of materials from<br />
the various homogeneous groups<br />
started, a preliminary, special campaign<br />
was conducted to exclude the<br />
presence of various isotopes among<br />
those given in Table 1, expecially in<br />
the most potentially activated or<br />
contaminated materials (category A).<br />
Surface<br />
(Bq/cm 2 )<br />
Mass<br />
(Bq/g)<br />
3 H 10,000 100,000 1 10,000 1 1<br />
14 C 1,000 1,000 1 10,000 1 1<br />
Mass<br />
(Bq/g)<br />
54 Mn 10 10 1 10 0.1 0.1<br />
55 Fe 1,000 10,000 1 10,000 1 1<br />
59 Ni 10,000 10,000 1 100,000 1 1<br />
60 Co 1 10 1 1 0.1 0.1<br />
63 Ni 1,000 10,000 1 100,000 1 1<br />
90 Sr 10 10 1 100 1 1<br />
125 Sb 10 100 1 10 1 1<br />
134 Cs 1 10 0.1 10 0.1 0.1<br />
137 Cs 10 100 1 10 1 1<br />
152 Eu 1 10 1 10 0.1 0.1<br />
154 Eu 1 10 1 10 0.1 0.1<br />
Generic Alfa 0.1 0.1 0.1 0.1 0.1 0.01<br />
241 Pu 10 10 1 100 1 1<br />
In particular 54Mn, 59Ni, 90Sr,<br />
125Sb, 134Cs, 137Cs, 239Pu, 240Pu<br />
and 241Pu were excluded from<br />
further analyses finalized to the unconditional<br />
release of materials. Then,<br />
for each homogeneous group, a precharacterization<br />
measurement campaign<br />
was con ducted with a three-fold<br />
aim: 1) to verify if the hypothesis on<br />
the homogeneity of activity for that<br />
given group held; 2) to evaluate the<br />
minimum number of samples to be<br />
analized<br />
for the subsequent characterization<br />
phase; 3) to evaluate the value of<br />
isotopic ratios of 55Fe to 60Co and<br />
of 63Ni to 60Co, so to limit the next<br />
analyses only to the research of 60Co<br />
contents. After that, and using typically<br />
13 multiple measurements for<br />
each batch of each homogeneous<br />
group, summations of the ratios<br />
between measured activity concentrations<br />
and limits (Table 1) over all<br />
the relevant isotopes were carried out.<br />
If these summations resulted
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
DECOMMISSIONING AND WASTE MANAGEMENT 244<br />
| | Fig. 4.<br />
Dismantling of the lower layers of the graphite reflector.<br />
in a 1:10 ratio with other similar<br />
metals of warranted non-nuclear<br />
provenance in order to be used again<br />
for various purposes. All the homogeneous<br />
groups were pre-characterized,<br />
characterized and released before the<br />
end of 2014. All the measurements<br />
were performed by trained ENEA staff<br />
and within qualified ENEA laboratories,<br />
with the exception of some 14C<br />
measurements of a small lot of rubbers<br />
which were performed, under special<br />
contract, by the LASE Laboratory<br />
of CEA in Saclay. Workmen for heavy<br />
or peculiar technological operations<br />
were hired from the Modena Fallimenti<br />
SaS, a private Italian company specialized<br />
in the dismantling of special<br />
plants. Further details about the plan<br />
for the characterization of materials<br />
and components can be found in<br />
[5,6].<br />
4 Radiological Characterization<br />
of the Building<br />
After the completion of all the dismantling<br />
activities, and after the release<br />
of all the batches of materials, a<br />
radiological characterization of the<br />
building of RB3 has been made. This<br />
consisted of two main steps. The first<br />
was the characterization of the activation<br />
status of the baritic concrete<br />
biological shielding of the core. This<br />
consisted in seven core drillings, (see<br />
Figure 5) each 16 cm long, so distributed:<br />
1 on the floor, 1 on the<br />
northern wall, 1 on the western wall,<br />
1 on the eastern wall, and 3 (at<br />
different heights) on the souther wall.<br />
All the drilling points were at positions<br />
where the neutron flux during<br />
operation was maximum. From each<br />
drilling, four aliquots, 4 cm long,<br />
where taken, so to cover the depth<br />
profile of any activation distribution<br />
inside the biological shielding. Each<br />
aliquot was subject to gamma spectrometry<br />
to search for the presence of<br />
60Co, 134Cs, 152Eu and 154Eu. All<br />
the 28 samples yielded results for all<br />
the four isotopes lower than a few<br />
mBq/g. Then, all the samples were<br />
subject to thorough statistical analysis,<br />
based on several Bartlett tests, to<br />
verify if they were all and altogether<br />
representative of the same statistical<br />
distribution of activity and therefore<br />
representative of a same “homogeneous<br />
group” constituted of the whole<br />
biological shielding. Once this condition<br />
has been verified, a Noether test,<br />
using 10 randomly chosen measurements,<br />
was put in place to verify the<br />
minimum number of samples to be<br />
used for the final characterization of<br />
the biological shielding. This resulted<br />
in 13 samples, randomly extracted<br />
from the complete set of all the 28<br />
available samples. However, ENEA decided<br />
to use all the 28 samples to verify<br />
the free release condition for the<br />
shielding, and for all the 28 samples<br />
the condition resulted verified, meaning<br />
that no significative activation of<br />
the shielding had been realized. As a<br />
further consequence, it could be<br />
proven that no activation of walls<br />
outside the biological shielding was<br />
in place, just because, due to its<br />
screening effect, the neutron flux<br />
outside the shielding itself was 6 to 7<br />
orders of magnitude lower.<br />
The second step of the characterization<br />
consisted in the assessment of<br />
the surface contamination of the various<br />
areas of the building. These were<br />
separated into three main surfaces:<br />
1) ceiling; 2) surfaces over +6.0 m<br />
level; 3) surfaces below +6.0 m level.<br />
The ceiling was indeed a false ceiling<br />
made of thin aluminium plates; these<br />
could have been contaminated by<br />
tritiated water vapours emerging<br />
from the core once open for refueling<br />
or fuel reshuffling. To investigate this,<br />
the aluminium plates were dismantled,<br />
taken to ground, and analyzed. It<br />
was assumed that, if no contamination<br />
was found, then also the real<br />
ceiling behind it was not contaminated.<br />
This proved indeed to be the<br />
case. Surfaces over +6.0 m were<br />
investigated randomly (Figure 6), by<br />
sampling a given number of points,<br />
quantified basing on statistical considerations.<br />
All surfaces below +6.0 m<br />
were completely measured, both walls<br />
and floors. The measurement technique<br />
consisted in using surface<br />
contamination meters (Berthold<br />
LB165 and LB124), properly cali brated<br />
with large area reference sources, to<br />
sum up count rates over 14C, 60Co,<br />
134Cs, 152Eu and 154Eu. A similar<br />
measurement methodology was successfully<br />
applied for the decommissioning<br />
of the ASTRA research reactor<br />
in Vienna [7]. Background contributions<br />
due to natural radionuclides in<br />
the different materials were subtracted<br />
after having made suitable averages<br />
from surely clean, similar materials<br />
to those which were to be measured<br />
inside the building. As a further, conservative<br />
penalization, it was decided<br />
to attribute to each of the 5 abovementioned<br />
nuclides the whole net<br />
counting over each surface portion<br />
being measured, counting time per<br />
surface element being about 30 seconds<br />
to reach a desired minimum<br />
detectable activity. LB124 hand held<br />
monitor was used for surfaces over<br />
+6.0 m, while LB165 (wheeled monitor<br />
as in Figure 7) was used over all<br />
other surface portions. A special automated<br />
vertical translational sledge<br />
(Figure 8) was used to carry LB165<br />
over the portions of the walls. In case<br />
a given measurement yielded values<br />
above the clearance limits, special<br />
cleaning procedures were to be<br />
adopted until subsequent measurements<br />
proved to be below the limits<br />
| | Fig. 5.<br />
Core drilling of the biological shielding.<br />
| | Fig. 6.<br />
LB124 measurements of selected portions of walls above +6.0 m level.<br />
Decommissioning and Waste Management<br />
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
| | Fig. 7.<br />
LB165 measurement of floor portions.<br />
themselves. No cleaning procedures<br />
were ever necessary and all the surface<br />
portions could be considered not<br />
contaminated and so able to be freely<br />
released.<br />
5 Radiological Characterization<br />
of the Site<br />
The third and last pillar of the decommissioning<br />
of RB3 is the radiological<br />
characterization of the areas of the<br />
site surrounding the building. In this<br />
regard, it is important to mention that<br />
during the operational life of the<br />
plant, given its low power and its<br />
construction features, no radiocativity<br />
discharges were in place and therefore<br />
no environmental analyses were prescribed<br />
by the Regulatory Authority.<br />
Another point worth of mentioning is<br />
that no radiological status of the site<br />
prior to the construction and exercise<br />
of RB3 is known. However, in light of<br />
the graded approach which is going to<br />
be taken for this third pillar by the<br />
Regulatory Authority, given the fact<br />
that no activated materials have been<br />
found and that no activation or contamination<br />
of the building has been<br />
detected, it is decided to base this<br />
characterization upon the measurement<br />
of some selected nuclides in<br />
certain terrain samples (soil) taken<br />
around the area of the RB3 site.<br />
In particular, 12 measurements of<br />
239+240Pu through alpha spectrometry<br />
will be done, together with<br />
25 gamma spectroscopy assessments<br />
for 54Mn, 60Co, 125Sb, 134Cs, 137Cs,<br />
| | Fig. 9.<br />
Collecting soil samples from the RB3 site.<br />
Radionuclide<br />
152Eu and 154Eu. Each terrain<br />
sample will be a parallelepiped of<br />
25x20x10 cm 3 corresponding roughly<br />
to 5 liters of humid soil (Figure 9).<br />
The site will be sampled considering<br />
both near-range and far-range positions<br />
in order to find patterns of radioactivity<br />
correlated with the distance<br />
from the RB3 building, if any at all.<br />
The obtained values will be confronted,<br />
through proper summations,<br />
with the limits for the free release<br />
of nuclear sites prescribed by the<br />
German national law, which correspond<br />
to the radiological nonrelevance<br />
value of 10 microSv/year<br />
to the public [8,9]. The limits for<br />
the above-mentioned isotopes are<br />
reported in Table 2.<br />
References<br />
[1] D.Lgs. 17 marzo 1995, n. 230,<br />
Attuazione delle direttive Euratom<br />
80/836, 84/467, 84/466, 89/618,<br />
90/64, 92/3, 96/29.<br />
[2] D. M. 29 Novembre 2010 Ministero<br />
dello Sviluppo Economico di<br />
Autorizzazione alla Disattivazione<br />
Impianto Nucleare di Ricerca Reattore<br />
RB-3 di Montecuccolino (BO) dell’ENEA.<br />
[3] Radiation Protection 89, Recommended<br />
radiological protection criteria for the<br />
recycling of metals from the<br />
dismantling of nuclear installations,<br />
European Commission, 1998.<br />
| | Fig. 8.<br />
LB165 and its translational sledge to measure wall portions.<br />
Concentration Limit<br />
(Bq/g)<br />
54 Mn 0.09<br />
60 Co 0.03<br />
125 Sb 0.08<br />
134 Cs 0.05<br />
137 Cs 0.06<br />
152 Eu 0.07<br />
154 Eu 0.06<br />
239 Pu 0.04<br />
240 Pu 0.04<br />
| | Tab. 2.<br />
Proposed clearance limits for the free release<br />
of the RB3 site.<br />
[4] Radiation Protection 113,<br />
Recommended radiological protection<br />
criteria for the clearance of buildings<br />
and building rubble from the<br />
dismantling of nuclear installations,<br />
European Commission, 2000.<br />
[5] I. Vilardi, C. M. Castellani, D. M.<br />
Castelluccio, F. Rocchi, Piano di<br />
Caratterizzazione Radiologica di Materiali<br />
provenienti dalla Disattivazione<br />
dell’impianto Nucleare di Ricerca Rb-3<br />
dell’enea sito in Bologna – Montecuccolino<br />
ai Fini del loro Allontanamento,<br />
Convegno Nazionale AIRP 2014, Aosta.<br />
[6] M. Capone, N. Cherubini, A. Compagno,<br />
A. Dodaro, F. Rocchi, The Dismantling of<br />
the Montecuccolino RB3 Research<br />
Reactor: Radiological Characterisation of<br />
Materials for Free Release, Proceedings<br />
of the European Reaserch Reactor<br />
Conference RRFM 2015, Bucharest<br />
19-23 April 2015, 528-537.<br />
[7] F. Meyer, F. Steger, R. Steininger,<br />
Decommissioning of the Astra Research<br />
Reactor – Dismantling the auxiliary<br />
Systems and Clearance and Reuse of the<br />
Buildings, Nuclear Technology &<br />
Radiation Protection, 1/2008, 54-62.<br />
[8] OECD/NEA Status Report, Releasing<br />
the Sites of Nuclear Installations,<br />
NEA Report 6187, 2006.<br />
[9] Bundesgesetzblatt G 5702 Teil I, Bonn<br />
26 July 2001, Nr. 38, 2001.<br />
Authors<br />
F. Rocchi<br />
ENEA FSN/SICNUC/SIN<br />
C. M. Castellani<br />
ENEA IRP<br />
A. Rizzo<br />
ENEA FSN/SICNUC/TNM<br />
Via Martiri di Monte Sole 4<br />
Bologna (BO), Italy<br />
A. Compagno<br />
ENEA FSN/FISS/CRGR<br />
I. Vilardi<br />
ENEA IRP/SFA<br />
Via Anguillarese, 301<br />
00123 S.Maria di Galeria (RM), Italy<br />
R. Lorenzelli<br />
ENEA FSN/SICNUC/SIN<br />
Località Brasimone<br />
40032 Camugnano (BO), Italy<br />
DECOMMISSIONING AND WASTE MANAGEMENT 245<br />
Decommissioning and Waste Management<br />
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
246<br />
RESEARCH AND INNOVATION<br />
Revised version of a<br />
paper presented at<br />
the Annual Meeting<br />
of Nuclear Technology<br />
(AMNT 2017), Berlin.<br />
Irradiation Tests of a Flat Vanadium Self-<br />
Powered Detector with 14 MeV Neutrons<br />
Prasoon Raj and Axel Klix<br />
Self-powered detector (SPD) represents a class of neutron and gamma monitoring instruments used in the fission<br />
reactor cores worldwide. This detector has inherent advantages of functioning without a bias voltage, simple measurement<br />
scheme, compactness, ease of maintenance, and high reliability. We are studying SPD for application as flux<br />
monitors in the European test blanket modules (TBM) of ITER, fusion reactor under construction in southern France.<br />
This paper presents results of experimental tests performed with 14 MeV neutrons for a flat SPD with vanadium emitter.<br />
Vanadium responds by beta emission from products of reactions (main routes: 51 V (n, γ) 52 V and 51 V (n, p) 51 Ti) with<br />
thermal and fast neutrons. Secondary electrons due to gammas from these reactions and neutron irradiation of<br />
surrounding materials are also important contributors to the signal. Thin foils of emitter, insulator and collector<br />
materials are used to construct the test SPD. The detector is irradiated with short and long pulses of neutrons and is<br />
found to respond in proportion with the incident neutron flux. Further experiments with simplified and better optimized<br />
design of detector are underway for thorough study of the signal-creation mechanism.<br />
1 Introduction<br />
ITER [1] is an experimental fusion<br />
reactor based on tokamak concept,<br />
under construction at St. Paul lez<br />
Durance in southern France. It is an<br />
international project aimed at proving<br />
feasibility of fusion as a large-scale<br />
and carbon-free source of energy. One<br />
of the main scientific goals of this<br />
project will be to test and prove the<br />
concepts of tritium breeding blankets.<br />
Tritium is an important fuel component<br />
for devices based on D-T reaction,<br />
which is being considered as<br />
main reaction for fusion power plants.<br />
Because tritium is a rare element, it is<br />
required to breed it in the fuel cycle of<br />
the reactor. A blanket with lithium<br />
compounds will cover the inner wall<br />
of the plasma vessel. Fusion neutrons<br />
from the plasma will be absorbed by<br />
lithium nuclei, causing reactions to<br />
produce tritium.<br />
There are multiple breeding<br />
blanket designs proposed by scientists.<br />
To determine their efficiencies in<br />
a real fusion environment, test blanket<br />
modules (TBM) based on different<br />
concepts will be inserted into equatorial<br />
ports of ITER for experimental<br />
tests in different operational phases of<br />
ITER. The European Union is going to<br />
test two such concepts, namely the<br />
Helium-Cooled Lead-Lithium (HCLL)<br />
and Helium-Cooled Pebble Bed<br />
(HCPB) TBMs [2]. In the neutronics<br />
experiments, nuclear responses like<br />
tritium production rate, material activation,<br />
nuclear heating etc. are to be<br />
measured and compared with the<br />
calculations. This step will validate<br />
the advanced computational tools<br />
and nuclear data utilized for nuclear<br />
analyses for fusion devices. The neutron<br />
and gamma fluxes are important<br />
quantities to be measured for these<br />
experiments, for which detectors like<br />
neutron activation system, fission<br />
chambers and self-powered detectors<br />
(SPD) are under study.<br />
An SPD is a multi-layered electrical<br />
device, which produces direct current<br />
(DC) signal on irradiation with<br />
neutrons and/or gammas. It can be<br />
preferentially responsive to neutrons<br />
(self- powered neutron detector,<br />
SPND) or gammas (SPGD), or as<br />
it is in most of the cases, to both.<br />
Figure 1 shows a rough sketch of the<br />
cross- section of a traditional detector.<br />
Central material, called emitter produces<br />
fast electrons on irradiation.<br />
These fast electrons can be betas<br />
from the decay of neutron activation<br />
products, or secondary electrons due<br />
to interaction of gammas in the bulk<br />
of the material. They slow down in a<br />
layer of insulation and stop in the<br />
outer electrode called collector. This<br />
electron-movement creates a potential<br />
difference and thus, produces a<br />
current signal proportional to the<br />
incident particle flux. The current due<br />
to beta electrons is “delayed” because<br />
of the half-life of beta-emitters, e.g.<br />
SPND based on Rh, V or Ag emitters.<br />
Whereas that due to gamma-initiated<br />
photoelectric or Compton electrons<br />
is “prompt”, for example Co-based<br />
SPND [3].<br />
An SPD responds in a sophisticated<br />
manner, with multiple factors<br />
contributing to the small current<br />
signals often totaling between 10 -12<br />
and 10 -3 Ampere. Due to its inherent<br />
advantages of simplicity, compactness<br />
and high-reliability, they are highly<br />
desirable for flux monitoring in areas<br />
with restricted access like reactor<br />
cores. At KIT, we are studying SPDs<br />
for application in ITER TBM [4].<br />
Vanadium based flat SPD is being<br />
tested with 14 MeV neutrons, to<br />
understand its behavior towards fast<br />
neutrons expected in fusion environment<br />
and ascertain the feasibility of<br />
its application as flux monitor for<br />
European ITER TBMs.<br />
2 Experimental details<br />
Vanadium is a common emitter for<br />
fission reactor SPNDs. The response<br />
of the detector towards thermal neutrons<br />
is understood well. The material<br />
is relatively inexpensive and easier to<br />
handle. However, due to lower cross<br />
sections the sensitivity of vanadium-<br />
SPND towards fast neutrons reduces<br />
(Figure 2). Commercially available<br />
SPND cannot be directly used for<br />
measurement of fusion neutron<br />
fluxes, going up to approx. 14 MeV in<br />
energy.<br />
Characteristics of the two main<br />
beta- emitters from 51 V (99.75 %<br />
isotopic abundance) in case of fast<br />
neutron irradiation, are reported in<br />
Table 1. Cross-sections of the fast<br />
neutron reactions in 51 V for a pure<br />
| | Fig. 1.<br />
Cross-sectional sketch of a cylindrical SPD showing emitter (green), insulator (dotted white) and<br />
collector (black) layers, with connection to the lead cable and current measurement device.<br />
Research and Innovation<br />
Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
| | Fig. 2.<br />
Cross sections of vanadium reactions and photon production under neutron irradiation.<br />
Reaction 51 V (n, p) 51 Ti 51 V (n, γ) 52 V<br />
Threshold Neutron Energy 1.72 MeV 0 MeV<br />
14 MeV Cross-section 30 mb (approx.) 0.6 mb (approx.)<br />
Beta Emitter, Half-life 51 Ti- 5.76 m 52 V- 3.74 m<br />
Average Beta Energy 51 Ti- 0.87 MeV 52 V- 1.07 MeV<br />
SPND Current (14 MeV) 3.46 × 10 -12 A 6.92 × 10 -14 A<br />
SPND Current (TBM) 7.97 × 10 -9 A 3.44 × 10 -8 A<br />
| | Tab. 1.<br />
Beta-emitters and corresponding currents from fast neutron reactions in vanadium based SPND.<br />
14 MeV source are shown. Neglecting<br />
the self-shielding of electrons in emitter<br />
material, effect of other materials<br />
and taking a saturation condition<br />
(considering the short half-lives of<br />
daughter nuclides), one can ascertain<br />
the orders of magnitude of currents<br />
possible with V-SPND, as reported.<br />
For this estimation, vanadium density<br />
of 6.1 g cm -3 , and a typical volume of<br />
1 cm 3 are assumed. For a 14 MeV<br />
neutron source, a flux intensity of<br />
1 × 10 10 cm -2 s -1 is considered, which<br />
is achievable with state of the art<br />
14 MeV neutron generators. For TBM,<br />
activation calculation was done [5]<br />
with the HCLL neutron spectrum<br />
and typical flux intensity (up to 1 ×<br />
10 14 cm -2 s -1 ) using EASY-2007 [6].<br />
With high-sensitivity ammeters,<br />
currents down to the order of 1 ×<br />
10 -14 A can be reliably measured [7].<br />
Values in Table 1 show that a vanadium<br />
emitter based SPND will produce<br />
measurable signals in TBM. Due<br />
to its high neutron threshold energy,<br />
the (n, p) reaction can be utilized to<br />
measure fast neutron flux exclusively.<br />
Fast neutron reactions lead to<br />
high-energy gamma production. This<br />
phenomenon competes with the neutron<br />
absorption reactions (Figure 2).<br />
Photoelectric and Compton electron<br />
emission from emitter causes a prompt<br />
current which is expected to form the<br />
major component of the signal of<br />
V-SPND towards 14 MeV neutrons.<br />
Secondly, vanadium being a medium-<br />
Z nucleus can be a potential<br />
emitter for SPGD also. With optimized<br />
dimensions and choice of collector<br />
material, a vanadium SPD can be<br />
envisaged for monitoring of photon<br />
flux in TBM.<br />
Instead of the usual coaxial type<br />
cylindrical geometry, we designed<br />
our test SPD in sandwich-type flat<br />
geometry. This provides a relatively<br />
higher cross section area to the incident<br />
neutrons, and ease of access for<br />
testing various materials in the same<br />
device. Thin foils (0.5 to 2 mm) of<br />
emitter, insulator and collector are<br />
arranged to form an assembly in an<br />
aluminum case, which also serves as<br />
an electromagnetic shield. Central<br />
conductor of the signal cable is linked<br />
to the emitter plates of the detector.<br />
The collector plates, case and the<br />
cable sheath are shorted and securely<br />
connected to the ground. Schematic<br />
sketch and photograph of the test<br />
detector are shown in Figure 3 (left).<br />
With comparable cross sections of<br />
reactions in different materials, the<br />
insulator and collector materials also<br />
play an important role in SPD<br />
response. Behaviors of different<br />
material combinations are experimentally<br />
tested. Alumina (Al 2 O 3 ) or<br />
beryllia (BeO) is used as insulator and<br />
Inconel-600 or graphite is used as<br />
collector in our experiments. Effects<br />
of the change of geometry and dimensions<br />
are also studied. A Keithley 6485<br />
Picoammeter (sensitivity range -20 fA<br />
to 20 mA) is used as the measuring<br />
device. A low-noise triax cable (Belden<br />
9222) is used to reduce the interferences<br />
in low-current measurement.<br />
The tests are conducted at the<br />
14 MeV neutron generator of Technical<br />
University of Dresden (TUD-NG),<br />
shown in Figure 3 (right). Here,<br />
deuteron beams are impinged on a<br />
tritiated titanium target causing D-T<br />
reaction which leads to production of<br />
neutrons with peak energy of approx.<br />
14 MeV. TUD-NG offers neutron flux<br />
intensities up to 1 × 10 10 cm -2 s -1 . The<br />
detector is placed in front of the<br />
tritium- target of TUD-NG and tested<br />
under different conditions by varying<br />
flux levels and irradiation times.<br />
3 Results<br />
The irradiation tests of flat sandwichtype<br />
vanadium SPD were performed<br />
at TUD-NG, with neutron flux intensities<br />
around 1 × 10 9 cm -2 s -1 . DC<br />
signals in the range of 100 fA to 100 pA<br />
were measured. In Figure 4, a plot<br />
shows variation of SPD signal with<br />
change in neutron flux. The detector<br />
was composed of 1 mm thick layers of<br />
vanadium emitter and Inconel-600<br />
collector. The signal was found to be<br />
proportional to the incident flux, with<br />
approx. 90 pA at the highest flux level.<br />
At low fluxes and low currents,<br />
the measurements have high uncertainties.<br />
Interference from electromagnetic<br />
sources of stray currents,<br />
| | Fig. 3.<br />
(Left) internal design of the sandwich-type flat SPD: (top)- an engineering sketch of the geometry<br />
having sandwich of foils of emitter (green), insulator (grey) and collector (red), and (below) a photograph<br />
of the assembly with vanadium SPD.<br />
(Right) experimental setup showing TUD-NG beamline, tritium target, mounted SPD, and the lead cable.<br />
RESEARCH AND INNOVATION 247<br />
Research and Innovation<br />
Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
RESEARCH AND INNOVATION 248<br />
| | Fig. 4.<br />
Vanadium SPD signal (left Y-axis, red curve) variation with change in neutron flux (right Y-axis, blue<br />
curve) plotted with respect to irradiation time.<br />
currents generated in coaxial cables,<br />
electrostatic effects at the contacts<br />
and degradation of insulation layer<br />
due to radiation, lead to background<br />
currents in the orders of 100 fA. This<br />
makes the measurement of low-level<br />
currents a very challenging task.<br />
The SPD response is often reported<br />
in terms of sensitivity, which is SPD<br />
current per unit of neutron (or<br />
gamma) flux intensity, reported in<br />
units of A cm 2 s. For the vanadium<br />
SPND signal in Figure 4, the sensitivity<br />
lies between 4.48 × 10 -20 A cm 2 s<br />
± 13.4 % (at flux intensity 2.04 ×<br />
10 9 cm -2 s -1 ) and 8.80 × 10 -19 A cm 2 s<br />
± 51.1 % (at flux intensity 6.40 ×<br />
10 5 cm -2 s -1 ).<br />
In another test, a constant-flux<br />
irradiation of around 15 minutes was<br />
done and the TUD-NG was switched<br />
off. This signal is shown in Figure 5. It<br />
was found that the detector current is<br />
dominated by a prompt component<br />
which appeared and disappeared with<br />
neutron flux. The delayed signal is<br />
usually less than 10% of the total<br />
signal. A decay of delayed current was<br />
observed as expected.<br />
There are parasitic beta emission<br />
reactions in insulator, collector and<br />
cable’s central conductor, e.g. 27 Al (n,<br />
p) 27 Mg reaction (half-life~ 9.46 min)<br />
in alumina insulation. Electrons<br />
emitted due to these reactions reduce<br />
the total delayed current. Due to this,<br />
the analysis of decay curve becomes<br />
very complex. After data reduction,<br />
subtraction of background contributions,<br />
and further analysis the major<br />
contribution was found to be from 51 Ti<br />
due to 51 V (n, p) 51 Ti reaction. For<br />
reduction of aforementioned effects<br />
materials with lower total cross<br />
sections of beta emission reactions,<br />
like graphite and beryllia were used as<br />
collector and insulator, respectively.<br />
The change in the signal characteristics<br />
was minimal with these alterations,<br />
leading us to conclude that the<br />
signal was mainly due to reactions in<br />
the vanadium emitter. A prompt current,<br />
makes the detector suitable for<br />
pulsed devices like ITER. However, it<br />
is important to understand the signal<br />
creation mechanism for calibration<br />
and application of the SPD.<br />
The high prompt signal is attributed<br />
to three main reasons. First is<br />
the interaction of photons in the<br />
emitter volume, which release high<br />
energy electrons producing high<br />
positive current. Unlike thermal neutrons,<br />
fast neutrons lead to emission<br />
of higher-energy photons with higher<br />
probability of secondary effects.<br />
Moreover, the photon production<br />
cross section is usually an order or two<br />
higher than the fast neutron reaction<br />
cross sections in materials of detector<br />
and surroundings (Figure 2). Secondly,<br />
the production of charged particles<br />
like protons and alphas in collector<br />
and insulator material (cross sections<br />
of (n, xp) and (n, xα) reactions are<br />
high for 14 MeV neutrons) lead to<br />
further difference of charge between<br />
electrodes and a prompt positive<br />
contribution to the signal. Finally, the<br />
electrical and nuclear effects in<br />
connecting wires and cables make a<br />
small fraction of the positive current<br />
signal<br />
Some of the contributing factors<br />
will be explicitly studied in future<br />
tests. To de-couple the effects of other<br />
materials, a detector with simplified<br />
geometry is under design. An air-insulated<br />
detector with box of collector<br />
material is being constructed. The<br />
material thicknesses are reduced in<br />
order to decrease the gamma interactions.<br />
Improved ways of making<br />
electrical contacts between cable and<br />
emitter are studied. Other less betaactive<br />
materials like niobium are<br />
being considered for collector. Vanadium<br />
detector is also planned to be<br />
optimized for photon response. To this<br />
end, thicker emitters and collectors<br />
with low gamma-activity will be used<br />
to make a test-device which will be<br />
irradiated with high-energy bremsstrahlung<br />
photon source.<br />
4 Conclusions<br />
A flat sandwich-type vanadium SPD<br />
has been constructed, for testing the<br />
feasibility of application of SPDs in<br />
ITER TBMs. Irradiation tests with<br />
14 MeV neutrons at TUD-NG resulted<br />
in current signals in range of 100 fA to<br />
100 pA. The signals are proportional<br />
to the incident neutron flux. Considering<br />
the higher flux intensities up<br />
to 1 × 10 14 cm -2 s -1 and a wider energy<br />
spectrum of neutrons in TBM, studies<br />
show that vanadium SPND is expected<br />
to produce measurable signals in ITER<br />
| | Fig. 5.<br />
Vanadium-SPD signal in a long constant-flux irradiation at TUD-NG showing (prompt and delayed)<br />
currents before, during and after the irradiation.<br />
Research and Innovation<br />
Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
TBM conditions. The high prompt<br />
component of the SPD signal is<br />
attributed to the interaction of high<br />
energy photons which are produced<br />
in the detector and surrounding<br />
materials. Charged particles emitted<br />
in fast neutron reactions and contributions<br />
from wires and signal cable<br />
contribute to the high positive signal.<br />
Parasitic reactions in non-emitter<br />
materials also play an important role.<br />
These effects need to be studied<br />
explicitly and compared for understanding<br />
of the overall currentgeneration<br />
mechanism. Optimization<br />
of design, dimensions and material<br />
combinations is underway to realize<br />
SPD flux monitors for application in<br />
European ITER TBMs.<br />
Acknowledgement<br />
The work leading to this publication<br />
has been funded partially by Fusion<br />
for Energy under the Specific<br />
Grant Agreement F4E-FPA-395-1.<br />
This publication reflects the views<br />
only of the authors, and Fusion for<br />
Energy cannot be held responsible for<br />
any use which may be made of the<br />
infor mation contained therein.<br />
References<br />
[1] ITER Organization – Homepage. [Online].<br />
Available: https://www.iter.org/.<br />
[2] P. Calderoni, Status of the HCLL and<br />
HCPB Test Blanket System instrumentation<br />
development, 21 st Top. Meet.<br />
Technol. Fusion Energy (TOFE), 9-13<br />
Nov. 2014, Anaheim, CA, USA.<br />
[3] N. P. Goldstein and W. H. Todt, A Survey<br />
of Self-Powered Detector - Present and<br />
Future, IEEE Trans. Nucl. Sci., vol. 26,<br />
no. 1, pp. 916–923, 1979.<br />
[4] P. Raj, M. Angelone, U. Fischer, and<br />
A. Klix, Self-powered detectors for test<br />
blanket modules in ITER, in 2016 IEEE<br />
Nuclear Science Symposium, Medical<br />
Imaging Conference and Room- Tem perature<br />
Semiconductor Detector Workshop<br />
(NSS/MIC/RTSD), 2016, pp. 1–4.<br />
[5] M. Angelone, A. Klix, M. Pillon, P.<br />
Batistoni, U. Fischer, and A. Santagata,<br />
Development of self-powered neutron<br />
detectors for neutron flux monitoring in<br />
HCLL and HCPB ITER-TBM, Fusion Eng.<br />
Des., vol. 89, no. 9–10, pp. 2194–2198,<br />
2014.<br />
[6] R. A. Forrest, FISPACT-2007: User<br />
manual, EASY Doc. Ser. UKAEA<br />
FUS 534, 2007.<br />
[7] Low Level Measurements Handbook –<br />
7 th Edition: Precision DC Current,<br />
Voltage, and Resistance Measurements.<br />
Keithley- A Tektronix Company.<br />
Authors<br />
Prasoon Raj<br />
Axel Klix<br />
Institute for Neutron Physics and<br />
Reactor Technology (INR)<br />
Karlsruhe Institute of Technology<br />
(KIT)<br />
Hermann von Helmholtz Platz 1<br />
76344 Eggenstein-Leopoldshafen<br />
(Germany)<br />
RESEARCH AND INNOVATION 249<br />
Nanofluid Applied Thermo-hydrodynamic<br />
Performance Analysis of Square<br />
Array Subchannel Under PWR Condition<br />
Jubair Ahmed Shamim and Kune Yull Suh<br />
1 Introduction Efficient engineered design of heat transfer and fluid flow with enhanced heating or cooling<br />
requires two pivotal aspects that must be taken into consideration for extracting thermal energy from nuclear fission<br />
reactions in order to save energy, reduce process time, raise thermal rating and increase the operating life of a reactor<br />
pressure vessel. Hence, one of the major challenges in designing a new nuclear power plant is the quantification of the<br />
optimal flow of coolant and distribution of pressure drop across the reactor core. While higher coolant flow rates will<br />
lead to better heat transfer and higher Departure from Nucleate Boiling (DNB) limits, it will also result in higher pressure<br />
drop across the core, therefore additional demand of pumping powers as well as larger dynamic loads on the core<br />
components. Thus, thermal hydraulic core analysis seeks to find proper working conditions with enhanced heat transfer<br />
and reduced pressure drop that will assure both safe and economical operation of nuclear plants.<br />
Recently, nanofluid has gained much<br />
renewed attention as a promising<br />
coolant for pressurized water reactors<br />
(PWRs) due to its enhanced thermal<br />
capabilities with least penalty in pressure<br />
drop. The improved heat transfer<br />
of nanofluids results from the fact that<br />
the nanoparticles increase the surface<br />
area and heat capacity of the fluid,<br />
improve the thermal conductivity of<br />
the fluid, cause more collisions and<br />
interactions between the fluid, particles<br />
and surfaces of the flow passages,<br />
and enhance turbulence and mixing<br />
of the fluid.<br />
Pak & Cho [1] experimentally<br />
observed the turbulent friction and<br />
heat transfer of dispersed fluids in a<br />
circular pipe using two different<br />
metallic oxide particles, γ-alumina<br />
(Al 2 O 3 ) and titanium dioxide (TiO 2 )<br />
with mean diameters of 13 and 27 nm,<br />
respectively. The results revealed<br />
that the Nusselt number Nu for the<br />
dispersed fluids increased with<br />
increasing volume concentration as<br />
well as the Reynolds number Re. But<br />
at constant average velocity, the<br />
convective heat transfer coefficient for<br />
the dispersed fluid was 12% less than<br />
that for pure water. They proposed a<br />
new correlation for Nu under their<br />
experimental ranges of volume concentration<br />
(0-3%), Re (10 4 -10 5 ), and<br />
the Prandtl number Pr (6.54-12.33)<br />
for the dispersed fluids γ-alumina<br />
(Al 2 O 3 ) and titanium dioxide (TiO 2 )<br />
particles as<br />
(1)<br />
Xuan and Li [2] observed the flow<br />
and convective heat transfer of the<br />
Cu-water nanofluid flowing through<br />
a straight brass tube of the inner<br />
diameter of 10 mm and the length of<br />
800 mm. They noted that suspended<br />
nanoparticles can remarkably enhance<br />
heat transfer given the velocities.<br />
For instance, the heat transfer<br />
coefficient of nanofluids containing<br />
2.0 vol % Cu nanoparticles was increased<br />
by as much as 40 % compared<br />
to that of water. The conventional<br />
Research and Innovation<br />
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<br />
Dittus–Boelter correlation failed to<br />
predict this augmented heat transfer<br />
data for nanofluids. They presented a<br />
new correlation for turbulent flow of<br />
nanofluids inside a tube as<br />
(2)<br />
Maïga et al. [3] numerically investigated<br />
fully-developed turbulent flow<br />
of water/Al 2 O 3 nanofluid through<br />
circular tube using different concentrations<br />
under the constant heat flux<br />
boundary condition. They proposed<br />
the following correlation for 10 4 ≤<br />
Re ≤ 5×10 5 , 6.6 ≤ Pr ≤ 13.9 and 0 ≤<br />
φ ≤ 10%<br />
(3)<br />
Asirvatham et al. [4] reviewed the<br />
published experimental investigations<br />
on convective heat transfer of different<br />
nanofluids.<br />
Despite numerous studies on both<br />
scaled experiments and numerical<br />
modeling on heat transfer enhancement<br />
of nanofluids proliferate over<br />
the past years, most of the test sections<br />
and computational domain were<br />
limited to round pipes. Their simulating<br />
parameters did not reflect the<br />
environment of a nuclear power reactor,<br />
either. Wu and Trupp [5] demonstrated<br />
that flow conditions inside the<br />
fuel rod assembly are quite different<br />
from those in typical pipes. There is<br />
so far no appropriate correlation in<br />
literature that can predict heat transfer<br />
characteristics of nanofluid in a<br />
fuel assembly under PWR operating<br />
condition. Therefore, numerical modeling<br />
has been performed in this study<br />
using a commercial computational<br />
fluid dynamic CFD tool “Star-CCM+<br />
(ver.9.06.011)” to predict heat transfer<br />
and pressure drop more precisely<br />
in a square array subchannel (1.25 ≤<br />
P/D ≤ 1.35) for different volume concentrations<br />
of water/alumina (Al 2 O 3 )<br />
nanofluid (0.5% ≤ φ ≤ 3.0%). Referring<br />
to the Advanced Power Reactor<br />
1400 MWe (APR1400).<br />
Properties<br />
Also, if the slip between the particles<br />
and the continuous phase is trifling,<br />
the flow inside the subchannel may as<br />
well be considered as single phase and<br />
incompressible with constant physical<br />
properties. Both the compression<br />
work and viscous dissipation are<br />
neglected. Under such conditions the<br />
general conservation equations for<br />
mass, momentum and energy can be<br />
written in vector notations:<br />
∇.(ρv) = 0 (4)<br />
∇.(ρvv) = -gradP+μΔ 2 v (5)<br />
∇.(ρvC P T) = ∇.(k gradT) (6)<br />
where v, P and T are fluid velocity<br />
vector, pressure and temperature,<br />
respectively.<br />
2.2 Determination of physical<br />
properties of nanofluid<br />
Determination of physical properties<br />
of nanofluid is key to any nanofluid<br />
research. If the nanoparticles are<br />
assumed to be well dispersed in the<br />
base fluid, the particle concentration<br />
can be considered as constant<br />
throughout the domain and effective<br />
physical properties of mixture can be<br />
evaluated using some classical formulas<br />
well known for two phase fluids<br />
[7]. The following formulas are used<br />
to determine such properties as density,<br />
specific heat, dynamic viscosity<br />
and thermal conductivity.<br />
ρ nf = (1-ϕ)ρ bf + ϕρ P (7)<br />
(C P ) nf = (1-ϕ)(C P ) bf + ϕ(C P ) P (8)<br />
μ nf = (1 + 7.3ϕ + 123ϕ 2 )μ bf (9)<br />
Base Fluid<br />
(Pure Water)<br />
Alumina<br />
Nanoparticles<br />
Density (kg/m 2 ) 734.928 3970<br />
Thermal Conductivity (W/m.K) 0.5701 40<br />
Specific Heat (J/kg. K) 5361.69 880<br />
Dynamics Viscosity (Pa. s) 9.01373E-05 -<br />
| | Tab. 1.<br />
Physical properties of base fluid and alumina nanoparticles.<br />
and later improved by Brinkman [10]<br />
and another by Batchelor [11], these<br />
formulas drastically underestimate<br />
the viscosity of nanofluids. Therefore,<br />
they performed a least-square curve<br />
fitting based on some scarce experimental<br />
data available [12, 13, 14]<br />
which leads to Equation (9). Equation<br />
(10) [7, 15] is introduced for the thermal<br />
conductivity as with the dynamic<br />
viscosity. However, the pressure and<br />
temperature of the above investigations<br />
sizably differ from the operating<br />
condition of a PWR. Since no such<br />
correlation exists for thermophysical<br />
properties of nanofluid applicable to<br />
the operating environment of a PWR it<br />
is assumed that the aforementioned<br />
correlations can also be utilized for<br />
nuclear reactors. Different properties<br />
of base fluid (pure water) and alumina<br />
nanoparticles that have been used in<br />
this study are tabulated in Table 1.<br />
3 Numerical modelling<br />
3.1 Computational domain<br />
The computational domain and<br />
boundaries considered in this study<br />
are shown in Figure 1, which represents<br />
a quarter of a 3-D square array<br />
subchannel created in Star-CCM+.<br />
The diameter of the fuel rod is taken<br />
as 9.5 mm and pitch-to-diameter ratio<br />
P/D of 1.25 and 1.35 are selected for<br />
simulation. The length of the subchannel<br />
is taken as 600 mm which<br />
is long enough to establish a fullydeveloped<br />
turbulent flow at the outlet<br />
under single phase forced convection<br />
condition up to Re = 6×10 5 according<br />
to the following criteria [16]<br />
2 Mathematical modelling<br />
k nf = (1 + 2.72ϕ + 4.97ϕ 2 )k bf (10)<br />
2.1 Governing equations<br />
The term “nanofluid” refers to a twophase<br />
mixture of saturated liquid and<br />
dispersed ultrafine particles of usual<br />
size below 40 nm. However, due to<br />
extremely tiny size of particles, it can<br />
be readily fluidized and thus may be<br />
considered to behave more like a fluid<br />
rather than heterogeneous fluid [6].<br />
Equations (7) and (8) are general<br />
relationships being used in literature<br />
[1, 7, 8] to compute the density and<br />
specific heat for a classical two phase<br />
mixture. Regarding the dynamic<br />
viscosity, Maïga et al. [9] showed that,<br />
albeit several correlations exist to<br />
calculate the dynamic viscosity of<br />
nanofluid as proposed by Einstein<br />
| | Fig. 1.<br />
Computational domain created in Star-CCM+.<br />
Research and Innovation<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
P/D = 1.25<br />
Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid<br />
(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />
6×10 5 7.829 7.963 8.351 9.196<br />
5.098×10 5 6.651 6.766 7.095 7.813<br />
4×10 5 5.219 5.309 5.568 6.130<br />
3×10 5 3.914 3.982 4.176 4.598<br />
| | Tab. 2.<br />
Different inlet velocities, v 0 (m/s) used in simulation.<br />
(11)<br />
l e = EI × D h (12)<br />
where l e is entrance length for fullydeveloped<br />
flow, EI is entrance length<br />
number and Dh is the channel hydraulic<br />
diameter.<br />
3.2 Boundary conditions &<br />
Physics set-up<br />
The coolant enters the subchannel<br />
with a uniform inlet velocity v 0 (m/s)<br />
at the inlet temperature 569 K. Different<br />
values of v 0 for different coolants<br />
that have been used in the simulation<br />
are listed in Table 2. Different properties<br />
of base fluid (pure water)<br />
have been calculated at temperature<br />
569 K and at pressure 155.1375 bar.<br />
At the outlet, a static pressure of<br />
155.1375 bar has been imposed. On<br />
the tube wall, the usual non-slip<br />
conditions with the standard wall<br />
function are considered with a constant<br />
heat flux of 600,000 W/m 2 . The<br />
above parameters and geometric configurations<br />
of the computational<br />
domain are based on the design<br />
features of the APR1400.<br />
The constant density model is chosen<br />
for the material. For turbulence<br />
modeling, the realizable k-ε model<br />
with high y + wall treatment is selected.<br />
Implicit coupled solver with secondorder<br />
upwind discretization scheme in<br />
conjunction with coupled energy<br />
model is implemented which solves<br />
the conservation equations for mass<br />
and momentum simultaneously using<br />
a pseudo time marching approach.<br />
3.3 Turbulence modeling<br />
By studying different literature on<br />
numerical simulation of flow through<br />
a rod bundle for nuclear applications,<br />
it can be concluded that no specific<br />
turbulence model can be regarded as<br />
superior to others for this sort of flow<br />
phenomena. Yadigaroglu et al. [17]<br />
carried out an exhaustive review of<br />
rod bundle numerical simulations<br />
and opined that the gradient transport<br />
models, like the standard k-ε<br />
model, are not capable of predicting<br />
turbulent flow in the narrow gap regions.<br />
Hàzi [18] had demonstrated<br />
that the Reynolds Stress Model (RSM)<br />
could be accurately applied in simulating<br />
the rod bundle geometry. Lee<br />
and Choi [19] also used the RSM turbulence<br />
model to compare the performance<br />
of grid designs between the<br />
small scale vortex flow (SSVF) mixing<br />
vane and the large scale vortex flow<br />
(LSVF) mixing vane. Liu and Ferng<br />
[20] have also adopted RSM turbulence<br />
model to numerically investigate<br />
the effects of different types of<br />
grid (standard grid and split-vane pair<br />
one) on the turbulence mixing and<br />
heat transfer. Palandi et al. [21] have<br />
successfully implemented SST k-ω<br />
model in comparing thermo-hydraulic<br />
performance of nanofluids and<br />
mixing vanes in VVER-440 triangular<br />
array fuel rod bundle. However, application<br />
of RSM turbulence model will<br />
require 50-60% more CPU time per<br />
iteration and 15-20% more memory<br />
usage compared to standard k-ε and<br />
k-ω model.<br />
Recently Conner et al. [22] have<br />
implemented renormalization group<br />
(RNG) k-ε model (Yakhot et al., [23])<br />
in simulation a 5×5 rod bundle with<br />
mixing-vane grid using Star-CCM+.<br />
The applicability of this model to<br />
simulate fuel rod bundles has been<br />
tested and validated by Westinghouse<br />
in their extensive research (Smith et<br />
al., [24]).<br />
Considering the established practice<br />
and computational time required<br />
as discussed above, it can be concluded<br />
that RNG k-ε model will be<br />
sufficient in modeling turbulence for<br />
flow through a rod bundle. However,<br />
in this study, realizable k-ε model<br />
(Shih et al., [25]) has been adopted<br />
for turbulence modeling inside a<br />
square array subchannel since it has<br />
been statistically proved that this<br />
model provides the best performance<br />
among all the k-ε model versions for<br />
separated flows and flows with complex<br />
secondary flow features [26].<br />
The term “realizable” means<br />
that the model satisfies certain mathematical<br />
constraints on the Reynolds<br />
stresses, consistent with the physics<br />
of turbulent flows. Neither the standard<br />
k-ε nor the RNG k-ε model is<br />
realizable.<br />
The modeled transport equation<br />
for k and ε in the realizable k-ε model<br />
are presented by Equation (13) and<br />
Equation (14) respectively:<br />
(13)<br />
and<br />
where,<br />
P/D = 1.35<br />
Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid<br />
(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />
6×10 5 5.826 5.926 6.215 6.843<br />
5.098×10 5 4.950 5.035 5.280 5.814<br />
4×10 5 3.884 3.951 4.143 4.562<br />
3×10 5 2.913 2.963 3.108 3.422<br />
(14)<br />
(15)<br />
(16)<br />
In above equations, G k represents<br />
the generation of turbulence kinetic<br />
energy due to mean velocity gradients,<br />
G b is the generation of turbulence<br />
kinetic energy due to buoyancy, Y M is<br />
the contribution of fluctuating dilatation<br />
in compressible turbulence to<br />
the overall dissipation rate, C 2 and C 1ε<br />
are constants, σ k and σ ε are the<br />
turbulent Prandtl numbers for k and ε<br />
respectively, S k and S ε are user- defined<br />
source terms.<br />
3.4 Convergence of numerical<br />
solution<br />
Another central criteria that must be<br />
satisfied in order to obtain proper<br />
numerical solution is convergence.<br />
The solver needs to be given adequate<br />
iterations so that the problem is converged<br />
and a solution can be treated<br />
as converged if the following criteria<br />
are satisfied [26]:<br />
• The solution no longer changes<br />
with subsequent iterations<br />
• Overall mass, momentum, energy<br />
and scalar balance are achieved<br />
• All equations (momentum, energy<br />
etc.) are obeyed in all cells to a<br />
specific tolerance<br />
RESEARCH AND INNOVATION 251<br />
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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 252<br />
In the present study, residuals for<br />
continuity, X & Y-momentum, Z-<br />
momentum and turbulence kinetic<br />
energy are decreased respectively to<br />
an order of 10 -2 , 10 -5 , 10 -2 and 10 -4<br />
| | Fig. 3.<br />
Distribution of wall y + values in case of pure water<br />
with Re=6×10 5 (P/D =1.35)<br />
| | Fig. 2.<br />
Convergence of mass flow averaged temperature at outlet (P/D = 1.35) for pure water at corresponding<br />
inlet Re = 6×10 5 .<br />
after 30,000 iterations and also a<br />
monitor is created to check how values<br />
for mass flow averaged temperature at<br />
outlet is converging and it is observed<br />
that after 30,000 iterations these<br />
values do not change significantly<br />
with further iterations. A typical plot<br />
of mass flow averaged temperature at<br />
outlet for pure water at inlet Re =<br />
6×10 5 is shown in Figure 2.<br />
3.5 Wall y + values<br />
The accurate calculations of y + value<br />
in the near-wall region, which is a<br />
measure of non-dimensional distance<br />
from the wall to the first mesh node<br />
(based on local cell fluid velocity), are<br />
of paramount importance to the success<br />
of any simulation. In order to use<br />
a wall function approach properly for<br />
a particular turbulence model with<br />
confidence, the y+ values should be<br />
within a certain range.<br />
In the present study, standard wall<br />
function is used in conjunction with<br />
realizable k-ε model and high-y + wall<br />
treatment in which the near-wall cell<br />
centroid are anticipated to be placed<br />
in the log-law region with a value<br />
30 ≤ y + ≤ 100. Results of performed<br />
simulations demonstrate that the<br />
wall y + values for different cases are<br />
within this specified range. A pictorial<br />
representation of wall y + in case<br />
of pure water with Re = 6×10 5<br />
(P/D = 1.35) is shown in Figure 3.<br />
4 Code validation<br />
4.1 Mesh convergence test<br />
Since the accuracy of finite volume<br />
method is directly related to the<br />
quality of discretization used, it is<br />
instrumental to select an optimized<br />
mesh size that will take into account<br />
both resolution of mesh structure and<br />
as well as computational time and<br />
cost.<br />
In the present study, different<br />
mesh settings are selected as presented<br />
in Table 3 and values of<br />
numerically obtained Nu are compared<br />
against an existing correlation<br />
for square array subchannel and for<br />
pure water as presented by Equation<br />
(17) through Equation (19) to check<br />
mesh convergence for computational<br />
domain with P/D =1.35. Results are<br />
plotted in Figure 4 which clearly<br />
states that a mesh setting with base<br />
size 0.7 mm, no. of prism layer 2,<br />
prism layer thickness 0.3mm and<br />
prism layer stretching 3.7 will be<br />
sufficient to produce Nu within<br />
reasonable deviation compared to<br />
the theoretical prediction made by<br />
correlation.<br />
Nu = ψ(Nu ∞ ) c.t. (17)<br />
where,<br />
(Nu ∞ ) c.t. = 0.023 Re 0.8 PR 0.4 (18)<br />
for square array with 1.05 ≤ P/D ≤<br />
1.9 and for pure water, Presser [27]<br />
suggested:<br />
(19)<br />
Base Size<br />
(mm)<br />
No.<br />
Prism Layers<br />
Stretching Thickness<br />
(mm)<br />
Nu<br />
(Star-CCM+)<br />
Nu<br />
(Presser)<br />
Deviation<br />
(%)<br />
0.5 5 1.5 0.7 742.940 -35.051<br />
0.6 4 1.5 0.5 862.627 -16.313<br />
0.7 3 3.8 0.4 933.92 1003.35 -7.434<br />
0.6 2 3.7 0.3 972.102 -3.214<br />
0.7 2 3.7 0.3 1010.57 0.714<br />
| | Fig. 4.<br />
Mesh convergence test with different mesh settings.<br />
| | Tab. 3.<br />
Different mesh settings used to check mesh convergence.<br />
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| | Fig. 5.<br />
Validation of numerical model against correlation for P/D =1.25.<br />
4.2 Validation of numerical<br />
model<br />
Since the ultimate test of any numerical<br />
simulation is the validation of<br />
results against well-known experimental<br />
data, the model under consideration<br />
in the present study has<br />
been validated against correlation of<br />
Presser for square array and pure<br />
water as presented by Equation (17)<br />
through Equation (19). Results are<br />
plotted in Figure 5 and Figure 6<br />
which demonstrates that there is<br />
an excellent agreement between<br />
numerical data and theoretical<br />
prediction for the specified range of<br />
inlet Re.<br />
4.3 Validation of turbulence<br />
model for nanofluid<br />
Despite in the present study it is<br />
assumed that nanofluid would behave<br />
as a single-phase homogeneous fluid<br />
and hence, all of the general conservation<br />
equations of mass, momentum<br />
and energy can directly be applied in<br />
case of nanofluid, however, a successful<br />
comparison of numerical Nu obtained<br />
realizable k-ε model has been<br />
carried out against both empirical<br />
correlation and experimental data of<br />
Pak & Cho [1] for turbulent flow<br />
inside a round pipe of inside diameter<br />
10.66 mm using alumina nanofluid<br />
(φ=2.78%) as coolant for inlet Re<br />
spanning from 5.03×10 4 to 1.48×10 4 .<br />
The results are plotted in Figure 7<br />
which clearly delineates that this<br />
model can perform quite satisfactorily<br />
with nanofluids.<br />
5 Numerical results<br />
and discussion<br />
5.1 Temperature<br />
Temperature profile along the centerline<br />
of subchannel (P/D =1.25) for<br />
different coolants at inlet Re = 6×10 5<br />
are illustrated in Figure 8 from which<br />
it is clear that there is a steady increase<br />
in the coolant temperature due to absorption<br />
of heat while flowing through<br />
the subchannel and bulk temperature<br />
of nanofluid is decreased with the increasing<br />
particle volume concentration.<br />
Numerically obtained fluid average<br />
temperature (in case<br />
of pure water at P/D =1.25 and<br />
inlet Re = 6×10 5 ) at different axial<br />
locations within the subchannel is<br />
compared against the theoretical<br />
predictions from energy balance<br />
according to equation (20) [28] and<br />
results are tabulated in Table 4.<br />
<br />
(20)<br />
The analogy shows that maximum<br />
deviation between numerically obtained<br />
axial temperature and theoretical<br />
prediction is less than 0.6%.<br />
5.2 Velocity<br />
Development of axial velocity along<br />
the centerline of subchannel (P/D<br />
| | Fig. 6.<br />
Validation of numerical model against correlation for P/D =1.35.<br />
=1.25) for different coolants at inlet<br />
Re = 6×10 5 is presented in Figure 9<br />
which clearly states that fullydeveloped<br />
velocity profile occurs<br />
approximately after z=0.3 m and if<br />
the current models are implemented<br />
to evaluate physical properties of<br />
nanofluid, development of velocity<br />
| | Fig. 7.<br />
Validation of turbulence model against Pak & Cho’s correlation.<br />
| | Fig. 8.<br />
Temperature along centerline of subchannel at Re = 6×10 5 .<br />
RESEARCH AND INNOVATION 253<br />
Axial Position<br />
(m)<br />
Average Bulk Fluid Temperature T m (K) %<br />
of Deviation<br />
Start-CCM+ Energy Balance<br />
0 569 569 0.000<br />
0.15 569.2431 570.6885 0.2532<br />
0.30 570.1277 572.3771 0.3929<br />
0.45 571.2205 574.0656 0.4956<br />
0.60 572.4116 575.7542 0.5805<br />
| | Tab. 4.<br />
Comparison of numerically obtained axial temperature against theoretical predictions for pure water<br />
(P/D =1.25 and inlet Re = 6×10 5 ).<br />
| | Fig. 9.<br />
Velocity along centerline of subchannel at Re = 6×10 5 .<br />
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RESEARCH AND INNOVATION 254<br />
| | Fig. 10.<br />
Pressure along centerline of subchannel at Re = 6×10 5 .<br />
profile is not affected by the inclusion<br />
of nanoparticles. From Figure 9, it can<br />
also be seen that there is an increase in<br />
the velocity magnitude due to growth<br />
of hydrodynamic boundary layer as<br />
coolant flows from inlet towards<br />
outlet. The inclusion of higher volume<br />
concentration of nanoparticles also<br />
augments the magnitude of axial<br />
velocity as seen in Figure 9. It can be<br />
explained from the fact that since<br />
with the rise of volume concentration<br />
the viscosity of the nanofluid is also<br />
aggrandized, hence to a maintain<br />
constant value of Reynolds number Re<br />
at the inlet of the channel, velocity<br />
magnitude should be increased too<br />
according to equation (21) if the other<br />
properties remain constant:<br />
<br />
(21)<br />
5.3 Pressure<br />
A plot of static pressure along the<br />
centerline of the subchannel (P/D<br />
=1.25) for different coolants at inlet<br />
Re = 6×10 5 is shown in Figure 10<br />
which depicts that there is an increase<br />
in axial pressure with the inclusion of<br />
nanoparticles which is expected due<br />
to higher viscosity and density as the<br />
particle volume concentration is increased.<br />
5.4 Nu and h Constant Inlet Re<br />
Convective heat transfer is studied<br />
with Star-CCM+ for pure water and<br />
different concentrations of alumina<br />
nanofluid according to Equation (22)<br />
and Equation (23) respectively. Values<br />
of Nu are evaluated at the outlet of the<br />
subchannel to assure fully-developed<br />
turbulent flow condition.<br />
<br />
<br />
(22)<br />
(23)<br />
where, q '' is the constant heat flux<br />
(W/m 2 ), k is thermal conductivity<br />
(W/m 2 .K), D h is hydraulic diameter<br />
(m), and T w and T m are wall and mean<br />
bulk fluid temperature (K) respectively.<br />
Numerical results of Nu and h for<br />
subchannel with different pitch-todiameter<br />
(P/D) ratio are presented<br />
through Figure 11 to Figure 14<br />
respectively and percentage of convective<br />
heat transfer increment for<br />
different nanofluid coolants are<br />
documented in Table 5.<br />
From the results, it is obvious that<br />
the convective heat transfer coefficient<br />
is remarkably increased with the<br />
increment of nanoparticle volume<br />
concentration and in case of 3.0 %<br />
volume concentration, convective<br />
heat transfer is increased above<br />
22.0 % compared to pure water.<br />
5.5 Comparison of Numerical<br />
Results against Correlations<br />
In case of nanofluid with volume<br />
concentration, φ =3.0% numerical<br />
results for Nu are compared against<br />
two well cited correlations of Pak &<br />
Cho [1] and Maïga et al. [3] as shown<br />
in Figure 15 (a) & (b) and an attempt<br />
has been made whether results of<br />
present study can be represented by<br />
either of these two correlations.<br />
The results revealed that Pak<br />
and Cho correlation severely underestimates<br />
the numerical results for<br />
Nu in subchannel and deviation lies<br />
between 17 to 22 percent subject to<br />
inlet Re and P/D.<br />
Regarding correlation of Maïga<br />
et al., it shows better approximation<br />
compared to correlation of Pak & Cho.<br />
Nevertheless, this correlation underestimates<br />
the numerical results for the<br />
| | Fig. 11.<br />
Comparison of Nu for different coolants in subchannel (P/D 1.25).<br />
| | Fig. 12.<br />
Comparison of Nu for different coolants in subchannel (P/D 1.35).<br />
| | Fig. 13.<br />
Comparison of h for different coolants in subchannel (P/D 1.25).<br />
| | Fig. 14.<br />
Comparison of h for different coolants in subchannel (P/D 1.35).<br />
Research and Innovation<br />
Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
range 5×10 5 ≤ Re ≤ 6×10 5 and overestimates<br />
for 3×10 5 ≤ Re ≤ 4×10 5<br />
and deviations are between 0.54 and<br />
6.66 percent depending on inlet Re<br />
and P/D.<br />
5.6 Pressure drop<br />
While nanofluid enhances the convective<br />
heat transfer, the fluid itself<br />
P/D = 1.25<br />
Inlet Re Increment of h (%)<br />
φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />
6×10 5 2.75 9.62 22.46<br />
5.098×10 5 2.75 9.58 22.37<br />
4×10 5 2.72 9.51 22.16<br />
3×10 5 2.74 9.42 21.89<br />
| | Tab. 5.<br />
Heat transfer increment (%) for different nanofluid coolants.<br />
also gets heavier compared to pure<br />
water. Hence, it is of utmost importance<br />
to determine the amount of<br />
pressure drop for the effective application<br />
of nanofluid coolant in nuclear<br />
reactors since it is directly related to<br />
the pumping power required. In this<br />
study, pressure drop along the center<br />
line of the subchannel is evaluated for<br />
different coolants and results are presented<br />
in Figure 16 (a) & (b). Percentage<br />
of pressure drop increment is<br />
documented in Table 6.<br />
The results shows that pressure<br />
drop is significantly increased with<br />
the augmentation of particle volume<br />
concentration which in turn increases<br />
the pumping power. For nanofluid<br />
P/D = 1.55<br />
Inlet Re Increment of h (%)<br />
φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />
6×10 5 2.72 9.56 22.35<br />
5.098×10 5 2.72 9.51 22.26<br />
4×10 5 2.71 9.44 22.01<br />
3×10 5 2.69 9.40 21.87<br />
RESEARCH AND INNOVATION 255<br />
(a) P/D = 1.25<br />
| | Fig. 15.<br />
Comparison of numerical Nu against different correlations.<br />
(b) P/D = 1.35<br />
(a) P/D = 1.25<br />
| | Fig. 16.<br />
Comparison of pressure drop for different coolant.<br />
(b) P/D = 1.35<br />
P/D = 1.25<br />
Inlet Re Increment of ∆p (%)<br />
φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />
6×10 5 6.22 21.53 56.60<br />
5.098×10 5 5.82 21.17 56.62<br />
4×10 5 5.79 21.79 56.02<br />
3×10 5 5.24 21.65 55.83<br />
P/D = 1.35<br />
Inlet Re Increment of ∆p (%)<br />
φ = 0.5 % φ = 1.5 % φ = 3.0 %<br />
6×10 5 5.82 20.94 56.37<br />
5.098×10 5 5.74 21.29 56.08<br />
4×10 5 5.46 20.90 55.10<br />
3×10 5 5.62 20.88 55.82<br />
| | Tab. 6.<br />
Pressure drop increment (%) for different nanofluid coolants.<br />
Research and Innovation<br />
Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
RESEARCH AND INNOVATION 256<br />
with φ=3.0%, pressure drop increment<br />
is about 56% higher compared<br />
to that of pure water.<br />
However, the typical nanoparticle<br />
loading in PWR coolant should be<br />
less than 1.0 vol %. At such lower<br />
con centration, nanofluid properties<br />
are almost similar to that of pure<br />
water and the rise in viscosity as well<br />
as pressure drop will be negligible too.<br />
The present study also portrays that<br />
pressure drop is approximately 20 %<br />
at 1.5 vol. % of nanoparticle concentration<br />
which can also be treated as<br />
tolerable.<br />
The convective heat transfer coefficient<br />
at such low concentration of<br />
nanofluid is yet to be improved due to<br />
higher turbulence produced near the<br />
grid spacers by the presence of nanoparticles<br />
in the base fluid. Since it is<br />
quite difficult to take into account<br />
such effects in numerical simulation,<br />
further experimental investigation is<br />
required for quantification of heat<br />
transfer increment aroused from the<br />
presence of nanoparticles near the<br />
spacer grids.<br />
6 Proposed new<br />
correction factor<br />
Finally, a multiple regression analysis<br />
is performed with numerical results to<br />
propose a new correction factor, β for<br />
the existing correlation of square<br />
array subchannel with pure water as<br />
suggested by Presser [27] so that Nu<br />
for nanofluid coolant can be approximated<br />
in such geometry. Based on<br />
regression results, β can be expressed<br />
as follows:<br />
β = 1 + 0.0247ϕ 1.39 (24)<br />
Nu for nanofluid can be calculated as<br />
follows:<br />
Nu nf = β*(Nu Presser ) Water (25)<br />
The validity of above correlation is for<br />
3×10 5 ≤ Re ≤ 6×10 5 ; 0.847 ≤ Pr ≤<br />
1.011; 1.25 ≤ P/D ≤ 1.35 and 0.5% ≤<br />
φ ≤ 3.0% in case of square array<br />
subchannel.<br />
7 Chemical and physical<br />
stability of nanofluid<br />
Albeit nnanofluid can readily boost<br />
the heat transfer capability of PWR<br />
coolant, there is still no satisfactory<br />
explanation proposed regarding the<br />
prevention of clustering in nanoparticle<br />
suspensions. Agglomeration<br />
in nanofluids containing oxide nanoparticles<br />
can be reduced remarkably<br />
by adjusting the pH to form electric<br />
changes on particle surface so that<br />
they repel each other [29]. However,<br />
the typical values of pH should be<br />
such that nanofluid itself becomes not<br />
corrosive and it should be agreeable<br />
with same allowable pH range of<br />
nuclear reactor, since altering the<br />
PWR coolant chemistry is not a viable<br />
option. Besides, use of surfactants are<br />
also not recommended since it may<br />
undergo severe radiolysis inside the<br />
reactor core during operation.<br />
Hence, issues concerning chemical<br />
and physical stabilities of nanofluid<br />
has yet to be resolved prior to utilizing<br />
nanofluid as a promising coolant in<br />
PWRs to achieve both extended life<br />
time of associated equipment and<br />
higher thermal efficiency.<br />
8 Conclusion<br />
Thermo- and hydrodynamic characteristics<br />
of water/alumina nanofluid<br />
have been studied in a square array<br />
subchannel featuring the pitch-todiameter<br />
ratios of 1.25 and 1.35 under<br />
the steady-state, incompressible,<br />
single- phase turbulent flow condition.<br />
Numerical results have been compared<br />
against correlations in the<br />
literature and the following conclusions<br />
can be drawn.<br />
• Convective heat transfer is increased<br />
with increasing volume<br />
concentration of water/alumina<br />
nanofluid given the inlet Reynolds<br />
number.<br />
• The convective heat transfer increment<br />
of nanofluid is obtained at<br />
the expense of increased pressure<br />
drop and hence, larger pumping<br />
power is required. Therefore,<br />
nano fluid as PWR coolant can be<br />
only be implemented in reality if<br />
the replacement of reactor coolant<br />
pump is a feasible option compared<br />
to higher power gained from<br />
increased nanofluid heat transfer.<br />
Acknowledgements<br />
This work was supported by the<br />
National Research Foundation of Korea<br />
(NRF) grant funded by the Korean<br />
Government (MSIP) under Grant No.<br />
2008-0061900 and partly supported<br />
by the Brain Korea 21 Plus under<br />
Grant No. 21A20130012821.<br />
Nomenclature<br />
∆p Pressure Drop Pa<br />
ρ Density kg/m 3<br />
v Flow Velocity m/s<br />
f Friction Factor -<br />
L Length of Flow Channel m<br />
le Entrance Length m<br />
EI Entrance Length Number -<br />
Dh Hydraulic Diameter m<br />
μ Dynamic Viscosity N.s/m 2<br />
Re Reynolds Number -<br />
Nu Nusselt Number -<br />
Pr Prandtl Number -<br />
Pe Peclet Number -<br />
h<br />
Convective Heat Transfer<br />
CoefficientW/m 2 .K<br />
k Thermal Conductivity W/m.K<br />
C p Specific Heat J/kg.K<br />
T m Bulk Temperature of Fluid K<br />
T w<br />
Surface Temperature<br />
of Heater Rod<br />
P Rod Pitch m<br />
D Rod Diameter m<br />
Q Total Heat Input W<br />
q” Heat Flux W/m 2<br />
φ<br />
ṁ Mass Flow Rate kg/sec<br />
Subscript<br />
nf<br />
bf<br />
P<br />
Volume Concentration<br />
of Nanoparticles %<br />
Nanofluid<br />
Basefluid<br />
Particle<br />
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Research and Innovation<br />
Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
9. Maïga S, Nguyen CT, Galanis N, Roy G,<br />
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Authors<br />
Jubair Ahmed Shamim<br />
Department of Nuclear<br />
Engineering<br />
Seoul National University<br />
Seoul 08826, ROK<br />
Kune Yull Suh<br />
Seoul National University<br />
1 Gwanak Ro, Gwanak Gu<br />
Seoul 08826, ROK<br />
257<br />
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Technik und dem Betrieb tätig. Meine Aufgaben bezogen sich auf<br />
den Bau, Betrieb und Rückbau von Kernkraftwerken. Während<br />
der nunmehr 27 Jahre Tätigkeit für PreussenElektra habe ich<br />
Aufgaben in den Kernkraftwerken und der Zentralorganisation wahrgenommen.<br />
Seit 2014 führe ich das Geschäftsführungsressort Technik<br />
und Betrieb. Ich war 13 Jahre im KTA tätig und 5 Jahre Mitglied<br />
der deutschen Reaktorsicherheitskommission. Seit 4 Jahren engagiere<br />
ich mich als Governor bei der WANO – World Association for<br />
Nuclear Operators – weltweit.<br />
Während meines bisherigen beruflichen Werdegangs war und<br />
ist der sichere, umweltverträgliche und wirtschaftliche Betrieb der<br />
Kernkraftwerke mein prioritäres Anliegen. Die Kernkraft hat mich in<br />
meinem ganzen Berufsleben fasziniert und die Faszination hält trotz<br />
aller Rückschläge und der teilweise schwierigen Randbedingungen<br />
für die Kernenergie in Deutschland an.<br />
Seit 1991 bin ich Mitglied in der KTG und nunmehr schon seit<br />
8 Jahren im Vorstand, jetzt als Schatzmeister. Ziel meiner erneuten<br />
Kandidatur ist, die KTG als Interessengemeinschaft aller in der<br />
Kerntechnik Tätigen und von ihr faszinierten Mitgliedern mit meinem<br />
Wissen und meiner beruflichen Erfahrung zu unterstützen sowie die<br />
Wissensübertragung und den Erfahrungsaustausch zu erhalten.<br />
Wie politisch gewollt, sollten wir Kerntechniker den sicheren Betrieb<br />
bis zum Laufzeitende und den Rückbau unserer Kernkraftwerke<br />
in Deutschland mit Ehre abschließen. Kein Grund mit Blick auf das<br />
Erreichte der letzten 50 Jahre nicht stolz sein zu dürfen!<br />
Jörg Starflinger<br />
Prof. Dr.-Ing. (51), Stuttgart<br />
Zur Person<br />
Nach dem Studium des Maschinenbaus an<br />
der Ruhr-Universität Bochum (RUB) mit<br />
Schwerpunkt Energietechnik Promotion<br />
im Jahr 1997 am Lehrstuhl für Nukleare<br />
und neue Energiesysteme der RUB, Prof.<br />
Dr.-Ing. H. Unger. 1998 Wechsel als Nachwuchswissenschaftler zum<br />
Forschungszentrum Karlsruhe, heute Karlsruhe Institut für Technologie.<br />
Themenschwerpunkte: Wasserstofferzeugung bei schweren<br />
Unfällen in Leichtwasserreaktoren und Kreislaufsimulation von<br />
innovativen Reaktorkonzepten. 2006 Leiter der Gruppe „Kraftwerkstechnik“<br />
am Institut für Kern- und Energietechnik (IKET), Prof. Dr.-Ing.<br />
T. Schulenberg, in der innovative Kernkraftwerkskonzepte mit überkritischem<br />
Wasser von mehreren Doktoranden untersucht wurden.<br />
2010 Ruf an die Universität Stuttgart zum ordentlichen Professor des<br />
Lehrstuhls für Kerntechnik und Reaktorsicherheit und Leiter des<br />
Instituts für Kernenergetik und Energiesysteme (IKE). Neben der<br />
Lehre im Bereich Kerntechnik Schwerpunkte in der Reaktorsicherheitsforschung,<br />
z.B. in der Modellentwicklung zur Beschreibung der<br />
späten Phase von Kernschmelzunfällen in Leichtwasserreaktoren<br />
KTG Inside
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
und auf dem Gebiet innovativer Sicherheitssysteme, z.B. der passiven<br />
Lagerbeckenkühlung mit Wärmerohren (Heat pipes) und nachrüstbaren<br />
Nachwärmeabfuhrsystemen mit überkritischem CO 2 als<br />
Arbeitsmittel.<br />
Zur Wahl als Vorstandsmitglied<br />
Ich engagiere mich in der KTG auf dem Gebiet des Kompetenzerhalts<br />
und der Kompetenzförderung. Den von Dr. Wolfgang Steinwarz ins<br />
Leben gerufenen, sehr erfolgreichen Workshop „Kompetenzerhalt in<br />
der Kerntechnik“ habe ich verantwortlich übernommen und möchte<br />
ihn in den kommenden Jahren weiterführen. Dr. Steinwarz steht uns<br />
auch als Ruheständler dankenswerterweise als Jurymitglied weiter<br />
zur Seite. Die Umbenennung in „Young Scientists Workshop“ soll eine<br />
Öffnung zu kerntechnisch verwandten Forschungsthemen, beispielsweise<br />
„Kerntechnik und Gesellschaft“, symbolisieren.<br />
Durch meine Mitarbeit im KTG-Vorstand als Vorstandsmitglied<br />
möchte ich einen Strategieentwicklungsprozess anstoßen, der<br />
mittelfristig eine genügende Anzahl an jungen hochqualifizierten<br />
und motivierten Personen für die zukünftigen spannenden und<br />
herausfordernden nationalen und internationalen kerntechnischen<br />
Aufgaben sicherstellt. Für den Kompetenzerhalt und die Nachwuchsförderung<br />
bieten die KTG und Ihre Mitglieder sowie unsere Tagung<br />
„Annual Meeting on Nuclear Technology“ die ideale Plattform.<br />
Walter Tromm<br />
Dr.-Ing. (58), Stutensee<br />
Zur Person<br />
Maschinenbaustudium an der Uni (TH)<br />
Karlsruhe mit dem Studienschwerpunkt<br />
Kerntechnik und dort Promotion zum Thema<br />
„Experimentelle Untersuchungen zum Nachweis<br />
der langfristigen Kühlbarkeit von<br />
Kernschmelzen“. Seit 1988 am damaligen Forschungszentrum<br />
Karlsruhe, heute Karlsruher Institut für Technologie, angestellt und<br />
schwerpunktmäßig mit Reaktorsicherheitsfragen bei auslegungsüberschreitenden<br />
Störfällen beschäftigt. Von 1998 bis 1999 Gastwissenschaftler<br />
am Europäischen Gemeinschaftsforschungszentrum<br />
in Ispra (Italien) tätig.<br />
Seit 2002 Programmbevollmächtigter in der Programmleitung<br />
Nukleare Sicherheitsforschung des FZK bzw. heute Nukleare<br />
Entsorgung, Sicherheit und Strahlenforschung des KIT; stellvertretender<br />
Leiter seit 2007 wurde seit 2010 Programmleiter. 2014 im<br />
geschäftsführenden Ausschuss des Bereichs Maschinenbau und<br />
Elektrotechnik des KIT berufen. Seit 2015 darüber hinaus Sprecher<br />
des vom KIT neu eingerichteten Kompetenzzentrums Rückbau und<br />
seit 2017 Vorsitzender des Kompetenzverbundes Kerntechnik.<br />
Tätig in nationalen und internationalen Gremien, bei der OECD/<br />
NEA der deutsche Repräsentant des Nuclear Science Committee, bei<br />
der IAEA in der Technical Working Group Light Water Reactors und<br />
Mitglied im Governing Board der EU-SNETP Plattform. Weiterhin<br />
innerhalb des VDI Vorsitzender des Fachausschusses Kraftwerkstechnik.<br />
Seit 2016 Leiter des neu gegründeten Kompetenz-Cluster<br />
Rückbau, der die Expertise im Rückbau mehrerer Länder zusammenführt.<br />
Zur Wahl als stellvertretender Vorsitzender<br />
Die Bundesregierung hat 2011 nach den Ereignissen in dem<br />
Kernkraftwerk Fukushima Daii-chi in Japan entschieden, aus der<br />
Stromproduktion mittels Kernkraft auszusteigen. In den nächsten<br />
4 Jahren werden die letzten Kernkraftwerke in Deutschland<br />
abgeschaltet. Diesen Ausstieg nach wie vor so sicher wie möglich<br />
mitzugestalten ist eine der Aufgaben, die die in der deutschen<br />
Kerntechnik arbeitenden Ingenieure und Naturwissenschaftler<br />
haben. International und auf europäischer Ebene wird jedoch Kernenergie<br />
langfristig weiterhin genutzt. Auch für den Industriestandort<br />
Deutschland und für den Erhalt von Arbeitsplätzen ist der Export von<br />
Komponenten für kerntechnische Anlagen nach wie vor bedeutsam.<br />
Ebenfalls werden der Rückbau der Kernkraftwerke und die Endlagerfrage<br />
die Gesellschaft noch über Jahrzehnte beschäftigen. Der<br />
Ausstieg aus der Stromproduktion durch Kernenergie darf daher<br />
nicht bedeuten, sich von den entsprechenden kerntechnischen<br />
Kompetenzen in der Industrie, den Behörden und den Universitäten<br />
und Forschungszentren zu verabschieden. In den Bereichen Reaktorsicherheit,<br />
Rückbau, Endlagerung, Strahlenschutz und Krisenmanagement<br />
sind diese Kompetenzen auch weiterhin gefragt. In<br />
Europa stammen 27 % der Stromproduktion aus Kernkraftwerken.<br />
Zur kompetenten Bewertung kerntechnischer Einrichtungen innerhalb<br />
Europas und zur kritischen Begleitung internationaler Entwicklungen<br />
sind eine enge Zusammenarbeit auf nationaler, europäischer<br />
und internationaler Ebene unerlässlich. Deshalb sehe ich als eine<br />
der Hauptaufgaben der KTG den Erhalt der kerntechnischen<br />
Kompetenzen in allen genannten Bereichen.<br />
259<br />
KTG INSIDE<br />
Herzlichen<br />
Glückwunsch<br />
April <strong>2018</strong><br />
97 Jahre wird<br />
2. Prof. Dr. Albert Ziegler, Karlsbad<br />
87 Jahre werden<br />
9. Dr. Klaus Penndorf, Geesthacht<br />
11. Hubert Bairiot, Mol/B<br />
19. Dr. Klaus Einfeld, Murnau<br />
28. Dipl.-Ing. Rudolf Eberhart, Burgdorf<br />
85 Jahre wird<br />
6. Ing. Reinhard Faulhaber, Köln<br />
84 Jahre wird<br />
22. Dipl.-Ing. Gert Slopianka,<br />
Gorxheimeral<br />
83 Jahre werden<br />
3. Dipl.-Psych. Georg Sieber,<br />
München<br />
5. Prof. Dr. Hans-Henning Hennies,<br />
Karlsruhe<br />
19. Dr. Ernst Müller, Rösrath<br />
19. Dr. Gottfried Class,<br />
Eggenstein-Leopoldshafen<br />
21. Dipl.-Ing. Walter Jansing,<br />
Bergisch Gladbach<br />
30. Dr. Friedrich-Wilhelm Heuser,<br />
Overath<br />
82 Jahre werden<br />
4. Helmut Kuhne, Neunkirchen<br />
6. Dipl.-Ing. Hans Pirk, Rottach-Egern<br />
10. Dipl.-Ing. Franz Stockschläder,<br />
Bad Bentheim<br />
11. Dipl.-Ing. Bernhard-F. Roth,<br />
Eggenstein-Leopoldshafen<br />
24. Dipl.-Ing. Horst Schott, Overath<br />
81 Jahre werden<br />
7. Dipl.-Ing. Helmut Adam, Neuenhagen<br />
13. Dr. Martin Peehs, Bubenreuth<br />
80 Jahre werden<br />
4. Prof. Dr. Klaus Kühn, Clausthal- Zellerfeld<br />
5. Dr. Hans Fuchs, Gelterkinden/CH<br />
9. Dr. Carl Alexander Duckwitz, Alzenau<br />
28. Prof. Dr. Georg-Friedrich Schultheiss,<br />
Lüneburg<br />
79 Jahre wird<br />
8. Dr. Siegbert Storch, Aachen<br />
78 Jahre wird<br />
18. Dipl.-Ing. Norbert Granner,<br />
Bergisch Gladbach<br />
77 Jahre werden<br />
17. Dipl.-Phys. Ernst Robinson, Gehrden<br />
28. Dr. Ludwig Richter, Hasselroth<br />
KTG Inside
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
260<br />
NEWS<br />
Wenn Sie keine<br />
Erwähnung Ihres<br />
Geburtstages in<br />
der <strong>atw</strong> wünschen,<br />
teilen Sie dies bitte<br />
rechtzeitig der KTG-<br />
Geschäftsstelle mit.<br />
KTG Inside<br />
Verantwortlich<br />
für den Inhalt:<br />
Die Autoren.<br />
Lektorat:<br />
Sibille Wingens,<br />
Kerntechnische<br />
Gesellschaft e. V.<br />
(KTG)<br />
Robert-Koch-Platz 4<br />
10115 Berlin<br />
T: +49 30 498555-50<br />
F: +49 30 498555-51<br />
E-Mail: s.wingens@<br />
ktg.org<br />
www.ktg.org<br />
76 Jahre werden<br />
9. Prof. Dr. Hans-Christoph Mehner,<br />
Dresden<br />
27. Dr. Dieter Sommer, Mosbach<br />
27. Dr. Jürgen Wunschmann, Eggenstein<br />
29. Dr. Klaus-Detlef Closs, Karlsruhe<br />
75 Jahre werden<br />
15. Dr. Werner Dander, Heppenheim<br />
18. Dipl.-Betriebsw. Uwe Janßen,<br />
Weinheim<br />
18. Dipl.-Ing. Victor Luster, Bamberg<br />
26. Ing. Helmut Schulz, Kürten<br />
70 Jahre werden<br />
6. Dr. Wolfgang Tietsch, Mannheim<br />
9. Ing. Herbert Moryson, Essen<br />
22. Dr. Heinz-Dietmar Maertens, Arnum<br />
26. Dr. Rainer Heibel, Ness Neston/GB<br />
27. Ulrich Wimmer, Erlangen<br />
65 Jahre werden<br />
10. Dipl.-Phys. Harold Rebohm, Berlin<br />
24. Dipl.-Phys. Michael Beczkowiak,<br />
Karben<br />
60 Jahre werden<br />
4. Dipl.-Ing. Holger Bröskamp,<br />
Höhnhorst<br />
4. Dipl.-Ing. (FH) Franz Xaver Pirzer,<br />
Schwandorf<br />
50 Jahre werden<br />
16. Rainer Bezold, Dormitz<br />
16. Dr. Matthias Messer, Tetbury/GB<br />
30. Dr. Christian Raetzke, Leipzig<br />
Mai <strong>2018</strong><br />
94 Jahre wird<br />
22. Prof. Dr. Fritz Thümmler, Karlsruhe<br />
90 Jahre wird<br />
10. Dr. Heinz Büchler, Sankt Augustin<br />
89 Jahre wird<br />
31. Dipl.-Ing. Werner-P. Kürsten,<br />
Mannheim<br />
88 Jahre wird<br />
9. Dr. Hans-Jürgen Hantke, Kempten<br />
85 Jahre werden<br />
4. Dr. Klaus Wiendieck, Baden-Baden<br />
25. Dr. Reinhold Mäule, Walheim<br />
25. Georg von Klitzing, Bonn<br />
84 Jahre werden<br />
11. Dr. Eckhart Leischner, Rodenbach<br />
14. Dr. Alexander Warrikoff, Frankfurt/M.<br />
26. Dr. Günter Kußmaul, Manosque/F<br />
83 Jahre werden<br />
1. Dr. Willi Bermel, Jülich<br />
8. Dipl.-Ing. Klaus Wegner, Hanau<br />
22. Dr. Heinz Vollmer, Lampertheim<br />
28. Dipl.-Ing. Anton Zimmermann,<br />
Hamburg<br />
29. Dipl.-Ing. Karlheinz Orth,<br />
Marloffstein<br />
82 Jahre werden<br />
3. Ewald Jurisch, Erlangen<br />
10. Dr. Peter Reinke, Röttenbach<br />
18. Dipl.-Ing. Gerhard Lorenz, Bochum<br />
81 Jahre werden<br />
1. Prof. Dr. Dietrich Munz,<br />
Graben-Neudorf<br />
3. Dipl.-Ing. Harald Enderlein, Karlsruhe<br />
6. Dr. Peter Strohbach, Mainaschaff<br />
7. Prof. Dr. Werner Lutze,<br />
Chevy Chase/USA<br />
20. Dr. Norbert Krutzik, Frankfurt/M.<br />
26. Dipl.-Ing. Rüdiger Müller, Heidelberg<br />
27. Dr. Johannes Wolters, Düren<br />
28. Dipl.-Ing. Heinz E. Häfner, Bruchsal<br />
80 Jahre werden<br />
12. Dr. Herbert Finnemann, Erlangen<br />
13. Dipl.-Ing. Otto A. Besch, Geesthacht<br />
13. Dr. Heinrich Werle,<br />
Karlsdorf-Neuthard<br />
16. Dr. Hans-Dieter Harig, Hannover<br />
21. Dr. Hans Spenke, Bergisch Gladbach<br />
79 Jahre werden<br />
4. Dipl.-Ing. Norbert Albert, Ettlingen<br />
5. Dr. Wolfgang Voigts, Linkenheim<br />
27. Prof. Dr. Dietrich Kirsch<br />
78 Jahre werden<br />
11. Dr. Andreas Hölzler, Schwaig<br />
15. Dipl.-Phys. Ludwig Aumüller,<br />
Freigericht<br />
18. Dr. Karl Schulte, Köln<br />
24. Dipl.-Ing. Herbert Krinninger,<br />
Bergisch Gladbach<br />
77 Jahre werden<br />
8. Prof. Dr. Helmut Alt, Aachen<br />
12. Dipl.-Ing. Dieter Rohde, Mannheim<br />
16. Dr. Jürgen Baier, Höchberg<br />
76 Jahre werden<br />
5. Hans-Bernd Maier, Aschaffenburg<br />
9. Dr. Egbert Brandau, Alzenau<br />
11. Dr. Erwin Lindauer, Köln<br />
17. Dr. Heinz-Peter Holley, Forchheim<br />
18. Dipl.-Ing. Josef Koban, Buckenhof<br />
28. Dipl.-Ing. Wolf-Dieter Krebs,<br />
Bubenreuth<br />
75 Jahre werden<br />
3. Dipl.-Ing. Hans Lettau, Effeltrich<br />
14. Dr. Helmut-K. Hübner, Bruchsal<br />
20. Dipl.-Ing. Dietmar Bittermann, Fürth<br />
22. Dr. Wolfgang Schütz, Bruchsal<br />
23. Dipl.-Ing. Max Heller, Uttenreuth<br />
24. Dipl.-Ing. Rudolf Weh,<br />
Stephanskirchen<br />
27. Dr. Kurt Fischer, Erlangen<br />
65 Jahre werden<br />
2. Dipl.-Ing. Marc Winter, Veitshöchheim<br />
3. Dipl.-Ing. Karl-Heinz Wiening,<br />
Herzogenaurach<br />
5. Michael Klein, Großenwörden<br />
16. Ing. grad. Eckhard Raabe, Geiselbach<br />
21. Dipl.-Ing. (FH) Reinhold Horstmann,<br />
Erlangen<br />
27. Dipl.-Ing. (FH) Ulrich Hudezeck,<br />
Nürnberg<br />
60 Jahre wird<br />
23. Dr. Hans-Josef Zimmer, Steinfeld<br />
50 Jahre werden<br />
10. Dr. Astrid Petersen, Hamburg<br />
20. Dipl.-Ing. (FH) Jürgen Bruder,<br />
Gundremmingen<br />
Die KTG gratuliert ihren Mitgliedern<br />
sehr herzlich zum Geburtstag und wünscht ihnen weiterhin alles Gute!<br />
Top<br />
IAEA Expands International<br />
Cooperation on Small,<br />
Medium Sized or Modular<br />
Nuclear Reactors<br />
(iaea) The International Atomic<br />
Energy Agency (IAEA) is launching an<br />
effort to expand international cooperation<br />
and coordination in the design,<br />
development and deployment of<br />
small, medium sized or modular<br />
reactors (SMRs), among the most promising<br />
emerging technologies in<br />
nuclear power.<br />
Significant advances have been<br />
made on SMRs, some of which will use<br />
pre-fabricated systems and components<br />
to shorten construction schedules<br />
and offer greater flexibility and<br />
affordability than traditional nuclear<br />
power plants. With some 50 SMR concepts<br />
at various stages of development<br />
around the world, the IAEA is forming<br />
a Technical Working Group (TWG) to<br />
guide its activities on SMRs and provide<br />
a forum for Member States to<br />
share infor mation and knowledge,<br />
IAEA Deputy Director General Mikhail<br />
Chudakov said.<br />
“Innovation is crucial for nuclear<br />
power to play a key role in de carbonising<br />
the energy sector,” Chudakov,<br />
who heads the IAEA Department of<br />
Nuclear Energy, said at a conference<br />
on SMRs in Prague on 15 February.<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
“Many Member States that are<br />
operating, expanding, introducing or<br />
considering nuclear power are quite<br />
keen on the development and<br />
deployment of SMRs.”<br />
Global interest in SMRs is growing.<br />
SMRs have the potential to meet the<br />
needs of a wide range of users and to<br />
be low carbon replacements for ageing<br />
fossil fuel fired power plants. They<br />
also display enhanced safety features<br />
and are suitable for non-electric applications,<br />
such as cooling, heating and<br />
water desalination. In addition, SMRs<br />
offer options for remote regions with<br />
less developed infrastructure and for<br />
energy systems that combine nuclear<br />
and alternative sources, including<br />
renewables.<br />
The first three advanced SMRs are<br />
expected to begin commercial operation<br />
in Argentina, China and the<br />
Russian Federation between <strong>2018</strong> and<br />
2020. SMR development is also well<br />
advanced in about a dozen other<br />
countries.<br />
The TWG, comprising some 20<br />
IAEA Member States and international<br />
organizations, is scheduled to<br />
meet for the first time on 23-26 April<br />
at the IAEA’s headquarters in Vienna.<br />
It is part of an expanding suite of<br />
services the IAEA offers Member<br />
States on this emerging nuclear power<br />
technology. These include an SMR<br />
computer simulation programme to<br />
help educate and train nuclear professionals;<br />
a methodology and related<br />
IT tool for training in assessing the<br />
reactor technology of different SMRs;<br />
and the SMR Regulators’ Forum.<br />
The forum, set up in 2015, enables<br />
discussions among Member States and<br />
other stakeholders to share SMR<br />
regulatory knowledge and experience.<br />
It contributes to enhancing safety by<br />
identifying and resolving issues that<br />
may challenge regulatory reviews of<br />
SMRs and by facilitating robust and<br />
thorough regulatory decisions.<br />
Responding to requests from<br />
Member States in Europe, the IAEA<br />
recently launched a project to build<br />
regional capacities for making knowledgeable<br />
decisions on SMRs, including<br />
technical assessments for SMRs<br />
that are commercially available for<br />
near term deployment. The two-year<br />
project seeks to contribute to meeting<br />
growing European demand for<br />
flexible sources of electricity that do<br />
not release greenhouse gases. Its first<br />
meeting will be held on 13-15 March<br />
at the IAEA in Vienna.<br />
An expeditious deployment of<br />
SMRs faces challenges, including the<br />
need to develop a robust regulatory<br />
| | IAEA Expands International Cooperation on Small, Medium Sized or Modular Nuclear Reactors.<br />
framework, new codes and standards,<br />
a resilient supply chain and human<br />
resources. And although SMRs require<br />
less upfront capital per unit, their<br />
electricity generating cost will<br />
probably be higher than that of large<br />
reactors. Their competitiveness must<br />
be weighed against alternatives and<br />
be pursued through economies of<br />
scale. Detailed technical information<br />
on SMRs under construction or design<br />
can be found at the IAEA’s Advanced<br />
Reactor Information System.<br />
“Realistically, we could expect the<br />
first commercial SMR fleet to start<br />
between 2025 and 2030,” said Hadid<br />
Subki, Scientific Secretary of the TWG<br />
and a Team Leader in SMR Technology<br />
Development at the IAEA. “We<br />
trust this new Technical Working<br />
Group will help further the advancement<br />
of SMR technology and guide<br />
the Agency in its programmes and<br />
projects in this field.”<br />
| | (18791436), www.iaea.org<br />
World<br />
Poll Shows Local Residents<br />
Support Poland’s Plans for<br />
First Nuclear Plant<br />
(nucnet) A poll carried out for Poland’s<br />
PGE EJ1, the company in charge of the<br />
country’s first nuclear power station<br />
project, has shown that 67% of residents<br />
in areas around the proposed<br />
site in northern Poland support the<br />
potential construction of a nuclear<br />
power station in their region.<br />
PGE said a poll was carried out<br />
in November and December 2017<br />
in three municipalities, Choczewo,<br />
Gniewino, Krokowa, all close to<br />
Poland’s Baltic coast in the northern<br />
province of Pomerania.<br />
According to PGE, local residents<br />
indicated they are in favour of the<br />
project because of the development<br />
and job opportunities it could bring to<br />
their regions. The poll showed 49% of<br />
respondents expect cheaper electricity<br />
to be one of the benefits from a<br />
nuclear station, while 35 % expect<br />
local infrastructure development.<br />
In April 2017, PGE began environmental<br />
and site selection surveys at<br />
two locations – Lubiatowo-Kopalino in<br />
the municipality of Choczewo and<br />
Żarnowiec in the municipality of<br />
Korkowa.<br />
The studies aim to determine the<br />
potential impact of the project on both<br />
the environment and local residents.<br />
An initial round of environmental<br />
studies has already been carried out at<br />
both locations.<br />
The Polish government has not<br />
made a final decision about the<br />
country’s nuclear programme, with<br />
the deadline being pushed back<br />
several times. According to latest<br />
reports, a decision is now expected in<br />
mid-<strong>2018</strong>.<br />
| | pgeej1.pl<br />
SKB, Sweden: Two Statements<br />
on the Spent Fuel Repository<br />
(skb) The answer was a clear yes in<br />
SSM’s statement to the Government<br />
on SKB’s system for final disposal of<br />
spent nuclear fuel. The Land and<br />
Environment Court was also positive<br />
| | Aerial photo of the planned site of the Spent Fuel Repository (centre)<br />
at Forsmark. The picture is a photomontage. Illustration: Phosworks.<br />
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*)<br />
Net-based values<br />
(Czech and Swiss<br />
nuclear power<br />
plants gross-based)<br />
1)<br />
Refueling<br />
2)<br />
Inspection<br />
3)<br />
Repair<br />
4)<br />
Stretch-out-operation<br />
5)<br />
Stretch-in-operation<br />
6)<br />
Hereof traction supply<br />
7)<br />
Incl. steam supply<br />
8)<br />
New nominal<br />
capacity since<br />
January 2016<br />
9)<br />
Data for the Leibstadt<br />
(CH) NPP will<br />
be published in a<br />
further issue of <strong>atw</strong><br />
BWR: Boiling<br />
Water Reactor<br />
PWR: Pressurised<br />
Water Reactor<br />
Source: VGB<br />
in several important respects but calls<br />
for more documentation on the<br />
copper canisters.<br />
The Swedish Radiation Safety<br />
Authority (SSM) has reviewed SKB’s<br />
applications under the Nuclear Activities<br />
Act and recommends the Government<br />
to grant a licence for a final<br />
repository for spent nuclear fuel in<br />
Forsmark and an encapsulation plant<br />
in Oskarshamn.<br />
The statement from the Land and<br />
Environment Court (MMD) is also<br />
positive in several important respects.<br />
The court says yes to the issues<br />
relating to the Forsmark site, the rock,<br />
the buffer and the environmental<br />
impact statement. The court also<br />
approves the encapsulation plant and<br />
increased capacity in the interim<br />
storage facility Clab. However, the<br />
court wants SKB to present more<br />
documentation on the properties of<br />
the canister and safety in the<br />
long term. Furthermore, it wants<br />
an investigation of the issue of responsibility<br />
after closure, which has also<br />
been requested by the munici pality.<br />
We can conclude that we have<br />
not been able to answer the court’s<br />
questions regarding the copper<br />
canister fully. At the same time, the<br />
Government’s expert authority SSM<br />
wrote in its statement that SKB has<br />
the potential to meet the legislative<br />
requirements on safe final disposal,<br />
says SKB’s managing director Eva<br />
Halldén in a comment.<br />
SKB will provide documentation<br />
That the two authorities have come<br />
to such different conclusions is in<br />
part due to the fact that they have<br />
tried the applications under different<br />
legislations, SSM under the Nuclear<br />
Activities Act and MMD under the<br />
Environmental Code. They also have<br />
different licensing procedures. SSM<br />
grants a licence in several steps with<br />
continuous updates of the safety<br />
analysis. But the court must say yes<br />
or no based on the currently available<br />
documentation.<br />
The issue now lies with the<br />
Ministry of the Environment and<br />
Energy for further investigation and<br />
SKB is working to develop the documentation<br />
on the canister required<br />
by the court.<br />
This is material that we have<br />
planned to produce for the preliminary<br />
safety analysis. The difference<br />
now is that we will prioritise the work<br />
Operating Results November 2017<br />
Plant name Country Nominal<br />
capacity<br />
Type<br />
gross<br />
[MW]<br />
net<br />
[MW]<br />
Operating<br />
time<br />
generator<br />
[h]<br />
Energy generated. gross<br />
[MWh]<br />
Month Year Since<br />
commissioning<br />
Time availability<br />
[%]<br />
Energy availability<br />
[%] *) Energy utilisation<br />
[%] *)<br />
Month Year Month Year Month Year<br />
OL1 Olkiluoto BWR FI 910 880 720 660 116 6 744 904 253 976 759 100.00 94.12 99.88 92.69 100.75 92.46<br />
OL2 Olkiluoto BWR FI 910 880 720 662 904 5 794 145 243 611 284 100.00 79.69 99.64 78.73 101.18 79.43<br />
KCB Borssele PWR NL 512 484 720 367 102 3 021 010 157 825 451 99.78 74.15 99.78 74.55 99.75 72.16<br />
KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0<br />
KKB 2 Beznau 7) PWR CH 380 365 720 276 072 2 646 900 130 879 056 100.00 87.20 100.00 86.71 100.93 86.16<br />
KKG Gösgen 7) PWR CH 1060 1010 720 768 486 7 788 300 304 398 935 100.00 92.37 99.99 91.99 100.69 91.66<br />
KKM Mühleberg BWR CH 390 373 720 278 340 2 838 690 124 050 935 100.00 92.24 99.85 91.61 99.12 90.80<br />
CNT-I Trillo PWR ES 1066 1003 720 764 776 7 740 744 238 234 461 100.00 91.36 99.95 91.07 99.24 90.09<br />
Dukovany B1 PWR CZ 500 473 720 362 651 2 456 677 108 267 051 100.00 62.82 100.00 62.46 100.74 61.29<br />
Dukovany B2 PWR CZ 500 473 720 360 040 2 950 413 104 273 041 100.00 75.24 100.00 74.70 100.01 73.61<br />
Dukovany B3 PWR CZ 500 473 655 314 334 2 623 607 102 248 463 90.97 75.75 86.99 65.95 87.32 65.46<br />
Dukovany B4 PWR CZ 500 473 361 174 635 2 371 933 102 900 084 50.14 69.14 48.36 59.29 48.51 59.18<br />
Temelin B1 PWR CZ 1080 1030 720 781 214 8 664 341 106 292 500 100.00 100.00 99.96 99.96 100.47 100.08<br />
Temelin B2 PWR CZ 1080 1030 720 787 897 6 819 241 100 683 563 100.00 78.34 100.00 78.01 101.32 78.77<br />
Doel 1 PWR BE 454 433 720 325 983 3 277 563 133 890 536 100.00 90.76 99.47 90.23 99.32 89.84<br />
Doel 2 PWR BE 454 433 720 328 778 3 268 119 131 921 768 100.00 91.40 99.71 91.03 100.27 89.30<br />
Doel 3 PWR BE 1056 1006 0 0 6 732 621 251 169 221 0 78.97 0 78.79 0 79.12<br />
Doel 4 PWR BE 1084 1033 720 773 286 7 054 678 253 727 128 100.00 83.08 97.81 82.31 98.06 80.49<br />
Tihange 1 PWR BE 1009 962 158 124 135 2 815 111 290 078 185 21.98 36.57 17.32 35.78 17.03 34.79<br />
Tihange 2 PWR BE 1055 1008 720 766 981 6 637 622 248 156 690 100.00 82.44 100.00 78.63 101.62 78.83<br />
Tihange 3 PWR BE 1089 1038 701 759 129 8 614 260 268 094 957 97.37 99.76 96.66 99.69 96.71 98.57<br />
Operating Results January <strong>2018</strong><br />
Plant name<br />
Type<br />
Nominal<br />
capacity<br />
gross<br />
[MW]<br />
net<br />
[MW]<br />
Operating<br />
time<br />
generator<br />
[h]<br />
Energy generated, gross<br />
[MWh]<br />
Time availability<br />
[%]<br />
Energy availability Energy utilisation<br />
[%] *) [%] *)<br />
Month Year Since Month Year Month Year Month Year<br />
commissioning<br />
KBR Brokdorf DWR 1480 1410 744 957 521 957 521 341 149 580 100.00 100.00 93.97 93.97 86.59 86.59<br />
KKE Emsland 4) DWR 1406 1335 744 1 010 637 1 010 637 336 333 920 100.00 100.00 100.00 100.00 96.55 96.55<br />
KWG Grohnde DWR 1430 1360 744 977 338 977 338 367 604 917 100.00 100.00 94.85 94.85 91.28 91.28<br />
KRB C Gundremmingen 4) SWR 1344 1288 744 982 159 982 159 321 562 051 100.00 100.00 100.00 100.00 97.58 97.58<br />
KKI-2 Isar DWR 1485 1410 744 1 082 908 1 082 908 342 681 231 100.00 100.00 99.98 99.98 97.73 97.72<br />
KKP-2 Philippsburg 1,2,4) DWR 1468 1402 744 1 062 603 1 062 603 356 230 119 100.00 100.00 99.92 99.92 96.06 96.06<br />
GKN-II Neckarwestheim DWR 1400 1310 744 1 006 200 1 006 200 321 129 334 100.00 100.00 99.40 99.39 96.80 96.80<br />
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differently and complete it faster<br />
than what was planned, says Helene<br />
Åhsberg, SKB’s project manager for<br />
the licensing process.<br />
No referendum<br />
Östhammar Municipality planned to<br />
hold a referendum on the final repository<br />
on March 4. But at a meeting in<br />
the municipal council in the end of<br />
January, it was decided to cancel the<br />
referendum.<br />
| | (18791534), www.skb.se<br />
Yucca Mountain:<br />
Can the US Finally End<br />
the $12 Billion Impasse?<br />
(nucnet) A US federal advisory panel<br />
recently took a step in what could be a<br />
lengthy process to determine if a deep<br />
geological nuclear waste repository<br />
should finally be built at Yucca Mountain,<br />
a project that has been on the<br />
drawing board since the 1970s at a<br />
cost of around $ 12 bn (€ 9.7 bn).<br />
The panel held a meeting to receive<br />
input on reconstructing an electronic<br />
library for documents needed to<br />
decide on the US Department of<br />
Energy’s Yucca licence application.<br />
The meeting, at the Nuclear Regulatory<br />
Commission’s headquarters in<br />
Maryland, came one week after<br />
another development: the White<br />
House pledged $120m of funding in<br />
its 2019 federal budget proposal to<br />
restart licensing for the Yucca site,<br />
north of Las Vegas in Nevada, and<br />
to establish an interim storage programme<br />
to address the growing<br />
stockpile of nuclear waste produced<br />
by nuclear plants across the nation.<br />
After decades of wrangling, could<br />
the US finally be on course to resolve<br />
the question of what to do with<br />
the high-level nuclear waste from<br />
the nation’s 99 commercial nuclear<br />
reactors?<br />
| | www..energy.gov<br />
US Nuclear Industry Calls<br />
for Advanced Reactor Fuel<br />
Cycle Infrastructure<br />
(nucnet) The US Nuclear Energy<br />
Institute has warned that preparations<br />
should begin now to develop a<br />
national fuel cycle infrastructure to<br />
support the operation of advanced<br />
reactors that are expected to begin<br />
deployment in the 2020s and 2030s.<br />
The Washington-based nuclear<br />
industry lobby group said interest in<br />
the development of advanced nuclear<br />
reactor designs has been increasing in<br />
recent years. Many of these designs<br />
will require uranium fuel that is<br />
enriched to a higher degree than<br />
in the current worldwide fleet of lightwater<br />
reactors. Fuel for advanced<br />
reactors, enriched in U-235 to<br />
between 5% and 20%, is called<br />
high-assay low-enriched uranium<br />
(HALEU).<br />
Some of the advanced-performance<br />
fuels being developed for use<br />
with the existing reactor fleet also will<br />
require HALEU. However, there are no<br />
US-based facilities that manufacture<br />
HALEU on a commercial scale. While<br />
small quantities of HALEU materials<br />
may be obtained on an interim basis<br />
by “blending down” existing government<br />
stocks of surplus high-enriched<br />
uranium (HEU), those HEU materials<br />
are limited in supply and not readily<br />
available, the NEI said.<br />
“Thus, for the long-term operation<br />
of advanced reactors, as well as for<br />
advanced fuels in existing reactors, a<br />
robust new infrastructure for HALEU<br />
fuel manufacture is needed.”<br />
An NEI white paper says establishing<br />
such a capability will better<br />
position the US to advance nuclear<br />
safety and non-proliferation policies<br />
around the world, while helping to<br />
ensure a robust commercial industry<br />
domestically in the decades ahead.<br />
On the other hand, “if the United<br />
States and its allies have to depend on<br />
foreign, state-owned enterprises to<br />
meet fuel needs, it will be in a much<br />
weaker position to influence these<br />
policies globally”, the paper says.<br />
| | Details online:<br />
http://bit.ly/2FnZwOF<br />
Reactors<br />
IAEA Sees Safety Commitment<br />
at Spain’s Almaraz<br />
Nuclear Power Plant<br />
(iaea) An International Atomic Energy<br />
Agency (IAEA) team of experts said<br />
the operator of Spain’s Almaraz<br />
Nuclear Power Plant demonstrated a<br />
commitment to the long-term safety of<br />
the plant and noted several good practices<br />
to share with the nuclear industry<br />
globally. The team also identified areas<br />
for further enhancement.<br />
The Operational Safety Review<br />
Team (OSART) today concluded an<br />
18-day mission to Almaraz, whose<br />
two 1,050-MWe pressurized-water<br />
reactors started commercial operation<br />
in 1983 and 1984, respectively.<br />
Centrales Nucleares Almaraz-Trillo<br />
(CNAT) operates the plant, located<br />
about 200 km southwest of Madrid.<br />
OSART missions aim to improve<br />
operational safety by objectively<br />
assessing safety performance using<br />
the IAEA’s safety standards and proposing<br />
recommendations for improvement<br />
where appropriate. Nuclear<br />
power generates more than 21 per<br />
cent of electricity in Spain, whose<br />
seven operating power reactors all<br />
began operation in the 1980s.<br />
“The team saw notable achievements<br />
made by Almaraz in recent<br />
years, such as implementing a comprehensive<br />
management system, as<br />
well as significant equipment renewal<br />
plans, to establish safety as the<br />
overriding priority at the plant,” said<br />
Team Leader Peter Tarren, Head of the<br />
IAEA’s Operational Safety Section.<br />
“We found that people at every<br />
level were willing to discuss their<br />
work and how they might learn from<br />
this OSART mission. They want to<br />
keep enhancing the safety and<br />
reliability of Almaraz.”<br />
The 14-member team comprised<br />
experts from Brazil, Bulgaria, France,<br />
Germany, Mexico, the Russian Federation,<br />
Sweden, United Arab Emirates,<br />
the United Kingdom and the United<br />
States of America, as well as three<br />
IAEA officials.<br />
The review was the 200th OSART<br />
mission conducted by the IAEA since<br />
the service was launched in 1982. It<br />
covered the areas of leadership and<br />
management for safety; training<br />
and qualification; operations; maintenance;<br />
technical support; operating<br />
experience; radiation protection;<br />
chemistry; emergency preparedness<br />
and response; accident management;<br />
human, technology and organizational<br />
interactions and long-term<br />
operation.<br />
The team identified a number of<br />
good practices that will be shared<br />
with the nuclear industry globally,<br />
including:<br />
The use of a film-forming amine<br />
compound to significantly reduce<br />
the transport of potential corrosive<br />
products to the steam generators.<br />
The use of a cross-functional<br />
indicator to show the cumulative<br />
effect of equipment status and<br />
planned activities for daily operations.<br />
The installation of a centralized<br />
vacuum system for cleaning, decontaminating<br />
and discharging liquid<br />
waste into the plant´s disposal system.<br />
The mission made a number of<br />
recommendations to improve operational<br />
safety, including:<br />
The plant should implement<br />
further actions related to management,<br />
staff and contractors to enforce<br />
standards and expectations related<br />
to industrial safety.<br />
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The plant should take measures<br />
to reinforce and implement standards<br />
to enhance the performance of reactivity<br />
manipulations in a deli berate<br />
and carefully-controlled manner.<br />
The plant should improve the<br />
support, training and documented<br />
guidance for Severe Accident Management<br />
Guideline users in order to<br />
mitigate complex severe accident<br />
scenarios.<br />
The team provided a draft report of<br />
the mission to the plant’s management.<br />
The plant management and the<br />
Nuclear Safety Council (CSN), which<br />
is responsible for nuclear safety<br />
oversight in Spain, will have the<br />
opportunity to make factual comments<br />
on the draft. These will be<br />
reviewed by the IAEA and the final report<br />
will be submitted to the<br />
Government of Spain within three<br />
months.<br />
The plant management said it<br />
would address the areas identified<br />
for enhancement and requested a<br />
follow-up OSART mission in about<br />
18 months.)<br />
| | (18791443), www.iaea.org<br />
Tianwan-3 Passes Commissioning<br />
Tests at 100% Power<br />
(nucnet) The Tianwan-3 nuclear<br />
reactor unit in Jiangsu province,<br />
northeastern China, has successfully<br />
operated for 100 hours at 100% of its<br />
design power level without interruption,<br />
Russian state nuclear corporation<br />
Rosatom said.<br />
Rosatom said the 990-MW VVER<br />
V-428M unit, which started to deliver<br />
electrical energy to the grid on<br />
30 December 2017, has undergone a<br />
series of tests during the 100-hour<br />
operation period required by regulators<br />
before giving green light for<br />
commercial operation.<br />
Construction of Tianwan-3 began<br />
in December 2012. The Tianwan<br />
| | Swiss regulator approves safety case for restart of Beznau-1 (Photo: Axpo).<br />
nuclear station is the largest economic<br />
cooperation project between Russia<br />
and China, an earlier statement had<br />
said.<br />
Tianwan-1 and -2, also VVER<br />
V-428M units, began commercial<br />
operation in 2007. The Tianwan-4<br />
VVER V-428M unit is also under construction<br />
by Russia while Tianwan-5<br />
and -6 will be indigenous Generation<br />
II+ CNP-1000 units.<br />
| | en.cnnc.com.cn<br />
Swiss Regulator Approves<br />
Safety Case for Restart of<br />
Beznau-1<br />
(nucnet) Switzerland’s Federal<br />
Nuclear Safety Inspectorate, ENSI,<br />
has given the go-ahead for the restart<br />
of the Beznau-1 nuclear unit after<br />
approving the safety case presented<br />
by owner Axpo following the discovery<br />
in 2015 of flaw indications in<br />
the reactor pressure vessel (RPV).<br />
ENSI said in a statement that<br />
Axpo had carried out “extensive<br />
investigations and analyses” to<br />
demonstrate that the RPV is safe.<br />
Materials testing has shown<br />
that agglomerates in the RPV do not<br />
affect its key properties and structural<br />
integrity analysis has shown that<br />
the RPV does not contain any flaws<br />
that could lead to its failure. “IRSN<br />
is satisfied that work has been done<br />
to all appropriate national and international<br />
standards,” the statement<br />
said.<br />
Axpo said the safety case for<br />
Beznau-1, the world’s oldest commercial<br />
nuclear plant still in operation,<br />
corroborates earlier assessments<br />
and investigations, and validates the<br />
existing safety margin for the safe<br />
operation of the plant for 60 years.<br />
Operator KKB will now begin the<br />
return to service process with the<br />
plant expected to be operating at full<br />
load by the end of March <strong>2018</strong>.<br />
In December 2015 Axpo submitted<br />
a roadmap ENSI detailing plans for<br />
further investigations of flaw indications<br />
in the RPV. During a scheduled<br />
outage that began in May 2015,<br />
inspections of the RPV registered<br />
findings at some points in the base<br />
material of the RPV indicating<br />
“ minimal irregularities in the fabrication<br />
process”, Axpo said. The company<br />
carried out further measurements<br />
and analyses and submitted a<br />
report to ENSI.<br />
In July 2015, Axpo announced<br />
that the restart of Beznau-1 had been<br />
postponed while the flaw indications<br />
were investigated further. Then in<br />
August, ENSI called for additional<br />
investigations.<br />
Beznau-2 was not affected by the<br />
flaw indications and was returned to<br />
service after its scheduled outage in<br />
2015.<br />
| | www.bkw.ch<br />
Kursk II Passed<br />
Construction Milestone<br />
(rosatom) Kursk II began reinforcing<br />
the foundation slab for the reactor<br />
building of Unit 1. This operation<br />
became the year’s key event on the<br />
construction site of the Kursk plant.<br />
On 21 December 2017, the first<br />
16-ton reinforced concrete block was<br />
installed on the rebar of the lower<br />
foundation belt. According to the<br />
project design, the foundation comprises<br />
105 reinforced concrete blocks<br />
with a total weight of 1,600 tons. This<br />
will enable the construction team<br />
to start concreting the foundation<br />
slab of the reactor building in the<br />
first half of <strong>2018</strong>.<br />
Prior to putting the first concrete<br />
block, a rebar coupler engraved with<br />
the words “The future is shaped today.<br />
The first coupling sleeve of the innovative<br />
VVER-TOI power unit” was<br />
ceremonially installed into the foundation<br />
reinforcement.<br />
VVER-TOI (which means ‘a standard<br />
optimized and automated power<br />
unit based on VVER technology’)<br />
reactors meet Russian and global<br />
safety requirements and have a longer<br />
service life and higher installed<br />
capacity than existing reactors of<br />
the Kursk Nuclear Power Plant.<br />
Alexander Mikhailov, Governor of<br />
the Kursk Region, noted that it was<br />
an honor for the region to build<br />
and commission one of the world’s<br />
first nuclear plants with advanced<br />
VVER-TOI reactors. “Construction of<br />
Kursk II designed to meet the latest<br />
global standards offers our region<br />
development prospects for the entire<br />
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<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Rosenergoatom had planned to<br />
build two BN-1200 units at Beloyarsk<br />
with commercial operation scheduled<br />
by 2025. But construction depended<br />
on the results of operating the pilot<br />
Beloyarsk-4 BN-800 plant, which<br />
began commercial operation in<br />
October 2016.<br />
There is another commercially<br />
operational sodium-cooled FBR at<br />
Beloyarsk, the BN-600. Both the<br />
BN-600 and the BN-800 are smaller<br />
versions of the BN-1200. There are<br />
also two permanently shut-down<br />
light-water reactors at the site.<br />
| | www.rosatom.ru<br />
265<br />
NEWS<br />
| | Kursk II passed construction milestone.<br />
21st century. Just a few Russian<br />
regions have such opportunities,” he<br />
stressed.<br />
Vyacheslav Fedyukin, Director<br />
of Kursk NPP, noted it was symbolic<br />
that the event happened on the<br />
25 th anniversary of RosEnergoAtom<br />
and 10 years after the foundation of<br />
Rosatom, the companies that shaped<br />
the newest history of Russia’s nuclear<br />
industry. “Construction of Russia’s<br />
first VVER-TOI-based power unit<br />
proves that the national nuclear<br />
power industry is always at the<br />
cutting edge of science and engineering.<br />
The new generation VVER-TOI<br />
units are state-of-the-art facilities<br />
made to the best of Russia’s nuclear<br />
engineering knowledge,” he added.<br />
At the moment, other operations<br />
are also underway at the construction<br />
site of Kursk II. Among them is excavation<br />
of 1.2 million cub m of soil<br />
to be completed in 2017, with over<br />
800,000 cub m of sand, gravel<br />
and aggregate already put in the<br />
foun dation of Kursk II buildings<br />
and structures. Construction of a<br />
330/10 kV substation and preparation<br />
of technical documents for its commissioning<br />
are also drawing to a<br />
close.<br />
For reference:<br />
Kursk II is designed to replace the<br />
existing Kursk Nuclear Power Plant<br />
that will be taken out of operation in<br />
the years to come. Its first two units<br />
with VVER-TOI, a new-type reactor,<br />
will be commissioned simultaneously<br />
with decommissioning of Units 1 and<br />
2 of the existing nuclear station.<br />
According to the master schedule of<br />
Kursk II, Unit 1 will be commissioned<br />
in late 2023 to be followed by Unit 2 in<br />
late 2024.<br />
| | (18791501),<br />
ww.rosatom.ru<br />
Russia Confirms Plans to<br />
Revive BN-1200 Fast Breeder<br />
Reactor Project<br />
(nucnet) Russia plans to begin construction<br />
of its first industrial-sized<br />
sodium-cooled fast neutron reactor in<br />
the 2020s after saying three years ago<br />
that the project had been postponed,<br />
the head of state nuclear corporation<br />
Rosatom Alexei Likhachev told president<br />
Vladimir Putin.<br />
According to a transcript of a<br />
meeting posted on the Kremlin’s<br />
website, Mr Likhachev told Mr Putin<br />
that fast breeder reactors (FBRs) have<br />
significant advantages over existing<br />
reactor types and Rosatom is proposing<br />
that Russia goes ahead<br />
with its plans for the Generation IV<br />
BN-1200.<br />
FBRs have been and are being<br />
explored or constructed in Russia,<br />
France, India, China, Japan and the<br />
US. They allow a significant increase<br />
in the amount of energy obtained<br />
from natural, depleted and recycled<br />
uranium. The technology also enables<br />
plutonium and other actinides to be<br />
used and recycled.<br />
Russia operates the BN-600 and<br />
BN-800 FBR units at Beloyarsk and<br />
the BOR-60 fast breeder research<br />
reactor at the Research Institute<br />
of Atomic Reactors (RIAR) site in<br />
Dimitrovgrad, southwest Russia.<br />
BOR-60 is used to test fuel cycle,<br />
sodium coolant technologies and a<br />
range of design concepts for fast<br />
breeder reactors.<br />
In 2015, Rosatom said construction<br />
of the planned BN-1200 at the<br />
Beloyarsk nuclear power station in<br />
central Russia had been postponed<br />
until at least 2020, with state<br />
nuclear operator Rosenergoatom<br />
citing the need to improve fuel<br />
for the reactor and questioning the<br />
project’s economic feasibility.<br />
Austria Begins Legal Action<br />
Against EC Over Hungary’s<br />
Paks Nuclear Project<br />
(nucnet) Austria has filed a legal<br />
complaint against the European Commission<br />
with the European Court of<br />
Justice in Luxembourg for allowing<br />
Hungary to expand its Paks nuclear<br />
power station.<br />
Austrian minister of sustainability<br />
and tourism Elisabeth Köstinger said<br />
in a statement that nuclear power<br />
“must have no place in Europe” and<br />
Austria will not “not budge one<br />
centimetre” from its anti-nuclear<br />
stance.<br />
The EC started an investigation<br />
into state aid given to the Paks 2<br />
project in November 2014. Last March<br />
it approved the project to build two<br />
new reactors, to be financed with the<br />
help of Russia’s state atomic energy<br />
corporation Rosatom, after regulators<br />
said Hungarian authorities had<br />
agreed to several measures to ensure<br />
fair competition.<br />
In January <strong>2018</strong>, Austria announced<br />
it planned to sue the EC over<br />
the decision. “EU assistance is only<br />
permissible when it is built on common<br />
interest. For us, nuclear energy is<br />
neither a sustainable form of energy<br />
supply, nor is it an answer to climate<br />
change”, a statement by the ministry<br />
of sustainability said at the time.<br />
The two planned units at Paks 2<br />
nuclear power station are expected<br />
to begin commercial operation in<br />
2026 and 2027, Attila Aszódi,<br />
the Hungarian government’s commissioner<br />
for the Paks 2 project,<br />
told a conference in Brussels late l<br />
ast year.<br />
An agreement signed in 2014<br />
will see Russia supply two VVER-<br />
1200 pressurised water reactors for<br />
Paks 2 and a loan of up to €10bn<br />
($12.3bn) to finance 80% of the<br />
€12bn project.<br />
| | www.bundeskanzleramt.gv.at<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
266<br />
NEWS<br />
Company News<br />
Framatome Completes<br />
Purchase of Schneider<br />
Electric’s Instrumentation and<br />
Control Nuclear Business<br />
(framatome) Framatome announced<br />
that it completed its purchase of<br />
Schneider Electric’s nuclear instrumentation<br />
and control offering. With<br />
this transaction, Framatome adds to<br />
its engineering expertise and expands<br />
its instrumentation and control (I&C)<br />
offerings.<br />
I&C systems are the central nervous<br />
system of a nuclear power plant,<br />
allowing operators to control reactor<br />
operations. Modernizations, upgrades<br />
and ongoing support are vital to manage<br />
economic, long-term operation of<br />
nuclear power plants, which provide<br />
reliable, low-carbon electricity.<br />
“With the integration of Schneider<br />
Electric’s nuclear instrumentation<br />
and control offering, we offer truly<br />
added value to our customers with<br />
a global technical expertise and<br />
market know-how on I&C solutions<br />
for the nuclear market,” said Bernard<br />
Fontana, Chairman of the Managing<br />
Board and Chief Executive Officer of<br />
Framatome. “We welcome our new<br />
colleagues to Framatome’s worldwide<br />
team of I&C engineers and experts.”<br />
This acquisition adds the nuclearqualified<br />
version of Tricon and the<br />
SPEC 200 platform to Framatome’s<br />
nuclear safety I&C offerings, which<br />
include the TELEPERM XS digital<br />
platform, and non-computerized<br />
analog solutions and instrumentation<br />
for nuclear power plants.<br />
This broadens the base of plants<br />
worldwide for which Framatome<br />
serves as the original equipment manufacturer<br />
for safety I&C systems. It also<br />
expands Framatome’s project and<br />
engineering capacities for non-safety<br />
I&C systems in the nuclear energy<br />
market, relying on Schneider Electric’s<br />
commercial TRICON and Foxboro<br />
platforms.<br />
Framatome also becomes the exclusive<br />
service provider to the nuclear energy<br />
market for the SPEC 200, nuclearqualified<br />
Tricon and Foxboro systems.<br />
| | www.framatome.com<br />
Framatome Continues<br />
Ramping up Production<br />
at Its Le Creusot Site<br />
(framatome) On January 25, <strong>2018</strong>,<br />
Framatome received the green light<br />
from the French Nuclear Safety<br />
Authority (ASN) and EDF to resume<br />
manufacture of forgings for the<br />
French nuclear fleet at its Le Creusot<br />
site. This decision allows the plant to<br />
continue ramping up its production<br />
with a target of 80 ingots per year.<br />
The authorization is an outcome of<br />
the improvement plan launched at the<br />
beginning of 2016 on the site following<br />
a series of quality audits. With the completion<br />
of all the actions necessary for<br />
the resumption of production for the<br />
French nuclear fleet and overall progress<br />
of 90% to date, the plan will be<br />
fully closed out in the first half of <strong>2018</strong>.<br />
The actions will then be incor porated<br />
into the site’s continuous improvement<br />
processes. Customers in France and<br />
abroad, as well as all the safety<br />
authorities concerned, have been kept<br />
regularly informed of the actions<br />
undertaken. Numerous reviews and<br />
inspections have been conducted in order<br />
to observe the progress of the plan<br />
and integrate stakeholders’ feedback.<br />
David Emond, Senior Executive<br />
Vice President of Framatome’s Component<br />
Manufacturing Business Unit,<br />
comments: “The authorization to<br />
resume manufacture of forgings for<br />
the French nuclear fleet is a very<br />
good news for the site that confirms<br />
the successful execution of its improvement<br />
plan. The 230 employees<br />
at the Le Creusot site are engaged<br />
in its deployment on a day to day basis<br />
so that we can supply our customers<br />
with equipment meeting the most<br />
stringent safety and quality requirements<br />
within agreed deadlines. I<br />
want to thank them for the substantial<br />
work they have accomplished on<br />
the site over the last two years.”<br />
Maintaining and developing the<br />
skills of the Le Creusot plant teams<br />
is a key element of the site’s improvement<br />
plan, with a particular focus<br />
on strengthening the nuclear safety<br />
culture.<br />
Framatome already invested<br />
7.5 million euros at the site in 2017<br />
to make the Le Creusot site a center<br />
of excellence for the manufacture<br />
of forgings for the nuclear industry,<br />
and will pursue this effort in <strong>2018</strong>.<br />
Major milestone reached<br />
in review of manufacturing<br />
records<br />
Moreover, a major milestone has<br />
been reached in the review of legacy<br />
manufacturing records at the Le<br />
Creusot site. The first stage in the<br />
inspection process which is being<br />
applied to all records relating to<br />
forgings produced for the nuclear<br />
industry, a key stage consisting in<br />
identifying findings, is now complete.<br />
The analysis of these findings and the<br />
processing of deviations will continue<br />
until the end of <strong>2018</strong>, in coordination<br />
with customers and safety authorities.<br />
Of the 6,000 records identified<br />
during the initial survey, 3,854 correspond<br />
to forgings installed on nuclear<br />
installations.<br />
At Framatome’s Jeumont and<br />
Saint-Marcel sites, the audit has been<br />
finalized since the summer of 2017 and<br />
no deviation impacting the safety of<br />
components has been brought to light.<br />
| | www.framatome.com<br />
JNFL and MHI Become<br />
Shareholders of<br />
Orano 2017 Revenue<br />
(orano) The Orano Board of Directors<br />
noted the completion of the capital<br />
increase reserved for Japan Nuclear<br />
Fuel Limited (JNFL) and Mitsubishi<br />
Heavy Industries, Ltd. (MHI) for<br />
a total of €500 million.<br />
Pursuant to the initial agreements<br />
signed with JNFL and MHI in<br />
March 2017, the funds corresponding<br />
to their total investment in Orano<br />
had been placed in trust on July 26,<br />
at the same time as the completion<br />
of the capital increase reserved for<br />
French State 2. These funds were<br />
released and used for the subscription<br />
of JNFL and MHI to Orano’s second<br />
capital increase.<br />
This transaction follows the<br />
completion on December 31, 2017 of<br />
the sale of the majority control of<br />
Framatome (formerly New NP) by<br />
AREVA SA to EDF as well as the<br />
fulfillment of the regulatory closing<br />
conditions related to the addition of<br />
an equity stake in Orano of both Japanese<br />
investors.<br />
Orano’s capital is now held by the<br />
French State (45.2%), the CEA<br />
(4.8%)3, AREVA SA (40%), JNFL<br />
(5%) and MHI (5%).<br />
This transaction is the last major<br />
step in the restructuring of the French<br />
nuclear industry, undertaken in 2015,<br />
and marks the end of the constitution<br />
phase of the Orano group. With a<br />
strengthened financial structure and<br />
sound strategic partnerships, Orano<br />
now has the means to grow and reach<br />
its goal of being a leading player in the<br />
production and recycling of nuclear<br />
materials, in waste management and<br />
dismantling.<br />
Appointment of a new<br />
independent director<br />
After completion of Orano’s second<br />
capital increase, the Orano General<br />
Meeting, also held on February 26,<br />
<strong>2018</strong>, appointed Patrick Pelata as<br />
independent director.<br />
| | www.orano.group<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
Westinghouse Electric<br />
Company Signs Cooperation<br />
Agreement for Lead-cooled<br />
Fast Reactor Development<br />
(westinghouse) Westinghouse Electric<br />
Company has signed a Cooperation<br />
Agreement for lead-cooled fast<br />
reactor (LFR) technology development<br />
with the Italian National Agency<br />
for New Technologies, Energy and<br />
Sustainable Economic Development<br />
(ENEA) and Ansaldo Nucleare. The<br />
agreement demonstrates each party’s<br />
commitment to collaborating toward<br />
the development of a next-generation<br />
nuclear plant based on LFR technology,<br />
which is both “walk-away”<br />
safe and economically competitive<br />
across global energy markets.<br />
“This agreement is an exciting<br />
step towards the development of a<br />
lead-cooled fast reactor for the<br />
marketplace,” said Ken Canavan,<br />
Westinghouse chief technology officer<br />
and vice president, Global Technology<br />
Office. “The LFR is game-changing<br />
technology for clean energy industries,<br />
and Westinghouse is pleased to<br />
be working with such experienced<br />
partners to bring this innovative<br />
concept to fruition.”<br />
Beyond baseload electricity<br />
generation, the high-temperature<br />
operation of the LFR will allow for<br />
a broad range of applications such<br />
as an effective load-following<br />
capability enabled by an innovative<br />
thermal energy storage system,<br />
delivery of process heat for industrial<br />
applications and water desalination.<br />
ENEA is a world leader in research<br />
and development on lead-based<br />
systems, and currently operates<br />
among the finest and largest experimental<br />
facilities for LFR research in<br />
the world.<br />
Ansaldo Nucleare has vast experience<br />
in nuclear power plant design,<br />
supply, service and decommissioning,<br />
and has played leading roles in<br />
multiple international LFR development<br />
programs for the past 15 years.<br />
| | www.westinghousenuclear.com<br />
BKW übernimmt Experten<br />
für Strahlenschutz<br />
(bkw) Die BKW Konzerngesellschaft<br />
Dienstleistungen für Nukleartechnik<br />
(DfN) übernimmt das ebenfalls<br />
auf den kerntechnischen Bereich<br />
spezia lisierte Unternehmen Technischer<br />
Strahlenschutz (TSS). Dadurch<br />
stärkt die BKW ihre Kompetenzen in<br />
diesem Gebiet und baut sie weiter aus.<br />
Dies vor dem Hintergrund der geplanten<br />
Stilllegung des Kernkraftwerks<br />
Mühleberg und zahlreicher weiterer<br />
Kernkraftwerke in Europa.<br />
Mit der Übernahme des Strahlenschutzunternehmens<br />
DfN hat die BKW<br />
bereits im letzten Jahr ihre bestehenden<br />
und bewährten Kom petenzen im hochspezialisierten<br />
Nukleartechnik-Bereich<br />
erweitert. Der Eintritt der TSS in den<br />
Unter nehmensverbund der BKW stellt<br />
nun einen weiteren Ausbau in diesem<br />
Gebiet dar. Die TSS ergänzt die<br />
Strahlenschutzkompetenzen innerhalb<br />
der BKW Gruppe und verstärkt diese<br />
auch im Hinblick auf die Still legung<br />
des Kernkraftwerks Mühleberg.<br />
In Europa ist ausserdem eine Vielzahl<br />
weiterer Stilllegungsprojekte in<br />
Planung oder bereits im Gang. Der<br />
Strahlenschutz spielt bereits beim<br />
Betrieb von Kernkraftwerken eine<br />
wichtige Rolle. Mit der Stilllegung<br />
und den dabei ausgeführten Demontage-<br />
und Freimessarbeiten nehmen<br />
die Strahlenschutzarbeiten zu. Für<br />
Strahlenschutzdienstleisterinnen wie<br />
TSS und DfN bietet der wachsende<br />
Stilllegungsmarkt daher ein grosses<br />
Potenzial und die Möglichkeit, sich<br />
weiterzuentwickeln.<br />
Die DfN und die TSS haben bereits<br />
verschiedentlich auf Projektbasis<br />
zusammengearbeitet. Die erfolgreiche<br />
Kooperation wird künftig<br />
weiter ausgebaut, was mit einer<br />
gegenseitigen Stärkung einhergeht.<br />
Um eine optimale Zusammenarbeit<br />
zu ermöglichen, wird die TSS in die<br />
DfN integriert.<br />
Die TSS mit Sitz in Geilenkirchen<br />
im deutschen Bundesland Nordrhein-<br />
Westfalen wurde 1979 gegründet<br />
und zählt 15 Mitarbeitende. Das<br />
Unternehmen bietet ein qualitativ<br />
hochwertiges und breites Angebot von<br />
Dienstleistungen im kerntechnischen<br />
Bereich. Dazu gehören neben dem<br />
Strahlenschutz die Dekontamination,<br />
die Abfallentsorgung, die Dosimetrie<br />
sowie die Abwicklung von Transporten<br />
radioaktiver Stoffe.<br />
| | (18791521), www.bkw.ch<br />
Companies<br />
China Approves $ 100 Billion<br />
Merger of Leading<br />
Nuclear Companies<br />
(nucnet) China has approved the<br />
merger of nuclear power producer<br />
China National Nuclear Corporation<br />
(CNNC) with nuclear plant builder<br />
China Nuclear Engineering and<br />
Construction Corporation (CNECC),<br />
the state-run China Daily news agency<br />
said.<br />
According to the China Daily, the<br />
combined assets of the new company<br />
will be worth about $100bn (€80bn),<br />
while its workforce will be about<br />
150,000 employees.<br />
CNNC is China’s number two<br />
nuclear power producer and CNECC<br />
the country’s top nuclear power plant<br />
builder.<br />
China Daily said the merger is in<br />
line with efforts by China to streamline<br />
the state-operated sector of its<br />
economy and reduce the number of<br />
state-owned companies administered<br />
by central government.<br />
Approval for the merger was<br />
confirmed by the State-Owned Assets<br />
Supervision and Administration<br />
Commission (SASAC) in a one-line<br />
statement posted on its website.<br />
| | (18800822), en.cnnc.com.cn<br />
267<br />
NEWS<br />
Research<br />
| | BKW übernimmt Experten für Strahlenschutz © BKW.<br />
NRG: Every Day,<br />
30,000 Patients Benefit From<br />
Medical Isotopes From Petten<br />
(nrg) Medical isotopes are indispensable<br />
for diagnosing and treating<br />
cancer. Demand for them is set to soar<br />
over the next 20 years, but supplies<br />
are diminishing. To put the spotlight<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
• Separative work: 58.00–92.00<br />
268<br />
NEWS<br />
2016<br />
January to June 2016<br />
• Uranium: 26.50–35.25<br />
• Conversion: 6.25–6.75<br />
• Separative work: 58.00–62.00<br />
July to December 2016<br />
• Uranium: 18.75–27.80<br />
• Conversion: 5.50–6.50<br />
• Separative work: 47.00–62.00<br />
| | NRG: Every day, 30000 patients benefit from medical isotopes from Petten View of the pool type reactor<br />
core. (Courtesy: JRC)<br />
on the world of medical isotopes, the<br />
platform 30000perdag.nl has been<br />
launched. The aim of the platform and<br />
the accompanying campaign is to<br />
boost awareness that the Netherlands<br />
must continue leading the field in<br />
cancer treatment.<br />
The future<br />
Over the next 20 years, the number of<br />
cancer diagnoses is expected to rise by<br />
70%. Fortunately, health care is<br />
constantly improving, partly through<br />
the use of medical isotopes. However,<br />
there are only 6 reactors worldwide<br />
which can produce medical isotopes,<br />
one of which is closing next year.<br />
This means that whilst demand for<br />
medical isotopes is growing worldwide,<br />
supplies are diminishing.<br />
30000perdag.nl<br />
An online information park for a<br />
wide audience has been built on<br />
30000perdag.nl. Visitors can learn all<br />
about medical isotopes here: from raw<br />
materials to the reactor in Petten to<br />
applications in the hospital. By opening<br />
up that world, NRG in Petten<br />
wants to show (former) cancer<br />
patients and their families and<br />
acquaintances what is needed to be<br />
able to treat cancer, and request<br />
support for medical isotopes and good<br />
cancer treatment in the Netherlands<br />
and abroad.<br />
Former cancer patients play<br />
starring role in campaign<br />
The campaign uses 3 video interviews<br />
with cancer survivors. The interviews<br />
were conducted by presenter Fien Vermeulen,<br />
herself a former lymphoma<br />
patient. Fien drives with former<br />
patients Anouk (26), Alexander (42)<br />
and Manon (34) to the research<br />
reactor in Petten, where they talk<br />
about their remarkable experiences in<br />
times of uncertainty. Each of them<br />
represents one of the 30,000 people<br />
who benefit or have benefitted from<br />
medical isotopes every day.<br />
Anyone can demonstrate their<br />
support by liking the Facebook page<br />
30.000perdag. Another very visible<br />
form of support is available through<br />
the T-shirts that can be ordered via<br />
30000perdag.nl. These enable former<br />
patients and supporters to show their<br />
backing for the campaign.<br />
| | (18800822), www.nrg.eu<br />
Market data<br />
(All information is supplied without<br />
guarantee.)<br />
Nuclear Fuel Supply<br />
Market Data<br />
Information in current (nominal)<br />
U.S.-$. No inflation adjustment of<br />
prices on a base year. Separative work<br />
data for the formerly “secondary<br />
market”. Uranium prices [US-$/lb<br />
U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =<br />
0.385 kg U]. Conversion prices<br />
[US-$/kg U], Separative work<br />
[US-$/SWU (Separative work unit)].<br />
January to December 2013<br />
• Uranium: 34.00–43.50<br />
• Conversion: 9.25–11.50<br />
• Separative work: 98.00–127.00<br />
January to December 2014<br />
• Uranium: 28.10–42.00<br />
• Conversion: 7.25–11.00<br />
• Separative work: 86.00–98.00<br />
January to December 2015<br />
• Uranium: 35.00–39.75<br />
• Conversion: 6.25–9.50<br />
2017<br />
January 2017<br />
• Uranium: 20.25–25.50<br />
• Conversion: 5.50–6.75<br />
• Separative work: 47.00–50.00<br />
February 2017<br />
• Uranium: 23.50–26.50<br />
• Conversion: 5.50–6.75<br />
• Separative work: 48.00–50.00<br />
March 2017<br />
• Uranium: 24.00–26.00<br />
• Conversion: 5.50–6.75<br />
• Separative work: 47.00–50.00<br />
April 2017<br />
• Uranium: 22.50–23.50<br />
• Conversion: 5.00–5.50<br />
• Separative work: 45.50–48.50<br />
May 2017<br />
• Uranium: 19.25–22.75<br />
• Conversion: 5.00–5.50<br />
• Separative work: 42.00–45.00<br />
June 2017<br />
• Uranium: 19.25–20.50<br />
• Conversion: 5.55–5.50<br />
• Separative work: 42.00–43.00<br />
July 2017<br />
• Uranium: 19.75–20.50<br />
• Conversion: 4.75–5.25<br />
• Separative work: 42.00–43.00<br />
August 2017<br />
• Uranium: 19.50–21.00<br />
• Conversion: 4.75–5.25<br />
• Separative work: 41.00–43.00<br />
September 2017<br />
• Uranium: 19.75–20.75<br />
• Conversion: 4.60–5.10<br />
• Separative work: 40.50–42.00<br />
October 2017<br />
• Uranium: 19.90–20.50<br />
• Conversion: 4.50–5.25<br />
• Separative work: 40.00–43.00<br />
November 2017<br />
• Uranium: 20.00–26.00<br />
• Conversion: 4.75–5.25<br />
• Separative work: 40.00–43.00<br />
December 2017<br />
• Uranium: 23.50–25.50<br />
• Conversion: 5.00–6.00<br />
• Separative work: 39.00–42.00<br />
<strong>2018</strong><br />
January <strong>2018</strong><br />
• Uranium: 21.75–24.00<br />
• Conversion: 6.00–7.00<br />
• Separative work: 38.00–42.00<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
February <strong>2018</strong><br />
• Uranium: 21.25–22.50<br />
• Conversion: 6.25–7.25<br />
• Separative work: 37.00–40.00<br />
| | Source: Energy Intelligence<br />
www.energyintel.com<br />
Cross-border Price<br />
for Hard Coal<br />
Cross-border price for hard coal in<br />
[€/t TCE] and orders in [t TCE] for<br />
use in power plants (TCE: tonnes of<br />
coal equivalent, German border):<br />
2012: 93.02; 27,453,635<br />
2013: 79.12, 31,637,166<br />
2014: 72.94, 30,591,663<br />
2015: 67.90; 28,919,230<br />
2016: 67.07; 29,787,178<br />
I. quarter: 56.87; 8,627,347<br />
II. quarter: 56.12; 5,970,240<br />
III. quarter: 65.03, 7.257.041<br />
IV. quarter: 88.28; 7,932,550<br />
| | Uranium spot market prices from 1980 to <strong>2018</strong> and from 2008 to <strong>2018</strong>. The price range is shown.<br />
In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.<br />
269<br />
NEWS<br />
2017:<br />
I. quarter: 95.75; 8,385,071<br />
II. quarter: 86.40; 5,094,233<br />
III. quarter: 88.07; 5,504,908<br />
| | Source: BAFA,<br />
some data provisional<br />
www.bafa.de<br />
EEX Trading Results<br />
February <strong>2018</strong><br />
(eex) In February <strong>2018</strong>, the European<br />
Energy Exchange (EEX) achieved a<br />
total volume of 274.3 TWh on its<br />
power derivatives markets (February<br />
2017: 200.8 TWh) which is a yearon-year<br />
increase of 37 %. In doing<br />
so, EEX was able to grow its power<br />
derivatives volumes across all market<br />
areas.<br />
In total, the German and Austrian<br />
markets (Phelix-DE, Phelix-AT and<br />
Phelix-DE/AT) increased by 12 % to<br />
169.7 TWh. This includes 153.4 TWh<br />
from the benchmark product Phelix-<br />
DE which achieved its highest volume<br />
since launch in April 2017. Volumes<br />
in the French market more than<br />
doubled to 26.9 TWh (February 2017:<br />
12.3 TWh) while Italian power<br />
volumes grew substantially to<br />
40.0 TWh (February 2017: 22.1 TWh).<br />
Furthermore, on the Spanish market,<br />
volumes increased by more than<br />
250 % to 8.3 TWh (February 2017:<br />
2.3 TWh).<br />
The February volume comprised<br />
173.5 TWh traded at EEX via Trade<br />
Registration with subsequent clearing.<br />
Clearing and settlement of all exchange<br />
transactions was executed by European<br />
Commodity Clearing (ECC).<br />
The Settlement Price for base<br />
load contract (Phelix Futures) with<br />
| | Separative work and conversion market price ranges from 2008 to <strong>2018</strong>. The price range is shown.<br />
)1<br />
In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.<br />
delivery in 2019 amounted to 33.85 €/<br />
MWh. The Settlement Price for peak<br />
load contract (Phelix Futures) with<br />
delivery in 2019 amounted to 42.40 €/<br />
MWh.<br />
On the EEX markets for emission<br />
allowances, the total trading volume<br />
increased by 57% to 144.2 million<br />
tonnes of CO 2 in February (February<br />
2017: 91.7, million tonnes of CO 2 ).<br />
Primary market auctions contributed<br />
75.1 million tonnes of CO 2 to the total<br />
volume. On the spot secondary<br />
market, volumes more than doubled<br />
to 4.5 million tonnes of CO 2 (February<br />
2017: 2.0 million tonnes of CO 2 ). On<br />
the EUA Futures market, EEX was able<br />
to increase volumes by 80% to<br />
37.3 million tonnes of CO 2 (February<br />
2017: 20.7 million tonnes of CO 2 ).<br />
Furthermore, 27.4 million tonnes of<br />
CO 2 were traded in EUA Options<br />
which is the highest monthly volume<br />
so far in this product.<br />
The EUA price with delivery in<br />
December <strong>2018</strong> amounted to<br />
8.80/10.15 €/ EUA (min./max.).<br />
| | www.eex.com<br />
MWV Crude Oil/Product Prices<br />
January 2017<br />
(mwv) According to information and<br />
calculations by the Association of the<br />
German Petroleum Industry MWV e.V.<br />
in January <strong>2018</strong> the prices for<br />
super fuel, fuel oil and heating oil<br />
noted inconsistent compared with<br />
the pre vious month December 2017.<br />
The average gas station prices for Euro<br />
super consisted of 136.84 €Cent<br />
( December 2017: 136.84 €Cent,<br />
approx. +-0.0 % in brackets: each<br />
information for pre vious month or<br />
rather previous month comparison),<br />
for diesel fuel of 120.48 €Cent<br />
(119.01; +1.24 %) and for heating oil<br />
(HEL) of 62.27 €Cent (60.65 €Cent,<br />
+2.67 %).<br />
The tax share for super with<br />
a consumer price of 136.84 €Cent<br />
(136.84 €Cent) consisted of<br />
65.45 €Cent (47.83 %, 65.45 €Cent)<br />
for the current constant mineral oil<br />
tax share and 21.85 €Cent (current<br />
rate: 19.0 % = const., 22.12 €Cent)<br />
for the value added tax. The product<br />
price (notation Rotterdam) consisted<br />
of 40.17 €Cent (29.36 %, 37.18 €Cent)<br />
and the gross margin consisted of<br />
9.37 €Cent (6.85 %; 12.36 €Cent).<br />
Thus the overall tax share for super<br />
results of 66.83 % (66.83 %).<br />
Worldwide crude oil prices<br />
(monthly average price OPEC/Brent/<br />
WTI, Source: U.S. EIA) were again<br />
significantly higher, approx. +8.36 %<br />
(+2.34 %) in January <strong>2018</strong> compared<br />
to December 2017.<br />
The market showed a stable<br />
development with higher prices; each<br />
in US-$/bbl: OPEC basket: 66.85<br />
(62.06); UK-Brent: 69.08 (64.37);<br />
West Texas Inter mediate (WTI): 63.7<br />
(57.88).<br />
| | www.mwv.de<br />
News
<strong>atw</strong> Vol. 63 (<strong>2018</strong>) | Issue 4 ı April<br />
270<br />
Czechs and Balances and Why ‘Ugly’<br />
Nuclear Deserves a Political Makeover<br />
NUCLEAR TODAY<br />
Author<br />
John Shepherd<br />
Shepherd<br />
Communications<br />
3 Brooklands<br />
West Sussex<br />
BN43 5FE<br />
Links to reference<br />
sources:<br />
Dana Drábová<br />
interview:<br />
http://bit.ly/2Ik7WaJ<br />
European Investment<br />
Bank announcement:<br />
http://bit.ly/2Ik7WaJ<br />
Yonhap News<br />
Agency report:<br />
http://bit.ly/2FyvZkw<br />
As if Europe does not have enough on its plate to deal with at the moment – politically and economically just for starters<br />
– could Brussels be on a collision course with the Czech government over the country's plans to expand nuclear energy?<br />
There is certainly friction over the issue between Prague and<br />
the European Commission (EC), to put it mildly. But why?<br />
The veteran head of the Czech Republic’s State Office<br />
for Nuclear Safety, Dana Drábová, last month accused<br />
other EU member states of “pressurising” Prague over the<br />
early closure of its oldest nuclear reactor units.<br />
Drábová reportedly told an energy conference in the<br />
country: “There is immense pressure developing that the<br />
operating life of nuclear reactors will be limited to 40 years.<br />
That means that our political representatives, whoever they<br />
might be, sometime around 2023 will face a battle over a<br />
further 10-year extension for Dukovany. The current State<br />
Energy Framework counts on the lifetime of the Dukovany<br />
reactors ending sometime between 2030 and 2040.”<br />
The nuclear safety chief later told Czech Radio the<br />
pressure was coming from “the 14 countries which are not<br />
using nuclear power and some of which regard it as<br />
something ugly”. If the pressure continued, she predicted<br />
there would be a concerted “willingness… to get rid of<br />
these nuclear plants in Europe as fast as possible”.<br />
Drábová’s comments came against a backdrop of the<br />
Czech government saying it would appoint an expert team<br />
to consider proposals to break up the majority state-owned<br />
electricity firm CEZ. The move was one of several options<br />
mooted to support financing of the construction of a new<br />
nuclear power plant at Dukovany.<br />
Analysts say the new nuclear plant could be built by the<br />
traditional energy unit, which would be fully state-owned<br />
and therefore in the best position to take on the risks of<br />
high costs that the utility could not if it were an entity with<br />
private owners.<br />
Czech prime minister Andrej Babiš is backing proposals<br />
to build the Dukovany reactor, around 50 kilometres north<br />
of the (anti nuclear) Austrian border, to replace a Soviet-era<br />
reactor. But this would mean persuading the EC to exempt<br />
the project from strict EU rules on government bids.<br />
If the Czech government fails in its quest, it could consider<br />
doing a deal with Russia, which would undoubtedly<br />
be very much along the lines of the nuclear construction<br />
and financing deal Moscow signed recently with EU<br />
member Hungary.<br />
If, dear reader, you now have a sense of déjà vu, you<br />
would be right. You may recall that Hungary went through<br />
a similar nuclear battle with the EC, despite Hungary’s parliament<br />
fully backing proposals to build two new nuclear<br />
reactor units in that country.<br />
Initially, the EC said in November 2015 it had started<br />
legal action against Hungary over a contract signed with<br />
Russia’s Rosatom to build two units at the existing Paks<br />
plant. Brussels expressed concern about the project’s<br />
compatibility with EU public procurement rules. However,<br />
the EC eventually cleared the issue and a state aid investigation<br />
into the project financing for the ‘Paks II’ project<br />
was subsequently dropped by the EC.<br />
There was a similar clash with the EC when the UK first<br />
unveiled plans to invest in building the Hinkley Point C<br />
nuclear plant.<br />
So is the latest tussle between Prague and Brussels<br />
really over concerns about state-aid rules or is it more a<br />
worrying trend of interference to stop nuclear in its tracks?<br />
And is the conflict really worth it…?<br />
Czech PM Babiš said following an official visit to<br />
Hungary last January, where he attended a summit of<br />
prime ministers of the Visegrad Group countries, that he<br />
and Hungarian counterpart Viktor Orbán discussed the<br />
potential for “further developing” cooperation in sectors<br />
such as the nuclear energy industry.<br />
But far more intriguing was what Babiš claimed was the<br />
attitude of Visegrad leaders about relations with the<br />
institutions of the EU. According to a statement issued by<br />
the Czech government, Babiš said the leaders agreed it was<br />
“necessary to depoliticise Brussels and the EC”. Apparently,<br />
the leaders believe that when it comes to EU affairs,<br />
“ member states, prime ministers and presidents, should<br />
have the main say”, according to Babiš.<br />
If there is behind-the-scenes pressure to stamp out<br />
nuclear wherever it might try to cling on or prosper in the<br />
EU, where is that effort coming from and why? Of course,<br />
it is no secret that Austria and Germany strongly oppose<br />
any expansion of nuclear power in Europe. Having lived<br />
and worked in Germany, I never understood that<br />
politically- inspired decision – but as a guest in the country<br />
for which I have great admiration I respect its decision.<br />
Austria’s approach has always puzzled me more – being<br />
willing as it is to host the International Atomic Energy<br />
Agency (IAEA) and enjoy all the ‘fruits’ that that privilege<br />
brings, not least in the economic benefit of having the<br />
agency based in Vienna.<br />
But back to the Czech project. As a possible fight with<br />
the EC shapes up, it is not only Moscow that is set to benefit<br />
from yet another new nuclear power order from an EU<br />
nation.<br />
South Korea is also reportedly circling – keen to tempt<br />
Prague to consider its nuclear technology, according to<br />
Seoul’s Yonhap News Agency.<br />
Can the EU really afford such a quarrel – again – with a<br />
member state over nuclear? And why should European<br />
skills, knowhow and investment not be channelled into the<br />
Czech nuclear project?<br />
I am struck by the EC’s approach to another industrial<br />
sector and how contrasting it is. The EC is currently working<br />
at full tilt to develop a European battery cell industry,<br />
with the goal of ensuring the EU is not overwhelmed by<br />
competition from Asian battery makers for products such as<br />
electric vehicles and energy storage devices.<br />
The EU’s vice-president for the energy union, Maroš<br />
Šefčovič, said in February “there are many extremely<br />
interesting actions that we need to pursue, including (a)<br />
simplification of approval procedures and permitting<br />
processes in the EU”. Indeed the European Investment Bank<br />
has already approved a loan for the construction and<br />
operation of what it said will be a first-of-a-kind demonstration<br />
plant in Sweden, for the manufacturing of lithium- ion<br />
batteries.<br />
The EC’s support for the development of such technology<br />
across EU member states is of course admirable, but one<br />
hears nothing of state-aid rules and complications here!<br />
Why is it that nuclear cannot win such favourable attention<br />
and support? Does it really have to be this way – and<br />
hasn’t the EC learned anything from the UK’s Brexit vote<br />
about treading carefully in issues that are seen by member<br />
states of national importance?<br />
Nuclear Today<br />
Czechs and Balances and Why ‘Ugly’ Nuclear Deserves a Political Makeover ı Jubair Ahmed Shamim and Kune Yull Suh
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