nucmag.com
2018
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217
Heat Transfer Systems
for Novel Nuclear
Power Plant Designs
221 ı Operation and New Build
Safety Research for GEN IV Reactors
226 ı Operation and New Build
Numerical Analysis for the MYRRHA Project
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atw Vol. 63 (2018) | Issue 4 ı April
Security of Supply ...
and the Clock is Ticking ...
Deal reader, More than one hundred years ago, around 1890, a conflict flared up between the two well-known
protagonists of electricity supply, Thomas Alva Edison and George Westinghouse, on the large-scale power supply and the
construction of power grids in the United States of America. While Edison technically preferred D.C. voltage, Westinghouse
counted instead on alternating voltage. In the end it was not a matter of the most suitable technique but of the anticipated
market shares of each company General Electric or Westinghouse Electric and the patents behind. At a breath taking pace,
the most important developments for the use of electricity were preceding: In the year 1866 Werner Von Siemens
discovered the dynamo- electrical principle, which enabled larger performance. The development of alternating voltage
in the year 1881 enabled generally technically and cost-effectively the transportation of electricity over long distances
– we are talking back then about distances of some ten kilometres. Alternating voltage enforced itself at that time due to
possible further transportation length enabled through higher trans mission voltage.
207
EDITORIAL
Both current types have something in common: generation
and use need to take place simultaneously. The grid fails if
both do not fit together. Neither alternating current grids
nor direct current grids offer storage possibilities. Thus, a
stable power system also requires a stable and reliable
generation, because if a larger system “fails”, the system
restoration is, from its task and process, a large-scale
project.
Different believes e.g. from politics or other interest
groups are simply wrong, power systems are – without any
further active establishments and plants- no accumulators.
A reliable power supplying system needs at any time
reliable generation. “Surpluses”, meaning potentials for a
higher generation than demand, when so ever, cannot be
shifted or stored “electrically” in the system at a later time.
It was not an inconspicuous message, which appeared
multiple times in the press at the end of February,
beginning of March 2018. Headlines such as “Time
synchronisation per power system: Energy shortages make
watches lose time”, described a phenomenon, of which,
according to the media “one became aware of – only
( editor’s note) - after weeks”: What happened?
As an indicator for the stability of alternating power
systems stand supply voltage a well as system frequency.
For the system frequency applies that she needs to be
identical at any point of the system. If generation and
consumption do not fit, deviations occur, leaking generation
leads among others to a perceived frequency decrease
among the entire connected system. As the system
frequency is defined for our alternating electricity net with
constant 50 Hertz, it is also qualified for watches, which
use the frequency as direct clock indicator.
We can for example – due to cost reasons – renounce to
a frequency stabilising quartz oscillator. Nevertheless, this
technical simplification is bought with failures in time, if
the frequency deviates from the standard over a longer
period. Only a few hundred Hertz is enough for days and
weeks in order to, as in the current case, generate a time
deviation of minus 360 seconds, 6 minutes, and those
inside the entire affected system of 25 West, Middle- and
South European countries.
The cause for this incident was later communicated by
the European Network of Transmission System Operators for
Electricity (short ENTSO-E) and the Swiss net operator
swissgrid, that in the control zone Serbia, Macedonia,
Montenegro (the so called SMM rule block), especially in
Kosovo and Serbia less energy was fed into the system. A
deficit of 113 gigawatt hours was shown, not much, in view
of a European daily production of around 8,000 gigawatt
hours. But especially this shows how delicate our power
system is and how sensitive it reacts to the smallest
malfunctions.
Reliable measures in power generation – meaning
currently only for conventional techniques, thus need,
with all considerations on the reconstruction of electricity
supply, to be reconsidered. Additionally and almost
simultaneously another alarming “availability message”
came in: At the beginning of March 2018 European gas
storage tanks were only filled with a quantity of 26.2 per
cent, Germany even on average only with 23.8 per cent.
Thus, according to an EU-conform proceeding an early
warning level was reached, because the filling level of
storage tanks may not be lower than around 20 % due to
reasons of guaranteeing mechanical stability. On top came
the message that more natural gas was imported to Europa
than in the previous years. All first hints, that there might
not be enough natural gas in Europe for dispose filling in as
a “reserve”?
In all, these are all important references that any,
especially neither direct market- nor technically driven,
interventions – where compensation factors can con tribute
– need to be well thought in our power system. Furthermore,
does the availability of a broad basis of conventional
generation not only gain more importance, she is even
more important than it is conceded for “conventionals”
vision wise in many places in terms of an „energy
transition“. To what extend “the clock” might tick on
possible severe supply shortfalls or even large-scale loss of
off-site power… one does not know…
Christopher Weßelmann
– Editor in Chief –
Editorial
Security of Supply ... and the Clock is Ticking ...
atw Vol. 63 (2018) | Issue 4 ı April
EDITORIAL 208
Versorgungssicherheit und die Uhr tickt ...
Liebe Leserin, lieber Leser, vor mehr als hundert Jahren, um 1890, entbrannte eine Auseinandersetzung
zwischen den beiden bekannten Protagonisten der Elektrizitätsversorgung, Thomas Alva Edison und George Westinghouse,
zur weiträumigen Versorgung der Vereinigten Staaten von Amerika mit Strom und dem Aufbau geeigneter Stromnetze.
Während Edison technisch die Gleichspannung favorisierte, setze Westinghouse die Wechselspannung dagegen.
Letztendlich ging es aber nicht wesentlich um die Frage der geeigneteren Technik, sondern um die avisierten Marktanteile
der jeweiligen Unternehmen General Electric bzw. Westinghouse Electric und die dahinter stehenden Patente. Vorangegangen
waren in atemberaubendem Tempo die wichtigsten Entwicklungen für die Nutzung der Elektrizität: Im Jahr
1866 entdeckte Werner von Siemens das dynamoelektrische Prinzip, das größere Leistungen ermöglichte. Die
Entwicklung des Wechselstromtransformators im Jahr 1881 ermöglichte technisch grundsätzlich und kostengünstiger
den Transport von Strom über längere Strecken – wir sprechen hier zu jener Zeit über Strecken im Bereich von einigen
zehn Kilometern. Durchgesetzt hatte sich aufgrund der durch höhere Übertragungsspannungen möglichen weiteren
Transportlängen zu jener Zeit die Wechselspannung.
Beiden Stromarten ist eines gemeinsam: Erzeugung und
Nutzung müssen exakt zeitgleich erfolgen. Sind Erzeugung
und Gebrauch nicht im Einklang, bricht das Netz
zusammen. Weder Wechsel- noch Gleichspannungsnetz
bieten „Speichermöglichkeiten“. Für ein stabiles Stromnetz
ist daher auch eine stabile und verlässliche Erzeugung
erforderlich, denn wenn einmal ein größeres Stromnetz
„zusammenbricht“, ist der Netzwiederaufbau ein von der
Aufgabe und dem zeitlichen Ablauf her aufwendiges
Vorhaben. Anderslautende Stimmen z.B. aus der Politik
oder von Interessengruppen sind schlichtweg falsch,
Strom netze sind – ohne weitere aktive Einrichtungen und
Anlagen – keine Speicher. Ein verlässliches Stromversorgungsnetz
benötigt eine jederzeit verlässliche Erzeugung.
„Überschüsse“, also Potenziale für eine höhere Erzeugung
als die vorhandene Nachfrage, wann und warum auch
immer, lassen sich „elektrisch“ im Netz nicht auf spätere
Zeiten verschieben, also speichern.
Es war eine nicht unscheinbare Nachricht, die Ende
Februar, Anfang März 2018 mehrfach durch die
Presse ging. Überschriften wie „Zeit-Synchronisation per
Stromnetz: Energieknappheit lässt Uhren nachgehen“,
beschrieben ein Phänomen, dessen man sich nach Angaben
in der Presse „nach Wochen – erst (Anm. der Red.) –
bewusst wurde“: Was war geschehen?
Für die Stabilität bzw. als Indikator für die Stabilität
von Wechselstromnetzen stehen die Netzspannung sowie
die Netzfrequenz. Für die Netzfrequenz gilt dabei, dass
diese an jedem Punkt in einem Netz identisch ist. Stimmen
Erzeugung und Verbrauch nicht überein, kommt es zu
Abweichungen, fehlende Erzeugung führt u.a. zu einer
im gesamten angebundenen Netz fühlbaren Frequenzabnahme.
Da die Netzfrequenz für unser Wechselstromnetz
mit konstant 50 Hertz vereinbart ist, eignet sich diese
auch für Uhren, die die Frequenz als direkten Taktgeber
nutzen. Auf z.B. einen frequenzstabilisierenden Quarzoszillator
kann – aus Kostengründen – verzichtet werden.
Diese technische Vereinfachung erkauft man sich allerdings
mit Fehlern in der Uhrzeit, wenn die Frequenz über
einen längeren Zeitraum vom Standard abweicht. Schon
wenige hundertstel Hertz reichen über Tage und Wochen
aus, um, wie im aktuellen Fall, eine kumulierte Zeitabweichung
von Minus 360 Sekunden, also 6 Minuten,
hervorzurufen; und dies im ganzen betroffenen Netz von
25 West-, mittel- und südosteuropäischen Ländern.
Als Ursache für dieses Ereignis wurde später vom
Verband Europäischer Übertragungsnetzbetreiber (kurz
ENTSO-E, European Network of Transmission System
Operators for Electricity) und dem Schweizer Netzbetreiber
swissgrid kommuniziert, dass in der Kontrollzone Serbien,
Mazedonien, Montenegro (dem sogenannten SMM Regelblock),
insbesondere in Kosovo und Serbien zu wenig
Energie ins Netz eingespeist wurde. Ein Fehlbetrag von
113 Gigawattstunden wurde ausgewiesen, nicht viel,
angesichts einer europaweiten Tagesproduktion von rund
8.000 Gigawattstunden. Aber gerade diese zeigt, wie
filigran unser Stromnetz ist und wie empfindlich es doch
auf kleinste Störungen reagiert.
Verlässliche Größen in der Stromerzeugung, sprich
derzeit letztendlich nur die konventionellen Techniken,
müssten von daher in allen Überlegungen zum Umbau der
Stromversorgung neu überdacht werden. Hinzu kam fast
zeitgleich eine weitere bedenkliche energiewirtschaftliche
„Verfügbarkeitsmeldung“: Europas Gasspeicher waren zu
Anfang März 2018 nur noch zu 26,2 Prozent gefüllt,
Deutschland gar im Schnitt nur zu 23,8 Prozent. Damit
war nach einem EU-einheitlichen Verfahren eine Frühwarnstufe
erreicht, denn die Speicher dürfen ihren Füllgrad
aus Gründen der Gewährleistung ihrer mechanischen
Stabilität nicht unter rund 20 % absenken. Hinzu kam die
Mitteilung, dass mehr Erdgas nach Europa importiert
wurde, als in den Vorjahren. Alles erste Hinweise darauf,
dass vielleicht in Zukunft doch nicht ausreichend Erdgas
in Europa zur Verfügung stehen wird, um als „Reserve“
einzuspringen?
In Summe sind dies alles wichtige Hinweise darauf,
dass jegliche, vor allem weder direkt markt- noch technisch
getriebenen Eingriffe – wo ausgleichende Faktoren wirken
können – in unser Stromversorgungssystem wohl überdacht
sein müssen. Zudem gewinnt die Verfügbarkeit einer
breiten Basis konventioneller Erzeugung damit nicht nur
an Bedeutung, sie ist bedeutungsvoller als vielerorts in
Visionen einer „Energiewende“ den Konventionellen
zugestanden wird.
Inwieweit „die Uhr“ möglicher schwerwiegender Versorgungsengpässe
oder gar großflächiger Netzausfälle
tickt ... man weis es nicht ...
Christopher Weßelmann
– Chefredakteur –
Editorial
Security of Supply... and the Clock is Ticking ...
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atw Vol. 63 (2018) | Issue 4 ı April
210
Issue 4
April
CONTENTS
217
Heat Transfer Systems
for Novel Nuclear
Power Plant Designs
| | The Swiss nuclear power plants generate up to 40 % of the country’s electricity production. At the Beznau site, two pressurised water
reactors are in operation with a gross capacity of 380 MW each and a net capacity of 365 MW. Switzerland’s Federal Nuclear Safety
Inspectorate, ENSI, gave the go-ahead for the restart of Beznau-1 after approving the safety case presented by operator Axpo following
the discovery in 2015 of flaw indications in the reactor pressure vessel. (Courtesy: Axpo)
Editorial
Security of Supply ... and the Clock is Ticking ... . . 207
Versorgungssicherheit und die Uhr tickt ... . . . . 208
Abstracts | English . . . . . . . . . . . . . . . . . . . 212
Abstracts | German . . . . . . . . . . . . . . . . . . . 213
Inside Nuclear with NucNet
Euratom: Industry Softens Stance
as Government Lays Out Plans for Transition . . . 214
NucNet
Calendar . . . . . . . . . . . . . . . . . . . . . . . 216
Operation and New Build
Heat Transfer Systems for Novel
Nuclear Power Plant Designs . . . . . . . . . . . . . 217
Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
Experimental and Analytical Tools
for Safety Research of GEN IV Reactors . . . . . . . 221
G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
DAtF Notes. . . . . . . . . . . . . . . . . . . . . .215
221
| | Centrum Výzkumu Řež facilities list.
217
Numerical Analysis of MYRRHA
Inter- wrapper Flow Experiment at KALLA . . . . . 226
| | Koeberg PWR steam generator and simulation model.
Abdalla Batta and Andreas G. Class
Contents
atw Vol. 63 (2018) | Issue 4 ı April
226
CONTENTS
211
| | Velocity magnitude within bundle showing flow distribution.
Heat Balance Analysis for
Energy Conversion Systems of VHTR . . . . . . . . 230
SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon
and Soyoung Park
Spotlight on Nuclear Law
Information Requirements Versus
Confidentiality Obligations – Extension of
the In-Camera Procedure Planned . . . . . . . . . . 235
Informationsbedarf versus
Geheimhaltungspflichten – Erweiterung
des In camera-Verfahrens geplant . . . . . . . . . . 235
Tobias Leidinger
Environment and Safety
CFD Modeling and Simulation of Heat and Mass
Transfer in Passive Heat Removal Systems . . . . . 238
Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
|241
249
| | Collecting soil samples.
Research and Innovation
Irradiation Tests of a Flat Vanadium Self- Powered
Detector with 14 MeV Neutrons . . . . . . . . . . . 246
Prasoon Raj and Axel Klix
Nanofluid Applied Thermo-hydro dynamic
Performance Analysis of Square Array
Subchannel Under PWR Condition. . . . . . . . . . 249
Jubair Ahmed Shamim and Kune Yull Suh
| Computational domain created in Star-CCM+.
KTG Inside . . . . . . . . . . . . . . . . . . . . . . 257
238
| | Liquid Volume fraction distribution.
Decommissioning and Waste Management
The Decommissioning of the ENEA RB3
Research Reactor in Montecuccolino . . . . . . . . 241
F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi,
R. Lorenzelli and A. Rizzo
News . . . . . . . . . . . . . . . . . . . . . . . . . 260
Nuclear Today
Czechs and Balances and Why ‘Ugly’
Nuclear Deserves a Political Makeover . . . . . . . 270
Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 236
AiNT. . . . . . . . . . . . . . . . . . . . . . . . . . . .Insert
AMNT 2018: Registration Form . . . . . . . . . . . Insert
Contents
atw Vol. 63 (2018) | Issue 4 ı April
212
ABSTRACTS | ENGLISH
Euratom: Industry Softens Stance as
Government Lays Out Plans for Transition
NucNet | Page 214
The UK’s nuclear industry has welcomed a government
commitment to continuity with existing
arrangements with Euratom, Europe’s nuclear safety
and research watchdog, a softening of its earlier
stance that the UK needed to stay in the group to
protect vital nuclear research and new-build projects,
and to make sure access to nuclear fuel and
medical isotopes is not disrupted. The next phase of
discussions will focus on the UK’s future relationship
with Euratom. Specific objectives include a close
association with the Euratom Research and Training
Programme, including the Joint European Torus
(JET) and the International Thermonuclear Experimental
Reactor (ITER) projects.
Heat Transfer Systems for Novel
Nuclear Power Plant Designs
Sebastian Vlach, Christoph Fischer and
Herman van Antwerpen | Page 217
This article focuses on designing or modifying heat
exchangers found in the auxiliary systems of any
power plant. The basic premise is to show that the
software provides a one-stop solution for designing
many types of heat transfer systems, where the
interaction bet ween various loops connected by
heat exchangers can be assessed. The nuclear power
plant industry is addressed as the quality control in
the development of the software makes it most
suitable for nuclear related applications. Moreover,
the software discussed has the capability to do
contaminant tracing, which could be very useful
for nuclear contamination studies in designing
specialized ventilation systems. To highlight the
versatility of the software network approach it will
be shown how to model any setup and kind of heat
exchanger such as plate, tube-in-tube, liquid/gas,
finned tube etc. Additionally, the Koeberg pressurized
water reactor steam generator comparison and
the THTR steam generator comparison are shown
as examples.
Experimental and Analytical Tools for
Safety Research of GEN IV Reactors
G. Mazzini, M. Kyncl, Alis Musa and
M. Ruscak | Page221
Current research on nuclear safety in the world, in
addition to supporting existing nuclear power
plants is focused on the more detailed aspects of the
new reactors. The new generation reactors are
expected inter alia to use innovative types of fuel
and new types of coolants, such as e.g. Super-
Critical Water (SCW), supercritical CO 2 , liquid
metals, fluoride salts or high-temperature Helium.
The paper will describe new experimental infrastructure
build recently in Research Centre Řež
under the SUSEN (Sustainable Energy) project and
available analytical tools for supporting safety
research of GEN IV reactors. Two experimental
loops – SCWL (Supercritical Water Loop) and HTHL
(High Temperature Helium Loop) will serve as
in-pile loops in the active core of the research
reactor LVR-15. The paper provides examples of
analyses made using codes ATHLET (supercritical
water) and TRACE (high temperature He) illustrating
process of their assessment and practical use.
Numerical Analysis of MYRRHA Inter-wrapper
Flow Experiment at KALLA
Abdalla Batta and Andreas G. Class | Page 226
The MYRRHA reactor, which is developed at
SCK-SCN in Belgium, represents a multi-purpose
irradiation facility. Its prominent feature is a pool
design with the nuclear core submerged in liquid
metal lead bismuth. During transients between
normal operation and accident conditions decay
heat removal is ensured by forced and natural
convection, respectively. The flow in the gap
between the fuel assemblies plays an important role
in limiting maximum temperatures which should
not be exceeded to avoid core damage. Due to the
scarce database, within the Horizon 2020 – research
and innovation framework program of the EU, the
SESAME project was established to develop and
validate advanced numerical approaches, to
achieve a new or extended validation base and to
establish best practice guidelines including verification
& validation and uncertainty quantification.
In particular the current work supports the
inter-wrapper flow experiment at KALLA.
Heat Balance Analysis for Energy
Conversion Systems of VHTR
SangIL Lee, YeonJae Yoo, Deok Hoon Kye,
Gyunyoung Heo, Eojin Jeon
and Soyoung Park | Page 230
VHTR(Very High Temperature Gas Reactor) with
helium used as a coolant can easily produce heat
required in high-temperature thermochemical process,
and because of low heat output density, the
possibility of core melting is low. In this study, provided
that VHTR is located in the primary system,
the heat conversion system will be discussed in
which hydrogen production and power supply are
possible. In order to control the ratio between power
and hydrogen production, the helium flowing
through nuclear reactor is made to pass through
heat exchanger for hydrogen production and steam
generator or heat exchanger. This study proposes
the whole heat conversion system model, and
carries out thermodynamic feasibility calculation
according to major design variable at each point
and sensitivity analysis for efficiency optimization.
Information Requirements Versus
Confidentiality Obligations – Extension
of the In-Camera Procedure Planned
Tobias Leidinger | Page 235
The justified right of the public to detailed information
on a project requiring nuclear licensing is
opposed by the state’s interest in effective protection
of sensitive data. This conflict is manifested
in licensing procedures but also at court. The differentiated
legal provisions that regulate the balancing
of these conflicting interests are now to be supplemented
by a further facet: An expanded in-camera
trial at court. According to the coalition agreement
of 7 February 2018, the regulation is to take place in
the current 18th legislative period.
CFD Modeling and Simulation of Heat
and Mass Transfer in Passive Heat
Removal Systems
Amirhosein Moonesi, Shabestary,
Eckhard Krepper and Dirk Lucas | Page 238
The CFD-modelling and simulation of condensation
inside passive heat removal systems are presented.
Designs of future nuclear boiling water reactor concepts
are equipped with emergency cooling systems
which are passive systems for heat removal. The
emergency cooling system consists of slightly
inclined horizontal pipes which are immersed in a
tank of subcooled water. The focus of the project is
on detection of different morphologies such as
annular flow, stratified flow, slug flow and plug flow
and also modeling of the laminar film which is
occurring during the condensation near the wall.
The Decommissioning of the ENEA RB3
Research Reactor in Montecuccolino
F. Rocchi, C. M. Castellani, A. Compagno,
I. Vilardi, R. Lorenzelli and A. Rizzo | Page 241
The ENEA RB3 reactor was a 100 Wth research
installation owned and operated by ENEA, in its
center of Montecuccolino near Bologna, from 1971
to 1989. In 1989, the RB3 reactor was shut down,
and in the late 2010 ENEA received by ministerial
decree the authorization to its dismantling, with the
aim of reaching the “green field” status. This paper
presents the three main pillars of the decommissioning
of RB3, namely the strategy and methods
for the dismantling, the strategy and methods for
the radiological characterization of the building,
and finally the strategy and methods for the radiological
characterization of the site.
Irradiation Tests of a Flat Vanadium
Self-Powered Detector with
14 MeV Neutrons
Prasoon Raj and Axel Klix | Page 246
Self-powered detector (SPD) represents a class of
neutron and gamma monitoring instruments used
in the fission reactor cores worldwide. This detector
has inherent advantages of functioning without a
bias voltage, simple measurement scheme, compactness,
ease of maintenance, and high reliability.
We are studying SPD for application as flux monitors
in the European test blanket modules (TBM) of
ITER, fusion reactor under construction in southern
France.
Nanofluid Applied Thermo-hydrodynamic
Performance Analysis of Square Array
Subchannel Under PWR Condition
Jubair Ahmed Shamim and Kune Yull Suh | Page 249
Efficient engineered design of heat transfer and
fluid flow with enhanced heating or cooling requires
two pivotal aspects that must be taken into consideration
for extracting thermal energy from
nuclear fission reactions in order to save energy,
reduce process time, raise thermal rating and
increase the operating life of a reactor pressure
vessel. Hence, one of the major challenges in
designing a new nuclear power plant is the quantification
of the optimal flow of coolant and distribution
of pressure drop across the reactor core.
Recently, nanofluid has gained much renewed
attention as a promising coolant for pressurized
water reactors (PWRs) due to its enhanced thermal
capabilities with least penalty in pressure drop.
Czechs and Balances and Why ‘Ugly’
Nuclear Deserves a Political Makeover
John Shepherd | Page 270
As if Europe does not have enough on its plate
to deal with at the moment – politically and
economically just for starters – could Brussels be on
a collision course with the Czech government over
the countries plans to expand nuclear energy?
There is certainly friction over the issue between
Prague and the European Commission (EC), to put
it mildly. But why?
The veteran head of the Czech Republic’s State
Office for Nuclear Safety, Dana Drábová, last month
accused other EU member states of “pressurising”
Prague over the early closure of its oldest nuclear
reactor units.
Abstracts | English
atw Vol. 63 (2018) | Issue 4 ı April
Euratom: Britische Industrie zufrieden
mit Übergangsplänen der Regierung
NucNet | Seite 214
Die britische Nuklearindustrie hat die Zusage der
Regierung begrüßt, die bestehenden Vereinbarungen
mit Euratom, dem europäischen Rahmen
für nukleare Sicherheit und Forschung, aufrechtzuerhalten
und ihren früheren Standpunkt, dass
das Vereinigte Königreich im Euratom-Vertrag
verbleiben müsse, um wichtige Forschungs- und
Neubauprojekte sicherzustellen, und den Zugang
zu Kernbrennstoffen und medizinischen Isotopen
zu gewährleisten, zu relativieren. Die nächste Phase
der Gespräche im Rahmen des Brexit wird sich auf
die künftigen Beziehungen des Vereinigten Königreichs
zu Euratom konzentrieren. Zu den spezifischen
Zielen gehört eine enge Zusammenarbeit
mit den Euratom-Forschungs- und Ausbildungsprogrammen,
einschließlich der Projekte Joint
European Torus (JET) und International Thermonuclear
Experimental Reactor (ITER).
Fortgeschrittene Wärmeübertragungssysteme
für zukünftige Kernkraftwerkskonzepte
Sebastian Vlach, Christoph Fischer und
Herman van Antwerpen | Seite 217
CFD-Systemsimulation mit FlownexSE ermöglicht
es Ingenieuren, einfache und komplexe strömungstechnische
und thermische Netzwerke schnell und
effizient aufzubauen und zu analysieren. Die
Simulation ermöglicht es Ingenieuren, Systeme
aufzubauen, effizient auszulegen und bereits frühzeitig
Schwachstellen in Entwürfen zu finden
sowie geeignete Änderungen und Maßnahmen zu
entwickeln und im Netzwerkmodell zu testen.
Besondere Aufmerksamkeit wird in diesem Artikel
den vielseitigen Möglichkeiten gewidmet, einfache
und komplexe Wärmetauschersysteme der verschiedensten
Arten (Plattenwärmetauscher, Rohrbündel
etc.) für moderne Kernkraftwerke anzuwenden. Als
praktische Beispiele werden gemessene Daten von
den Kraftwerken Koeberg und Hamm-Uentrop mit
den Ergebnissen aus der Simulation verglichen.
Experimentelle und analytische
Werkzeuge für die Sicherheitsforschung
zu GEN-IV-Reaktoren
G. Mazzini, M. Kyncl, Alis Musa und
M. Ruscak | Seite 221
Die aktuelle Forschung zur Sicherheit von
Kernkraftwerken konzentriert sich neben den
Aktivitäten für bestehende Kernkraftwerke auf die
detaillierteren Aspekte neuer Reaktorkonzepte.
Hier werden u.a. innovative Brennstoffe und Kühlmittel
wie z.B. überkritisches Wasser, überkritisches
CO 2 , Flüssigmetalle, Salzschmelzen oder Helium
eingesetzt. Vorgestellt wird dazu die neue experimentelle
Infrastruktur, die im Forschungszentrum
Řež im Rahmen des SUSEN-Projekts (Sustainable
Energy) aufgebaut wurde, sowie die verfügbaren
Analyseinstrumente zur Unterstützung der Sicherheitsforschung
zu GEN IV-Reaktoren.
Numerische Analyse der Zwischenspaltströmung
im MYRRHA-Reaktor mit Ergebnissen
des Strömungsexperiment KALLA
Abdalla Batta und Andreas G. Class | Seite 226
Der am SCK-SCN in Belgien entwickelte MYRRHA-
Reaktor ist eine Mehrzweck-Bestrahlungsanlage.
Sein herausragendes Merkmal ist eine Reaktorkonstruktion
mit einer Kernkühlung aus flüssigem Blei-
Wismut. Bei Transienten zwischen Normalbetrieb
und Unfallbedingungen wird die Wärmeabfuhr
durch erzwungene bzw. natürliche Konvektion
sichergestellt. Die Strömung im Spalt zwischen den
Brennelementen spielt eine wichtige Rolle bei der
Begrenzung von Maximaltemperaturen, die zur
Vermeidung von Kernschäden nicht überschritten
werden sollten. Im Rahmenprogramm Horizon
2020 – Forschung und Innovation der EU wurde
dazu das Projekt SESAME initiiert, um fortgeschrittene
numerische Ansätze zu entwickeln und
zu validieren, die eine neue oder erweiterte
Validierungsbasis für damit verbundene Fragestellungen
zur Verfügung stellen.
Wärmebilanzanalyse für
Energieumwandlungssysteme von VHTR
SangIL Lee, YeonJae Yoo, Deok Hoon Kye,
Gyunyoung Heo, Eojin Jeon und
Soyoung Park | Seite 230
VHTR (Very High Temperature Gas Reactor) mit
Helium als Kühlmittel können Wärme bereit stellen,
die bei thermochemischen Hochtemperaturprozessen
benötigt wird. In Bezug auf die Sicherheit ist
aufgrund der geringen Wärmeleistungsdichte das
Risiko einer Kernschmelze minimiert. Diskutiert
werden Voraussetzungen für die Nutzung von
VHTR für eine Wasserstofferzeugung und Stromversorgung.
Vorgestellt wird ein Gesamtmodell des
Wärmeumwandlungssystems mit einer thermodynamischen
Machbarkeitsberechnung.
Informationsbedarf versus
Geheimhaltungspflichten – Erweiterung
des In camera-Verfahrens geplant
Tobias Leidinger | Seite 235
Dem berechtigten Anspruch der Öffentlichkeit auf
detaillierte Informationen über ein atomrechtlich
genehmigungsbedürftiges Vorhaben steht das
staatliche Interesse an einem effektiven Geheimnisschutz
sensibler Daten gegenüber. Dieser Konflikt
tritt regelmäßig im Genehmigungsverfahren aber
auch vor Gericht zu Tage. Die differenzierten
Gesetzesbestimmungen, die den Ausgleich dieser
widerstreitenden Interessen regeln, sollen nun
durch eine weitere Facette ergänzt werden: Ein
erweitertes In-camera-Verfahren bei Gericht. Nach
dem Koalitionsvertrag vom 7. Februar 2018 soll die
Regelung in der schon laufenden 18. Legislaturperiode
erfolgen.
CFD-Modellierung und Simulation
von Wärme- und Stoffaustausch
in passiven Wärmeabfuhrsystemen
Amirhosein Moonesi, Shabestary,
Eckhard Krepper und Dirk Lucas | Seite 238
Die CFD-Modellierung und Simulation der Kondensation
in passiven Wärmeabfuhrsystemen wird vorgestellt.
Zukünftige Siedewasserreaktorkonzepte
werden mit Notkühlsystemen ausgestattet, die eine
passive Wärmeabfuhr gewährleisten. Das Notkühlsystem
besteht aus leicht geneigten horizontalen
Rohren in einem Wasserbehälter. Der Schwerpunkt
des vorgestellten Projektes liegt auf der Identifikation
verschiedener Morphologien wie Ringströmung,
Schichtenströmung, Schwallströmung
und Pfropfenströmung sowie der Modellierung des
laminaren Films, der bei der Kondensation in
Wandnähe auftritt.
Die Stilllegung der ENEA RB3
Forschungsreaktor in Montecuccolino
F. Rocchi, C. M. Castellani, A. Compagno,
I. Vilardi, R. Lorenzelli und A. Rizzo | Seite 241
Der ENEA RB3-Reaktor war eine 100-Watt-Forschungsanlage,
die von 1971 bis 1989 im Zentrum
von Montecuccolino bei Bologna, Italien betrieben
wurde. 1989 wurde der RB3-Reaktor abgeschaltet
und Ende 2010 erhielt ENEA per Ministerialerlass
die Genehmigung zu seinem Rückbau mit dem Ziel,
den Status „Grünen Wiese“ zu erreichen. Vorgestellt
werden die drei wesentlichen Fragestellungen für
die Stilllegung des RB3: Strategie und Methoden
für den Rückbau, Strategie und Methoden für die
radiologische Charakterisierung des Gebäudes und
schließlich die Strategie und Methoden für die
radiologische Charakterisierung des Standortes.
Bestrahlungstests eines Vanadium-
Detektors mit 14 MeV Neutronen
Prasoon Raj und Axel Klix| | Seite 246
Self-powered Detektoren (SPD) sind eine Klasse
von Neutronen- und Gamma-Überwachungsgeräten,
die weltweit in Kernreaktoren eingesetzt
werden. Diese Detektoren besitzen die Vorteile,
dass keine Spannungsversorgung erforderlich ist,
das Messverfahren einfach und die Detektoreinheit
kompakt, wartungsfreundlich und zuverlässig ist.
SPDs werden im Rahmen des vorgestellten Projektes
für den Einsatz als Flussmonitor in den Blanketmodulen
des in Bau befindlichen Fusionsreaktors
ITER .
Einsatz von Nanofluiden und
thermohydraulische Analyse
für Druckwasserreaktoren
Jubair Ahmed Shamim und Kune Yull Suh | Seite 249
Eine effiziente Auslegung von Wärmeübertragung
und Flüssigkeitsströmung mit verbessertem
Wärme übergang, -transport oder Kühlung bedingt
zwei zentrale Aspekte, die in Kernkraftwerken
berücksichtigt werden müssen: Leistungsdichte
und technische Lebensdauer des Reaktordruckbehälters.
Eine Herausforderung für die Auslegung
neuer Kernkraftwerkskonzepte ist daher die
Quantifizierung einer optimalen Kühlmittelverteilung
und die Verteilung des Druckverlustes über
den Reaktorkern. In jüngster Zeit werden „Nanofluide“
als vielversprechendes Kühlmittel für Druckwasserreaktoren
(DWR) aufgrund verbesserter
thermischer Eigenschaften mit geringst möglichem
Druckabfall diskutiert, die auch Thema dieser
Arbeit sind..
Tschechien und Ausgewogenheit und
warum es die „hässliche“ Kernenergie verdient,
politisch neu bewertet zu werden
John Shepherd | Seite 270
Als ob Europa derzeit nicht genug zu tun hätte, mit
sich selbst – politisch und wirtschaftlich, nur um
zwei Themenbereiche zu nennen – ... könnte jetzt
Brüssel auf Kollisionskurs mit der tschechischen
Regierung zu den Plänen des Landes zum Ausbau
der Kernenergie gehen?
In der Frage zwischen Prag und der Europäischen
Kommission (EC) geht es, um es milde auszudrücken,
sicherlich um Differenzen. Aber warum?
Die langjährige Leiterin der tschechischen
Aufsichtsbehörde für nukleare Sicherheit, Dana
Drábová, warf zudem im vergangenen Monat
anderen EU-Mitgliedstaaten vor, die Regierung in
Prag unter inakzeptablem Druck zu setzen hinsichtlich
der Forderung einer vorzeitigen Stilllegung
ihrer ältesten Kernkraftwerke.
213
ABSTRACTS | GERMAN
Abstracts | German
atw Vol. 63 (2018) | Issue 4 ı April
214
INSIDE NUCLEAR WITH NUCNET
Euratom: Industry Softens Stance as
Government Lays Out Plans for Transition
NucNet
The UK’s nuclear industry has welcomed a government commitment to continuity with existing
arrangements with Euratom, Europe’s nuclear safety and research watchdog, a softening of its earlier stance
that the UK needed to stay in the group to protect vital nuclear research and new-build projects, and to make
sure access to nuclear fuel and medical isotopes is not disrupted.
Energy secretary Greg Clark said in a written statement to
parliament on 11 January 2018 that the government wants
to include Euratom in any implementation period agreed
as part of wider discussions on Brexit and plans to put in
place “all the necessary measures” to ensure that the UK
can operate as an independent and responsible nuclear
state from day one of Brexit and its separation from the
Euratom Treaty, which regulates the nuclear industry and
the movement of nuclear material across Europe.
According to Mr Clark’s statement, the government has
made good progress on separation issues in the last few
months as part of phase one of negotiations with the EU.
Negotiations have covered a set of legal and technical
issues related to nuclear material and waste, and safeguards
obligations and equipment.
The next phase of discussions will focus on the UK’s
future relationship with Euratom. Specific objectives
include a close association with the Euratom Research and
Training Programme, including the Joint European Torus
(JET) and the International Thermonuclear Experimental
Reactor (ITER) projects.
For the nuclear industry, rapid departure from Euratom
without a clear replacement spells disaster. Scientists have
warned that British nuclear stations may not be able to
source nuclear fuel if it cannot be legally transported
across borders. The shipment of medical isotopes used in
scans and cancer treatment might be jeopardised.
European workers on shared research projects, such as
experimental fusion reactors, face an equally uncertain
future without Euratom’s separate guarantees of freedom
of movement.
But the London-based Nuclear Industry Association
(NIA), which represents more than 260 nuclear companies,
cautiously welcomed Mr Clark’s statement, calling it a
“useful and welcome step” in setting out the government’s
approach in seeking to secure equivalent arrangements to
those the UK benefits from as a member of Euratom. The
NIA also welcomed clarity on the government’s intention
to negotiate an implementation period to ensure a smooth
transition from the current to new arrangements.
It warned, however, that there is much still to do in
equipping the UK’s regulator to take on Euratom’s safeguarding
activities. The UK needs to reach post-Euratom
agreements with the International Atomic Energy Agency,
the US, Canada, Australia, Japan and others. It needs to
agree new trading arrangements with the Euratom
community and conclude a new funding agreement for the
UK to continue its work in Euratom’s fusion R&D activities.
“It is vital government continues to prioritise these issues
in the period ahead if there is to be a successful outcome,”
the NIA said.
Unlike the dozens of other regulatory arrangements for
industries such as aviation or pharmaceuticals, Euratom
has been singled out for special treatment through the
Brexit process because it is not technically part of the EU.
Instead, the treaty that established this body to coordinate
Europe’s civil nuclear energy industry was born in parallel
with the birth of the European economic community in
1957. The UK’s participation in Euratom therefore required
a separate legal relationship with the European court of
justice to enforce it.
The nuclear industry had been hoping that because of
this separation from the “mainstream Brexit,” the UK
might decide to remain part of Euratom.
The NIA and the Brussels-based trade body Foratom
both said the UK should maintain its membership. They
argued that the nuclear industry is global, and the ease of
movement of nuclear goods, people and services enables
new build, decommissioning, R&D and other programmes
of work to continue without interruption.
The government insists that leaving Euratom is an
inevitable consequence of Brexit – a position shared by the
European negotiators. But is says it wants continuity of
open trade arrangements for nuclear goods and products
to ensure the nuclear industry is able to continue to trade
across EU borders without disruption.
Support for remaining in Euratom had come not only
from within the industry, but also from politicians.
Conservative MPs said they would for the government to
fight harder for the UK to stay in Euratom. The opposition
Labour Party said Britain should remain in Euratom,
adding it is increasingly clear that the government acted
“recklessly” by giving up on membership.
Scientists said leaving Euratom will cause widespread
confusion and have a potentially devastating impact
on the nuclear industry. They warned of potential problems
related to the transportation of nuclear materials, including
nuclear fuel; research, especially fusion research; and
overseas investment in development of British nuclear
power stations.
Mr Clark’s statement addressed another concern for the
industry – the issue of accessing a skilled pan-European
workforce for the sector once Brexit is complete.
Mr Clark said the nuclear sector needs the workforce for
decommissioning, operation of existing facilities and
new-build projects. He said proposals for the UK’s future
immigration system will be set out shortly and “we will
ensure that those businesses and communities, and
parliament have the opportunity to contribute their views
before making any decisions about the future system”.
Whatever the outcome of negotiations with the EU,
it is vital that the civil nuclear industry has a safeguards
regime that meets international standards. But this
is not dependent on the EU negotiations and the UK
government is well advanced in delivering this plan, the
statement said.
Inside Nuclear with NucNet
Euratom: Industry Softens Stance as Government Lays Out Plans for Transition ı NucNet
atw Vol. 63 (2018) | Issue 4 ı April
Advertisement
The UK is establishing a legislative and regulatory
framework for a domestic safeguards regime which will
provide legal powers to establish a domestic regime which
the Office for Nuclear Regulation will regulate. It is also
negotiating bilateral safeguards agreements with the
International Atomic Energy Agency and putting in place
bilateral nuclear cooperation agreements with key third
countries.
NIA chief executive Tom Greatrex said the UK industry
and research facilities have been consistently clear with
government about the importance of these issues since
the referendum. “Even with a suitable transition, there
remains much work for the government to do to prevent
the significant disruption that industry is concerned
about.”
Mr Clark’s statement is online:
http://bit.ly/2CQ1wwQ
Fachseminar Nuklearhaftung
Haftung und Deckung im Nuklearbereich
q Das Haftungssystem des Pariser Übereinkommens
und des Atomgesetzes
q Besitzerhaftung nach § 26 AtG
q Deckungsvorsorge und Versicherung
q Aktuelle Aspekte, z.B. Haftung für Anlagen
in Stilllegung und Rückbau sowie Stand
der Umsetzung des PÜ-Änderungsprotokolls
Zielgruppe
q Projektleiter, Führungskräfte, Fachleute in den
Bereichen Versicherung, Vertrag, Genehmigung,
Projekte, Juristen wie Nichtjuristen
Dozenten
2 Dr. Christian Raetzke ı Rechtsanwalt
2 Achim Jansen-Tersteegen ı Geschäftsführer,
Deutsche Kernreaktor-Versicherungsgemeinschaft
DATF EDITORIAL NOTES
215
Author
NucNet
The Independent Global Nuclear News Agency
David Dalton
Editor in Chief, NucNet
Avenue des Arts 56
1000 Brussels, Belgium
www.nucnet.org
Am 5. September 2018 in Leipzig
Ï Information und Anmeldung: www.conlar.de
Rechtsanwaltskanzlei Dr. Christian Raetzke
Beethovenstraße 19 · 04107 Leipzig
Tel. 0341 – 9999 1444
christian.raetzke@conlar.de · www.conlar.de
Notes
Conlar atw 18-04 75x124.indd 1 18.03.18 15:02
Grafik des Monats
Bundesministerium für Umwelt, Naturschutz, Bau und Reaktorsicherheit (BMUB)
Fachaufsicht
Beteiligungsverwaltung
Fachaufsicht
Bundesaufsicht
Zusammenarbeit im
Länderausschuss für
Atomkernenergie (LAA)
Organigramm der Behördenstruktur
im Rahmen des
Standortauswahlverfahrens
für das Endlager für
hochradioaktive Abfälle.
Bundesamt
für kerntechnische
Entsorgungssicherheit
(BfE)
Bundesgesellschaft
für Endlagerung (BGE)
Bundesamt
für Strahlenschutz (BfS)
Landesministerien
Regulierung
von Endlagern
Planfeststellung
und Genehmigung
von Endlagern
Aufsicht von Endlagern
Regulierung
Private Rechtsform –
100% öffentliche Hand
Nicht an öffentliche
Haushalte gebunden
Vorhabenträger:
• Standortsuche
• Bau
• Betrieb
• Stilllegung
von Endlagern
Wissenschaftliche
Bundesbehörde
für Aspekte
des Strahlenschutzes
Atomrechtliche
Vollzugsaufgaben
Bergrechtliche
Zulassungen und
wasserrechtliche
Erlaubnisse im
Benehmen mit BfE
| | Quelle: DAtF in Anlehnung
an Endlagerkommission
For further details
please contact:
Nicolas Wendler
DAtF
Robert-Koch-Platz 4
10115 Berlin
Germany
E-mail: presse@
kernenergie.de
www.kernenergie.de
DAtF Notes
atw Vol. 63 (2018) | Issue 4 ı April
216
CALENDAR
Calendar
2018
08.04.-11.04.2018
International Congress on Advances in Nuclear
Power Plants – ICAPP 18. Charlotte, NC, USA,
American Nuclear Society (ANS), www.ans.org
08.04.-13.04.2018
11 th International Conference on Methods and
Applications of Radioanalytical Chemistry –
MARC XI. Kailua-Kona, HI, USA, American Nuclear
Society (ANS), www.ans.org
12.04.2018
Desalination Powered by Nuclear Energy. Essen,
Germany, Deutsche Meerwasser Entsalzung GmbH
in cooperation with International Atomic Energy
Agency (IAEA) and PowerTech Training Center
( Kraftwerksschule, KWS), www.dme-gmbh.de,
www.iaea.org, www.kraftwerksschule.de
16.04.-19.04.2018
Einführung in die Kerntechnik. Mannheim,
Germany, TÜV SÜD, nucleartraining@tuev-sued.de
16.04.-17.04.2018
VdTÜV Forum Kerntechnik – Sicherheit im Fokus.
Berlin, Germany, VdTÜV mit Unterstützung des
TÜV NORD, des TÜV SÜD und des TÜV Rheinland,
www.tuev-sued.de/tagungen
17.04.-19.04.2018
World Nuclear Fuel Cycle 2018. Madrid, Spain,
World Nuclear Association (WNA),
www.world-nuclear.org
18.04.-19.04.2018
9. Symposium zur Endlagerung radioaktiver
Abfälle. Vorbereitung auf KONRAD – Wege zum
G2-Gebinde. Hanover, Germany, TÜV NORD
Akademie, www.tuev-nord.de/tk-era
22.04.-26.04.2018
Reactor Physics Paving the Way Towards More
Efficient Systems – PHYSOR 2018. Cancun, Mexico,
www.physor2018.mx
24.04.-25.04.2018
Integrated Waste Management Conference.
Penrith, Cumbria, United Kingdom, The Nuclear
Institute, www.iwmeurope.com
08.05.-10.05.2018
29 th Conference of the Nuclear Societies in Israel.
Herzliya, Israel. Israel Nuclear Society and Israel
Society for Radiation Protection, ins-conference.com
13.05.-19.05.2018
BEPU-2018 – ANS International Conference on
Best-Estimate Plus Uncertainties Methods. Lucca,
Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,
ANS, IAEA, NEA, www.nineeng.com/bepu/
13.05.-18.05.2018
RadChem 2018 – 18th Radiochemical Conference.
Marianske Lazne, Czech Republic,
www.radchem.cz
14.05.-16.05.2018
ATOMEXPO 2018. Sochi, Russia,
atomexpo.ru
15.05.-17.05.2018
11 th International Conference on the Transport,
Storage, and Disposal of Radioactive Materials.
London, United Kingdom, Nuclear Institute,
www.nuclearinst.com
20.05.-23.05.2018
5 th Asian and Oceanic IRPA Regional Congress
on Radiation Protection – AOCRP5. Melbourne,
Australia, Australian Radiation Protection Society
(ARPS) and International Radiation Protection
Association (IRPA), www.aocrp-5.org
29.05.-30.05.2018
49 th Annual Meeting on Nuclear Technology
AMNT 2018 | 49. Jahrestagung Kerntechnik.
Berlin, Germany, DAtF and KTG,
www.nucleartech-meeting.com
03.06.-07.06.2018
38 th CNS Annual Conference and 42nd CNS-CNA
Student Conference. Saskotoon, SK, Canada,
Candian Nuclear Society CNS, www.cns-snc.ca
03.06.-06.06.2018
HND2018 12 th International Conference of the
Croatian Nuclear Society. Zadar, Croatia, Croatian
Nuclear Society, www.nuklearno-drustvo.hr
04.06.-05.06.2018
13 th European Nuclear Energy Forum. Bratislava,
Slova Republic, European Commission, ec.europa.eu
04.06.-07.06.2018
10 th Symposium on CBRNE Threats. Rovaniemi,
Finland, Finnish Nuclear Society, ats-fns.fi
04.06.-08.06.2018
5 th European IRPA Congress – Encouraging
Sustainability in Radiation Protection.
The Hague, The Netherlands, Dutch Society
for Radiation Protection (NVS), local organiser,
irpa2018europe.com
06.06.-08.06.2018
2 nd Workshop on Safety of Extended Dry Storage
of Spent Nuclear Fuel. Garching near Munich,
Germany, GRS, www.grs.de
25.06.-26.06.2018
index2018 – International Nuclear Digital
Experience. Paris, France, Société Française d’Energie
Nucléaire, www.sfen.org, www.sfen-index2018.org
27.06.-29.06.2018
EEM – 2018 15 th International Conference on the
European Energy Market. Lodz, Poland, Lodz
University of Technology, Institute of Electrical Power
Engineering, Association of Polish Electrical
Engineers (SEP), www.eem18.eu
24.06.-30.06.2018
ANNETTE Summer School on Nuclear Technology,
Nuclear Waste Management and Radiation
Protection. Turku, Finland, Advanced Networking
for Nuclear Education, Training and Transfer of
Expertise, annettesummerschool.org, www.enen.eu
29.07.-02.08.2018
International Nuclear Physics Conference 2019.
Glasgow, United Kingdom, www.iop.org
22.08.-31.08.2018
Frédéric Joliot/Otto Hahn (FJOH) Summer School
FJOH-2018 – Maximizing the Benefits of
Experiments for the Simulation, Design and
Analysis of Reactors. Aix-en-Provence, France,
Nuclear Energy Division of Commissariat à l’énergie
atomique et aux énergies alternatives (CEA)
and Karlsruher Institut für Technologie (KIT),
www.fjohss.eu
28.08.-31.08.2018
TINCE 2018 – Technological Innovations in
Nuclear Civil Engineering. Paris Saclay, France,
Société Française d’Energie Nucléaire, www.sfen.org,
www.sfen-tince2018.org
05.09.-07.09.2018
World Nuclear Association Symposium 2018.
London, United Kingdom, World Nuclear Association
(WNA), www.world-nuclear.org
09.09.-14.09.2018
21 st International Conference on Water
Chemistry in Nuclear Reactor Systems.
San Francisco, CA, USA, EPRI – Electric Power
Research Institute, www.epri.com
17.09.-21.09.2018
62 nd IAEA General Conference. Vienna, Austria.
International Atomic Energy Agency (IAEA),
www.iaea.org
17.09.-20.09.2018
FONTEVRAUD 9. Avignon, France,
Société Française d’Energie Nucléaire (SFEN),
www.sfen-fontevraud9.org
17.09.-19.09.2018
4 th International Conference on Physics and
Technology of Reactors and Applications –
PHYTRA4. Marrakech, Morocco, Moroccan
Association for Nuclear Engineering and Reactor
Technology (GMTR), National Center for Energy,
Sciences and Nuclear Techniques (CNESTEN) and
Moroccan Agency for Nuclear and Radiological
Safety and Security (AMSSNuR), phytra4.gmtr.ma
26.09.-28.09.2018
44 th Annual Meeting of the Spanish Nuclear
Society. Avila, Spain, Sociedad Nuclear Española,
www.sne.es
30.09.-04.10.2018
TopFuel 2018. Prague, Czech Republic, European
Nuclear Society (ENS), American Nuclear Society
(ANS). Atomic Energy Society of Japan, Chinese
Nuclear Society and Korean Nuclear Society,
www.euronuclear.org
02.10.-04.10.2018
7 th EU Nuclear Power Plant Simulation ENPPS
Forum. Birmingham, United Kingdom, Nuclear
Training & Simulation Group, www.enpps.tech
14.10.-18.10.2018
12 th International Topical Meeting on Nuclear
Reactor Thermal-Hydraulics, Operation and
Safety – NUTHOS-12. Qingdao, China, Elsevier,
www.nuthos-12.org
14.10.-18.10.2018
NuMat 2018. Seattle, United States,
www.elsevier.com
16.10.-17.10.2018
4 th GIF Symposium at the 8 th edition of Atoms
for the Future. Paris, France, www.gen-4.org
22.10.-24.10.2018
DEM 2018 Dismantling Challenges: Industrial
Reality, Prospects and Feedback Experience. Paris
Saclay, France, Société Française d’Energie Nucléaire,
www.sfen.org, www.sfen-dem2018.org
22.10.-26.10.2018
NUWCEM 2018 Cement-based Materials for
Nuclear Waste. Avignon, France, French
Commission for Atomic and Alternative Energies
and Société Française d’Energie Nucléaire,
www.sfen-nuwcem2018.org
24.10.-25.10.2018
Chemistry in Power Plant. Magdeburg, Germany,
VGB PowerTech e.V., www.vgb.org
05.11.-08.11.2018
International Conference on Nuclear
Decom missioning – ICOND 2018. Aachen,
Eurogress, Germany, achen Institute for Nuclear
Training GmbH, www.icond.de
2019
07.05.-08.05.2019
50 th Annual Meeting on Nuclear Technology
AMNT 2019 | 50. Jahrestagung Kerntechnik.
Berlin, Germany, DAtF and KTG,
www.nucleartech-meeting.com
Calendar
atw Vol. 63 (2018) | Issue 4 ı April
Heat Transfer Systems for Novel Nuclear
Power Plant Designs
Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
This article focuses on work that involves designing or modifying heat exchangers that usually can be found in the auxiliary
systems of any power plant. The basic premise of the article is to show that the software provides a one-stop solution for
designing many types of heat transfer systems, where the interaction between various loops connected by heat exchangers can
be assessed. This article especially addresses the audience among nuclear power plants as the quality control in the development
of the software makes it most suitable for nuclear related work. Moreover, the software discussed in this article has the
capability to do contaminant tracing, which could be very useful for nuclear contamination studies in designing specialized
ventilation systems. To highlight the versatility of the software network approach it will be shown how to model any setup and
kind of heat exchanger such as plate, tube-in-tube, liquid/gas, finned tube etc. Additionally, the Koeberg pressurized water
reactor (PWR) steam generator comparison and the Hamm-Uentrop thorium high temperature reactor (THTR) steam
generator comparison are shown as practical examples.
Introduction “Every type of technology benefits from advances inspired by new knowledge and understanding.
Although nuclear energy has operated mostly safely in the past, nuclear engineers do continue to devise new ideas for
making nuclear energy even safer and more secure. The future of reliable nuclear energy requires scientific research to
verify that new types of advanced nuclear fuels and materials are robust enough to withstand the conditions inside a
nuclear reactor during normal and abnormal conditions.” (Idaho National Laboratory).
217
OPERATION AND NEW BUILD
Based on the laws of thermodynamics 1D system
simulation is extremely robust, fast, and reliable. One
software package for 1D system simulation that gains
more and more attention recently was developed in the
early 1990ies by a South African company, namely M-Tech
Industrial. Initially, Flownex® Simulation Environment
was developed for aerospace applications and the energy
sector. Moreover, nuclear validation and verification were
supervised by the governmental ESKOM institution
through its subsidiary PBMR Ltd., who developed a
high-temperature gas-cooled (pebble-bed) reactor in
cooperation with Jülich Research Centre at that time.
Specifically for the nuclear safety analyses required by
PBMR, the software has Nuclear Quality Assurance
( NQA-1) Certification and its development process is
based on ISO 9001.
System simulation programmes provide engineers and
designers a fast and efficient way to set up simulation
models for simple as well as complex fluid dynamic
networks. Such networks commonly contain several
components such as fans, pumps, heat exchangers etc. that
can be computed almost instantly. Furthermore, dynamics
and the control of such networks can be investigated by
running different operation scenarios, such as start-up,
shut down, and various loading conditions, where steady
state and transient effects are taken into account. Thus,
weak spots within a system can be eliminated during the
design process prior to manufacturing as literally any
modification can be tested virtually.
Subsequently, the user is able to analyse the results very
quickly.
Material data that the software supports can be
gaseous, gas mixtures, as well as incompressible pure
fluids and two-phase pure fluids. The user is able to access
a vast library based on the NIST data base. Hence, complex
flows can be modelled using temperature and pressure
dependent material data as well as multiphase effects like
conden sation, evaporation, and cavitation.
The software is equipped with a vast array of components
that cover most required simulation scenarios.
Those components can be used as single components or as
building blocks of components found in thermal fluid
systems or subsystems.
Building blocks, with various levels of detail are
available to model heat transfer phenomena as shown in
Figure 1. Some of the simple heat exchanger models
utilises the Number of Transfer Units (NTU) Method while
other more complex versions employ a fully discretised
approach to heat exchanger modelling. The heat exchanger
types range from tube to plate heat exchangers that can be
modelled as parallel, counter, or cross flow types. Other
components can be vessels, reactors, tube systems, valves,
pumps, fans, compressors, seals etc. Moreover, a whole
library of com ponents for dynamics and control is available
within the software.
1D System Simulation
Flownex® Simulation Environment includes all the
necessary numerical formulations for solving all important
thermo-fluid physical phenomena and moreover, a modern
Windows-GUI that enables an intuitive and easy interaction
for the user. Therefore, the user can concentrate on
design and optimisation rather than on the complexities
usually associated with operating such calculation software.
Typical simulations are run in real time or in the
order of seconds, which makes parameter studies and
optimisation loops extremely fast and very efficient.
| | Fig. 1.
Library for heat exchangers [1].
| | Fig. 2.
Heat transfer library [1].
Operation and New Build
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
atw Vol. 63 (2018) | Issue 4 ı April
OPERATION AND NEW BUILD 218
| | Fig. 3.
Plate heat exchanger model with a two-pass hot side and a single-pass cold side (left) and
schematic (right) [2].
If one is looking into thermo dynamic analyses, simple
components would be used to represent radiation,
conduction, or convection as shown in Figure 2. Thus,
heat exchangers can be custom-built to answer the
question at hand.
Figure 3 shows a simple custom made plate heat
exchanger consisting of composite heat transfer components
and pipe components. The flow path is represented
with a hydraulic diameter and the flow area. The plates are
represented with heat transfer area and actual metal
thickness. User-specified correlations according to the
plate corrugation profiles are defined allowing for full
discretisation along the flow path that results in accurate
pinch-point calculation and transient response.
Another heat exchanger example is shown in Figure 4
where a finned tube air-water heat exchanger can be seen.
The air-side is modelled as a straight-through flow path
(left to right) whereas the water-side is modelled as an
up-down overall counter flow (right to left) configuration
according to the design of the header box plates. The fully
discretised flow path provides an accurate transient
response. The fin-side pressure drop and heat transfer
correlations can be specified with Chilton-Colburn J-factor
tables.
Heat exchangers are crucial for any power plant design.
Figure 5 shows the schematic of the Koeberg PWR steam
generator and the equivalent model built in the software.
For the dryer/separator a complete phase separation is
assumed. The recirculation flow rate is calculated from
buoyancy-driven flow (red circle) that is dependent on
heat transfer coefficient and flow resistance. The model
also assumes a homogeneous two-phase flow. The Chen
correlation for the shell side was implemented with
scripting. Specific material properties can be implemented
via Engineering Equation Solver (EES) coupling or
scripting if necessary.
Table 1 shows the comparison of measured and
simulated data of the Koeberg PWR (South Africa) steam
generator at 60 % and 100 % power load. The software
shows reasonably good agreement to the measured data,
especially when looking at the recirculation ratio R circ
which is a good indication of the overall calculation
accuracy.
Another power plant example is the Hamm-Uentrop
thorium high temperature reactor (THTR-300, Germany)
power plant. One challenge in modelling the THTR is that
at certain combinations of flow rate and heat input, the
flow could be oscillatory. Several types of oscillation are
possible: density wave, pressure wave, and critical heat
flux (dryout)-related oscillations. Fundamental fluiddynamic
modelling is crucial to detect this, which is
provided in the software. Furthermore, this capability is
critical to determine the minimum flow through a steam
generator because it is typically at low power levels that
the steam flow becomes oscillatory. Figure 6 shows the
schematic of the THTR-300 and an equivalent model built
in the software.
The THTR steam generator plant was modelled to
verify the steady-state performance of the assembled
steam generator model. Figure 7 shows the comparison of
measured and simulated data of the THTR-300 at 40 %
and 100 % power load. The software shows very good
agreement to the measured data. The simulation revealed
| | Fig. 4.
Model of a finned tube air-water heat exchanger with multiple water-side passes and a single air pass (top) and schematic (bottom) [2].
T pi
[C]
T po
p so
[kPa]
p si T si x so ṁ s
[kg/s]
Q boiler
[MW]
R circ
60 % power Koeberg 294 273 4889 5055 195 1.0 341 670 7.0
Simulation 294 273 4889 4919 195 1.0 341 666 6.4
100 % power Koeberg 312 279 4911 5277 220 1.0 618 1143 3.8
Simulation 312 280 4911 4951 220 1.0 618 1092 3.8
| | Tab. 1.
Koeberg PWR steam generator comparison.
Operation and New Build
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
atw Vol. 63 (2018) | Issue 4 ı April
that the helium-side heat transfer correlation needed to
have an appropriate Reynolds-number dependence as the
error became quite large at lower power or flow levels
neglecting this.
As aforementioned, heat exchangers are crucial for
any power plant design, especially when designing new
power plants. In addition to the heat transfer modelling
capabilities and with respect to nuclear power generation
the software has recently expanded the Generic Nuclear
Reactor model to simulate the latest nuclear reactor
designs of any geometry. Novel nuclear reactor designs
include liquid fuel reactors, liquid-metal-cooled reactors,
and high temperature gas-cooled reactors (HTGR). In
more detail, there are six reactor types that have gained
researches interest all over the world:
• Very High Temperature Reactor,
• Molten Salt Reactor,
• Sodium-Cooled Fast Reactor,
• Supercritical-Water-Cooled Reactor,
• Gas-Cooled Fast Reactor, and
• Lead-Cooled Fast Reactor.
The new “generalized fuel zone” in the GNR model that is
shown in Figure 8 is capable of handling any fuel geometry
and any fluid type. It expands the geometry capability to
plate fuel, cylindrical fuel rods, spherical fuel elements,
irregular cross-section fuel (like the four-lobe cross-shape
produced by the Lightbridge Corporation), as well as
prismatic block fuel used in some HTGRs.
Appropriate pressure drop and heat transfer correlations
can be selected from the built-in library or defined by
the user. For neutronic calculations, the generalized fuel
zone can provide temperature feedback, as well as heat
generation in all solids and in the core coolant.
The default neutronics model that is supplied with the
software is the point kinetic model which requires the
following inputs:
• Temperature feedback coefficients,
• Heat distribution map, and
• Control rod worth vs. position.
This point kinetic model is provided in a user-editable C#
script, which makes it possible to replace the point kinetic
model by linking the simulation model to an external
neutronics code. The scripted neutronics model also makes
it possible for the user to define one’s own feedback
mechanisms based on the design of the specific reactor.
| | Fig. 5.
Koeberg PWR steam generator schematic (left) [2] and simulation model (right).
| | Fig. 6.
Hamm-Uentrop THTR schematic (left) [2] and simulation model (right).
OPERATION AND NEW BUILD 219
| | Fig. 7.
Hamm-Uentrop THTR-300 steam generator comparison experiment (Exp) [3] and simulation (FNX).
Operation and New Build
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
atw Vol. 63 (2018) | Issue 4 ı April
OPERATION AND NEW BUILD 220
| | Fig. 8.
Schematic of the General Nuclear Reactor (GNR) model [1].
Being able to model all kinds of heat transfer accurately
and to include fission physics makes the software a
valuable tool for every nuclear engineer and power cycle
developer. Figure 9 shows an integrated simulation model
that includes a reactor, steam generator, heat exchange,
and some turbomachinery.
Summary
In order to size control valves or determine the control
strategy for a loop, it is necessary to have the pump
performance curve, the heat exchanger pressure drop and
heat transfer characteristics as well as reactor dynamic
behaviour in one simulation model. In this article, a fast
and efficient solution for designing many types of heat
transfer systems is presented. It was shown how to model
any setup and kind of heat exchanger such as plate, tubein-tube,
liquid/gas, finned tube etc. Flownex® Simulation
Environment offers a straight-forward workflow for
engineers who are involved in designing auxiliary systems
that usually contain one or more heat exchangers, such as
in the power plant industry. The software is a specialized
software (e.g. used by ITER, X Energy, BATAN, Hyundai
Heavy Industries) for sizing specific types of heat
exchangers or for doing basic steady-state and transient
mass-and-energy balances. The value of the software in
this area is that one can really integrate the information
from all available sources into a single representative
model, where one can size all kind of devices, test control
strategies, and do integrated system-level analysis and
design. Furthermore, examples from the nuclear power
plant industry, namely the Koeberg PWR steam generator
and the Hamm-Uentrop THTR-300 steam generator which
demonstrated the software’s usability for nuclear related
work were shown. In addition, the lately incorporated
Generic Nuclear Reactor model was introduced.
Further Reading
| | Flownex® SE: www.flownex.de
| | M-Tech Industrial: www.mtechindustrial.com
| | Idaho National Laboratory: www.inl.gov
References
[1] Flownex (2017) User Manual.
[2] Van Antwerpen, H.: Design and Optimization of Advanced
Nuclear Technologies with 1-d Simulation. 7 th Annual
International SMR and Advanced Reactor Summit 2017,
30-31 March, Atlanta, GA, USA.
[3] Esch, M., Hurtado, A., Knoche, D., and Tietsch, W.: Analysis of the
Influence of Different Heat Transfer Correlations for HTR Helical
Coil Tube Bundle Steam Generators with the System Code TRACE.
Nuclear Engineering and Design, 251, 374-380, 2012.
[4] Van Antwerpen, H., Chi, H., Brits, Y., and Botha, F.: Plant-Wide
Simulation Model for Transient Studies on the Xe-100. 2016 ANS
Winter Meeting and Nuclear Technology Expo, 6-10 November
2016, Las Vegas, NV, USA.
Authors
Sebastian Vlach
Leiter Marketing & Vertrieb
Christoph Fischer (PhD)
CFX Berlin Software GmbH
Berlin, Germany
Herman van Antwerpen (PhD)
M-Tech Industrial (Pty) Ltd
South Africa
| | Fig. 9.
Layout of a complete plant power cycle with an example reactor geometry input map (left) [4].
Operation and New Build
Heat Transfer Systems for Novel Nuclear Power Plant Designs ı Sebastian Vlach, Christoph Fischer and Herman van Antwerpen
atw Vol. 63 (2018) | Issue 4 ı April
Experimental and Analytical Tools
for Safety Research of GEN IV Reactors
G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
Current research on nuclear safety in the world, in addition to supporting existing nuclear power plants (PLEX,
mitigation of severe accidents, the development of accident tolerant fuel, decommissioning, etc.), is focused on the
more detailed aspects of the new reactors. The new generation reactors are expected inter alia to use innovative types
of fuel and new types of coolants, such as e.g. Super-Critical Water (SCW), supercritical CO 2 , liquid metals, fluoride
salts or high-temperature Helium. The paper will describe new experimental infrastructure build recently in Research
Centre Řež under the SUSEN (Sustainable Energy) project and available analytical tools for supporting safety research
of GEN IV reactors. Two experimental loops - SCWL (Supercritical Water Loop) and HTHL (High Temperature Helium
Loop) will serve as in-pile loops in the active core of the research reactor LVR-15. The loops insertion in the reactor
LVR-15 requires performing additional safety analyses studying the mutual interference of the loops and the reactor,
especially in conditions of abnormal operation or accident conditions of the loops. The paper will provide examples of
these analyses made using codes ATHLET (supercritical water) and TRACE (high temperature He) illustrating process
of their assessment and practical use. These activities provide significant opportunity for TSO team in building its new
competencies.
Revised version
of a paper presented
at the Eurosafe,
Paris, France, 6 and
7 November 2017.
OPERATION AND NEW BUILD 221
1 Introduction
The Centrum Výzkumu Řež (CVŘ) and
its partners in the Czech Republic and
abroad are supporting the development
[1] of the Generation IV and
Fusion concepts as well as demonstrators
of these technologies such
as ALLEGRO, ALFRED, DEMO and
others. For this reason, the CVŘ has
had a large R&D program financed
from SUStainable Energy (SUSEN)
project and from its continuation
Research 4 Sustenibility (R4S) [2].
The construction and the operation of
the new SUSEN infrastructure was
supported by the grant of the Ministry
of Education, Youth and Sports as the
part of state help for the large research
infrastructure in the Czech Republic
dedicated to the period 2011–2019.
The SUSEN project consists of 4
programs:
1. Technological Experimental Circuits
(TEO)
2. Structural and System Diagnostics
(SSD)
3. Nuclear Fuel Cycle (NFC)
4. Material Research (MAT)
Within this program, several facilities
were designed and built in order to
study and to address new challenges
of such new technologies. In particular,
the paper focuses on two new
loops which are going to be inserted
inside the LVR-15 research reactor
existing in Řež. The LVR-15 is a light
water tank-type research reactor in
operation since 1957. It is placed in a
stainless steel vessel under a shielding
cover, has forced cooling, uses IRT-4M
type fuel and an has an operational
power level of 10 MWt. The reactor
operations run in campaigns that
usually last for 3 weeks, followed by
an outage lasting for 10 to 14 days
necessary for maintenance and fuel
reloading. There can be also other
campaigns which can operate for
‘short-time’ experiments. Some of the
LVR-15 applications are in the field of
material irradiation research and services,
neutron physics, development
and production of new radiopharmaceuticals
[3]. The loops in concern are
the High Temperature Helium Loop
(HTHL) and the Super Critical Water
Loop (SCWL) and their main scope
are to analyse the cladding behaviour
and structural materials under different
pressure, temperature and coolant
media conditions different from the
standard Light Water Reactors (LWR)
technology [2].
In order to get the regulatory
permit for in-pile operation of these
loops in LVR-15, CVŘ has to prepare
an amendment to the Final Safety
Analyses Report (FSAR) containing
safety analyse of the loops under
| | Fig. 1.
CVR Facilities list.
operational and accidental conditions.
Aim of this paper is to present
the methodology and the analyses
done in support of this process, starting
from code benchmarking/assessment
and the methods adopted in preparing
the safety case.
2 Facilities description
The map of experimental facilities put
into operation in 2016 and those
under preparation to be finalized in
2017 is shown in Figure 1 in the
technology – knowledge map.
In particular, the SCWL and HTHL
represent a pioneer and unique experimental
facility for Gen. IV and Fusion.
2.1 SCWL
The SCWL is going to be a part of a
research facility dedicated to GIV
technologies which will focus on
obtaining data in several areas of the
supercritical fields like: corrosion
processes of construction materials in
supercritical water, with influence of
Operation and New Build
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
atw Vol. 63 (2018) | Issue 4 ı April
OPERATION AND NEW BUILD 222
radiation field, supercritical water
radiolysis and its influence on materials
and water chemistry, development
and testing of sensors, mostly for
measuring of electrochemical potential
(ECP), testing and optimization of
supercritical water regimes [2]. The
specimens being tested will be placed
into the test chamber located in the
active channel where high pressure/
temperature of SCW flow parameters
will be reached.
The SCWL heart is the active
channel, where water reaches required
parameters (pressure of 25 MPa;
temperature of 600 °C; very clean
demineralised water. After successful
out-of-pile (i.e. non active, without
presence of radiation field) operation,
the active channel will be inserted into
the LVR-15 research reactor core. The
bottom part of the active channel is
then submerged between the core’s
fuel assemblies and will face a neutron
flux of up to 1.5 × 10 18 n/m 2 s (thermal
neutrons) and 3 × 10 18 n/m 2 s (fast
neutrons).
The fluid flows in the SCWL is
shown in Figure 2a while the CAD
sketches is shown in Figure 2b.
The active channel has been
modelled with the use of ATHLET
code in two different configurations
see S3.1:
• the out-of-pile configuration that
takes into consideration only pressure
and temperature conditions;
• the in-pile configuration, with the
channel placed inside the LVR-15
active core, that takes into account
also the gamma heating.
2.2 HTHL
HTHL test facility is designed for the
material testing under the simulation
of Gas-cooled Fast Reactor (GFR)
and/or Very High Temperature Reactor
(VHTR) operational conditions.
The specimens being tested will be
placed into the test chamber located
in the active channel where high
pressure/temperature helium flow
parameters will be reached. In addition
to that exposure, during the
in-pile operation, with the active
channel placed into predefined position
of LVR-15 active core rectangular
grid the irradiation effects on the
samples will be studied. The scheme
of the flows in the HTHL is shown in
Figure 3a while the CAD sketch can
be seen Figure 3b.
The active channel has been
modelled with the use of TRACE
code in two different configurations
see S3.2:
• the out-of-pile configuration that
takes into consideration only pressure
and temperature conditions;
• the in-pile configuration, with the
channel placed inside the LVR-15
active core, that takes into account
also the gamma heating.
Views of the channel and of the
coolant flow pattern can be seen in
Figure 3a and Figure 3b.
The temperature inside the channel
is reached through electrical
heater and the coolant flow circulation
is maintained using a two stages
compressor.
3 Methodology
The methodology used to select the
codes and to perform the analyses for
the amendment for the LVR-15 FSAR
consisted in 3 – independent steps:
• Searching and assessing the codes
ability to simulate helium and SCW
during steady-state and transients
conditions.
• Creating the loops model to be
used for the TH analyses and
developing it based on the steadystate
thermohydraulic parameters
• Performing analyses of the selected
scenarios in order to verify the
safety criteria and obtaining the
necessary data for the structural
analyses.
The present methodology complies
with the [4] IAEA standard in introducing
new research facilities inside
nuclear research installations such as
the LVR-15 reactor.
3.1 ATHLET 3.1A code
ATHLET 3.1 patch A code [5] is a
thermal hydraulic system code
developed by the GRS for simulating
time-dependent phenomena in the
PWRs and BWRs. Furthermore, the
code can also simulate GEN IV working
fluids like helium, liquid metals
and supercritical water.
The heat transfer behaviour in
supercritical water represents a
challenging task mainly connected
with ensuring safety and reliable
operation. Nowadays, the understanding
of the supercritical water
regimes is rather limited, specifically
regarding the close proximity of the
critical point.
For the simulation of supercritical
water, a range of properties approximation
has been extended up to a
pressure of 100 MPa. An additional
module cover the pressure range from
22.5 to 100 MPa. The transition
between subcritical and the supercritical
properties is performed by a
suitable interpolation between these
packages for pressures between 22.0
and 22.5 MPa [5].
In ATHLET 3.1A the selection of
correlations for supercritical water is
performed by switching a built in flag
found in the heat structure module.
A number of six correlations are
available which were tested against
the results obtained by IAEA-benchmark
exercise [6] and three of them
were selected for the purpose of the
certification and further use.
3.2 TRACE 5 Patch 4 codes
The TRACE code has been used as an
alternative to the RELAP5/Mod3.3
code, since US NRC decided to stop
| | Fig. 2a.
SCWL Flow.
| | Fig. 2b.
CAD Sketches.
| | Fig. 3a.
Flow in HTHL.
| | Fig. 3b.
HTHL CAD Sketches.
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Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
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| | Fig. 4.
HTHL TRACE Nodalization.
the development of RELAP starting
with next year.
TRACE has been designed to perform
best-estimate analyses of loss- ofcoolant
accidents (LOCAs), operational
transients, and other accident
scenarios in pressurized light-water
reactors (PWRs) and boiling lightwater
reactors (BWRs). It can also
model phenomena occurring in
experimental facilities designed to
simulate transients in reactor systems.
Models used include multidimensional
two-phase flow, none quilibrium
thermo-dynamics, generalized heat
transfer, reflood, level tracking, and
reactor kinetics. In addition, TRACE is
able to simulate several other coolants
such as helium and water in subcooled
condition and atmospheric pressure
(LVR-15 conditions). [7], [8]
For this reason, TRACE code was
selected and used for the simulation in
the Helium at 7 MPa with a temperature
rise from 200 °C up to 900 °C
(nominal parameters for HTHL). The
correlation adopted for simulating the
heat transfer from heat structures to
the helium coolant and vice versa
implemented in TRACE are Gnielinsky
and El Genk [7-9].
3.3 Codes assessment
The code assessment was done by
benchmarking of the codes with
available experimental results done in
different facilities around the world.
One of the most important steps
was selecting the code that can
perform the heat transfer calculation
under the high temperature He or
SCW conditions along with adequate
correlations [10], [11]. In the case of
ATHLET, the code was carefully assed
and benchmarked with experimental
results of a project coordinated by
IAEA [6] for steady state and with
Chinese SWAMUP facility [12] for
| | Fig. 5.
SCWL ATHLET Nodalization.
the transition from supercritical to
subcritical condition.
The aim of this analyses was to
simulate the deterioration phenomenon
[9] of heat transfer with fluid
transiting between subcritical and
supercritical condition. According to
Ref. [6], Mokry, Gupta and Watts-
Chou correlations show acceptable
prediction capabilities of the Heat
Transfer Coefficient (HTC). Both our
analyses and IAEA CRP program
concluded that an uncertainty for
calculating HTC is about ±25% while
the calculating wall temperature was
between ±10 to 15 %. As a result of
this exercise, the code certification
was obtained from SONS (State of
Office for Nuclear Safety) in March
2017 for using the code in simulating
the CVŘ SCWL.
The TRACE assessment was done
with the data available from the
project GoFastR [13] financed by the
EC in the Framework Program 7, in
particular with data related to the
HE-FUS3 facility [14], [15]. The
facility operational parameters are
similar to the HTHL.
The TRACE HE-FUS3 thermal hydraulic
model was developed and
compared with experimental data
from steady state loop operation and
selected transients. The comparison
showed that the TRACE T/H model
can simulate the helium temperatures
as well as the piping wall temperatures
along the different sections
of the facility accurately. After a sensitivity
analysis, the electrical heater
power has been lowered to 10.76 kW.
The certification for TRACE code was
obtained from SONS in December
2016 by CVŘ for simulating water in
PWR condition, sub-cooled water at
atmospheric pressure (such as LVR-15
operational condition) and helium
behaviour in the range of 7 MPa for a
temperature range between 200 to
900 °C. [16]
3.4 Model description for
HTHL and SCWL
The HTHL and SCWL are similar
experimental facilities characterized
by 2 steps upward and downward
flows, although some major differences
exist in the design. In particular,
the HTHL active channel contains
all necessary components for heat
transfer inside except of the compressor
and the main compensator, which
are located in the chemical control
system. The Figure 4 shows the
TRACE nodalization containing simulated
components.
The SCWL is different in such way
that it needs some extra components
larger than the HTHL to help the sub
critical water to become gas. For this
reason additional axillary facilities,
such as a recuperator, cooler, pump,
compensator and other 4 sections of
electrical heater are located in a
different building along with the
chemical control system.
The ATHET SCWL loop model
shown in Figure 5 is focused mainly
on the active channel from inlet to
outlet, although all the previous
components are also simulated as a
part of the primary and the secondary
circuits. In addition to the primary
and the secondary circuits of the
SCWL, there is the third open loop
representing the active channel position
into the LVR-15 core and providing
additional heat transfer between
active channel and reactor coolant.
3.5 Analysed Scenarios
The planned in-pile operation of both
loops requires an amendment of the
LVR-15 Final Safety Report providing
thermohydraulic and structural integrity
analyses during normal operation
OPERATION AND NEW BUILD 223
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Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
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OPERATION AND NEW BUILD 224
Normal operating
conditions
Steady State
LVR-15 Start up
LVR-15 Shutdown
Loops Start up
Loops Shutdown
and during Loss of Flow Accident
( LOFA) and Loss of Coolant Accident
(LOCA) accident conditions. In particular,
the structural integrity analyses
required the temperature profile
inside the Pressure Envelope (PE) as
boundary condition. For this reason
the normal operation and abnormal
operation conditions were calculated
using TRACE and ATHLET codes with
very narrow mesh nodal distribution
in the PE. For structural integrity
following criteria and limitation due
to the non-boiling condition in LVR-15
were used:
1. PE maximum temperature during
normal/abnormal transients is less
than 450 °C.
2. PE maximum temperature during
accident conditions is less than
500 °C.
3. Aluminium surface of the Receiver
maximum temperature in contact
with LVR-15 coolant less than 45 °C
during normal/abnormal conditions.
4. Aluminium surface of the Receiver
maximum temperature in contact
with LVR-15 coolant less than 60 °C
during accident conditions.
In the case of accident conditions,
both active channels of HTHL and
SCWL will have to be replaced. The
analysed scenarios are described in
the Table 1.
4 Illustrative results
The results described in the paper
refer to the simulations of SCWL and
Pressure
tests
(not simulated)
| | Tab. 1.
Operational and Accident Scenarios Description.
Abnormal
conditions
Switch off Loops Electrical
Heater for 1 min.
LVR-15 SCRAM and switch off
of Loops Electrical Heater
at t = 0 s + pump trip after 1 min.
Switch off Loops Electrical Heater
at t = 0 s + LVR15 SCRAM and
Pump Trip after 3 min.
Parameter Value Unit
Pressure 25 MPa
Inlet Flow
Temperature
Outlet Flow
Temperature
Max Flow
Temperature
Sample Area
Mass flow
| | Tab. 2.
SCWL main parameters calculated during
steady state.
HTHL during the steady state operation
with continuing in LOFA condition.
The results represent an extract
of the large number of calculations of
various combinations of operational
transients with the aim to demonstrate
the capabilities of the codes to
simulate behaviour the loops.
4.1 SCWL steady state and
LOFA analyses
The main parameters for the steady
state are shown in Table 2. The whole
steady state calculation was rather
long due to some inertia of the system.
The computer model simulated
behaviour during the transient of all
heat structures representing the
complete piping system. In the calculation
some numerical instability
complicated the steady state due to
Accident
conditions
385 ºC
406 ºC
600 ºC
35 %
By pass flow 65 %
Mass flow 200 kg/h
Loss of Flow Accident
(LOFA)
Loss of Coolant Accident
(LOCA)
small dimensions of the component
facing the deterioration flow phenomenon
during the heating up process.
For these reasons, the whole steady
state was completed in 25,000 s,
where 15,000 to 20,000 s were needed
to adjust the steady state and the
rest 5,000 s were used to verify the
steady behaviour of the main parameters.
After this period the model simulated
the accident scenario – LOFA
without the reactor SCRAM in order
to maximize the consequences and to
calculate the time to reach temperature
of PE (AC) 500 °C.
The scenario is described in the
following steps:
1. Pump stops in 1 s after the initialization
event (25,001 s)
2. Active channel internal electrical
heaters shut down to 0 % on the
nominal power in 7s (25,007 s)
3. The LVR-15 SCRAM starts at 40 s
when the maximum temperature
in the PE rises above the 500 °C.
(25,040 s)
4. The whole transient is completed
in 15,000 s (40,000 s), when the
SCWL and LVR-15 are in the
controlled cold state.
The Figure 6 and Figure 7 represent
the SCW maximum temperature
calculated in the sample area and the
outlet temperature from the active
channel, while the Figure 8 shows the
maximum temperature of the PE,
where there is the neutron flux peak
in the Boltzmann distribution.
4.2 HTHL steady state and
LOFA analyses
The design conditions calculated for
the active channel are described in
Table 3. And they are mainly summarized
as reported:
1. Mass flow rate of 0.0105 kg/s
2. Design pressure of 7 MPa
3. Design electrical heater power of
11.85 kW
4. Cold helium temperature of 210 °C
| | Fig. 6.
SCWL Coolant Maximum Temperature in LOFA.
| | Fig. 7.
SCWL Active Channel Outlet Temperature in LOFA.
Operation and New Build
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
atw Vol. 63 (2018) | Issue 4 ı April
Location of the Thermocouples
Thermocouple
| | Fig. 8.
SCWL Maximum EP Temperature in LOFA.
Parameter Value Unit
Pressure 7 MPa
Inlet Flow
Temperature
Max Flow
Temperature
Maximum AC
Pressure Envelop (PE)
Temperature
210 ºC
900 ºC
450 ºC
Mass flow 40 kg/h
Inlet into the interpiping space of the reheater
Output from the interpiping space of the reheater
Entry into the test chamber
Inlet to the reheater piping space
Output from the reheater piping space
Output from the primary side of the heat exchanger
Maximum helium temperature
| | Tab. 4.
Thermocouples position and description.
T1
T2
T3
T4
T5
T6
Tmax
OPERATION AND NEW BUILD 225
| | Tab. 3.
HTHL main parameters calculated during
steady state.
The steady state simulation was
run in null transient mode for 5,000 s
and the stabilized conditions were
reached after 3,500 s. The LOFA
transients was characterized by an
immediate safety shutdown of the
reactor due to the loss of power. As a
result of the SCRAM, the temperature
went immediately down following the
heat generated by decay gamma flux.
Figure 9 shows the calculated
temperature for various thermocouples
positions (according to
Table 4), while Figure 10 represents
the maximum temperatures in the
HTHL PE.
| | Fig. 9.
HTHL Helium temperatures during LOFA.
5 Conclusions
The article provides a brief introduction
about the SUSEN project and the
experimental facilities built in CVŘ in
the Czech Republic for research and
development in support of the safe,
reliable and long‐term sustainable
operation of existing energy facilities
and in development of GIF IV and
fusion technologies. The SUSEN
R&D activities include four complementary
programmes, mentioned in
the introduction, which are focused
on material science, thermal hydraulics,
neutronics, radiation protection,
nuclear chemistry, waste management
and environmental studies. A
significant part of the research programme
is devoted to HTH and SCW
experimental loops, which are going
to be installed into the active core of
the research reactor LVR-15. Both of
| | Fig. 10.
HTHL PE temperature during LOFA.
these unique facilities are challenging
to model and the selection of appropriate
codes was a demanding process.
A special methodology was used for
assessing the abilities of the codes to
simulate these advanced coolants and
to obtain regulatory certificate/ permit
for their use in operational and accident
conditions and for preparation of
the amendment of the LVR-15 FSAR.
These presented activities represent
only starting steps for the further
codes validation which will be based
on benchmarking of the codes with
experimental data provided by the
SCWL and HTHL loops in their
experimental campaigns.
Aknoledgment
The authors would like to thank
Mr. Miroslav Hrehor and Dr. Vincenzo
Romanello for their kind revisions and
suggestions.
The presented work was financially
supported by the Project CZ.02.1.01/
0.0/0.0/15_008/0000293: Sustainable
energy (SUSEN) – 2 nd phase,
realized in the framework of the
Operation and New Build
Experimental and Analytical Tools for Safety Research of GEN IV Reactors ı G. Mazzini, M. Kyncl, Alis Musa and M. Ruscak
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OPERATION AND NEW BUILD 226
European Structural and Investment
Funds.
This work has been supported
by the Project CZ.02.1.01/0.0/0.0/
15_008/0000293: Sustainable energy
(SUSEN) – 2 nd phase realized in the
framework of the European Structural
and Investment Funds.
References
[1] CVR Annual Report 2016.
[2] http://susen2020.cz/
[3] http://cvrez.cz/en/infrastructure/
research-reactor-lvr-15
[4] IAEA, Standards Safety in the Utilization
and Modification of Research Reactors”,
Safety Standard n° SSG-24, VIENNA,
2012.
[5] ATHLET 3.1A, 2016 User manual:
ATHLET Mod 3.1 Cycle a, G. Lerchl,
H. Austregesilo, P. Schoffel, D. von
der Cron, F. Weyermann, March 2016.
[6] Heat Transfer Behaviour and Thermohydraulics
Code Testing for Supercritical
Water Cooled Reactors (SCWRs),
IAEA. http://www-pub.iaea.org/
books/IAEABooks/10731/Heat-
Transfer-Behaviour-and-Thermo-
hydraulics-Code-Testing-for-
Supercritical-Water-Cooled-R
[7] TRACE V5.840 Theory Manual,
U.S. Nuclear Regulatory Commission,
Washington DC, March 2013.
[8] TRACE V5.840 User’s Manual, Volume 1:
Input Specification, U.S. Nuclear
Regulatory Commission, Washington
DC, February 2014.
[9] TRACE V5.840 User’s Manual, Volume 2:
Modelling Guidelines, U.S. Nuclear
Regulatory Commission, Washington
DC, February 2014.
[10] G. Mazzini et al., ATHLET 3.1A
SIMULATION CAPABILITIES FOR SUPER-
CRITICAL STATE, CVR 1581, 1.1.2017.
[11] G. Mazzini et al., ATHLET 3.1A HEAT
TRANSFER ASSESMENT FOR SUPER-
CRITICAL WATER, CVR 1582, 1.1.2017.
[12] G. Mazzini et al., ATHLET 3.1A
CAPABILITIES IN SIMULATING SWAMUP
FACILITY IN SCW CONDITIONS, CVR
1583, 1.1.2017.
[13] M. Polidori, HE-FUS3 Benchmark
Specifications, GoFastR-DEL-1.5-01,
Rev. 0, ENEA, July 2011.
[14] M. Polidori, HE-FUS3 Experimental
Campaign for the Assessment of
Thermal-Hydraulic Codes: Pre-Test
Analysis and Test Specifications,
Report RSE/2009/88.
[15] M. Polidori et al, HE-FUS3 Benchmark
Results, GoFastR-DEL-1.5-6, Rev. 0,
November 2012.
[16] Miloš Kynčl, Development and Assessment
of TRACE HTHL-2 Facility Thermal
Hydraulic Model, Internal Project Status
Report, CVŘ 1334, March 2017.
Authors
G. MazziniM. Kyncl
Alis Musa
M. Ruscak
Centrum Vyzkumu Rez (CVŘRez)
Hlavní 130
250 68 Husinec – Řež,
Czech Republic
Numerical Analysis of MYRRHA Interwrapper
Flow Experiment at KALLA
Abdalla Batta and Andreas G. Class
Introduction The MYRRHA reactor, which is developed at SCK-SCN in Belgium, represents a multi-purpose
irradiation facility. Its prominent feature is a pool design with the nuclear core submerged in liquid metal lead bismuth.
During transients between normal operation and accident conditions decay heat removal is ensured by forced and
natural convection, respectively. The flow in the gap between the fuel assemblies plays an important role in limiting
maximum temperatures which should not be exceeded to avoid core damage. The term inter-wrapper flow (IWF)
describes the convection in the small gap between the wrapper tubes of neighbouring fuel assemblies (FAs). It plays an
important role for passive decay heat removal (DHR).
Based on numerous experiments
several correlations have been proposed
for the flow within wirewrapped
rod bundles. However, for
the flow within the gap between
neighbouring bundles only few
studies are reported. Recently [1]
reviewed the existing correlations by
Rheme [2], Baxi & Dalle Donne [3]
Cheng and Tordreras [4], and Kirillov
[5] for the pressure-drop in wirewrapped
rod bundles. The existing
correlations were compared to all the
available experimental data and
showed that agreement of approximately
±20 % can be expected. For
the inter-wrapper flow within the
gap only few studies exist, see [6].
Due to the scarce database, within the
Horizon 2020 – research and innovation
framework program of the EU,
the SESAME project was established
to develop and validate advanced
numerical approaches, to achieve a
new or extended validation base and
to establish best practice guidelines
including verification & validation
and uncertainty quantification, see
[7]. In particular the current work
supports the inter-wrapper flow
experiment at KALLA. Three fuel
assemblies including the gap flow are
studied covering the full range of
thermo- hydraulic conditions expected
in the reactor application. For this
purpose, an experimental test matrix
has been established which covers
relevant scenarios. The aim of our
numerical pre-test study is to help the
design of the experiment. The current
study applied RANS-CFD methods for
design support of the experiment. In
the body of this compact the experiment,
the corresponding numerical
model, and preliminary numerical
results are provided.
1 Experimental setup
The KALLA experiment investigates
IWF between three bundles which
are thermally connected by a gap.
Figure 1 shows a cross-sectional view
of the test section which consists of
three ducts representing the fuel
assemblies. Each duct contains 7 wirewrapped
electrically-heated pins
representing the fuel rods. The gap
between the channels, i.e. assemblies,
is filled with liquid metal, so
that strong thermal coupling exists
between neighbouring assemblies.
The test matrix covers independent
variation of flow and thermal conditions
in both the gap and the bundles.
Detailed description of the experiment
is reported in [8]. The geometrical
parameters of the bundle and the
nomenclature are also shown in
Figure 1. The experimental loop
facility THESYS at KALLA and the
Operation and New Build
Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
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| | Fig. 1.
SCWL Coolant Maximum Temperature in LOFA.
| | Fig. 2.
Left: Experimental loop facility THESYS at KALLA showing location where the inter wrapper flow
experiment (see Figure 3) will be installed; right: flow diagram for the IWF tests with four parallel
channels; the valves V2.1-V2.3 control the flow through the assemblies Q1-Q3. V.2.4 controls the
flow in the gap [8].
location where the IWF experiment
will be installed is shown in Figure 2
left. Figure 2 right shows the flow
diagram of the IWF tests with four
parallel channels representing the
three assemblies (Q1-Q3) and the gap
( illustrated by the box containing
Q1-Q3). The flow and temperature
within each assembly and the gap can
be set individually by choosing valve
openings (V2.1-V2.4) and heating
rates according to the KALLA test
matrix. Figure 3 shows the geometry
of the IWF test section.
and mesh resolution for the thermoshydraulic
investigation of the gap and
the bundle. In particular, we include
the upstream components to verify
their influence on the flow field within
the test section. We employ the k-ε
turbulence model and the commercial
CFD-code Star CCM+. Our first
studied case (i) focuses on the gap
| | Fig. 3.
Geometry of the IWF test section, dimensions are in mm, the heated part
of the bundle is marked red on the left side of the figure, 600 mm, [8].
flow and our second case (ii) on the
fuel assembly. For the study of case (i)
a computational domain including
the lower flow distributer, riser pipe
( including venture tube), upper flow
vessel, and the gap are considered (for
corresponding technical drawings of
components refer to Figure 3). For the
study of case (ii) the computational
domain includes the lower flow distributer,
riser pipe (including venture
tube), one inlet expansion and a single
7-pin bundle. Flow properties of the
liquid metal Lead-Bismuth eutectic at
200 °C are employed. Note that corresponding
upstream pipes and flow
conditioners are modelled so that
all relevant geometric details are
captured. Quantifying the effect of
the flow conditioning sections is
important for future simulations, as it
would enable the use of a simpler
computational domain, which still
provides accurate results. In the future
post-test analysis, the smallest representative
computational domain (e.g.,
potentially without flow conditioner
etc.) will be used to compose a fully
coupled thermos-hydraulic simulation
of the three bundles including
the IWF in the gap. Figures 4 left
and right show the computational
domains for the pre-test studies
OPERATION AND NEW BUILD 227
2 Numerical study
A comprehensive analysis of the
experiment requires efficient simulations.
In the pre-test analysis of the
hydraulics separate simulations of the
gap region and the fuel assembly are
performed. In a first step, we determine
suitable computational domains
| | Fig. 4.
Computational domain for IWF-gap (left) and bundle (right) including the upstream domains.
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OPERATION AND NEW BUILD 228
of cases (i) and (ii), respectively.
Obviously, a substantial effort was
undertaken to include the upstream
flow domain, so that the inflow into
the fuel assembly and the gap are
properly represented in the flow
simulations.
Since we have less experience with
the gap region, and in particular, the
applicable turbulence regime we have
considered 3 cases corresponding to
laminar flow, transitional flow, and
fully developed turbulence, respectively.
This covers the flow range 0.17
to 0.86 kg/s (Re = 1,250 to 6,250),
proposed by the test matrix. For the
investigation of the fuel assembly, we
consider the nominal flow rate, i.e. the
maximum flow rate planned in the
test matrix. This corresponds to a flow
rate of 3.58 kg/s and Re = 8,910
where Re is based on the bundle
hydraulic diameter. All cases considered
in the experimental test matrix
are within the range of transitional
flow according to Cheng and Todreas
[4] (see next section on correlations)
so that no distinction of various flow
regimes is needed for the comparison
to correlations.
2.1 Inter-wrapper flow gap
region
The objective of case study (i) is to
investigate the effects of all upstream
components on the flow distribution
entering the gap, i.e. the inter wrapper
flow region. This study employs the
computational domain shown in
Figure 4 left. For the simulation, a
mesh with approximately 0.72 million
cells has been generated. The investigated
range of flow rates results in
turbulent flow in all components
upstream of the gap, since the
Reynolds- numbers based on pipe
diameter varies between 5,200 and
26,000. However, within the gap
the Reynolds-number based on gapwidth
equates to 1,250 to 6,250 corresponding
to the transitional regime of
turbulence. The pressure drop along
the gap accounts for about 20 % of the
total pressure drop. Since we are
interested in accurately predicting the
upstream flow in the gap region the
use of a turbulent model is mandatory.
Moreover, in order to judge the uniformity
of the flow entering the gap
there is no need to use a very accurate
result within the gap. Thus, a high-
Reynolds-number turbulence model
using automatic wall functions is
used. Figure 5 shows the velocity
vectors in the gap entrance region for
the case where the Reynolds-number
is 5,200 based on pipe diameter
(Reynolds-number is 1,250 based on
the gap hydraulic diameter). We
observe that the flow within the gap
becomes near uniform after a short
length, which does not exceed 10 %
of the length of the gap region. The
heated zone starts further downstream
approximately at half the
length of the gap region. For higher
Reynolds-number a qualitative similar
result is obtained. For future simulations
aiming at accurately simulating
the temperature field, we conclude
that the effect of upstream components
is negligible. In Table 1 the
pressure drop across the simulated
region is compared to design values
for three selected cases covering
the full range of flow rates. Design
values are calculated using lumped
parameter models. Both results agree
reasonably well, indicating that
lumped parameter models well
describe the flow in the gap.
2.2 Flow within a single
wire-wrapped rod bundle
As in the previous study, we aim at
investigating whether the upstream
region that conditions the flow
entering the wire-wrapped bundle
influences the flow in the heated
section of the bundle. Here, i.e. in case
study (ii), a single Reynolds-number
of 8,900 based on the bundle hydraulic
diameter is considered. This
corresponds to the nominal flow
rate as well as the maximum flow
rate intended in the experimental
tests. The computational domain of
Figure 4 right uses approximately
1 million cells. Figure 6 shows the
velocity magnitude within the bundle.
At the entrance, we still observe pronounced
non-uniformities of the flow
distribution. These quickly equilibrate
so that a more-uniformly distributed
flow is observed well before the
heated section of the bundle is
reached (for the location of the heated
region refer to Figure 3).
This result suggests that inflow
effects are negligible for the intended
thermal analysis of the bundle. Thus
in a second simulation we remove the
flow-conditioning region to reduce
the size of the considered flow
domain. To validate our simulation
results we use higher mesh resolution
within the smaller domain. Figure 7
shows the pressure along two selected
axial lines, which are depicted in the
small inset. The influence of the wirewrap
manifests in the periodical
modulation of the pressure profile.
Obviously, development effects have
decayed at a length of approximately
100 mm. We compute the pressure
| | Fig. 5.
Velocity vectors within the gap upstream region.
Flow rate
[kg/s]
design Δp tot ,
[Pa]
CFD Δp tot , [Pa]
1. 0.86 15350 13500
2. 0.688 9964 -
3. 0.516 5667 5500
4. 0.344 2586 -
5. 0.172 663 850
| | Tab. 1.
Comparison of design values evaluated by lumped parameter model
versus computed pressure drop across the test section including the flowconditioning
components.
| | Fig. 6.
Velocity magnitude within bundle showing non-uniformities of flow distribution at leftmost plane and
more-uniformly distributed flow in subsequent planes.
Operation and New Build
Numerical Analysis of MYRRHA Inter- wrapper Flow Experiment at KALLA ı Abdalla Batta and Andreas G. Class
atw Vol. 63 (2018) | Issue 4 ı April
| | Fig. 7.
Pressure along two selected axial lines in the wire-wrapped rod bundle.
The inset specifies location of lines.
drop using data at corresponding
wire-wrap positions, i.e. from axial
positions 0.065 m to 1.268 m. The
mean pressure drop is 946 Pa/m.
2.3 Model validation
In this subsection, results of our
numerical study are compared to the
simplified Cheng and Todreas [1986]
correlation. The correlation was
recently recommended in (1) to
predict pressure drop (Δp) in bundles
with an accuracy of ±20 %. It applies
for a wide range of Reynolds- numbers.
The friction factor (f) is defined in
eq. 1, where d h,bdl , L, and u b
2
are
hydraulic diameter, length, and
average axial bundle velocity, respectively.
(1)
The correlation for f reads
for Re < Re L
for Re L ≤ Re ≤ Re T
for Re > Re T (2)
where
Re L = 300 x 10 1.7(P/D−1.0) (3)
Re T = 10,000 x 10 0.7(P/D−1.0) (4)
ψ = log(Re/Re L ) / log(Re T /Re L ) (5)
C fL = (-974.6 + 1612.0(P/D) −
598.5(P/D) 2 )(H/D) .06-0.085(P/D)
(6)
C fT = (0.8063 − 0.9022(log(H/D)) +
0.3526(log(H/D)) 2 ) ×
(P/D) 9.7 (H/D) 1.78-2.0(P/D) (7)
We compare the nominal flow case
of 3.580 kg/s which corresponds to a
velocity of 0.2 m/s and Re is 8910,
which is in the transient region.
According to eqns (3) and (4), Re L and
Re T are 902 and 15735, respectively.
The calculated friction factor f
equates to 0.0557. This corresponds
to a pressure drop in the bundle of
1407.2 Pa. The predicted pressure
drop resulting from the CFD study is
1,138 Pa. The difference is near 19 %,
which lays within the accuracy limits.
In future thermos-hydraulic simulations,
the current model can be
applied. For posttest analysis, additional
sensitive studies might be
necessary to further reduce the
uncertainty.
Conclusions
The flow in the gap between neighbouring
fuel assemblies plays an
important role in transients between
forced and natural convection. At
KALLA an experiment on the interwrapper
flow is currently setup and
accompanied by pre-test numerical
CFD studies. These proof that both
the flow in the gap region and the
fuel bundle are not influenced by the
upstream flow-conditioning region.
Moreover, development length are
much shorter than the unheated
length of the test section, so that
the thermal field is uninfluenced by
flow non-uniformities. Preliminary
comparison of pressure losses computed
by CFD and correlation provide
reasonable agreement for both the
gap and bundle. The result of our
study enters pre-test studies of the
thermal field within the EU-H2020
SESAME project. There complete
simulation of the test section consisting
of three bundles connected
by the gap region including conjugate
heat transfer is performed.
Acknowledgement
This project has received funding from
the Euratom research and training
programme 2014-2018 under grant
agreement No 654935 and from the
AREVA Nuclear Professional School.
References:
[1] Chen, S.; Todreas, N.; Nguyan, N.
(2014). Evaluation of existing correlations
for the prediction of pressure drop
in wire-wrapped hexagonal array pin
bundles. Nuclear Engineering and
Design 267, pp. 109 – 131
[2] Rehme, K. (1973). Pressure drop
correla tions for fuel element spacers.
Nuclear Technology 17, 15–23.
[3] Baxi, C.B., Dalle Donne, M., (1981).
Helium cooled systems, the gas cooled
fast breeder reactor. In: Fenech, H. (Ed.),
Heat Transfer and Fluid Flow in Nuclear
Systems. Pergamon Press Inc.,
pp. 410–462.
[4] Cheng, S.-K.; Todreas, N. (1986). Hydrodynamic
models and correlations for
bare and wire-wrapped hexagonal rod
bundles - Bundle friction factors,
subchannel friction factors and mixing
parameters. Nuclear Engineering and
Design 92 (2), 227 – 251.
[5] Kirillov, P.L., Bobkov, V.P., Zhukov, A.V.,
Yuriev, Y.S., (2010). Handbook on
Thermo hydraulic Calculations in
Nuclear Engineering. Thermohydraulic
Processes in Nuclear Power Facilities,
vol. 1. Energoatomizdat, Moscow.
[6] Kamide, H.; Hayashi, K.; Toda, S. (1998).
An experimental study of intersubassembly
heat transfer during
natural circulation decay heat removal
in fast breeder reactors. Nuclear
Engineering and Design 183, 97 – 106.
[7] http://sesame-h2020.eu/
[8] Pacio, J, et. al. (2016), Deliverable 2.10 –
KALLA Inter- wrapper flow setup for
SESAME (thermal hydraulics Simulations
and Experiments for the Safety
Assessment of MEtal cooled reactors)
project, activity: NFRP-01-2014
Improved safety design and operation
of fission reactors, H2020 Grant
Agreement Number: 654935.
Authors
Abdalla Batta
Andreas G. Class
AREVA Nuclear Professional School
Karlsruhe Institute of Technology
Karlsruhe, Germany
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OPERATION AND NEW BUILD 230
Heat Balance Analysis for Energy
Conversion Systems of VHTR
SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park
VHTR (Very High Temperature gas Reactor) which helium is used as a coolant can easily produce heat required in
high-temperature thermochemical process, and because of low heat output density, the possibility of core melting is
low. Helium has the advantage of safety, and the coolant can become super high temperature, thereby power production
as well as hydrogen production application is possible. In this study, provided that VHTR is located in the primary
system, the heat conversion system will be discussed in which hydrogen production and power supply are possible. In
order to control the ratio between power and hydrogen production, the helium flowing through nuclear reactor is made
to pass through heat exchanger for hydrogen production and steam generator or heat exchanger. Power production was
made to be composed of ultra-super critical steam cycle (USC) and supercritical CO 2 (SCO 2 ) cycle so that efficient
operation condition can be selected. This study proposed the whole heat conversion system model, and carried out
thermodynamic feasibility calculation according to major design variable at each point and sensitivity analysis for
efficiency optimization.
1 Introduction
Recently, an interest on hydrogen as a
clean energy source and a fossil fuel
substitute has been increasing. From
the viewpoint that hydrogen utilizes
the energy system which uses the
existing fossil fuel without the
emission of environmental pollution
material, contrary to fossil fuels,
hydrogen is emerging as a promising
future clean energy. Among hydrogen
production methods, high-temperature
pyrolysis hydrogen production
method using heat chemical process is
considered as a proper method for
mass hydrogen production. Heat is
required much for high-temperature
heat chemical process, and lightwater
reactor that uses water as coolant
does not produce heat required
for high-temperature heat chemical
process. VHTR (Very High Temperature
gas Reactor) which uses helium
as coolant can easily produce heat
required for high-temperature thermochemical
process, so recently the
study of the use of high temperature
gas for hydrogen production has been
the research trend [1, 2].
VHTR has no possibility of core
melting due to low heat output
density, and it does not use water, so
there is no risk of explosion danger
due to hydrogen generation in the
case of coolant loss accident. Besides,
it has the advantage that high-temperature
coolant can be made compared
to water-cooled reactor, so it has the
advantage of power production and
process heat supply [3]. Nuclear
reactor is in charge of heat supply, and
this can be converted variously to be
used as the production of hydrogen or
power. In this study, by borrowing
general name in the atomic power
field, VHTR is called as a primary
system, the part which hydrogen production
and power supply are possible
through heat conversion, is defined as
the secondary system. Helium flowing
in nuclear reactor delivers the heat of
the primary system to the secondary
system through HX (Heat Exchanger).
Helium flowing through the secondary
system passes first through heat
exchanger where hydrogen production
occurs, and secondly and thirdly
passes through steam generator and
heat exchanger composed of ultrasuper
critical cycle (Ultra- supercritical
steam cycle: USC) and super critical
carbon dioxide (Supercritical CO 2 :
SCO 2 ) cycle, respectively, producing
process heat and power. In this study,
the authors proposed the overall heat
conversion system model, and performed
the thermodynamic feasibility
calculation in accordance with major
design variable at each point and
sensitivity analysis for efficiency
optimization.
2 Research methodology
2.1 Concept and methodology
of hydrogen production
equipment
As a method of hydrogen production
which uses water as a raw material by
using 900 °C heat, high temperature
electrolysis using heat energy simultaneously
and the mixed method of
using thermochemistry process method
and electrolytic method. Recently,
research has been focused on Sulfur-
Iodine thermochemical cycle where
iodide and sulfuric acid were used to
break down water. This is because the
required equipment can be scaled up
and process handling material is only
composed of gas and liquid so that
continuous operation is possible.
Besides, it is advantageous to use
nuclear reactor where the safety of
load change is demanded as heat
source [2].
In the hydrogen production equipment
where high temperature heat is
used, according to the Reaction 1
below, sulfuric acid (H 2 SO 4 ) can be
broken down into water vapor
(H 2 O(g)), oxygen (O 2 (g)), and sulfur
dioxide (SO 2 (g)).
Reaction 1:
2H 2 SO 4 + Heat 2H 2 O + 2SO 2 + O 2
After decomposition, oxygen(O 2 (g))
is removed, and water vapor(H 2 O(g))
and sulfur dioxide (SO 2 (g)) are cooled
down, reacting with iodide (I).
According to Reaction 2 below,
sulfuric acid (H 2 SO 4 ) and hydrogen
iodide (HI) are formed.
Reaction 2:
4H 2 O + 2SO 2 + 2I 2 2H 2 SO 4 + 4HI
+ Heat
Finally, by using high temperature
heat, hydrogen Iodide (HI) can be
separated into hydrogen (H 2 ) and
iodide (I) according to the reaction 3
below.
Reaction 3: 4HI + heat 2I 2 + 2H 2
2.2 The concept and status
of USC and S-CO 2 cycle
USC power plant means the power
plant where vapor pressure is 254 kg/
cm 2 or higher, and main vapor’s
or reheated vapor’s temperature is
593 °C or higher. The reasons why
pressure and temperature of the
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evaporator are elevated are that the
efficiency of power plant is increased.
When the temperature of main evaporator
and reheating steam increases
by 10 °C, the efficiency increases
by 0.5 %; and pressure increases by
10 kg/cm 2 , the efficiency increases by
about 0.2 %. Domestically, in 1990’s,
500 MW-grade standard coal thermal
power plant was designed and built,
and its operation condition was pressure
246 kg/cm 2 and temperature
538 °C.
In the case of Dangjin Thermal
Power No. 9, No. 10 and Samcheok
Thermal Power No. 1, No .2 that have
been being built, the pressure of
250 kg/cm 2 , temperature of 600 °C
were accomplished [4].
SCO 2 cycle is the power generation
technology of the Gas Brayton Cycle
method where pressurized carbon
dioxide is heated by the pressure
greater than critical condition to high
temperature and turbine is driven.
Presently, CO 2 power generation cycle
can be applied to most heat sources
used, and also it can be used for large
power plant, small scale distribution
power supply, or power supply for
marine plant.
Super critical condition means the
conditions for temperature and pressure
greater than critical point in the
general material state where liquid-gas
phase change occurs, and the
temperature and pressure at the lower
pressure part is greater than 32 °C, 74
atm, and all parts of cycle are maintained
over critical condition. While
operation is carried out at high
pressure, volumetric flow decreases,
so the size of overall heat conversion
cycle can be decreased; accordingly,
construction period and production
unit price can be lowered to secure
high economic feasibility.
Besides, compared to water vapor,
the compatibility with existing material
is excellent, so it can be supplied
to turbine at the temperature higher
than evaporator cycle. From this, the
increase of additional power generation
efficiency can be possible [5].
2.3 Heat Conversion Model
Design
IHX loop of VHTR that is studied in
the present study is the system where
the high temperature heat generated
in the reactor by connecting hydrogen
generation equipment and power
generation equipment in series can be
supplied in the same manner.
IHX loop nuclear reactor shown in
Figure 1 provides 350 MWt heat output,
and the heat generated from
| | Fig. 1.
IHX Loop Modelling.
nuclear fission is supplied to helium
fluid. For heat transfer to produce
hydrogen, heat exchanger, steam generator
for the power generation via
USC cycle, and in the power generation
via SCO 2 , one heat exchanger is
provided. In order to utilize the result
of the study regarding the existing
VHTR, the major principle and
variable if heat conversion model
were set as follows. Temperature and
pressure at No. 1, 2, 3, 4, 10 were
presumed by reference literature [8].
Temperature and pressure of ultrasuper
critical cycle No. 5, 6 and SCO 2
cycle, No. 8 were assumed by using
reference literature [9]. The model to
be explained below was defined as
reference model, and then the present
authors will plan to develop a model
that considers a variety of heat
efficiency improvement method. In
the present study, in the concept
similar to general Rankine cycle’s
reheating cycle, bypass mode was
proposed.
To begin with, the reference model
is as follows. After 910 °C helium fluid
discharging from VHTR carries out
heat exchange with heat exchanger 1,
hydrogen is produced by receiving
heat from high temperature helium
fluid in the heat exchanger 1. 846 °C
helium fluid passing heat exchanger 1
enters into steam generator 2 and go
through heat exchange. The fluid of
this steam generator is ultra-super
critical state water, and produces
power. The temperature of helium
fluid that passes through steam
generator 2 is 614.8 °C, this helium
fluid enters into heat exchanger 3
where heat exchange is carried out.
The fluid of this heat exchanger is
super critical-state carbon dioxide,
and it produces power by the heat
supplied. The temperature of helium
fluid coming out of heat exchanger 3
is 450 °C. The heat output that is produced
in heat exchanger 1 producing
hydrogen is 37.37 MWt. The mass flow
of helium from IHX is m 1 , and the
mass flow of water flowing in heat
exchanger 1 is m 2 , the mass flow of
water flowing in steam generator 2 is
m 3 , and the mass flow of CO 2 flowing
in heat exchanger 3 is m 4 . In this
study, the temperatures and pressures
from No.1 to No.10 in Figure 1 were
assumed, and m 1 and m 2 were calculated
by using the Equation (1), and
m 3 and m 4 were calculated by using
the Equation (2). Besides, considering
the characteristics of general longitudinal
temperature difference of heat
exchanger, the temperature at No. 6
and No. 9 was assumed to decrease by
10 °C compared to the temperature at
No. 4 and No. 7 of the steam generator
inlet.
Major equation or relationship for
heat equilibrium analysis is as follows:
• Equation used for calculating m 1
and m 2
: W = m∆h = m(h out – h in )... (1)
Here,W : Thermal power (MWt)
m : Mass flow (kg/hr)
h : Enthalpy (kJ/kg)
in : Entrance of the equipment
out : Outlet of equipment
• Equation used for calculating m 5
and m 8
∑m in h in = ∑m out h out ... (2)
In the case of hydrogen production, it
was assumed that all heat was
converted to work required, and in
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Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park
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OPERATION AND NEW BUILD 232
the case of power production, it was
assumed that only a part of the heat
delivered was converted to electricity.
Besides, it was considered that the
pumping power was consumed due to
the flow in the power generation,
therefore it was considered in the
calculation of efficiency.
The general efficiency of USC cycle
and SCO 2 cycle was 43 % and 45 %,
respectively. Using Equation (3),
efficiency was corrected, and more
realistic calculation was carried out
[6].
• efficiency correction equation
η Oper = [1.0+{(T h,oper – T h,des ) × C}]
× η Des … (3)
Here, η Des : standard efficiency
according to reference
literature
η Oper : Standard efficiency’s
correction efficiency according
to high temperature
T h,des : Standard exit temperature in
Steam Generator tube according
to reference literature
T h,oper : Exit temperature within
specified range at Steam
Generator or Heat exchanger
tube
C : Efficiency correction factor;
USC : 0.3 % / 5 °C [6],
SCO 2 : 1.0 % / 5 °C applied
(assumption)
The output of steam generator 2 and
heat exchanger 3 is as follows, and
total output W gross is the sum of all the
values.
W 1 = m 2 × (h 3 – h 2 ) kJ/hr
W 2 = η 2,Operator × m 3 ×(h 4 – h 7 ) kJ/hr
W 3 = η 3,Operator × m 4 ×(h 7 – h 10 ) kJ/hr
Here, w pump : Work used
in the pump (MWt)
η pump : Pump efficiency
v : Specific volume (m 3 /kg)
P : pressure (kPa)
In order to simulate the above model,
the flow of helium gas, water, and
carbon dioxide was calculated by
using thermodynamic system analysis
software, EES (Engineering Equation
Solver).
The following is regarding IHX
loop model to which Bypass mode was
added. Bypass mode was added to the
existing IHX loop, and the efficiency
improvement of overall heat conversion
cycle was studied. The temperature
of the entrance of evaporator 2
and heat exchanger 3 was reheated
by using high temperature helium
coming from IHX, and the output
change was studied.
In the same way as the existing IHX
loop, VHTR supplies heat generated
by nuclear fission in 350 MWt nuclear
reactor. The fluid coming from IHX is
helium, and the fluid flowing in heat
exchanger 1 is water, the fluid flowing
in steam generator 2 is ultra-supercritical-state
water, and the fluid
flowing in heat exchanger 3 is super
critical-state carbon dioxide.
The mass flow of helium from
IHX is m 1 , and m 1 is divided into m 2
and m 3 , and m 3 enters into heat
exchanger 1, and do heat exchange
with the water flowing in heat exchanger
1. At this time, the mass flow
of water flowing in heat exchanger 1
is m 5 . and m 2 is divided into m 4 and
m 9 , and m 4 enters into No. 7 in order
to reheat helium that went through
heat exchange in heat exchanger 1,
and the reheated temperature is that
of No. 8. m 9 enters No. 12 in order to
reheat helium that went through heat
exchange in the steam generator 2,
and the reheated temperature is the
temperature of No. 13. The mass flow
in steam generator 2 is m 10 , and mass
flow of CO 2 in heat exchanger 3 flowing
through heat exchanger 3 is m 14 .
The temperature and pressure at
every point except No. 12 were assumed,
and m 1 value was obtained by
using Equation (1) in the same as m 1
of the existing IHX Loop.
At this time, the temperature of
No.7 and No.12 were assumed to be
that of No. 4 and No. 7 of the existing
IHX loop. Besides, it was assumed that
the temperatures of No. 8 and No. 13
increased to 860 °C and 620 °C,
respectively due to m 4 and m 9 . By
this, the change of the output and
efficiency on the cycle of steam generator
2 and heat exchanger 3. Heat
exchanger 1 in accordance with the
addition of Bypass mode was assumed
to produce the same output, 37.37
MWt, as the existing IHX loop, and
fixed m 5 value.
Considering the characteristics of
the general longitudinal temperature
difference of the heat exchanger as
the existing IHX loop, it was assumed
that the temperatures of No. 11 and
No. 15 decrease by 10 °C compared to
that of No. 8 (at Steam Generator
entrance) and No. 13 (at Heat
Exchanger). With the obtained m 1
value, m 14 value was calculated by
using Equation (2). After that, m 5
value was fixed to the value which can
make the output as the existing IHX
loop, then m 3 value was obtained
by Equation (2). m 2 was calculated by
m 2 = m 1 – m 3 , and m 4 was obtained by
using Equation (2). m 9 was obtained
Here, W 1 : heat exchanger 1
heat output (MWt)
W 2 : Steam generator 2
heat output (MWt)
W 3 : Heat exchanger 3
heat output (MWt)
In steam generator 2 and heat
exchanger 3 in order to consider
pumping power in accordance with
mass flow, pump’s efficiency (η pump )
was assumed to be 0.9, and Equation
(4) was used.
w pump = η pump × m × v out
× (P out – P in ) kJ/hr … (4)
W net = W gross – w pump
| | Fig. 2.
Bypass mode-added IHX loop Modelling.
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by m 9 = m 2 – m 4 , and the temperature
of No. 12 can be calculated by using
Equation (2). Finally, m 10 was also
obtained by using Equation (2). Like
the existing IHX loop, in the bypass
mode-added IHX loop, correction
efficiency and pumping power in
accordance with mass flow were
considered, and pumping power used
the above Equation (4). For the
simulation for this, thermodynamic
system analysis software, EES, was
used, in the same way with the
existing IHX loop model obtained
before, and the flow of helium gas
and fluid was analyzed.
3 Result
Table 1 shows the result of physical
value at each point by simulating the
existing IHX loop EES [5]. Physical
value of each point was assumed in
accordance with reference [8], [9]
literature, and the assumed values
were colored.
The temperature of helium fluid
that leaves from the first heat exchanger
after producing the hydrogen
decreases to 846 °C from 910 °C, and
the temperature of helium fluid that
leaves from the second steam generator
is 614.8 °C, and the temperature
of helium fluid that leaves from the
last heat exchanger is 450 °C. The
temperature of helium fluid decreases
steadily, but because the fluid flowing
each steam generator and heat
exchanger is different, efficient electricity
can be produced by using each
characteristics. The existing IHX
Loop’s m 5 and m 8 are in inverse
proportion, as more mass flow moves
toward high efficiency, the amount of
overall electricity output increases.
Although the efficiency of heat
conversion cycle connected to each
steam generator may be influenced by
various causes, but in the present
study, correction factor presumed
about high temperature was used, so
the detailed design for this part would
be needed.
If heat conversion cycle connected
to each steam generator should be
operated simultaneously by a specific
objective, considering the inverse proportion
relationship between m 5 and
m 8 , the output must be distributed.
Besides, the exit temperature at the
tube part of steam generator 2 is in
inverse proportion with the exit
temperature at the shell part. This will
eventually influence on the exit
temperature of the tube part of the
heat exchanger 3. When operating
heat conversion cycle connected to
each steam generator, it is necessary
to find balanced point on the temperature
between steam generators.
In the steam generator 2 and heat
exchanger 3, exit temperature and
mass flow are in inverse proportion.
This is because if high exit enthalpy is
maintained in order to deliver the
same heat energy, less mass flow is
needed, and if a large amount of mass
flow is needed, exit enthalpy should
be maintained low. Maximum output
would be in the parabolic form as exit
enthalpy and temperature change, so
if maximum output is needed, proper
exit temperature must be selected. Or
in case there is a requirement for exit
temperature, it is possible that output
would be determined according to
that.
Table 2 shows the physical value at
each point where IHX loop added by
bypass mode is simulated with EES.
No. Fluid Temperature
(°C)
| | Tab. 1.
IHX loop Simulation Result.
In the case of IHX loop to which
bypass mode was added, the helium
fluid that passed through the first
hydrogen-producing heat exchanger
is 910 °C~ 846 °C, which is the same
as the existing IHX loop, but here by
reheating high-temperature helium
fluid, the temperature increases to
860 °C. The temperature of the helium
fluid that passed through the second
steam generator is 614.8 °C, which
is the same as that of helium fluid
that passed through the second evaporator.
However, since the temperature at
the entrance reheated, and returned,
the amount of electricity output produced
increases. When helium fluid
enters the third heat exchanger, it is
reheated from 614.8 °C to 620 °C, the
temperature of helium fluid is 450 °C,
and the amount of electricity output
Pressure
(kPa)
| | Tab. 2.
Result of IHX Loop to which Bypass Mode was added.
Enthalpy
(kJ/kg)
Mass flow
(kg/hr)
1 Helium 910.0 4000 6,161.00 527,662
2 Water 193.0 18,000 828.70 49,839
3 Water 585.0 16,500 3,528.00 49,839
4 Helium 846.0 4,000 5,829.00 527,662
5 Water 260.2 20,790 1,134.00 208,359
6 Water 836.0 16,475 4,174.00 208,359
7 Helium 614.8 4,000 4,628.00 527,662
8 CO 2 203.5 19,760 96.59 902,043
9 CO 2 604.8 19,290 597.00 902,043
10 Helium 450.0 4,000 3,773.00 527,662
No. Fluid Temperature
(°C)
Pressure
(kPa)
Enthalpy
(kJ/kg)
Mass flow
(kg/hr)
1 Helium 910.0 4,000 6,161.00 527,662
2 Helium 910.0 4,000 6,161.00 122,749
3 Helium 910.0 4,000 6,161.00 404,913
4 Helium 910.0 4,000 6,161.00 113,375
5 Water 195.0 18,000 837.50 50,000
6 Water 585.0 16,500 3,528.00 50,000
7 Helium 846.0 4,000 5,829.00 404,913
8 Helium 860.0 4,000 5,901.00 518,288
9 Helium 910.0 4,000 6,161.00 9,374
10 Water 260.2 20,790 1,134.00 214,580
11 Water 850.0 16,475 4,209.00 214,580
12 Helium 614.8 4,000 4,628.00 518,288
13 Helium 620.0 4,000 4,655.00 527,662
14 SCO 2 203.5 19,760 96.59 918,581
15 SCO 2 610.0 19,290 603.50 918,581
16 Helium 450.0 4,000 3,773.00 527,662
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OPERATION AND NEW BUILD 234
increases compared to the existing
IHX loop.
The following is the major comparison
of the result of reference
model and bypass mode model.
Mass flow IHX loop Bypass Mode Loop
Heat Exchanger 1 49839 50000
Steam Generator 2 208359 214580
Heat Exchanger 3 902043 918581
| | Tab. 3.
IHX Loop and Bypass Mode IHX Loop: Mass Flow Comparison.
Pumping power
When IHX loop and bypass modeadded
IHX loop were compared, mass
flow of m 10 and m 14 in the bypass
mode-added loop was greater compared
to the mass flow of m 5 and m 8 in
the existing loop, as shown in Table 3.
IHX loop
(MWt)
| | Tab. 4.
IHX Loop and Bypass Mode IHX loop: Pumping Power Comparison.
As shown in Table 4, depending on
the difference of mass flow value,
pumping power used in the pump also
can be high in steam generator 2 and
heat exchanger 3. However, although
pumping power is higher in the bypass
IHX loop, by reheating, efficiency of
steam generator 2 increased from
53.79 % to 54.4 %, and that of heat
exchanger 3 increased from 45.83 %
to 45.87 %; accordingly, it is seen that
the value of Power increased. As a
result, Net Power that considered
pumping power in Total Power was
178.6 MWt in the present IHX
loop, but the IHX loop to which
bypass mode was added increased
to 185.3 MWt, as shown in Table 5.
| | Tab. 5.
IHX loop and Bypass mode IHX loop: Power Comparison.
Bypass mode loop
(MWt)
Steam Generator 2 1.091 1.123
Heat Exchanger 3 9.82 10
Power
IHX loop
(MWt)
Bypass mode loop
(MWt)
Heat Exchanger 1 37.37 37.37
Steam Generator 2 94.63 99.7
Heat Exchanger 3 57.46 59.34
Net Power 178.6 185.3
Since the assumption was that
constant heat was supplied from the
primary system, it is seen that the
efficiency of the bypass model where
net power is high, and it is judged that
efficiency optimization model can be
formulated by detailed design.
4 Conclusion
In this study, VHTR system was
modelled for supplying high temperature
heat, by distribution, produced in
the high temperature gas furnace to
hydrogen producing equipment and
power generation equipment.
Provided that high temperature
gas- cooled reactor is located in
primary system, the secondary system
where hydrogen production and
power supply are possible were
explained. The helium that flows in
the nuclear reactor first passes
through the HX (heat exchanger)
whose purpose is the production of
hydrogen, and secondly and thirdly
pass through the steam generator
composed of super critical carbon
dioxide cycle, and heat exchanger,
respectively, producing the process
heat and power. In order to analyze
existing IHX loop model and bypass
mode-added IHX loop model, the
present authors studied the input &
output conditions and output change
of each steam generator and heat
exchanger, and based on this result,
by designing IHX loop in the power
production part in detail, the authors
performed the calculation of thermodynamic
physical value and efficiency
at each point. Additionally, the
authors studied the change regarding
electricity output and efficiency
according to bypass mode, when
reheating cycle is added, the possibility
on the efficiency optimization
was proposed.
References
[1] Kim. Y. W., 2015, Nuclear Hydrogen
Production Technology development
Using Very High Temperature Reactor,
Trans. Korean Soc. Mech. Eng. C, Vol. 3,
No. 4, pp. 299~305.
[2] Chang. J. H., 2006, Current Status of
Nuclear Hydrogen Development,
Journal of Energy Engineering, Vol.15,
No.2, pp. 127~137.
[3] Lee. S. I., 2015, Heat Balance Study
on Integrated Cycles for Hydrogen
and Electricity Generation in VHTR,
Transaction of the KNS Spring Meeting.
[4] Sung. H. C., 2012, Development of
Ultra-Supercritical (USC) Power Plant,
Trans. Korean Soc. Mech. Eng. B,
Vol. 36, No.2, pp.205~210.
[5] Yeom Chung-seop, Im Dong-ryeol,
Lee Jung-ik, 2014, Trend of Electricity
Generation Technology using supercritical
CO 2 , Institute for Advanced
Engineering, KIC News, Volume 17,
No.1.
[6] K.C.Cotton,1998, Evaluating and
Improving Steam Turbine Performance,
2 nd edition, Cotton Fact Inc.
[7] F-Chart Software, 2016,
Engineering Equation Solver,
http://www.fchart.com/ees/
[8] NGNP Conceptual Design Report/Steam
Cycle Modular Helium Reactor
(SC-MHR) Demonstration Plant,
Table 3-6 SC-MHR Conceptual Design
Point Design Parameter.
[9] SangIL Lee, Yeon Jae Yoo, Gyunyoung
Heo, Soyoung Park, Yeon Kwan Kang,
Heat Balance Study on Integrated
Cycles for Hydrogen and Electricity
Generation in VHTR-Part 2, Korean
Nuclear Society Autumn Meeting,
Oct 28-30, 2015.
Authors
SangIL Lee
YeonJae Yoo
Deok Hoon Kye
Department of Nuclear Team
Power & Energy Plant Division
Hyundai Engineering Company
Seoul, Korea
Gyunyoung Heo
Eojin Jeon
Soyoung Park
Department of Nuclear
Engineering
Kyung Hee University
Yongin Korea
Operation and New Build
Heat Balance Analysis for Energy Conversion Systems of VHTR ı SangIL Lee, YeonJae Yoo, Deok Hoon Kye, Gyunyoung Heo, Eojin Jeon and Soyoung Park
atw Vol. 63 (2018) | Issue 4 ı April
Informationsbedarf versus Geheimhaltungspflichten –
Erweiterung des In-camera-Verfahrens geplant
235
Tobias Leidinger
Dem berechtigten Anspruch der Öffentlichkeit auf detaillierte Informationen über ein atomrechtlich genehmigungsbedürftiges
Vorhaben steht das staatliche Interesse an einem effektiven Geheimnisschutz sensibler Daten
gegenüber. Dieser Konflikt tritt regelmäßig im Genehmigungsverfahren aber auch vor Gericht zu Tage. Die differenzierten
Gesetzesbestimmungen, die den Ausgleich dieser widerstreitenden Interessen regeln, sollen nun durch eine
weitere Facette ergänzt werden: Ein erweitertes In-camera-Verfahren bei Gericht. Nach dem Koalitionsvertrag vom
12. März 2018 soll die Regelung in der schon laufenden 18. Legislaturperiode erfolgen.
I Grundkonflikt Informationsbedarf vs.
Geheimhaltungspflicht
In atomrechtlichen Genehmigungsverfahren zeigt sich
regelmäßig ein Grundkonflikt: Dem Interesse der Öffentlichkeit
an möglichst vertieften Informationen über
alle sicherheits- und sicherungsrelevanten Aspekte des
Vorhabens steht das Erfordernis eines effektiven Geheimnisschutzes
in Bezug auf sensible Daten gegenüber.
Genauer betrachtet lassen sich für beide Pole Grundrechtspositionen
anführen: Einerseits ist Information Voraussetzung
für Transparenz und Teilhabe der Öffentlichkeit
am Genehmigungsverfahren. Das Verfahren dient der
Gewährleistung materieller Schutzansprüche Dritter.
Ohne Information ist Kontrolle gegenüber der Verwaltung
kaum realisierbar. Information ist die Grundlage für
Partizipation und Teilhabe der Öffentlichkeit an einem
Verfahren. Das BVerfG bringt dies mit der Formel „Grundrechtsschutz
durch Verfahren und Teilhabe an Information“
auf den Punkt.
Für die andere Seite, dem Interesse an Geheimhaltung
sensibler Daten, lassen sich aber nicht minder gewichtige
Grundrechtsinteressen anführen: Die Geheimhaltung
dient ebenfalls zum Schutz der Grundrechtsträger: Ist der
Staat zum Schutz der Grundrechte („Leben, Gesundheit“)
seiner Bürger verpflichtet, bedarf es des Geheimnisschutzes
in Bezug auf sensible Daten, damit eine effektive
Terrorabwehr – gerade zum Schutz der Bürger – gewährleistet
bleibt. Die Nicht-Preisgabe sicherheits- und
sicherungsrelevanter Informationen ist mithin nicht
minder essentielle Voraussetzung für einen effektiven
Grundrechtsschutz der Bürger.
II Interessenausgleich durch differenzierte
Gesetzesregelungen
Der Gesetzgeber trägt zur Lösung dieser widerstreitenden
Interessen im atomrechtlichen Genehmigungsverfahren
bereits heute durch eine ganze Reihe differenzierter
Regelungen bei. Nach § 6 der Atomrechtlichen Verfahrensordnung
(AtVfV) sind nicht nur der Antrag, der Sicherheitsbericht
und eine Kurzbeschreibung des jeweils zu
genehmigenden Vorhabens für die Öffentlichkeit auszulegen,
sondern es besteht nach § 6a Abs. 2 Satz 1 und
Abs. 3 AtVfV die Möglichkeit, in Bezug auf das Vorhaben –
im Interesse der Sicherheit und Sicherung – geheimhaltungsbedürftige
Informationen durch eine Beschreibung
oder Inhaltsdarstellung zu ersetzen. Anstelle einer
„Schwärzung“ von Unterlagen – die letztlich eine „Verweigerung“
von Information bedeutete –, tritt so die
Möglichkeit, geheimhaltungsbedürftige Informationen zu
umschreiben, so dass der Dritte in der Lage bleibt, seine
Betroffenheit durch das Vorhaben gleichwohl erkennen
und beurteilen zu können.
Eine Einschränkung von Informationsansprüchen ist
auch jenseits dieser Regelung möglich: Während eines
atomrechtlichen Verfahrens besteht der Anspruch auf
Akteneinsicht gemäß § 6 Abs. 4 AtVfV i.V.m. § 29 Abs. 1
S. 3, Abs. 2 und 3 des Verwaltungsverfahrensgesetzes
(VwVfG) nur nach Ermessen der Behörde (also nicht
„ unbedingt“). Informationen, die sicherheits- oder
sicherungsrelevant sind, weil sie den Ansatz für die Ausschaltung
von Sicherheits- und Sicherungsmaßnahmen
oder für die Identifizierung/Lokalisierung von Schwachstellen
eröffnen könnten, können – soweit durch ihre
Preisgabe ein „Nachteil zum Wohl des Bundes oder
Landes“ zu befürchten ist – von der Offenlegung ausgeschlossen
werden. Spezialgesetzlich ist die Geheimhaltung
von sensiblen Informationen im Sicherheitsüberprüfungsgesetz
(SÜG) geregelt. Besteht danach die Gefahr
eines „Nachteils“ oder wäre die Preisgabe der Information
sogar „schädlich“ für Bund oder Land, kann sie
nach Maßgabe der Verschlusssachen-Anweisung (VS-
Anweisung) durch den Geheimschutzbeauftragten der
Behörde als „Verschlusssache – Nur für den Dienstgebrauch“
oder sogar als „Verschlusssache – Vertraulich“
eingestuft und ihre Offenlegung verweigert werden. Was
nach Maßgabe des SÜG i.V.m. VS-Anweisung geheim zu
halten ist, darf auch nicht in anderem Zusammenhang
preisgegeben werden: So bestehen – auch außerhalb eines
atomrechtlichen Verfahrens – Informationsansprüche
Dritter, z.B. auf Herausgabe von umweltrelevanten
Informationen gegen die Genehmigungsbehörde nach
Umweltinformationsgesetz (UIG) oder – soweit die
Informationen nicht umweltrelevant sind – nach Maßgabe
des Informationsfreiheitsgesetzes (IFG). Pressevertreter
können sich darüber hinaus auch auf das jeweilige
Landes-Pressegesetz stützen. In all diesen Fällen besteht
indes die Möglichkeit – mit oder ohne ausdrücklichen
Bezug auf das SÜG –, dass sicherheits- und sicherungsrelevante
Informationen im Ergebnis nicht offenbart
werden müssen, wenn die materiellen Schutzvoraussetzungen
nach SÜG i.V.m. der VS-Anweisung vorliegen.
III In-camera-Verfahren de lege lata und
de lege ferenda
Verweigert die atomrechtliche Genehmigungsbehörde die
Herausgabe sensibler Informationen unter Verweis auf
den Geheimschutz auch im Gerichtsverfahren, – in dem
z.B. über die Rechtmäßigkeit einer atomrechtlichen
Genehmigung gestritten wird – so sieht die bislang
existierende Gesetzesregelung zum sog. In-camera-
Verfahren in § 99 Verwaltungsgerichtsordnung (VwGO)
vor, dass über die Frage der Geheimhaltungsbedürftigkeit
ein speziell besetzter Fachsenat vorab entscheidet. Ihm
sind ausschließlich die geheimhaltungsbedürftigen Akten
vorzulegen („in camera“), um zu prüfen, ob die Einstufung
als „geheim“ zurecht erfolgt ist und daher die Verweigerung
der Aktenvorlage durch die Behörde Bestand hat
oder nicht. Nur wenn die Geheimhaltungsbedürftigkeit
verneint wird, ist die vorenthaltene Information dem
Verwaltungsgericht zugänglich zu machen. Nur dann
kann es darauf zugreifen und seine Entscheidung darauf
stützen.
SPOTLIGHT ON NUCLEAR LAW
Spotlight on Nuclear Law
Information Requirements Versus Confidentiality Obligations – Extension of the In-Camera Procedure Planned ı Tobias Leidinger
atw Vol. 63 (2018) | Issue 4 ı April
SPOTLIGHT ON NUCLEAR LAW 236
Der Koalitionsvertrag vom 12. März 2018 (vgl. Seite
141) sieht nun vor, dass die Regelungen für das In-camera-
Verfahren für das Atomrecht dahingehend erweitert
werden sollen, dass geheimhaltungsbedürftige Unter lagen
auch zum Zwecke des Nachweises der Genehmigungsvoraussetzungen
in ein verwaltungsgerichtliches Hauptsacheverfahren
– bei gleichzeitiger Wahrung des Geheimschutzes
– eingeführt werden können. Das In-camera-
Verfahren dient dann nicht (mehr allein) zur Klärung der
Frage der Geheimhaltungsbedürftigkeit einer Unterlage
(wie bisher), sondern ermöglicht darüber hinaus eine
weitergehende Prüfung in der Sache durch das Gericht.
Das Gericht prüft dann auch, ob der erforderliche Schutz
gegen Störmaßnahmen Dritter (SEWD) als gegeben unterstellt
werden darf oder nicht. Die Gewährleistung des
SEWD-Schutzes ist eine wesentliche Tatbestandsvoraussetzung,
die erfüllt sein muss, damit eine atomrechtliche
Genehmigung erteilt werden kann. Dabei ist aber auch in
einem erweiterten In-camera- Verfahren sicherzustellen,
dass die behördliche Ein schätzungsprärogative in Bezug
auf genehmigungs relevante Wertungen bei Sicherheit und
Sicherung beachtet werden. Das bedeutet, dass das Gericht
sich nicht an die Stelle der Behörde setzen darf, also eine
eigene Entscheidung anstelle der Behörde trifft, sondern
bei seiner Nachprüfung auf eine Vertretbarkeitskontrolle
beschränkt bleibt.
die Frage, ob im Ergebnis davon ausgegangen werden darf,
dass die erforderliche Schadensvorsorge und der gebotene
Schutz gegen SEWD-Ereignisse gewährleistet ist oder
nicht, könnte auf diese Weise weitergehend als bisher entschärft
werden. Idealerweise bliebe der Geheimnisschutz
auch so gewahrt, zugleich aber wäre dem Interesse
der Drittbetroffenen an einer Überprüfung essentieller
Genehmigungsvoraussetzungen unter Berücksichtigung
geheimhaltungsbedürftiger Informationen weitergehend
als bisher entsprochen. Das wäre als konstruktiver Beitrag
zur Stärkung eines effektiven Grundrechtsschutzes zu
bewerten: Ein erweitertes In-camera-Verfahren diente so
in besonderer Weise zur Gewährleistung der dem Dritten
zustehenden Schutzansprüche und wahrte dabei
gleichwohl den erforderlichen Geheimschutz, der nicht
minder einem effektiven Grundrechtsschutz der Bürger
geschuldet ist.
Allerdings bleiben die konkrete Ausgestaltung und der
Vollzug dieser Regelung in der Praxis abzuwarten: Folgt
einer guten Absicht des Gesetzgebers eine in der Praxis
tatsächlich und rechtlich brauchbare Lösung? Ziel müsste
sein, dadurch nicht neue Fragen zur Anwendung und
Reichweite eines erweiterten In-camera-Verfahrens
aufzuwerfen, sondern eine inhaltlich klare und hinreichend
bestimmte Norm zu schaffen, die das Versprechen
des Koalitionsvertrages vollzugsfähig einlöst.
IV Erweiterung des In-camera-Verfahrens:
Bedenkenswerter Schritt
Die Absicht, das In-camera-Verfahren in Bezug auf die
Prüfung materieller Genehmigungsvoraussetzungen zu
erweitern, ist ein bedenkenswerter Ansatz. Der Streit über
Autor
Prof. Dr. Tobias Leidinger
Rechtsanwalt und Fachanwalt für Verwaltungsrecht
Luther Rechtsanwaltsgesellschaft
Graf-Adolf-Platz 15
40213 Düsseldorf
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Prof. Dr. Marco K. Koch
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Ulf Kutscher
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Dr. Jens Schröder
Dr. Wolfgang Steinwarz
Prof. Dr. Bruno Thomauske
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Dr. Hannes Wimmer
Ernst Michael Züfle
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atw Vol. 63 (2018) | Issue 4 ı April
238
ENVIRONMENT AND SAFETY
CFD Modeling and Simulation
of Heat and Mass Transfer in
Passive Heat Removal Systems
Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
This paper is presenting the CFD-modelling and simulation of condensation inside passive heat removal systems.
Designs of future nuclear boiling water reactor concepts are equipped with emergency cooling systems which are
passive systems for heat removal. The emergency cooling system consists of slightly inclined horizontal pipes which are
immersed in a tank of subcooled water. At normal operation conditions, the pipes are filled with water and no heat
transfer to the secondary side of the condenser occurs. In the case of some accident scenarios the water level may
decrease in the core, steam enters the emergency pipes and due to the subcooled water around the pipe, this steam
condenses. The emergency condenser acts as a strong heat sink which is responsible for a quick depressurization of the
reactor core. This procedure acts passive i.e. without any additional external measures. The actual project is defined to
model the phenomena which are occurring inside the emergency condensers. The focus of the project is on detection of
different morphologies such as annular flow, stratified flow, slug flow and plug flow and also modeling of the laminar
film which is occurring during the condensation near the wall.
The condensation procedure inside the
pipe is determined by two important
phenomena. The first one is wall
condensation and the second one is the
direct contact condensation (DCC).
The Algebraic Interfacial Area Density
(AIAD) concept is used in order to
model the interface between liquid
and steam. In the next steps the Generalized
Two-Phase Flow ( GENTOP)
model will be used to model also the
dispersed phases which are occurring
inside the pipe. Finally, the results of
the simulations will be validated by
experimental data which will be available
in HZDR. In this paper the results
of the first part are presented.
1 Introduction
Condensation plays a crucial role in
the emergency condenser of passive
heat removal systems of nuclear power
plants. Passive safety systems do not
need any external power supplies and
they mostly depend on physical phenomena
such as natural circulation
and gravity driven flows. In order to
assess the performance of passive safety
systems and their efficiency mostly
one-dimensional codes are used such
as ATHLET, RELAP and TRACE. These
codes are able to calculate most of the
phe nomena in power plants; however,
they cannot reflect the 3D phenomena.
Therefore, Computational Fluid
Dynamics (CFD) methods should be
used to simulate and predict the
complex multiphase flow structure.
Despite the previous research being
done on the two-phase flow behavior,
this phenomenon needs much more
investigations. The two-phase flow
patterns and transition between vapor
and liquid are studied by Thome and
Hajal et al. [1, 2]. They introduced a
logarithmic mean void fraction (LMe)
method in order to calculate the vapor
void fractions which change from the
low pressure up to the critical pressure
point. Moreover, they proposed a new
heat transfer model based on the same
simplified flow structures that have
been used in the flow boiling model
of Kattan et al. [3]. The model can
predict the local condensation heat
transfer coefficient for different flow
regimes such as annular, intermittent,
stratified-wavy fully stratified and
wavy flow.
Many attempts have been done to
investigate the mass transfer between
liquid and gas phase in condensation.
Lee et al. [4] introduced a model for
prediction of the mass transfer. They
assumed that the interface between
liquid and steam is on saturation
temperature and introduced an
iterative technique in order to reach to
desired boundary condition inside
each cell. This model depends on a
relaxation factor which needs to be
tuned. The tuning needs many trial
and error simulations which is
time-consuming and doesn’t have any
predictive capabilities.
Moreover, there are empirical or
semi-empirical methods to calculate
the mass transfer in the interface.
Strubelj et al. [5] by using ANSYS CFX
and NEPTUNE_CFD [6] code tried to
simulate Direct Contact Condensation
(DCC) in stratified flows. In DCC the
phase change occurs due to the direct
contact interaction of subcooled water
and saturated steam. The defined
phase change mass flux depends on
thermal conductivity of the liquid and
Nusselt number of the liquid. The
Nusselt number was calculated
by Coste et al. [7] based on Surface
Renewal Theory (SRT) [8]. The SRT
theory calculates the mass transfer
according to the renewal period of
eddies and the liquid turbulent
properties. Hughes and Duffey [9]
used the surface renewal theory and
the Kolmogorov turbulent length
scale theory to define a correlation for
the heat transfer coefficient. They
considered that the heat removal from
interface occurs by smallest turbulent
scales. This model will be introduced
more detailed in the next sections.
This correlation is validated for
Pressurized Thermal Schock (PTS)
phenomenon by Egorov [10] and
Apanasevich [11]. Further to Hughes
correlation, Shen et al. [12] developed
another correlation for calculation of
heat transfer coefficient based on the
surface renewal theory. Ceuca et al.
[13] used both of these correlations
in order to simulate the direct contact
condensation for the LAOKOON
facility [14]. By comparison of Hughes
and Duffey correlation with Shen
correlation, Ceuca et al. [13] concluded
that both of the models provide
accurate results for the horizontal
stratified quasi-steady state.
Evidently, many attempts have been
done in the modeling of con densation
inside the pipes. The goal of the current
work is modeling of the transition
between different mor phologies which
are occurring during the condensation
inside the pipe ( Figure 1). In order to
do that, several CFD models such as
IMUSIG, AIAD and GENTOP which
have been developed in HZDR in cooperation
with ANSYS are available. The
Inhomo geneous MUSIG model considers
the bubble size distribution and
is used for modeling the small-scaled
dispersed gas phase [15]. The AIAD
Environment and Safety
CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems
ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
atw Vol. 63 (2018) | Issue 4 ı April
| | Fig. 1.
Schematic representation of horizontal flow and different morphologies.
(Algebraic Interfacial Area Density
Model) is developed for detection of
the local morphology and corresponding
switch between them [16]. The
recently developed GENTOP- model
combines both concepts. GENTOP
(Generalized Two-Phase Flow) approach
is able to simulate co-existing
large-scaled (continuous) and smallscaled
(polydispersed) structures [17].
All these models are validated for adiabatic
cases without any phase change.
Therefore, the start point of the current
work project is using the available
models and integrating phase transition
and con densation models into
them. In the current work as initial
stages the AIAD model has been used
since in this model 2 continues phases
should be considered and it is less complicated
compare to GENTOP model
which also considers a poly- dispersed
phase. In the proceeding sections a
more detail explanation of AIAD model
will be given.
2 CFD model formulation
In the current work a multi-field twophase
CFD approach is used with
ANSYS CFX 17.2 in order to simulate
the condensation inside horizontal
pipe flows. The mass, momentum and
energy equations can be defined,
respectively, as follow:
• Mass conservation equation:
(1)
where S Mi describes user specified
mass source.
χ iβ the mass flow rate per unit volume
from phase β to phase i.
• Momentum conservation equation:
(2)
where S mi is the momentum source
caused by external body forces
and user defined momentum
sources.
M i is the interfacial forces acting
on phase i due to the presence
of other phases.
χ + iβ v β – χ + βi v i is the momentum
transfer induced
by mass transfer.
• The total energy equation:
(3)
where: h tot is the total enthalpy
related to static enthalpy by:
(4)
T i , λ i represents the temperature
and the thermal conductivity
of phase i.
S Ei describes external heat sources.
Q i is interphase heat transfer
to phase i across interfaces
with the other phase.
χ + iβ h βs – χ + βi h is denotes the interphase
mass transfer.
In ANSYS CFX in order to describe the
phase change which occurs due to the
interphase heat transfer, the Thermal
Phase Change Model has been introduced
[30]. This model is particularly
useful in simulation of the condensation
of saturated vapor. The heat
flux from the interface to phase i and
phase β is:
q i = h i (T sat – T i ) (5)
q β = h β (T sat – T β ) (6)
where h i , h β and T sat are heat transfer
coefficients of the phase i and phase
β and the saturation temperature,
respectively. ṁ iβ is the mas flux from
phase β to phase i. H is and H βs are the
interfacial enthalpy values which
come into and out of the phase due
to phase change which occurs. By
usage of the total heat balance
equation the interphase mas flux can
be determined as follow:
| | Fig. 2.
3D geometry of the pipe and mesh of the cross section.
(7)
ṁ iβ > 0 → H is = H i,sat , H βs = H β (8)
ṁ iβ < 0 → H is = H i , H βs = H β,sat (9)
In the current work, the steam
con sidered to be in saturation temperature.
Therefore, the heat flux
from the steam to the interface equals
zero since both are in saturation
temperature. As a result, the interphase
mass flux formula can be
written as:
(10)
In this work in order to model the heat
transfer coefficient the Hughes and
Duffy model has been used which is
based on the SRT model [9]. They
used the Surface Renewal Theory
(SRT) and the Kolmogorov turbulent
length scale theory to find a correlation
for heat transfer coefficient.
Therefore, the heat transfer coefficient
was derived as:
(11)
where ε is the turbulent dissipation, v l
is the kinematic viscosity and λ is the
thermal conductivity.
3 Computational grid and
boundary conditions
In Figure 2 the pipe and the boundary
conditions are shown. The pipe is
horizontal and has 1 m length
and 0.043 m diameter. In order to
define a mesh for the pipe ANSYS
ICEM software is used. Due to the
higher importance of the wall region
compare to the middle of the pipe,
the mesh near the wall needs to be
finer than the mesh in the pipe
center. The number of nodes is
1,250,000.
Mass flow rate
[Kg/s]
Temperature
(k)
Inlet 0.5 537.1
Wall - 312.18
outlet outflow -
ENVIRONMENT AND SAFETY 239
CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems
Environment and Safety
ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
atw Vol. 63 (2018) | Issue 4 ı April
ENVIRONMENT AND SAFETY 240
| | Fig. 3.
(a) Area averaged liquid volume fraction in different cross sections over the pipe length,
(b) Temperature distribution in the outlet of the pipe for 5 different radial lines.
4 Results
The results are obtained with the AIAD
approach for modeling the free surface
and morphologies. Moreover, the
Hughes correlation is used for the heat
transfer coefficient. Figure 3 represents
the qualitative profiles of liquid
volume fraction and tem perature. In
Figure 3 (a) the volume fraction profile
in the vertical cross section in the
middle of the pipe and in the streamwise
direction is represented. As it can
be seen at the inlet the pure steam
exists and by going further in the pipe,
due to the film condensation a liquid
film starts to generate near the wall.
The liquid film is growing and leads to
the thicker film. In a cross section
500 mm far from the inlet the liquid
film is falling down gradually and
gathering at lower part of the pipe.
The liquid film always exists near the
wall because as soon as the liquid is
falling down the steam becomes in the
direct contact with the wall and condenses
and again new film generates.
Moreover, in Figure 3 (d) the temperature
profile is shown for different
cross sections along the pipe. As
mentioned before, the steam is fixed
at the satu ration temperature, but
further along the pipe by generating
the liquid the temperature of the liquid
is decreasing because of the heat
flux to the wall. Moreover, the wall
heat flux is cooling the liquid which
causes the direct contact condensation
between liquid and steam interface. As
the steam is on the saturation temperature
there is no heat flux between
the interface which is also on the saturation
temperature and the steam.
Therefore, just the phase is changing
and the steam turns into the liquid.
| | Fig. 4.
(a) Liquid Volume fraction distribution on a cross section along the pipe, (b) temperature distribution on a cross section along the pipe,
(c) Volume fraction distribution on different cross sections, (d) temperature distribution on different cross sections.
Figure 4 (a) shows the change of
cross section averaged liquid volume
fraction along the pipe. According to
the figure the average liquid volume
fraction at the inlet is 0.0 and due to
the mass transfer it’s increasing along
the pipe and it reaches to around 0.1
at end of the pipe. Therefore, in a
horizontal pipe with one meter length
the total condensation rate is around
10 percent. In Figure 4 (b) the temperature
distribution for the five
radial lines on the outlet of the pipe
is presented. This plot shows the
temperature difference from the
center of the pipe towards the wall.
As the plot shows, in the center the
temperature is equal to the saturation
temperature. As far as getting closer to
the wall which is in subcooled
tem perature, the temperature gradient
is increasing. In other words, in
the region near the wall the temperature
difference from the saturation
temperature is higher. Moreover,
slope of the plot for L5 is higher than
L1. The reason is in lower part of the
pipe (which is showed by L5) the
amount of cooled liquid is higher
which causes higher temperature
gradient in the lower parts of the pipe.
As the pipe is symmetric and the
boundary conditions for both sides of
the pipe are same, there is no need to
plot the temperature distribution in
another half of the cross section.
5 Conclusion
The ANSYS CFX 17.2 has been used in
order to simulate the condensation
inside horizontal tubes. In order to
model the two phase flow, heat
transfer and phase change are included
in the available AIAD concept
which was developed for adiabatic
cases. Moreover, the Hughes heat
transfer coefficient correlation is
implemented for the modeling of the
direct contact condensation in the
interface. The changes of the flow
structure inside the pipe and the
volume fraction and the temperature
profiles have been studied in detail.
The liquid film which is generated
near the wall due to the wall condensation
is modeled and it can be seen in
the volume fraction profiles. By generating
the liquid film near the wall both
wall condensation and direct contact
condensation are occurring inside
the pipe at the same time. Whereas in
the actual paper only the test for
plausibility of the AIAD model was
done, in the near future the comparison
to the experiment is planned.
The next step which is an ongoing
part of the project is simulation of the
Environment and Safety
CFD Modeling and Simulation of Heat and Mass Transfer in Passive Heat Removal Systems
ı Amirhosein Moonesi, Shabestary, Eckhard Krepper and Dirk Lucas
atw Vol. 63 (2018) | Issue 4 ı April
whole condensation phenomena and
flow morphologies by using GENTOP
concept. Further to the AIAD concept
which considers two continuous
fluids, the GENTOP approach is a
three field two fluid model and considers
also a poly dispersed phase.
Acknowledgments
This project is an ongoing project in
Helmholtz-Zentrum Dresden Rossendorf
(HZDR), which is funded by Bundesministerium
für Bildung und Forschung
(BMBF) under grant number
02NUK041B in Germany.
References
[1] Hajal, J.El.; Thome, J.; Cavallini, A.
Condensation inside horizontal tubes,
part 1: two phase flow pattern map.
International Journal of Heat and Mass
Transfer 46: 3349-3363(2003).
[2] Thome, J.; Hajal, J.El; Cavallini, A.
Condensation inside horizontal tubes,
part 2: New heat transfer model based
on flow regimes. International Journal
of Heat and Mass Transfer 46: 3365-
3387(2003).
[3] Kattan, N.; Thome, J.R.; Favrat, D. Flow
boiling in horizontal tubes:part2-New
heat transfer data for five refrigerants.
J. Heat Transfer 120: 148-155 (1998).
[4] Lee, W. H. A Pressure Iteration Scheme
for Two-Phase Flow Modeling. Multiphase
Transport Fundamentals,
Reactor Safety, Applications: 407–432,
(1980).
[5] Štrubelj, L.; Ézsöl, Gy. ; Tiselj, I. Direct
Contact Condensation Induced
Transition from Stratified to Slug Flow.
Nuclear Engineering and Design 240:
266–274 (2010).
[6] Lavieville, J.; Quemerais, E.; Boucker, M.;
Maas, L., NEPTUNE CFD V1.0 User Guide
(2005).
[7] Coste, P. ; Pouvreau, J. ; Lavieville, J.;
Boucker, M. A Two-phase CFD approach
to the PTS problem evaluated on COSI
experiment. Proceedings of the 16 th
International Conference on Nuclear
Engineering ICONE16, USA, (2008).
[8] Banerjee, S.; A surface renewal model
for interfacial heat and mass transfer in
transient two-phase flow. International
Journal of Multiphase Flow, Vol.4:
571-573 (1978).
[9] Hughes, E. D.; Duffey, R. B. Direct
Contact Condensation and Momentum
Transfer in Turbulent Separated Flows.
Internal Journal of Multiphase Flow 17:
599–619 (1991).
[10] Egorov, Y. Validation of CFD codes with
PTS relevant test cases. Technical Report
EVOL-ECORA-D07, ANSYS, Germany
(2004).
[11] Apanasevich, P. ; Lucas, D.; Beyer, M.;
Szalinski, L. CFD based approach for
modeling direct contact condensation
heat transfer in two-phase turbulent
stratified flows. International Journal of
Thermal Sciences 95: 123-135(2015).
[12] Shen, L.; Triantafyllou, G.S.; Yue. D.K.P.
Turbulent diffusion near a free surface
Journal of Fluid Mechanics 407:
145–166 (2000).
[13] Ceuca, S. C. ; Macián-Juan R. CFD
Simulation of Direct contact Condensation
with ANSYS CFX using Locally
defined Heat Transfer Coefficient.
In ICONE-20, Anaheim, California, USA,
No. 54347 (2012).
[14] Goldbrunner, M.; Karl, J. ; Hein, D.
Experimental Investigation of Heat
Transfer Phenomena During Direct
Contact Condensation in the Presence
of Noncondensable gas by means of
Linear Raman Spectroscopy. In 10 th Int.
Symp. on Laser Techniques Applied to
Fluid Mechanics, Lisbon (2000).
[15] Krepper, E.; Frank, Th.; Lucas, D.; Prasser,
H.-M.; Zwart, P.J. The Inhomogeneous
MUSIG model for the simulation of
poly-dispersed flow. Nuclear Engineering
Design 238: 1690-1702 (2008).
[16] Höhne, T.; Deendarlianto; Lucas, D.
Numerical simulations of countercurrent
two-phase flow experiments in
a PWR hot leg model using an area
density model. International Journal
of Heat and Fluid Flow 31 (5):
1047-1056 (2011).
[17] Hänsch, S.; Lucas, D.; Krepper, E.;
Höhne, T. A multi-field two-fluid
concept for transitions between
different scales of interfacial structures.
International Journal of Multiphase
Flow 47:171-182(2012).
Authors
Amirhosein Moonesi Shabestary,
Eckhard Krepper,
Dirk Lucas
Helmholtz-Zentrum
Dresden-Rossendorf
P.O.Box 510119
01314 Dresden, Germany
241
DECOMMISSIONING AND WASTE MANAGEMENT
The Decommissioning of the ENEA RB3
Research Reactor in Montecuccolino
F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
The ENEA RB3 reactor was a 100 Wth research installation owned and operated by ENEA, in its center of Montecuccolino
near Bologna, from 1971 to 1989. It consisted of a cylindrical aluminium vessel, about 4.3 m high and 2.9 m in diameter,
which could host various types of fuel elements suspended from the top of a special adjustable rack and submerged into
moderating and cooling heavy water. Principal aim of the reactor was to provide neutronics data for the CIRENE NPP, a
SGHWR that was being designed and then partially built in Latina starting from 1979. The specific RB3 core, surrounded
by a graphite reflector and housed inside a concrete biological shielding, allowed to test easily very different fuel
element configurations by changing their pitches and by regulating the heavy water level inside the vessel. The reactor
design, similar to that of the ZED-II Canadian research facility, was originally developed by CEA for its Aquilon facility
in Saclay in 1956; in fact, through a special arrangement between ENEA and CEA, parts of the Aquilon facility were
ultimately donated to ENEA at the end of the 60s for the construction of RB3. In 1989, the RB3 reactor was shut down,
and in the late 2010 ENEA received by ministerial decree the authorization to its dismantling, with the aim of reaching
the “green field” status and with the unconditional release of its building, which is actually owned by the University of
Bologna. The dismantling activities started in May 2013 and were concluded at the end of 2014; after that, a campaign
for the radiological characterization of the building was initiated and concluded in June 2015. Now, all the necessary
site characterization activities are being conducted with the aim to present the results declaring the “green field” status
before the end of 2017. This paper will present the three main pillars of the decommissioning of RB3, namely the
strategy and methods for the dismantling, the strategy and methods for the radiological characterization of the building,
and finally the strategy and methods for the radiological characterization of the site. The radionuclide limits imposed
by the Italian Regulatory Body, together with the challenges encountered so far will be likewise shown and described.
Revised version of
a paper presented
at the Eurosafe,
Paris, France, 6 and 7
November 2017.
Decommissioning and Waste Management
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
atw Vol. 63 (2018) | Issue 4 ı April
DECOMMISSIONING AND WASTE MANAGEMENT 242
1 Introduction
The ENEA RB3 (Reattore Bologna 3)
reactor was a 100 Wth research installation
owned and operated by ENEA
in its center of Montecuccolino, near
Bologna, from 1971 to 1989. It consisted
of a cylindrical aluminium vessel,
about 4.3 m high and 2.9 m in diameter,
which could host various types of
fuel elements suspended from the top
of a special adjustable rack, and submerged
into heavy water serving both
as moderator and coolant. Principal
aim of the reactor was to provide
neutronics data for the CIRENE NPP, a
SGHWR that was being designed, and
then partially built in Latina, starting
from 1979. The specific RB3 core, surrounded
by a graphite reflector and
housed inside a concrete biological
shielding, allowed to test easily very
different fuel element configurations
by changing their pitches and by
regulating the heavy water level inside
the vessel. The reactor design, similar
to that of the ZED-II Canadian
research facility, was originally developed
by CEA for its Aquilon facility in
Saclay in 1956; in fact, through a
special arrangement between ENEA
and CEA, parts of the Aquilon facility
were ultimately donated to ENEA at
the end of the 60s for the construction
of RB3. In 1989, after more than 18
years of operation, the RB3 reactor
was shut down, and in the late 2010,
after waiting for the entry into force of
Legislative Decree (L.D.) 230/1995
[1], which introduced new laws for
the decommissioning of NPPs, ENEA
received by ministerial decree the
authorization to its dismantling, with
the aim of reaching the “green field”
status and with the unconditional
release of its building, including the
reactor concrete biological shielding,
which is actually owned by the
University of Bologna. In fact the site
of Montecuccolino, some 3.5 km to
the South of downtown Bologna,
hosted three research reactors: RB1,
owned and operated by the University
of Bologna, RB2, owned and operated
by AGIP Nucleare, and RB3, owned
and operated by ENEA. RB1 and RB2
were decommissioned up to the green
field status well before the entry into
force of L.D. 230/1995.
Figure 1 shows an aerial view of
the Montecuccolino research center,
with the area hosting RB3 contoured
in red. Figure 2 shows a plan of the
main reactor hall, with in red the
area once occupied by the reactor
vessel, surrounded by the hectagonal
graphite reflector and encased within
a thick concrete biological shielding.
Figure 3 shows a vertical section of
the RB3 building; the lowermost floor
hosted 4 large tanks for a total of
20,000 L (in red) for the storage of the
heavy water which was daily pumped
up into the vessel to reach criticality
and then drained after the conclusion
of the experiments. Three floors are
present in the building: floor +6.0 m
corresponding to the ground level,
floor +0.0 m, corresponding to the
level of the reactor vessel, and floor
-3.0 m, with the heavy water storage
tanks, heating and cooling systems,
and other auxiliaries. The control
| | Fig. 2.
Plan of main hall of RB3.
room was located at floor +0.0 m.
While allowed to operate up to 100
Wth, operations at RB3 were always
conducted at 50 Wth.
Between 1991 and 1992, all the
fuel elements used at RB3 were either
restituted at their owner (JRC Euratom
Ispra) or sent to the ENEA Research
Center of Saluggia or to the fuel fabrication
plant of Fabbricazioni Nucleari
at Bosco Marengo. Between 1992 and
1993 all the heavy water was transferred
to the ENEA Research Center of
Borgo Sabotino, and before the end of
1997 all the sealed radioactive sources
used at the plant were disposed of.
2 Regulatory Requirements
and Classification
of Components and
Materials
In the late 2010, ENEA received, by
decree of the Italian Ministry of
Economic Development, the authorization
[2] to proceed with the dismantling
of RB3; included in the
| | Fig. 1.
Aerial view of the Montecuccolino site; the RB3 building is inside the red square.
| | Fig. 3.
Section of the RB3 building.
Decommissioning and Waste Management
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
atw Vol. 63 (2018) | Issue 4 ı April
Radionuclide Metals Concrete Other
materials
Reused Recycle Both reuse and recycle Demolition
Surface
(Bq/cm 2 )
decree were also the requirements
imposed by the Italian Nuclear Regulatory
Body ISPRA. According to these
requirements, the components and
materials of RB3 were classified by
ENEA, taking into account the various
areas of the plant and the history of
its functioning, into four main
categories:
A) materials and components which
were both in contact with possibly
contaminated or activated process
fluids and subject to neutron flux;
B) materials and components which
were in contact with possibly
contaminated or activated process
fluids but not directly irradiated by
neutrons;
C) materials and components which
were irradiated by the neutron flux
but which never went into contact
with possibly contaminated or
activated process fluids;
D) s.c. “exempt” materials, which
were never irradiated and never
went into contact with possibly
contaminated or activated process
fluids.
The only component classified in the
A category was the aluminium vessel;
the only components in the B category
were the heavy water distribution
pipings. Exempt materials, given their
unirradiated and uncontaminated
status, were subject only to a general
screening through CANBERRA In Situ
Object Counting Systems (ISOCS) to
estimate any possible level of presence
of 60Co and 137Cs; if the measured
levels were below the decision threshold
of the measuring system in terms
of mass concentration levels, then
Surface
(Bq/cm 2 )
| | Tab. 1.
Surface or mass activity concentration levels for clearance.
Mass
(Bq/g)
these materials were automatically
discarded from the plant without any
further radiological analysis. This
demonstrates the “instrumental” zero
of this category of materials hence the
“exempt” classification. All materials
which had been classified as “exempt”
were released unconditionately, for a
total mass of about 30 tons, between
March 2013 and May 2015. For all the
other three categories, the clearance
levels imposed by the Regulatory
Authority are summarized in Table 1.
These were derived either from the
Italian L.D. n. 230/95 or from RP 89
[3] and RP 113 [4] publications. In
presence of more than one radionuclide,
the sum of the ratios of
the measured concentrations to the
respective levels must be lower than 1.
The components and materials
were further grouped by ENEA into 12
s.c. “homogeneous groups” using
material and historic criteria; homogeneous
groups are therefore constituted
by components (or parts of
them) made by the same material and
possibly with a homogeneous and
uniform activity content.
3 Radiological Characterization
of Homogeneous
Groups
Before the radiological characterization
of the batches of materials from
the various homogeneous groups
started, a preliminary, special campaign
was conducted to exclude the
presence of various isotopes among
those given in Table 1, expecially in
the most potentially activated or
contaminated materials (category A).
Surface
(Bq/cm 2 )
Mass
(Bq/g)
3 H 10,000 100,000 1 10,000 1 1
14 C 1,000 1,000 1 10,000 1 1
Mass
(Bq/g)
54 Mn 10 10 1 10 0.1 0.1
55 Fe 1,000 10,000 1 10,000 1 1
59 Ni 10,000 10,000 1 100,000 1 1
60 Co 1 10 1 1 0.1 0.1
63 Ni 1,000 10,000 1 100,000 1 1
90 Sr 10 10 1 100 1 1
125 Sb 10 100 1 10 1 1
134 Cs 1 10 0.1 10 0.1 0.1
137 Cs 10 100 1 10 1 1
152 Eu 1 10 1 10 0.1 0.1
154 Eu 1 10 1 10 0.1 0.1
Generic Alfa 0.1 0.1 0.1 0.1 0.1 0.01
241 Pu 10 10 1 100 1 1
In particular 54Mn, 59Ni, 90Sr,
125Sb, 134Cs, 137Cs, 239Pu, 240Pu
and 241Pu were excluded from
further analyses finalized to the unconditional
release of materials. Then,
for each homogeneous group, a precharacterization
measurement campaign
was con ducted with a three-fold
aim: 1) to verify if the hypothesis on
the homogeneity of activity for that
given group held; 2) to evaluate the
minimum number of samples to be
analized
for the subsequent characterization
phase; 3) to evaluate the value of
isotopic ratios of 55Fe to 60Co and
of 63Ni to 60Co, so to limit the next
analyses only to the research of 60Co
contents. After that, and using typically
13 multiple measurements for
each batch of each homogeneous
group, summations of the ratios
between measured activity concentrations
and limits (Table 1) over all
the relevant isotopes were carried out.
If these summations resulted
atw Vol. 63 (2018) | Issue 4 ı April
DECOMMISSIONING AND WASTE MANAGEMENT 244
| | Fig. 4.
Dismantling of the lower layers of the graphite reflector.
in a 1:10 ratio with other similar
metals of warranted non-nuclear
provenance in order to be used again
for various purposes. All the homogeneous
groups were pre-characterized,
characterized and released before the
end of 2014. All the measurements
were performed by trained ENEA staff
and within qualified ENEA laboratories,
with the exception of some 14C
measurements of a small lot of rubbers
which were performed, under special
contract, by the LASE Laboratory
of CEA in Saclay. Workmen for heavy
or peculiar technological operations
were hired from the Modena Fallimenti
SaS, a private Italian company specialized
in the dismantling of special
plants. Further details about the plan
for the characterization of materials
and components can be found in
[5,6].
4 Radiological Characterization
of the Building
After the completion of all the dismantling
activities, and after the release
of all the batches of materials, a
radiological characterization of the
building of RB3 has been made. This
consisted of two main steps. The first
was the characterization of the activation
status of the baritic concrete
biological shielding of the core. This
consisted in seven core drillings, (see
Figure 5) each 16 cm long, so distributed:
1 on the floor, 1 on the
northern wall, 1 on the western wall,
1 on the eastern wall, and 3 (at
different heights) on the souther wall.
All the drilling points were at positions
where the neutron flux during
operation was maximum. From each
drilling, four aliquots, 4 cm long,
where taken, so to cover the depth
profile of any activation distribution
inside the biological shielding. Each
aliquot was subject to gamma spectrometry
to search for the presence of
60Co, 134Cs, 152Eu and 154Eu. All
the 28 samples yielded results for all
the four isotopes lower than a few
mBq/g. Then, all the samples were
subject to thorough statistical analysis,
based on several Bartlett tests, to
verify if they were all and altogether
representative of the same statistical
distribution of activity and therefore
representative of a same “homogeneous
group” constituted of the whole
biological shielding. Once this condition
has been verified, a Noether test,
using 10 randomly chosen measurements,
was put in place to verify the
minimum number of samples to be
used for the final characterization of
the biological shielding. This resulted
in 13 samples, randomly extracted
from the complete set of all the 28
available samples. However, ENEA decided
to use all the 28 samples to verify
the free release condition for the
shielding, and for all the 28 samples
the condition resulted verified, meaning
that no significative activation of
the shielding had been realized. As a
further consequence, it could be
proven that no activation of walls
outside the biological shielding was
in place, just because, due to its
screening effect, the neutron flux
outside the shielding itself was 6 to 7
orders of magnitude lower.
The second step of the characterization
consisted in the assessment of
the surface contamination of the various
areas of the building. These were
separated into three main surfaces:
1) ceiling; 2) surfaces over +6.0 m
level; 3) surfaces below +6.0 m level.
The ceiling was indeed a false ceiling
made of thin aluminium plates; these
could have been contaminated by
tritiated water vapours emerging
from the core once open for refueling
or fuel reshuffling. To investigate this,
the aluminium plates were dismantled,
taken to ground, and analyzed. It
was assumed that, if no contamination
was found, then also the real
ceiling behind it was not contaminated.
This proved indeed to be the
case. Surfaces over +6.0 m were
investigated randomly (Figure 6), by
sampling a given number of points,
quantified basing on statistical considerations.
All surfaces below +6.0 m
were completely measured, both walls
and floors. The measurement technique
consisted in using surface
contamination meters (Berthold
LB165 and LB124), properly cali brated
with large area reference sources, to
sum up count rates over 14C, 60Co,
134Cs, 152Eu and 154Eu. A similar
measurement methodology was successfully
applied for the decommissioning
of the ASTRA research reactor
in Vienna [7]. Background contributions
due to natural radionuclides in
the different materials were subtracted
after having made suitable averages
from surely clean, similar materials
to those which were to be measured
inside the building. As a further, conservative
penalization, it was decided
to attribute to each of the 5 abovementioned
nuclides the whole net
counting over each surface portion
being measured, counting time per
surface element being about 30 seconds
to reach a desired minimum
detectable activity. LB124 hand held
monitor was used for surfaces over
+6.0 m, while LB165 (wheeled monitor
as in Figure 7) was used over all
other surface portions. A special automated
vertical translational sledge
(Figure 8) was used to carry LB165
over the portions of the walls. In case
a given measurement yielded values
above the clearance limits, special
cleaning procedures were to be
adopted until subsequent measurements
proved to be below the limits
| | Fig. 5.
Core drilling of the biological shielding.
| | Fig. 6.
LB124 measurements of selected portions of walls above +6.0 m level.
Decommissioning and Waste Management
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
atw Vol. 63 (2018) | Issue 4 ı April
| | Fig. 7.
LB165 measurement of floor portions.
themselves. No cleaning procedures
were ever necessary and all the surface
portions could be considered not
contaminated and so able to be freely
released.
5 Radiological Characterization
of the Site
The third and last pillar of the decommissioning
of RB3 is the radiological
characterization of the areas of the
site surrounding the building. In this
regard, it is important to mention that
during the operational life of the
plant, given its low power and its
construction features, no radiocativity
discharges were in place and therefore
no environmental analyses were prescribed
by the Regulatory Authority.
Another point worth of mentioning is
that no radiological status of the site
prior to the construction and exercise
of RB3 is known. However, in light of
the graded approach which is going to
be taken for this third pillar by the
Regulatory Authority, given the fact
that no activated materials have been
found and that no activation or contamination
of the building has been
detected, it is decided to base this
characterization upon the measurement
of some selected nuclides in
certain terrain samples (soil) taken
around the area of the RB3 site.
In particular, 12 measurements of
239+240Pu through alpha spectrometry
will be done, together with
25 gamma spectroscopy assessments
for 54Mn, 60Co, 125Sb, 134Cs, 137Cs,
| | Fig. 9.
Collecting soil samples from the RB3 site.
Radionuclide
152Eu and 154Eu. Each terrain
sample will be a parallelepiped of
25x20x10 cm 3 corresponding roughly
to 5 liters of humid soil (Figure 9).
The site will be sampled considering
both near-range and far-range positions
in order to find patterns of radioactivity
correlated with the distance
from the RB3 building, if any at all.
The obtained values will be confronted,
through proper summations,
with the limits for the free release
of nuclear sites prescribed by the
German national law, which correspond
to the radiological nonrelevance
value of 10 microSv/year
to the public [8,9]. The limits for
the above-mentioned isotopes are
reported in Table 2.
References
[1] D.Lgs. 17 marzo 1995, n. 230,
Attuazione delle direttive Euratom
80/836, 84/467, 84/466, 89/618,
90/64, 92/3, 96/29.
[2] D. M. 29 Novembre 2010 Ministero
dello Sviluppo Economico di
Autorizzazione alla Disattivazione
Impianto Nucleare di Ricerca Reattore
RB-3 di Montecuccolino (BO) dell’ENEA.
[3] Radiation Protection 89, Recommended
radiological protection criteria for the
recycling of metals from the
dismantling of nuclear installations,
European Commission, 1998.
| | Fig. 8.
LB165 and its translational sledge to measure wall portions.
Concentration Limit
(Bq/g)
54 Mn 0.09
60 Co 0.03
125 Sb 0.08
134 Cs 0.05
137 Cs 0.06
152 Eu 0.07
154 Eu 0.06
239 Pu 0.04
240 Pu 0.04
| | Tab. 2.
Proposed clearance limits for the free release
of the RB3 site.
[4] Radiation Protection 113,
Recommended radiological protection
criteria for the clearance of buildings
and building rubble from the
dismantling of nuclear installations,
European Commission, 2000.
[5] I. Vilardi, C. M. Castellani, D. M.
Castelluccio, F. Rocchi, Piano di
Caratterizzazione Radiologica di Materiali
provenienti dalla Disattivazione
dell’impianto Nucleare di Ricerca Rb-3
dell’enea sito in Bologna – Montecuccolino
ai Fini del loro Allontanamento,
Convegno Nazionale AIRP 2014, Aosta.
[6] M. Capone, N. Cherubini, A. Compagno,
A. Dodaro, F. Rocchi, The Dismantling of
the Montecuccolino RB3 Research
Reactor: Radiological Characterisation of
Materials for Free Release, Proceedings
of the European Reaserch Reactor
Conference RRFM 2015, Bucharest
19-23 April 2015, 528-537.
[7] F. Meyer, F. Steger, R. Steininger,
Decommissioning of the Astra Research
Reactor – Dismantling the auxiliary
Systems and Clearance and Reuse of the
Buildings, Nuclear Technology &
Radiation Protection, 1/2008, 54-62.
[8] OECD/NEA Status Report, Releasing
the Sites of Nuclear Installations,
NEA Report 6187, 2006.
[9] Bundesgesetzblatt G 5702 Teil I, Bonn
26 July 2001, Nr. 38, 2001.
Authors
F. Rocchi
ENEA FSN/SICNUC/SIN
C. M. Castellani
ENEA IRP
A. Rizzo
ENEA FSN/SICNUC/TNM
Via Martiri di Monte Sole 4
Bologna (BO), Italy
A. Compagno
ENEA FSN/FISS/CRGR
I. Vilardi
ENEA IRP/SFA
Via Anguillarese, 301
00123 S.Maria di Galeria (RM), Italy
R. Lorenzelli
ENEA FSN/SICNUC/SIN
Località Brasimone
40032 Camugnano (BO), Italy
DECOMMISSIONING AND WASTE MANAGEMENT 245
Decommissioning and Waste Management
The Decommissioning of the ENEA RB3 Research Reactor in Montecuccolino ı F. Rocchi, C. M. Castellani, A. Compagno, I. Vilardi, R. Lorenzelli and A. Rizzo
atw Vol. 63 (2018) | Issue 4 ı April
246
RESEARCH AND INNOVATION
Revised version of a
paper presented at
the Annual Meeting
of Nuclear Technology
(AMNT 2017), Berlin.
Irradiation Tests of a Flat Vanadium Self-
Powered Detector with 14 MeV Neutrons
Prasoon Raj and Axel Klix
Self-powered detector (SPD) represents a class of neutron and gamma monitoring instruments used in the fission
reactor cores worldwide. This detector has inherent advantages of functioning without a bias voltage, simple measurement
scheme, compactness, ease of maintenance, and high reliability. We are studying SPD for application as flux
monitors in the European test blanket modules (TBM) of ITER, fusion reactor under construction in southern France.
This paper presents results of experimental tests performed with 14 MeV neutrons for a flat SPD with vanadium emitter.
Vanadium responds by beta emission from products of reactions (main routes: 51 V (n, γ) 52 V and 51 V (n, p) 51 Ti) with
thermal and fast neutrons. Secondary electrons due to gammas from these reactions and neutron irradiation of
surrounding materials are also important contributors to the signal. Thin foils of emitter, insulator and collector
materials are used to construct the test SPD. The detector is irradiated with short and long pulses of neutrons and is
found to respond in proportion with the incident neutron flux. Further experiments with simplified and better optimized
design of detector are underway for thorough study of the signal-creation mechanism.
1 Introduction
ITER [1] is an experimental fusion
reactor based on tokamak concept,
under construction at St. Paul lez
Durance in southern France. It is an
international project aimed at proving
feasibility of fusion as a large-scale
and carbon-free source of energy. One
of the main scientific goals of this
project will be to test and prove the
concepts of tritium breeding blankets.
Tritium is an important fuel component
for devices based on D-T reaction,
which is being considered as
main reaction for fusion power plants.
Because tritium is a rare element, it is
required to breed it in the fuel cycle of
the reactor. A blanket with lithium
compounds will cover the inner wall
of the plasma vessel. Fusion neutrons
from the plasma will be absorbed by
lithium nuclei, causing reactions to
produce tritium.
There are multiple breeding
blanket designs proposed by scientists.
To determine their efficiencies in
a real fusion environment, test blanket
modules (TBM) based on different
concepts will be inserted into equatorial
ports of ITER for experimental
tests in different operational phases of
ITER. The European Union is going to
test two such concepts, namely the
Helium-Cooled Lead-Lithium (HCLL)
and Helium-Cooled Pebble Bed
(HCPB) TBMs [2]. In the neutronics
experiments, nuclear responses like
tritium production rate, material activation,
nuclear heating etc. are to be
measured and compared with the
calculations. This step will validate
the advanced computational tools
and nuclear data utilized for nuclear
analyses for fusion devices. The neutron
and gamma fluxes are important
quantities to be measured for these
experiments, for which detectors like
neutron activation system, fission
chambers and self-powered detectors
(SPD) are under study.
An SPD is a multi-layered electrical
device, which produces direct current
(DC) signal on irradiation with
neutrons and/or gammas. It can be
preferentially responsive to neutrons
(self- powered neutron detector,
SPND) or gammas (SPGD), or as
it is in most of the cases, to both.
Figure 1 shows a rough sketch of the
cross- section of a traditional detector.
Central material, called emitter produces
fast electrons on irradiation.
These fast electrons can be betas
from the decay of neutron activation
products, or secondary electrons due
to interaction of gammas in the bulk
of the material. They slow down in a
layer of insulation and stop in the
outer electrode called collector. This
electron-movement creates a potential
difference and thus, produces a
current signal proportional to the
incident particle flux. The current due
to beta electrons is “delayed” because
of the half-life of beta-emitters, e.g.
SPND based on Rh, V or Ag emitters.
Whereas that due to gamma-initiated
photoelectric or Compton electrons
is “prompt”, for example Co-based
SPND [3].
An SPD responds in a sophisticated
manner, with multiple factors
contributing to the small current
signals often totaling between 10 -12
and 10 -3 Ampere. Due to its inherent
advantages of simplicity, compactness
and high-reliability, they are highly
desirable for flux monitoring in areas
with restricted access like reactor
cores. At KIT, we are studying SPDs
for application in ITER TBM [4].
Vanadium based flat SPD is being
tested with 14 MeV neutrons, to
understand its behavior towards fast
neutrons expected in fusion environment
and ascertain the feasibility of
its application as flux monitor for
European ITER TBMs.
2 Experimental details
Vanadium is a common emitter for
fission reactor SPNDs. The response
of the detector towards thermal neutrons
is understood well. The material
is relatively inexpensive and easier to
handle. However, due to lower cross
sections the sensitivity of vanadium-
SPND towards fast neutrons reduces
(Figure 2). Commercially available
SPND cannot be directly used for
measurement of fusion neutron
fluxes, going up to approx. 14 MeV in
energy.
Characteristics of the two main
beta- emitters from 51 V (99.75 %
isotopic abundance) in case of fast
neutron irradiation, are reported in
Table 1. Cross-sections of the fast
neutron reactions in 51 V for a pure
| | Fig. 1.
Cross-sectional sketch of a cylindrical SPD showing emitter (green), insulator (dotted white) and
collector (black) layers, with connection to the lead cable and current measurement device.
Research and Innovation
Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix
atw Vol. 63 (2018) | Issue 4 ı April
| | Fig. 2.
Cross sections of vanadium reactions and photon production under neutron irradiation.
Reaction 51 V (n, p) 51 Ti 51 V (n, γ) 52 V
Threshold Neutron Energy 1.72 MeV 0 MeV
14 MeV Cross-section 30 mb (approx.) 0.6 mb (approx.)
Beta Emitter, Half-life 51 Ti- 5.76 m 52 V- 3.74 m
Average Beta Energy 51 Ti- 0.87 MeV 52 V- 1.07 MeV
SPND Current (14 MeV) 3.46 × 10 -12 A 6.92 × 10 -14 A
SPND Current (TBM) 7.97 × 10 -9 A 3.44 × 10 -8 A
| | Tab. 1.
Beta-emitters and corresponding currents from fast neutron reactions in vanadium based SPND.
14 MeV source are shown. Neglecting
the self-shielding of electrons in emitter
material, effect of other materials
and taking a saturation condition
(considering the short half-lives of
daughter nuclides), one can ascertain
the orders of magnitude of currents
possible with V-SPND, as reported.
For this estimation, vanadium density
of 6.1 g cm -3 , and a typical volume of
1 cm 3 are assumed. For a 14 MeV
neutron source, a flux intensity of
1 × 10 10 cm -2 s -1 is considered, which
is achievable with state of the art
14 MeV neutron generators. For TBM,
activation calculation was done [5]
with the HCLL neutron spectrum
and typical flux intensity (up to 1 ×
10 14 cm -2 s -1 ) using EASY-2007 [6].
With high-sensitivity ammeters,
currents down to the order of 1 ×
10 -14 A can be reliably measured [7].
Values in Table 1 show that a vanadium
emitter based SPND will produce
measurable signals in TBM. Due
to its high neutron threshold energy,
the (n, p) reaction can be utilized to
measure fast neutron flux exclusively.
Fast neutron reactions lead to
high-energy gamma production. This
phenomenon competes with the neutron
absorption reactions (Figure 2).
Photoelectric and Compton electron
emission from emitter causes a prompt
current which is expected to form the
major component of the signal of
V-SPND towards 14 MeV neutrons.
Secondly, vanadium being a medium-
Z nucleus can be a potential
emitter for SPGD also. With optimized
dimensions and choice of collector
material, a vanadium SPD can be
envisaged for monitoring of photon
flux in TBM.
Instead of the usual coaxial type
cylindrical geometry, we designed
our test SPD in sandwich-type flat
geometry. This provides a relatively
higher cross section area to the incident
neutrons, and ease of access for
testing various materials in the same
device. Thin foils (0.5 to 2 mm) of
emitter, insulator and collector are
arranged to form an assembly in an
aluminum case, which also serves as
an electromagnetic shield. Central
conductor of the signal cable is linked
to the emitter plates of the detector.
The collector plates, case and the
cable sheath are shorted and securely
connected to the ground. Schematic
sketch and photograph of the test
detector are shown in Figure 3 (left).
With comparable cross sections of
reactions in different materials, the
insulator and collector materials also
play an important role in SPD
response. Behaviors of different
material combinations are experimentally
tested. Alumina (Al 2 O 3 ) or
beryllia (BeO) is used as insulator and
Inconel-600 or graphite is used as
collector in our experiments. Effects
of the change of geometry and dimensions
are also studied. A Keithley 6485
Picoammeter (sensitivity range -20 fA
to 20 mA) is used as the measuring
device. A low-noise triax cable (Belden
9222) is used to reduce the interferences
in low-current measurement.
The tests are conducted at the
14 MeV neutron generator of Technical
University of Dresden (TUD-NG),
shown in Figure 3 (right). Here,
deuteron beams are impinged on a
tritiated titanium target causing D-T
reaction which leads to production of
neutrons with peak energy of approx.
14 MeV. TUD-NG offers neutron flux
intensities up to 1 × 10 10 cm -2 s -1 . The
detector is placed in front of the
tritium- target of TUD-NG and tested
under different conditions by varying
flux levels and irradiation times.
3 Results
The irradiation tests of flat sandwichtype
vanadium SPD were performed
at TUD-NG, with neutron flux intensities
around 1 × 10 9 cm -2 s -1 . DC
signals in the range of 100 fA to 100 pA
were measured. In Figure 4, a plot
shows variation of SPD signal with
change in neutron flux. The detector
was composed of 1 mm thick layers of
vanadium emitter and Inconel-600
collector. The signal was found to be
proportional to the incident flux, with
approx. 90 pA at the highest flux level.
At low fluxes and low currents,
the measurements have high uncertainties.
Interference from electromagnetic
sources of stray currents,
| | Fig. 3.
(Left) internal design of the sandwich-type flat SPD: (top)- an engineering sketch of the geometry
having sandwich of foils of emitter (green), insulator (grey) and collector (red), and (below) a photograph
of the assembly with vanadium SPD.
(Right) experimental setup showing TUD-NG beamline, tritium target, mounted SPD, and the lead cable.
RESEARCH AND INNOVATION 247
Research and Innovation
Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix
atw Vol. 63 (2018) | Issue 4 ı April
RESEARCH AND INNOVATION 248
| | Fig. 4.
Vanadium SPD signal (left Y-axis, red curve) variation with change in neutron flux (right Y-axis, blue
curve) plotted with respect to irradiation time.
currents generated in coaxial cables,
electrostatic effects at the contacts
and degradation of insulation layer
due to radiation, lead to background
currents in the orders of 100 fA. This
makes the measurement of low-level
currents a very challenging task.
The SPD response is often reported
in terms of sensitivity, which is SPD
current per unit of neutron (or
gamma) flux intensity, reported in
units of A cm 2 s. For the vanadium
SPND signal in Figure 4, the sensitivity
lies between 4.48 × 10 -20 A cm 2 s
± 13.4 % (at flux intensity 2.04 ×
10 9 cm -2 s -1 ) and 8.80 × 10 -19 A cm 2 s
± 51.1 % (at flux intensity 6.40 ×
10 5 cm -2 s -1 ).
In another test, a constant-flux
irradiation of around 15 minutes was
done and the TUD-NG was switched
off. This signal is shown in Figure 5. It
was found that the detector current is
dominated by a prompt component
which appeared and disappeared with
neutron flux. The delayed signal is
usually less than 10% of the total
signal. A decay of delayed current was
observed as expected.
There are parasitic beta emission
reactions in insulator, collector and
cable’s central conductor, e.g. 27 Al (n,
p) 27 Mg reaction (half-life~ 9.46 min)
in alumina insulation. Electrons
emitted due to these reactions reduce
the total delayed current. Due to this,
the analysis of decay curve becomes
very complex. After data reduction,
subtraction of background contributions,
and further analysis the major
contribution was found to be from 51 Ti
due to 51 V (n, p) 51 Ti reaction. For
reduction of aforementioned effects
materials with lower total cross
sections of beta emission reactions,
like graphite and beryllia were used as
collector and insulator, respectively.
The change in the signal characteristics
was minimal with these alterations,
leading us to conclude that the
signal was mainly due to reactions in
the vanadium emitter. A prompt current,
makes the detector suitable for
pulsed devices like ITER. However, it
is important to understand the signal
creation mechanism for calibration
and application of the SPD.
The high prompt signal is attributed
to three main reasons. First is
the interaction of photons in the
emitter volume, which release high
energy electrons producing high
positive current. Unlike thermal neutrons,
fast neutrons lead to emission
of higher-energy photons with higher
probability of secondary effects.
Moreover, the photon production
cross section is usually an order or two
higher than the fast neutron reaction
cross sections in materials of detector
and surroundings (Figure 2). Secondly,
the production of charged particles
like protons and alphas in collector
and insulator material (cross sections
of (n, xp) and (n, xα) reactions are
high for 14 MeV neutrons) lead to
further difference of charge between
electrodes and a prompt positive
contribution to the signal. Finally, the
electrical and nuclear effects in
connecting wires and cables make a
small fraction of the positive current
signal
Some of the contributing factors
will be explicitly studied in future
tests. To de-couple the effects of other
materials, a detector with simplified
geometry is under design. An air-insulated
detector with box of collector
material is being constructed. The
material thicknesses are reduced in
order to decrease the gamma interactions.
Improved ways of making
electrical contacts between cable and
emitter are studied. Other less betaactive
materials like niobium are
being considered for collector. Vanadium
detector is also planned to be
optimized for photon response. To this
end, thicker emitters and collectors
with low gamma-activity will be used
to make a test-device which will be
irradiated with high-energy bremsstrahlung
photon source.
4 Conclusions
A flat sandwich-type vanadium SPD
has been constructed, for testing the
feasibility of application of SPDs in
ITER TBMs. Irradiation tests with
14 MeV neutrons at TUD-NG resulted
in current signals in range of 100 fA to
100 pA. The signals are proportional
to the incident neutron flux. Considering
the higher flux intensities up
to 1 × 10 14 cm -2 s -1 and a wider energy
spectrum of neutrons in TBM, studies
show that vanadium SPND is expected
to produce measurable signals in ITER
| | Fig. 5.
Vanadium-SPD signal in a long constant-flux irradiation at TUD-NG showing (prompt and delayed)
currents before, during and after the irradiation.
Research and Innovation
Irradiation Tests of a Flat Vanadium Self- Powered Detector with 14 MeV Neutrons ı Prasoon Raj and Axel Klix
atw Vol. 63 (2018) | Issue 4 ı April
TBM conditions. The high prompt
component of the SPD signal is
attributed to the interaction of high
energy photons which are produced
in the detector and surrounding
materials. Charged particles emitted
in fast neutron reactions and contributions
from wires and signal cable
contribute to the high positive signal.
Parasitic reactions in non-emitter
materials also play an important role.
These effects need to be studied
explicitly and compared for understanding
of the overall currentgeneration
mechanism. Optimization
of design, dimensions and material
combinations is underway to realize
SPD flux monitors for application in
European ITER TBMs.
Acknowledgement
The work leading to this publication
has been funded partially by Fusion
for Energy under the Specific
Grant Agreement F4E-FPA-395-1.
This publication reflects the views
only of the authors, and Fusion for
Energy cannot be held responsible for
any use which may be made of the
infor mation contained therein.
References
[1] ITER Organization – Homepage. [Online].
Available: https://www.iter.org/.
[2] P. Calderoni, Status of the HCLL and
HCPB Test Blanket System instrumentation
development, 21 st Top. Meet.
Technol. Fusion Energy (TOFE), 9-13
Nov. 2014, Anaheim, CA, USA.
[3] N. P. Goldstein and W. H. Todt, A Survey
of Self-Powered Detector - Present and
Future, IEEE Trans. Nucl. Sci., vol. 26,
no. 1, pp. 916–923, 1979.
[4] P. Raj, M. Angelone, U. Fischer, and
A. Klix, Self-powered detectors for test
blanket modules in ITER, in 2016 IEEE
Nuclear Science Symposium, Medical
Imaging Conference and Room- Tem perature
Semiconductor Detector Workshop
(NSS/MIC/RTSD), 2016, pp. 1–4.
[5] M. Angelone, A. Klix, M. Pillon, P.
Batistoni, U. Fischer, and A. Santagata,
Development of self-powered neutron
detectors for neutron flux monitoring in
HCLL and HCPB ITER-TBM, Fusion Eng.
Des., vol. 89, no. 9–10, pp. 2194–2198,
2014.
[6] R. A. Forrest, FISPACT-2007: User
manual, EASY Doc. Ser. UKAEA
FUS 534, 2007.
[7] Low Level Measurements Handbook –
7 th Edition: Precision DC Current,
Voltage, and Resistance Measurements.
Keithley- A Tektronix Company.
Authors
Prasoon Raj
Axel Klix
Institute for Neutron Physics and
Reactor Technology (INR)
Karlsruhe Institute of Technology
(KIT)
Hermann von Helmholtz Platz 1
76344 Eggenstein-Leopoldshafen
(Germany)
RESEARCH AND INNOVATION 249
Nanofluid Applied Thermo-hydrodynamic
Performance Analysis of Square
Array Subchannel Under PWR Condition
Jubair Ahmed Shamim and Kune Yull Suh
1 Introduction Efficient engineered design of heat transfer and fluid flow with enhanced heating or cooling
requires two pivotal aspects that must be taken into consideration for extracting thermal energy from nuclear fission
reactions in order to save energy, reduce process time, raise thermal rating and increase the operating life of a reactor
pressure vessel. Hence, one of the major challenges in designing a new nuclear power plant is the quantification of the
optimal flow of coolant and distribution of pressure drop across the reactor core. While higher coolant flow rates will
lead to better heat transfer and higher Departure from Nucleate Boiling (DNB) limits, it will also result in higher pressure
drop across the core, therefore additional demand of pumping powers as well as larger dynamic loads on the core
components. Thus, thermal hydraulic core analysis seeks to find proper working conditions with enhanced heat transfer
and reduced pressure drop that will assure both safe and economical operation of nuclear plants.
Recently, nanofluid has gained much
renewed attention as a promising
coolant for pressurized water reactors
(PWRs) due to its enhanced thermal
capabilities with least penalty in pressure
drop. The improved heat transfer
of nanofluids results from the fact that
the nanoparticles increase the surface
area and heat capacity of the fluid,
improve the thermal conductivity of
the fluid, cause more collisions and
interactions between the fluid, particles
and surfaces of the flow passages,
and enhance turbulence and mixing
of the fluid.
Pak & Cho [1] experimentally
observed the turbulent friction and
heat transfer of dispersed fluids in a
circular pipe using two different
metallic oxide particles, γ-alumina
(Al 2 O 3 ) and titanium dioxide (TiO 2 )
with mean diameters of 13 and 27 nm,
respectively. The results revealed
that the Nusselt number Nu for the
dispersed fluids increased with
increasing volume concentration as
well as the Reynolds number Re. But
at constant average velocity, the
convective heat transfer coefficient for
the dispersed fluid was 12% less than
that for pure water. They proposed a
new correlation for Nu under their
experimental ranges of volume concentration
(0-3%), Re (10 4 -10 5 ), and
the Prandtl number Pr (6.54-12.33)
for the dispersed fluids γ-alumina
(Al 2 O 3 ) and titanium dioxide (TiO 2 )
particles as
(1)
Xuan and Li [2] observed the flow
and convective heat transfer of the
Cu-water nanofluid flowing through
a straight brass tube of the inner
diameter of 10 mm and the length of
800 mm. They noted that suspended
nanoparticles can remarkably enhance
heat transfer given the velocities.
For instance, the heat transfer
coefficient of nanofluids containing
2.0 vol % Cu nanoparticles was increased
by as much as 40 % compared
to that of water. The conventional
Research and Innovation
Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh
atw Vol. 63 (2018) | Issue 4 ı April
RESEARCH AND INNOVATION 250
Dittus–Boelter correlation failed to
predict this augmented heat transfer
data for nanofluids. They presented a
new correlation for turbulent flow of
nanofluids inside a tube as
(2)
Maïga et al. [3] numerically investigated
fully-developed turbulent flow
of water/Al 2 O 3 nanofluid through
circular tube using different concentrations
under the constant heat flux
boundary condition. They proposed
the following correlation for 10 4 ≤
Re ≤ 5×10 5 , 6.6 ≤ Pr ≤ 13.9 and 0 ≤
φ ≤ 10%
(3)
Asirvatham et al. [4] reviewed the
published experimental investigations
on convective heat transfer of different
nanofluids.
Despite numerous studies on both
scaled experiments and numerical
modeling on heat transfer enhancement
of nanofluids proliferate over
the past years, most of the test sections
and computational domain were
limited to round pipes. Their simulating
parameters did not reflect the
environment of a nuclear power reactor,
either. Wu and Trupp [5] demonstrated
that flow conditions inside the
fuel rod assembly are quite different
from those in typical pipes. There is
so far no appropriate correlation in
literature that can predict heat transfer
characteristics of nanofluid in a
fuel assembly under PWR operating
condition. Therefore, numerical modeling
has been performed in this study
using a commercial computational
fluid dynamic CFD tool “Star-CCM+
(ver.9.06.011)” to predict heat transfer
and pressure drop more precisely
in a square array subchannel (1.25 ≤
P/D ≤ 1.35) for different volume concentrations
of water/alumina (Al 2 O 3 )
nanofluid (0.5% ≤ φ ≤ 3.0%). Referring
to the Advanced Power Reactor
1400 MWe (APR1400).
Properties
Also, if the slip between the particles
and the continuous phase is trifling,
the flow inside the subchannel may as
well be considered as single phase and
incompressible with constant physical
properties. Both the compression
work and viscous dissipation are
neglected. Under such conditions the
general conservation equations for
mass, momentum and energy can be
written in vector notations:
∇.(ρv) = 0 (4)
∇.(ρvv) = -gradP+μΔ 2 v (5)
∇.(ρvC P T) = ∇.(k gradT) (6)
where v, P and T are fluid velocity
vector, pressure and temperature,
respectively.
2.2 Determination of physical
properties of nanofluid
Determination of physical properties
of nanofluid is key to any nanofluid
research. If the nanoparticles are
assumed to be well dispersed in the
base fluid, the particle concentration
can be considered as constant
throughout the domain and effective
physical properties of mixture can be
evaluated using some classical formulas
well known for two phase fluids
[7]. The following formulas are used
to determine such properties as density,
specific heat, dynamic viscosity
and thermal conductivity.
ρ nf = (1-ϕ)ρ bf + ϕρ P (7)
(C P ) nf = (1-ϕ)(C P ) bf + ϕ(C P ) P (8)
μ nf = (1 + 7.3ϕ + 123ϕ 2 )μ bf (9)
Base Fluid
(Pure Water)
Alumina
Nanoparticles
Density (kg/m 2 ) 734.928 3970
Thermal Conductivity (W/m.K) 0.5701 40
Specific Heat (J/kg. K) 5361.69 880
Dynamics Viscosity (Pa. s) 9.01373E-05 -
| | Tab. 1.
Physical properties of base fluid and alumina nanoparticles.
and later improved by Brinkman [10]
and another by Batchelor [11], these
formulas drastically underestimate
the viscosity of nanofluids. Therefore,
they performed a least-square curve
fitting based on some scarce experimental
data available [12, 13, 14]
which leads to Equation (9). Equation
(10) [7, 15] is introduced for the thermal
conductivity as with the dynamic
viscosity. However, the pressure and
temperature of the above investigations
sizably differ from the operating
condition of a PWR. Since no such
correlation exists for thermophysical
properties of nanofluid applicable to
the operating environment of a PWR it
is assumed that the aforementioned
correlations can also be utilized for
nuclear reactors. Different properties
of base fluid (pure water) and alumina
nanoparticles that have been used in
this study are tabulated in Table 1.
3 Numerical modelling
3.1 Computational domain
The computational domain and
boundaries considered in this study
are shown in Figure 1, which represents
a quarter of a 3-D square array
subchannel created in Star-CCM+.
The diameter of the fuel rod is taken
as 9.5 mm and pitch-to-diameter ratio
P/D of 1.25 and 1.35 are selected for
simulation. The length of the subchannel
is taken as 600 mm which
is long enough to establish a fullydeveloped
turbulent flow at the outlet
under single phase forced convection
condition up to Re = 6×10 5 according
to the following criteria [16]
2 Mathematical modelling
k nf = (1 + 2.72ϕ + 4.97ϕ 2 )k bf (10)
2.1 Governing equations
The term “nanofluid” refers to a twophase
mixture of saturated liquid and
dispersed ultrafine particles of usual
size below 40 nm. However, due to
extremely tiny size of particles, it can
be readily fluidized and thus may be
considered to behave more like a fluid
rather than heterogeneous fluid [6].
Equations (7) and (8) are general
relationships being used in literature
[1, 7, 8] to compute the density and
specific heat for a classical two phase
mixture. Regarding the dynamic
viscosity, Maïga et al. [9] showed that,
albeit several correlations exist to
calculate the dynamic viscosity of
nanofluid as proposed by Einstein
| | Fig. 1.
Computational domain created in Star-CCM+.
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P/D = 1.25
Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid
(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %
6×10 5 7.829 7.963 8.351 9.196
5.098×10 5 6.651 6.766 7.095 7.813
4×10 5 5.219 5.309 5.568 6.130
3×10 5 3.914 3.982 4.176 4.598
| | Tab. 2.
Different inlet velocities, v 0 (m/s) used in simulation.
(11)
l e = EI × D h (12)
where l e is entrance length for fullydeveloped
flow, EI is entrance length
number and Dh is the channel hydraulic
diameter.
3.2 Boundary conditions &
Physics set-up
The coolant enters the subchannel
with a uniform inlet velocity v 0 (m/s)
at the inlet temperature 569 K. Different
values of v 0 for different coolants
that have been used in the simulation
are listed in Table 2. Different properties
of base fluid (pure water)
have been calculated at temperature
569 K and at pressure 155.1375 bar.
At the outlet, a static pressure of
155.1375 bar has been imposed. On
the tube wall, the usual non-slip
conditions with the standard wall
function are considered with a constant
heat flux of 600,000 W/m 2 . The
above parameters and geometric configurations
of the computational
domain are based on the design
features of the APR1400.
The constant density model is chosen
for the material. For turbulence
modeling, the realizable k-ε model
with high y + wall treatment is selected.
Implicit coupled solver with secondorder
upwind discretization scheme in
conjunction with coupled energy
model is implemented which solves
the conservation equations for mass
and momentum simultaneously using
a pseudo time marching approach.
3.3 Turbulence modeling
By studying different literature on
numerical simulation of flow through
a rod bundle for nuclear applications,
it can be concluded that no specific
turbulence model can be regarded as
superior to others for this sort of flow
phenomena. Yadigaroglu et al. [17]
carried out an exhaustive review of
rod bundle numerical simulations
and opined that the gradient transport
models, like the standard k-ε
model, are not capable of predicting
turbulent flow in the narrow gap regions.
Hàzi [18] had demonstrated
that the Reynolds Stress Model (RSM)
could be accurately applied in simulating
the rod bundle geometry. Lee
and Choi [19] also used the RSM turbulence
model to compare the performance
of grid designs between the
small scale vortex flow (SSVF) mixing
vane and the large scale vortex flow
(LSVF) mixing vane. Liu and Ferng
[20] have also adopted RSM turbulence
model to numerically investigate
the effects of different types of
grid (standard grid and split-vane pair
one) on the turbulence mixing and
heat transfer. Palandi et al. [21] have
successfully implemented SST k-ω
model in comparing thermo-hydraulic
performance of nanofluids and
mixing vanes in VVER-440 triangular
array fuel rod bundle. However, application
of RSM turbulence model will
require 50-60% more CPU time per
iteration and 15-20% more memory
usage compared to standard k-ε and
k-ω model.
Recently Conner et al. [22] have
implemented renormalization group
(RNG) k-ε model (Yakhot et al., [23])
in simulation a 5×5 rod bundle with
mixing-vane grid using Star-CCM+.
The applicability of this model to
simulate fuel rod bundles has been
tested and validated by Westinghouse
in their extensive research (Smith et
al., [24]).
Considering the established practice
and computational time required
as discussed above, it can be concluded
that RNG k-ε model will be
sufficient in modeling turbulence for
flow through a rod bundle. However,
in this study, realizable k-ε model
(Shih et al., [25]) has been adopted
for turbulence modeling inside a
square array subchannel since it has
been statistically proved that this
model provides the best performance
among all the k-ε model versions for
separated flows and flows with complex
secondary flow features [26].
The term “realizable” means
that the model satisfies certain mathematical
constraints on the Reynolds
stresses, consistent with the physics
of turbulent flows. Neither the standard
k-ε nor the RNG k-ε model is
realizable.
The modeled transport equation
for k and ε in the realizable k-ε model
are presented by Equation (13) and
Equation (14) respectively:
(13)
and
where,
P/D = 1.35
Inlet Re Pure Water Alumina (Al 2 O 3 ) Nanofluid
(φ = 0 %) φ = 0.5 % φ = 1.5 % φ = 3.0 %
6×10 5 5.826 5.926 6.215 6.843
5.098×10 5 4.950 5.035 5.280 5.814
4×10 5 3.884 3.951 4.143 4.562
3×10 5 2.913 2.963 3.108 3.422
(14)
(15)
(16)
In above equations, G k represents
the generation of turbulence kinetic
energy due to mean velocity gradients,
G b is the generation of turbulence
kinetic energy due to buoyancy, Y M is
the contribution of fluctuating dilatation
in compressible turbulence to
the overall dissipation rate, C 2 and C 1ε
are constants, σ k and σ ε are the
turbulent Prandtl numbers for k and ε
respectively, S k and S ε are user- defined
source terms.
3.4 Convergence of numerical
solution
Another central criteria that must be
satisfied in order to obtain proper
numerical solution is convergence.
The solver needs to be given adequate
iterations so that the problem is converged
and a solution can be treated
as converged if the following criteria
are satisfied [26]:
• The solution no longer changes
with subsequent iterations
• Overall mass, momentum, energy
and scalar balance are achieved
• All equations (momentum, energy
etc.) are obeyed in all cells to a
specific tolerance
RESEARCH AND INNOVATION 251
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RESEARCH AND INNOVATION 252
In the present study, residuals for
continuity, X & Y-momentum, Z-
momentum and turbulence kinetic
energy are decreased respectively to
an order of 10 -2 , 10 -5 , 10 -2 and 10 -4
| | Fig. 3.
Distribution of wall y + values in case of pure water
with Re=6×10 5 (P/D =1.35)
| | Fig. 2.
Convergence of mass flow averaged temperature at outlet (P/D = 1.35) for pure water at corresponding
inlet Re = 6×10 5 .
after 30,000 iterations and also a
monitor is created to check how values
for mass flow averaged temperature at
outlet is converging and it is observed
that after 30,000 iterations these
values do not change significantly
with further iterations. A typical plot
of mass flow averaged temperature at
outlet for pure water at inlet Re =
6×10 5 is shown in Figure 2.
3.5 Wall y + values
The accurate calculations of y + value
in the near-wall region, which is a
measure of non-dimensional distance
from the wall to the first mesh node
(based on local cell fluid velocity), are
of paramount importance to the success
of any simulation. In order to use
a wall function approach properly for
a particular turbulence model with
confidence, the y+ values should be
within a certain range.
In the present study, standard wall
function is used in conjunction with
realizable k-ε model and high-y + wall
treatment in which the near-wall cell
centroid are anticipated to be placed
in the log-law region with a value
30 ≤ y + ≤ 100. Results of performed
simulations demonstrate that the
wall y + values for different cases are
within this specified range. A pictorial
representation of wall y + in case
of pure water with Re = 6×10 5
(P/D = 1.35) is shown in Figure 3.
4 Code validation
4.1 Mesh convergence test
Since the accuracy of finite volume
method is directly related to the
quality of discretization used, it is
instrumental to select an optimized
mesh size that will take into account
both resolution of mesh structure and
as well as computational time and
cost.
In the present study, different
mesh settings are selected as presented
in Table 3 and values of
numerically obtained Nu are compared
against an existing correlation
for square array subchannel and for
pure water as presented by Equation
(17) through Equation (19) to check
mesh convergence for computational
domain with P/D =1.35. Results are
plotted in Figure 4 which clearly
states that a mesh setting with base
size 0.7 mm, no. of prism layer 2,
prism layer thickness 0.3mm and
prism layer stretching 3.7 will be
sufficient to produce Nu within
reasonable deviation compared to
the theoretical prediction made by
correlation.
Nu = ψ(Nu ∞ ) c.t. (17)
where,
(Nu ∞ ) c.t. = 0.023 Re 0.8 PR 0.4 (18)
for square array with 1.05 ≤ P/D ≤
1.9 and for pure water, Presser [27]
suggested:
(19)
Base Size
(mm)
No.
Prism Layers
Stretching Thickness
(mm)
Nu
(Star-CCM+)
Nu
(Presser)
Deviation
(%)
0.5 5 1.5 0.7 742.940 -35.051
0.6 4 1.5 0.5 862.627 -16.313
0.7 3 3.8 0.4 933.92 1003.35 -7.434
0.6 2 3.7 0.3 972.102 -3.214
0.7 2 3.7 0.3 1010.57 0.714
| | Fig. 4.
Mesh convergence test with different mesh settings.
| | Tab. 3.
Different mesh settings used to check mesh convergence.
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| | Fig. 5.
Validation of numerical model against correlation for P/D =1.25.
4.2 Validation of numerical
model
Since the ultimate test of any numerical
simulation is the validation of
results against well-known experimental
data, the model under consideration
in the present study has
been validated against correlation of
Presser for square array and pure
water as presented by Equation (17)
through Equation (19). Results are
plotted in Figure 5 and Figure 6
which demonstrates that there is
an excellent agreement between
numerical data and theoretical
prediction for the specified range of
inlet Re.
4.3 Validation of turbulence
model for nanofluid
Despite in the present study it is
assumed that nanofluid would behave
as a single-phase homogeneous fluid
and hence, all of the general conservation
equations of mass, momentum
and energy can directly be applied in
case of nanofluid, however, a successful
comparison of numerical Nu obtained
realizable k-ε model has been
carried out against both empirical
correlation and experimental data of
Pak & Cho [1] for turbulent flow
inside a round pipe of inside diameter
10.66 mm using alumina nanofluid
(φ=2.78%) as coolant for inlet Re
spanning from 5.03×10 4 to 1.48×10 4 .
The results are plotted in Figure 7
which clearly delineates that this
model can perform quite satisfactorily
with nanofluids.
5 Numerical results
and discussion
5.1 Temperature
Temperature profile along the centerline
of subchannel (P/D =1.25) for
different coolants at inlet Re = 6×10 5
are illustrated in Figure 8 from which
it is clear that there is a steady increase
in the coolant temperature due to absorption
of heat while flowing through
the subchannel and bulk temperature
of nanofluid is decreased with the increasing
particle volume concentration.
Numerically obtained fluid average
temperature (in case
of pure water at P/D =1.25 and
inlet Re = 6×10 5 ) at different axial
locations within the subchannel is
compared against the theoretical
predictions from energy balance
according to equation (20) [28] and
results are tabulated in Table 4.
(20)
The analogy shows that maximum
deviation between numerically obtained
axial temperature and theoretical
prediction is less than 0.6%.
5.2 Velocity
Development of axial velocity along
the centerline of subchannel (P/D
| | Fig. 6.
Validation of numerical model against correlation for P/D =1.35.
=1.25) for different coolants at inlet
Re = 6×10 5 is presented in Figure 9
which clearly states that fullydeveloped
velocity profile occurs
approximately after z=0.3 m and if
the current models are implemented
to evaluate physical properties of
nanofluid, development of velocity
| | Fig. 7.
Validation of turbulence model against Pak & Cho’s correlation.
| | Fig. 8.
Temperature along centerline of subchannel at Re = 6×10 5 .
RESEARCH AND INNOVATION 253
Axial Position
(m)
Average Bulk Fluid Temperature T m (K) %
of Deviation
Start-CCM+ Energy Balance
0 569 569 0.000
0.15 569.2431 570.6885 0.2532
0.30 570.1277 572.3771 0.3929
0.45 571.2205 574.0656 0.4956
0.60 572.4116 575.7542 0.5805
| | Tab. 4.
Comparison of numerically obtained axial temperature against theoretical predictions for pure water
(P/D =1.25 and inlet Re = 6×10 5 ).
| | Fig. 9.
Velocity along centerline of subchannel at Re = 6×10 5 .
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RESEARCH AND INNOVATION 254
| | Fig. 10.
Pressure along centerline of subchannel at Re = 6×10 5 .
profile is not affected by the inclusion
of nanoparticles. From Figure 9, it can
also be seen that there is an increase in
the velocity magnitude due to growth
of hydrodynamic boundary layer as
coolant flows from inlet towards
outlet. The inclusion of higher volume
concentration of nanoparticles also
augments the magnitude of axial
velocity as seen in Figure 9. It can be
explained from the fact that since
with the rise of volume concentration
the viscosity of the nanofluid is also
aggrandized, hence to a maintain
constant value of Reynolds number Re
at the inlet of the channel, velocity
magnitude should be increased too
according to equation (21) if the other
properties remain constant:
(21)
5.3 Pressure
A plot of static pressure along the
centerline of the subchannel (P/D
=1.25) for different coolants at inlet
Re = 6×10 5 is shown in Figure 10
which depicts that there is an increase
in axial pressure with the inclusion of
nanoparticles which is expected due
to higher viscosity and density as the
particle volume concentration is increased.
5.4 Nu and h Constant Inlet Re
Convective heat transfer is studied
with Star-CCM+ for pure water and
different concentrations of alumina
nanofluid according to Equation (22)
and Equation (23) respectively. Values
of Nu are evaluated at the outlet of the
subchannel to assure fully-developed
turbulent flow condition.
(22)
(23)
where, q '' is the constant heat flux
(W/m 2 ), k is thermal conductivity
(W/m 2 .K), D h is hydraulic diameter
(m), and T w and T m are wall and mean
bulk fluid temperature (K) respectively.
Numerical results of Nu and h for
subchannel with different pitch-todiameter
(P/D) ratio are presented
through Figure 11 to Figure 14
respectively and percentage of convective
heat transfer increment for
different nanofluid coolants are
documented in Table 5.
From the results, it is obvious that
the convective heat transfer coefficient
is remarkably increased with the
increment of nanoparticle volume
concentration and in case of 3.0 %
volume concentration, convective
heat transfer is increased above
22.0 % compared to pure water.
5.5 Comparison of Numerical
Results against Correlations
In case of nanofluid with volume
concentration, φ =3.0% numerical
results for Nu are compared against
two well cited correlations of Pak &
Cho [1] and Maïga et al. [3] as shown
in Figure 15 (a) & (b) and an attempt
has been made whether results of
present study can be represented by
either of these two correlations.
The results revealed that Pak
and Cho correlation severely underestimates
the numerical results for
Nu in subchannel and deviation lies
between 17 to 22 percent subject to
inlet Re and P/D.
Regarding correlation of Maïga
et al., it shows better approximation
compared to correlation of Pak & Cho.
Nevertheless, this correlation underestimates
the numerical results for the
| | Fig. 11.
Comparison of Nu for different coolants in subchannel (P/D 1.25).
| | Fig. 12.
Comparison of Nu for different coolants in subchannel (P/D 1.35).
| | Fig. 13.
Comparison of h for different coolants in subchannel (P/D 1.25).
| | Fig. 14.
Comparison of h for different coolants in subchannel (P/D 1.35).
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range 5×10 5 ≤ Re ≤ 6×10 5 and overestimates
for 3×10 5 ≤ Re ≤ 4×10 5
and deviations are between 0.54 and
6.66 percent depending on inlet Re
and P/D.
5.6 Pressure drop
While nanofluid enhances the convective
heat transfer, the fluid itself
P/D = 1.25
Inlet Re Increment of h (%)
φ = 0.5 % φ = 1.5 % φ = 3.0 %
6×10 5 2.75 9.62 22.46
5.098×10 5 2.75 9.58 22.37
4×10 5 2.72 9.51 22.16
3×10 5 2.74 9.42 21.89
| | Tab. 5.
Heat transfer increment (%) for different nanofluid coolants.
also gets heavier compared to pure
water. Hence, it is of utmost importance
to determine the amount of
pressure drop for the effective application
of nanofluid coolant in nuclear
reactors since it is directly related to
the pumping power required. In this
study, pressure drop along the center
line of the subchannel is evaluated for
different coolants and results are presented
in Figure 16 (a) & (b). Percentage
of pressure drop increment is
documented in Table 6.
The results shows that pressure
drop is significantly increased with
the augmentation of particle volume
concentration which in turn increases
the pumping power. For nanofluid
P/D = 1.55
Inlet Re Increment of h (%)
φ = 0.5 % φ = 1.5 % φ = 3.0 %
6×10 5 2.72 9.56 22.35
5.098×10 5 2.72 9.51 22.26
4×10 5 2.71 9.44 22.01
3×10 5 2.69 9.40 21.87
RESEARCH AND INNOVATION 255
(a) P/D = 1.25
| | Fig. 15.
Comparison of numerical Nu against different correlations.
(b) P/D = 1.35
(a) P/D = 1.25
| | Fig. 16.
Comparison of pressure drop for different coolant.
(b) P/D = 1.35
P/D = 1.25
Inlet Re Increment of ∆p (%)
φ = 0.5 % φ = 1.5 % φ = 3.0 %
6×10 5 6.22 21.53 56.60
5.098×10 5 5.82 21.17 56.62
4×10 5 5.79 21.79 56.02
3×10 5 5.24 21.65 55.83
P/D = 1.35
Inlet Re Increment of ∆p (%)
φ = 0.5 % φ = 1.5 % φ = 3.0 %
6×10 5 5.82 20.94 56.37
5.098×10 5 5.74 21.29 56.08
4×10 5 5.46 20.90 55.10
3×10 5 5.62 20.88 55.82
| | Tab. 6.
Pressure drop increment (%) for different nanofluid coolants.
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RESEARCH AND INNOVATION 256
with φ=3.0%, pressure drop increment
is about 56% higher compared
to that of pure water.
However, the typical nanoparticle
loading in PWR coolant should be
less than 1.0 vol %. At such lower
con centration, nanofluid properties
are almost similar to that of pure
water and the rise in viscosity as well
as pressure drop will be negligible too.
The present study also portrays that
pressure drop is approximately 20 %
at 1.5 vol. % of nanoparticle concentration
which can also be treated as
tolerable.
The convective heat transfer coefficient
at such low concentration of
nanofluid is yet to be improved due to
higher turbulence produced near the
grid spacers by the presence of nanoparticles
in the base fluid. Since it is
quite difficult to take into account
such effects in numerical simulation,
further experimental investigation is
required for quantification of heat
transfer increment aroused from the
presence of nanoparticles near the
spacer grids.
6 Proposed new
correction factor
Finally, a multiple regression analysis
is performed with numerical results to
propose a new correction factor, β for
the existing correlation of square
array subchannel with pure water as
suggested by Presser [27] so that Nu
for nanofluid coolant can be approximated
in such geometry. Based on
regression results, β can be expressed
as follows:
β = 1 + 0.0247ϕ 1.39 (24)
Nu for nanofluid can be calculated as
follows:
Nu nf = β*(Nu Presser ) Water (25)
The validity of above correlation is for
3×10 5 ≤ Re ≤ 6×10 5 ; 0.847 ≤ Pr ≤
1.011; 1.25 ≤ P/D ≤ 1.35 and 0.5% ≤
φ ≤ 3.0% in case of square array
subchannel.
7 Chemical and physical
stability of nanofluid
Albeit nnanofluid can readily boost
the heat transfer capability of PWR
coolant, there is still no satisfactory
explanation proposed regarding the
prevention of clustering in nanoparticle
suspensions. Agglomeration
in nanofluids containing oxide nanoparticles
can be reduced remarkably
by adjusting the pH to form electric
changes on particle surface so that
they repel each other [29]. However,
the typical values of pH should be
such that nanofluid itself becomes not
corrosive and it should be agreeable
with same allowable pH range of
nuclear reactor, since altering the
PWR coolant chemistry is not a viable
option. Besides, use of surfactants are
also not recommended since it may
undergo severe radiolysis inside the
reactor core during operation.
Hence, issues concerning chemical
and physical stabilities of nanofluid
has yet to be resolved prior to utilizing
nanofluid as a promising coolant in
PWRs to achieve both extended life
time of associated equipment and
higher thermal efficiency.
8 Conclusion
Thermo- and hydrodynamic characteristics
of water/alumina nanofluid
have been studied in a square array
subchannel featuring the pitch-todiameter
ratios of 1.25 and 1.35 under
the steady-state, incompressible,
single- phase turbulent flow condition.
Numerical results have been compared
against correlations in the
literature and the following conclusions
can be drawn.
• Convective heat transfer is increased
with increasing volume
concentration of water/alumina
nanofluid given the inlet Reynolds
number.
• The convective heat transfer increment
of nanofluid is obtained at
the expense of increased pressure
drop and hence, larger pumping
power is required. Therefore,
nano fluid as PWR coolant can be
only be implemented in reality if
the replacement of reactor coolant
pump is a feasible option compared
to higher power gained from
increased nanofluid heat transfer.
Acknowledgements
This work was supported by the
National Research Foundation of Korea
(NRF) grant funded by the Korean
Government (MSIP) under Grant No.
2008-0061900 and partly supported
by the Brain Korea 21 Plus under
Grant No. 21A20130012821.
Nomenclature
∆p Pressure Drop Pa
ρ Density kg/m 3
v Flow Velocity m/s
f Friction Factor -
L Length of Flow Channel m
le Entrance Length m
EI Entrance Length Number -
Dh Hydraulic Diameter m
μ Dynamic Viscosity N.s/m 2
Re Reynolds Number -
Nu Nusselt Number -
Pr Prandtl Number -
Pe Peclet Number -
h
Convective Heat Transfer
CoefficientW/m 2 .K
k Thermal Conductivity W/m.K
C p Specific Heat J/kg.K
T m Bulk Temperature of Fluid K
T w
Surface Temperature
of Heater Rod
P Rod Pitch m
D Rod Diameter m
Q Total Heat Input W
q” Heat Flux W/m 2
φ
ṁ Mass Flow Rate kg/sec
Subscript
nf
bf
P
Volume Concentration
of Nanoparticles %
Nanofluid
Basefluid
Particle
References
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transfer study of dispersed fluids with
submicron metallic oxide particles.
Experimental Heat Transfer an International
Journal. 1998;11:151-70.
2. Xuan Y, Li Q. Investigation on
convective heat transfer and flow
features of nanofluids. Journal of Heat
transfer. 2003;125:151-5.
3. El Bécaye Maïga S, Tam Nguyen C,
Galanis N, Roy G, Maré T, Coqueux M.
Heat transfer enhancement in turbulent
tube flow using Al2O3 nanoparticle
suspension. International Journal of
Numerical Methods for Heat & Fluid
Flow. 2006;16:275-92.
4. Asirvatham LG, Vishal N, Gangatharan
SK, Lal DM. Experimental study on
forced convective heat transfer with low
volume fraction of CuO/water nanofluid.
Energies. 2009;2:97-119.
5. Wu X, Trupp AC. Experimental study on
the unusual turbulence intensity
distributions in rod-to-wall gap regions.
Experimental Thermal and Fluid
Science. 1993;6:360-70.
6. Xuan Y, Roetzel W. Conceptions for
heat transfer correlation of nanofluids.
International Journal of heat and Mass
transfer. 2000;43:3701-7.
7. Maïga SEB, Palm SJ, Nguyen CT, Roy G,
Galanis N. Heat transfer enhancement
by using nanofluids in forced convection
flows. International Journal of
Heat and Fluid Flow. 2005;26:530-46.
8. Bianco V, Chiacchio F, Manca O,
Nardini S. Numerical investigation of
nanofluids forced convection in circular
tubes. Applied Thermal Engineering.
2009;29:3632-42.
K
Research and Innovation
Nanofluid Applied Thermo-hydro dynamic Performance Analysis of Square Array Subchannel Under PWR Condition ı Jubair Ahmed Shamim and Kune Yull Suh
atw Vol. 63 (2018) | Issue 4 ı April
9. Maïga S, Nguyen CT, Galanis N, Roy G,
Heat transfer enhancement in forced
convection laminar tube flow by using
nanofluids. 2004: Publisher.
10. Brinkman H. The viscosity of
concentrated suspensions and
solutions. The Journal of Chemical
Physics. 1952;20:571-.
11. Batchelor G. The effect of Brownian
motion on the bulk stress in a
suspension of spherical particles. Journal
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12. Lee S, Choi S-S, Li S, and, Eastman J.
Measuring thermal conductivity of fluids
containing oxide nanoparticles. Journal
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13. Masuda H, Ebata A, Teramae K.
Alteration of thermal conductivity and
viscosity of liquid by dispersing ultrafine
particles. Dispersion of Al 2 O 3 , SiO 2
and TiO 2 ultra-fine particles. 1993.
14. Wang X, Xu X, S. Choi SU. Thermal
conductivity of nanoparticle-fluid
mixture. Journal of thermophysics and
heat transfer. 1999;13:474-80.
15. Maïga SEB, Nguyen CT, Galanis N, Roy
G. Heat transfer behaviours of nanofluids
in a uniformly heated tube.
Superlattices and Microstructures.
2004;35:543-57.
16. Häfeli R. Fluid dynamic characterization
of single-and multiphase flow in
structured porous media: Master Thesis
ETH Zurich, 2010; 2010.
17. Yadigaroglu G, Andreani M, Dreier J,
Coddington P. Trends and needs in
experimentation and numerical simulation
for LWR safety. Nuclear Engineering
and Design. 2003;221:205-23.
18. Házi G. On turbulence models for rod
bundle flow computations. Annals of
Nuclear Energy. 2005;32:755-61.
19. Lee C, Choi Y. Comparison of thermohydraulic
performances of large scale
vortex flow (LSVF) and small scale vortex
flow (SSVF) mixing vanes in 17× 17
nuclear rod bundle. Nuclear Engineering
and Design. 2007;237:2322-31.
20. Liu CC, Ferng YM. Numerically simulating
the thermal-hydraulic characteristics
within the fuel rod bundle using CFD
methodology. Nuclear Engineering and
Design. 2010;240:3078-86.
21. Palandi SJ, Rahimi-Esbo M, Vazifeshenas
Y. Comparison of thermo- hydraulic
performance of nanofluids and mixing
vanes in a triangular fuel rod bundle.
Journal of the Brazilian Society of
Mechanical Sciences and Engineering.
2015;37:173-86.
22. Conner ME, Baglietto E, Elmahdi AM.
CFD methodology and validation for
single-phase flow in PWR fuel assemblies.
Nuclear Engineering and Design.
2010;240:2088-95.
23. Yakhot V, Orszag S, Thangam S, Gatski
T, Speziale C. Development of turbulence
models for shear flows by a
double expansion technique. Physics of
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24. Smith III L, Conner M, Liu B, Dzodzo M,
Paramonov D, Beasley D, Langford H,
Holloway M. Benchmarking computational
fluid dynamics for application to
PWR fuel. Proceedings of ICONE.
2002;10.
25. Shih T-H, Liou W, Shabbir A, Yang Z,
Zhu J. A new k-epsilon eddy viscosity
model for high Reynolds number
turbulent flows: Model development
and validation. 1994.
26. Introduction to ANSYS FLUENT:
Customer Training Material, Release 13,
December 2010. .
27. Presser KH. Wärmeübergang und
Druckverlust an Reaktorbrennelementen
in Form längsdurchströmter
Rundstabbündel: Kernforschungsanlage,
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28. Azari A, Kalbasi M, Derakhshandeh M,
Rahimi M. An Experimental Study on
Nanofluids Convective Heat Transfer
Through a Straight Tube under Constant
Heat Flux. Chinese Journal of Chemical
Engineering. 2013;21:1082-8.
29. Buongiorno J, Hu L-w. Nanofluids for
Enhanced Economics and Safety of
Nuclear Reactors [Published 2007
[cited November 29].
Authors
Jubair Ahmed Shamim
Department of Nuclear
Engineering
Seoul National University
Seoul 08826, ROK
Kune Yull Suh
Seoul National University
1 Gwanak Ro, Gwanak Gu
Seoul 08826, ROK
257
KTG INSIDE
Inside
KTG-Vorstandswahl 2018
Liebe Mitglieder, gemäß unserer Satzung stehen mit Wirksamkeit zur Mitgliederversammlung am 28. Mai 2018
Wahlen zum Vorstand der KTG an. Die Wahlunterlagen mit dem vom Beirat aufgestellten funktionsbezogenen
Gesamtwahlvorschlag gehen Ihnen in den nächsten Wochen auf postalischem Wege zu. Wie gewohnt möchten sich die
Kandidaten zur KTG-Vorstandswahl 2018 Ihnen nachstehend (noch einmal) vorstellen.
Herzlichst,
Ihre KTG-Geschäftsstelle
Kandidaten
Frank Apel
Dipl.-Ing. (54), Heidelberg
Zur Person
Seit Februar 2017 Geschäftsführer bei der
Kraftanlagen Heidelberg GmbH. Vorher
leitend für den Bereich „Back-End“ bei der
AREVA (heute ORANO) und davor Leiter von
Vertrieb und Marketing der AREVA (heute
FRAMATOME) für die Region „Zentral- und Nord europa“. Darüber
hinaus verantwortlich in der Internationalen Vertriebsorganisation
für die weltweite Zusammenarbeit mit AREVAs wichtigsten Kunden
(Key Accounts). Mehr als 30 Jahre Erfahrung in der Energiewirtschaft
und in dieser Zeit in ver schiedenen Führungspositionen bei Siemens
und den Nachfolgefirmen tätig. Dabei unter anderem für alle
Service- und Wartungsarbeiten von AREVA in den deutschen
Kernkraftwerken verantwortlich. Abschluss als Diplombauingenieur
für Kernkraftwerke der Moskauer Staat lichen Universität für
Bauingenieurwesen.
Zur Wahl als Vorsitzender der KTG
In unserer letzten Vorstandsklausur haben wir erneut sehr intensiv
das Thema der Mitgliedschaft in unserem Verband diskutiert. Die
Frage, was die Mitglieder der KTG verbindet, beantworten wir
KTG Inside
atw Vol. 63 (2018) | Issue 4 ı April
258
KTG INSIDE
mit der „Faszination Kerntechnik“. Unsere Mitglieder, das sind
aktive oder ehemalige Mitarbeiter bei den Betreibern, Herstellern,
Behörden und Gutachtern, der Lehre und der Forschung oder
Menschen in anderen Berufsgruppen, die die Kerntechnik spannend
finden und für die der Austausch in unserer (kerntechnischen)
Gesellschaft wichtig ist. Sie werden überrascht sein: Wenn Sie
nach „Faszination Kerntechnik“ googeln, ist der erste Treffer:
Faszination Kerntechnik |
Kerntechnische Gesellschaft e.V.
https://www.ktg.org/ktg/faszination-kerntechnik/
Darüber können wir uns freuen, darauf können wir stolz sein; in
puncto Kommunikation sind wir besser geworden, das zeigt auch
unsere neuer Internet-Auftritt. Wir Kerntechniker haben in Deutschland
– dem Land der Bedenkenträger – (eigentlich) häufig Grund zur
Freude:
• Die am Netz befindlichen deutschen Kernkraftwerke erzeugen im
sicheren Leistungsbetrieb umweltfreundlichen Strom, eine
Verantwortung, die wir bis zum letzten Tag des Jahres 2022 und
darüber hinaus haben.
• Der (leider größtenteils politische verordnete) Rückbau geht
voran: Im Februar dieses Jahres hat Unterweser als 5. Anlage
in der „Post-Fukushima-Ära“ die 1. Stilllegungs- und Abbaugenehmigung
erhalten.
• In vielen stillgelegten Anlagen wurden und werden die
abgebrannten Brennelemente aus den Lagerbecken in
Castoren geladen und in die standortnahen Zwischenlager
verbracht.
• Im letzten Jahr hat die Deutsche Rechtsprechung „ideologiefrei“
z. B. Urteile zur „Durchsetzung von Castor-Transporten auf dem
Neckar“ oder die „Nichtrechtmäßigkeit der Brennelement-Steuer“
verkündet.
• In unserem Nachbarland Schweiz darf das Kraftwerk Beznau den
Block 1 nach einer dreijährigen Betriebsunterbrechung wieder
in Betrieb nehmen. Der Betreiber Axpo konnte nachweisen, dass
die Einschlüsse im Stahl des Reaktordruckbehälters keinen
negativen Einfluss auf die Sicherheit haben. Die Schweizer
Gutachter und Behörden haben – ideologiefrei und gestützt auf
die Meinung internationaler Experten – die entsprechenden
technischen Nachweise geprüft und akzeptiert.
An diesen Entwicklungen haben auch Sie mit Ihren hervorragenden
kerntechnischen Kompetenzen beigetragen. Die deutschen Kerntechnikerinnen
und Kerntechniker verfügen über ein weltweit
anerkanntes und nachgefragtes Know-how, was wir in Deutschland
erhalten wollen, um unter anderem:
• den verbleibenden Leistungsbetrieb, den Nachbetrieb, die
Stilllegung und den Rückbau deutscher Anlagen sicherzustellen
und die Entsorgungsfrage nachhaltig zu lösen,
• das Exportgeschäft deutscher Anbieter und Dienstleister zu
sichern,
• nationale und internationale Sicherheitsbewertungen durchführen
zu können und
• auch in Zukunft den Beitrag deutscher Standards und Innovationen
zu internationalen Entwicklungen für neue Technologien
erhalten zu können.
Das ist – nach wie vor – die Perspektive, die ich als KTG-Vorstand
und Vorsitzender in die kerntechnische Gesellschaft einbringen
möchte. Die KTG muss als Verband der Beschäftigten der Kerntechnik
eine führende Rolle spielen, diesen Kompetenzerhalt zu
sichern. Neben wirtschaftlich erfolgreichen und weiterhin innovativen
Unternehmen brauchen wir auch eine leistungsfähige
wissenschaftliche Landschaft bei den Forschungseinrichtungen und
Universitäten. Dies alles kann bei einem absehbar schrumpfenden
Heimatmarkt durch Wachstum im Ausland gewährleistet werden,
das sowohl die wirtschaftliche Zukunft der Unternehmen als auch
die persönliche Perspektive der Kolleginnen und Kollegen unserer
Branche sichert.
Erwin Fischer
Dr.-Ing. (61), Rodenberg
Zur Person
Nach einer praktischen Ausbildung zum
Maschinenschlosser Studienabschlüsse im
zweiten Bildungsweg als Ingenieur (grad.)
für Maschinentechnik nach einem Fachhochschulstudium
und als Dipl.-Ing. für
Maschinenbau mit Vertiefungsrichtung Energie technik an der Ruhr-
Universität (RUB), Bochum. Promotion, ebenfalls an der RUB, am
Lehrstuhl für Reaktortechnik/Neue und Nukleare Energie systeme im
Fachgebiet Reaktortechnik 1991.
Zur Wahl als Schatzmeister
Seit meiner Promotion bin ich bei der PreussenElektra GmbH und
ihren Vorläuferunternehmen im Kernenergiebereich in der Nuklearen
Technik und dem Betrieb tätig. Meine Aufgaben bezogen sich auf
den Bau, Betrieb und Rückbau von Kernkraftwerken. Während
der nunmehr 27 Jahre Tätigkeit für PreussenElektra habe ich
Aufgaben in den Kernkraftwerken und der Zentralorganisation wahrgenommen.
Seit 2014 führe ich das Geschäftsführungsressort Technik
und Betrieb. Ich war 13 Jahre im KTA tätig und 5 Jahre Mitglied
der deutschen Reaktorsicherheitskommission. Seit 4 Jahren engagiere
ich mich als Governor bei der WANO – World Association for
Nuclear Operators – weltweit.
Während meines bisherigen beruflichen Werdegangs war und
ist der sichere, umweltverträgliche und wirtschaftliche Betrieb der
Kernkraftwerke mein prioritäres Anliegen. Die Kernkraft hat mich in
meinem ganzen Berufsleben fasziniert und die Faszination hält trotz
aller Rückschläge und der teilweise schwierigen Randbedingungen
für die Kernenergie in Deutschland an.
Seit 1991 bin ich Mitglied in der KTG und nunmehr schon seit
8 Jahren im Vorstand, jetzt als Schatzmeister. Ziel meiner erneuten
Kandidatur ist, die KTG als Interessengemeinschaft aller in der
Kerntechnik Tätigen und von ihr faszinierten Mitgliedern mit meinem
Wissen und meiner beruflichen Erfahrung zu unterstützen sowie die
Wissensübertragung und den Erfahrungsaustausch zu erhalten.
Wie politisch gewollt, sollten wir Kerntechniker den sicheren Betrieb
bis zum Laufzeitende und den Rückbau unserer Kernkraftwerke
in Deutschland mit Ehre abschließen. Kein Grund mit Blick auf das
Erreichte der letzten 50 Jahre nicht stolz sein zu dürfen!
Jörg Starflinger
Prof. Dr.-Ing. (51), Stuttgart
Zur Person
Nach dem Studium des Maschinenbaus an
der Ruhr-Universität Bochum (RUB) mit
Schwerpunkt Energietechnik Promotion
im Jahr 1997 am Lehrstuhl für Nukleare
und neue Energiesysteme der RUB, Prof.
Dr.-Ing. H. Unger. 1998 Wechsel als Nachwuchswissenschaftler zum
Forschungszentrum Karlsruhe, heute Karlsruhe Institut für Technologie.
Themenschwerpunkte: Wasserstofferzeugung bei schweren
Unfällen in Leichtwasserreaktoren und Kreislaufsimulation von
innovativen Reaktorkonzepten. 2006 Leiter der Gruppe „Kraftwerkstechnik“
am Institut für Kern- und Energietechnik (IKET), Prof. Dr.-Ing.
T. Schulenberg, in der innovative Kernkraftwerkskonzepte mit überkritischem
Wasser von mehreren Doktoranden untersucht wurden.
2010 Ruf an die Universität Stuttgart zum ordentlichen Professor des
Lehrstuhls für Kerntechnik und Reaktorsicherheit und Leiter des
Instituts für Kernenergetik und Energiesysteme (IKE). Neben der
Lehre im Bereich Kerntechnik Schwerpunkte in der Reaktorsicherheitsforschung,
z.B. in der Modellentwicklung zur Beschreibung der
späten Phase von Kernschmelzunfällen in Leichtwasserreaktoren
KTG Inside
atw Vol. 63 (2018) | Issue 4 ı April
und auf dem Gebiet innovativer Sicherheitssysteme, z.B. der passiven
Lagerbeckenkühlung mit Wärmerohren (Heat pipes) und nachrüstbaren
Nachwärmeabfuhrsystemen mit überkritischem CO 2 als
Arbeitsmittel.
Zur Wahl als Vorstandsmitglied
Ich engagiere mich in der KTG auf dem Gebiet des Kompetenzerhalts
und der Kompetenzförderung. Den von Dr. Wolfgang Steinwarz ins
Leben gerufenen, sehr erfolgreichen Workshop „Kompetenzerhalt in
der Kerntechnik“ habe ich verantwortlich übernommen und möchte
ihn in den kommenden Jahren weiterführen. Dr. Steinwarz steht uns
auch als Ruheständler dankenswerterweise als Jurymitglied weiter
zur Seite. Die Umbenennung in „Young Scientists Workshop“ soll eine
Öffnung zu kerntechnisch verwandten Forschungsthemen, beispielsweise
„Kerntechnik und Gesellschaft“, symbolisieren.
Durch meine Mitarbeit im KTG-Vorstand als Vorstandsmitglied
möchte ich einen Strategieentwicklungsprozess anstoßen, der
mittelfristig eine genügende Anzahl an jungen hochqualifizierten
und motivierten Personen für die zukünftigen spannenden und
herausfordernden nationalen und internationalen kerntechnischen
Aufgaben sicherstellt. Für den Kompetenzerhalt und die Nachwuchsförderung
bieten die KTG und Ihre Mitglieder sowie unsere Tagung
„Annual Meeting on Nuclear Technology“ die ideale Plattform.
Walter Tromm
Dr.-Ing. (58), Stutensee
Zur Person
Maschinenbaustudium an der Uni (TH)
Karlsruhe mit dem Studienschwerpunkt
Kerntechnik und dort Promotion zum Thema
„Experimentelle Untersuchungen zum Nachweis
der langfristigen Kühlbarkeit von
Kernschmelzen“. Seit 1988 am damaligen Forschungszentrum
Karlsruhe, heute Karlsruher Institut für Technologie, angestellt und
schwerpunktmäßig mit Reaktorsicherheitsfragen bei auslegungsüberschreitenden
Störfällen beschäftigt. Von 1998 bis 1999 Gastwissenschaftler
am Europäischen Gemeinschaftsforschungszentrum
in Ispra (Italien) tätig.
Seit 2002 Programmbevollmächtigter in der Programmleitung
Nukleare Sicherheitsforschung des FZK bzw. heute Nukleare
Entsorgung, Sicherheit und Strahlenforschung des KIT; stellvertretender
Leiter seit 2007 wurde seit 2010 Programmleiter. 2014 im
geschäftsführenden Ausschuss des Bereichs Maschinenbau und
Elektrotechnik des KIT berufen. Seit 2015 darüber hinaus Sprecher
des vom KIT neu eingerichteten Kompetenzzentrums Rückbau und
seit 2017 Vorsitzender des Kompetenzverbundes Kerntechnik.
Tätig in nationalen und internationalen Gremien, bei der OECD/
NEA der deutsche Repräsentant des Nuclear Science Committee, bei
der IAEA in der Technical Working Group Light Water Reactors und
Mitglied im Governing Board der EU-SNETP Plattform. Weiterhin
innerhalb des VDI Vorsitzender des Fachausschusses Kraftwerkstechnik.
Seit 2016 Leiter des neu gegründeten Kompetenz-Cluster
Rückbau, der die Expertise im Rückbau mehrerer Länder zusammenführt.
Zur Wahl als stellvertretender Vorsitzender
Die Bundesregierung hat 2011 nach den Ereignissen in dem
Kernkraftwerk Fukushima Daii-chi in Japan entschieden, aus der
Stromproduktion mittels Kernkraft auszusteigen. In den nächsten
4 Jahren werden die letzten Kernkraftwerke in Deutschland
abgeschaltet. Diesen Ausstieg nach wie vor so sicher wie möglich
mitzugestalten ist eine der Aufgaben, die die in der deutschen
Kerntechnik arbeitenden Ingenieure und Naturwissenschaftler
haben. International und auf europäischer Ebene wird jedoch Kernenergie
langfristig weiterhin genutzt. Auch für den Industriestandort
Deutschland und für den Erhalt von Arbeitsplätzen ist der Export von
Komponenten für kerntechnische Anlagen nach wie vor bedeutsam.
Ebenfalls werden der Rückbau der Kernkraftwerke und die Endlagerfrage
die Gesellschaft noch über Jahrzehnte beschäftigen. Der
Ausstieg aus der Stromproduktion durch Kernenergie darf daher
nicht bedeuten, sich von den entsprechenden kerntechnischen
Kompetenzen in der Industrie, den Behörden und den Universitäten
und Forschungszentren zu verabschieden. In den Bereichen Reaktorsicherheit,
Rückbau, Endlagerung, Strahlenschutz und Krisenmanagement
sind diese Kompetenzen auch weiterhin gefragt. In
Europa stammen 27 % der Stromproduktion aus Kernkraftwerken.
Zur kompetenten Bewertung kerntechnischer Einrichtungen innerhalb
Europas und zur kritischen Begleitung internationaler Entwicklungen
sind eine enge Zusammenarbeit auf nationaler, europäischer
und internationaler Ebene unerlässlich. Deshalb sehe ich als eine
der Hauptaufgaben der KTG den Erhalt der kerntechnischen
Kompetenzen in allen genannten Bereichen.
259
KTG INSIDE
Herzlichen
Glückwunsch
April 2018
97 Jahre wird
2. Prof. Dr. Albert Ziegler, Karlsbad
87 Jahre werden
9. Dr. Klaus Penndorf, Geesthacht
11. Hubert Bairiot, Mol/B
19. Dr. Klaus Einfeld, Murnau
28. Dipl.-Ing. Rudolf Eberhart, Burgdorf
85 Jahre wird
6. Ing. Reinhard Faulhaber, Köln
84 Jahre wird
22. Dipl.-Ing. Gert Slopianka,
Gorxheimeral
83 Jahre werden
3. Dipl.-Psych. Georg Sieber,
München
5. Prof. Dr. Hans-Henning Hennies,
Karlsruhe
19. Dr. Ernst Müller, Rösrath
19. Dr. Gottfried Class,
Eggenstein-Leopoldshafen
21. Dipl.-Ing. Walter Jansing,
Bergisch Gladbach
30. Dr. Friedrich-Wilhelm Heuser,
Overath
82 Jahre werden
4. Helmut Kuhne, Neunkirchen
6. Dipl.-Ing. Hans Pirk, Rottach-Egern
10. Dipl.-Ing. Franz Stockschläder,
Bad Bentheim
11. Dipl.-Ing. Bernhard-F. Roth,
Eggenstein-Leopoldshafen
24. Dipl.-Ing. Horst Schott, Overath
81 Jahre werden
7. Dipl.-Ing. Helmut Adam, Neuenhagen
13. Dr. Martin Peehs, Bubenreuth
80 Jahre werden
4. Prof. Dr. Klaus Kühn, Clausthal- Zellerfeld
5. Dr. Hans Fuchs, Gelterkinden/CH
9. Dr. Carl Alexander Duckwitz, Alzenau
28. Prof. Dr. Georg-Friedrich Schultheiss,
Lüneburg
79 Jahre wird
8. Dr. Siegbert Storch, Aachen
78 Jahre wird
18. Dipl.-Ing. Norbert Granner,
Bergisch Gladbach
77 Jahre werden
17. Dipl.-Phys. Ernst Robinson, Gehrden
28. Dr. Ludwig Richter, Hasselroth
KTG Inside
atw Vol. 63 (2018) | Issue 4 ı April
260
NEWS
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KTG Inside
Verantwortlich
für den Inhalt:
Die Autoren.
Lektorat:
Sibille Wingens,
Kerntechnische
Gesellschaft e. V.
(KTG)
Robert-Koch-Platz 4
10115 Berlin
T: +49 30 498555-50
F: +49 30 498555-51
E-Mail: s.wingens@
ktg.org
www.ktg.org
76 Jahre werden
9. Prof. Dr. Hans-Christoph Mehner,
Dresden
27. Dr. Dieter Sommer, Mosbach
27. Dr. Jürgen Wunschmann, Eggenstein
29. Dr. Klaus-Detlef Closs, Karlsruhe
75 Jahre werden
15. Dr. Werner Dander, Heppenheim
18. Dipl.-Betriebsw. Uwe Janßen,
Weinheim
18. Dipl.-Ing. Victor Luster, Bamberg
26. Ing. Helmut Schulz, Kürten
70 Jahre werden
6. Dr. Wolfgang Tietsch, Mannheim
9. Ing. Herbert Moryson, Essen
22. Dr. Heinz-Dietmar Maertens, Arnum
26. Dr. Rainer Heibel, Ness Neston/GB
27. Ulrich Wimmer, Erlangen
65 Jahre werden
10. Dipl.-Phys. Harold Rebohm, Berlin
24. Dipl.-Phys. Michael Beczkowiak,
Karben
60 Jahre werden
4. Dipl.-Ing. Holger Bröskamp,
Höhnhorst
4. Dipl.-Ing. (FH) Franz Xaver Pirzer,
Schwandorf
50 Jahre werden
16. Rainer Bezold, Dormitz
16. Dr. Matthias Messer, Tetbury/GB
30. Dr. Christian Raetzke, Leipzig
Mai 2018
94 Jahre wird
22. Prof. Dr. Fritz Thümmler, Karlsruhe
90 Jahre wird
10. Dr. Heinz Büchler, Sankt Augustin
89 Jahre wird
31. Dipl.-Ing. Werner-P. Kürsten,
Mannheim
88 Jahre wird
9. Dr. Hans-Jürgen Hantke, Kempten
85 Jahre werden
4. Dr. Klaus Wiendieck, Baden-Baden
25. Dr. Reinhold Mäule, Walheim
25. Georg von Klitzing, Bonn
84 Jahre werden
11. Dr. Eckhart Leischner, Rodenbach
14. Dr. Alexander Warrikoff, Frankfurt/M.
26. Dr. Günter Kußmaul, Manosque/F
83 Jahre werden
1. Dr. Willi Bermel, Jülich
8. Dipl.-Ing. Klaus Wegner, Hanau
22. Dr. Heinz Vollmer, Lampertheim
28. Dipl.-Ing. Anton Zimmermann,
Hamburg
29. Dipl.-Ing. Karlheinz Orth,
Marloffstein
82 Jahre werden
3. Ewald Jurisch, Erlangen
10. Dr. Peter Reinke, Röttenbach
18. Dipl.-Ing. Gerhard Lorenz, Bochum
81 Jahre werden
1. Prof. Dr. Dietrich Munz,
Graben-Neudorf
3. Dipl.-Ing. Harald Enderlein, Karlsruhe
6. Dr. Peter Strohbach, Mainaschaff
7. Prof. Dr. Werner Lutze,
Chevy Chase/USA
20. Dr. Norbert Krutzik, Frankfurt/M.
26. Dipl.-Ing. Rüdiger Müller, Heidelberg
27. Dr. Johannes Wolters, Düren
28. Dipl.-Ing. Heinz E. Häfner, Bruchsal
80 Jahre werden
12. Dr. Herbert Finnemann, Erlangen
13. Dipl.-Ing. Otto A. Besch, Geesthacht
13. Dr. Heinrich Werle,
Karlsdorf-Neuthard
16. Dr. Hans-Dieter Harig, Hannover
21. Dr. Hans Spenke, Bergisch Gladbach
79 Jahre werden
4. Dipl.-Ing. Norbert Albert, Ettlingen
5. Dr. Wolfgang Voigts, Linkenheim
27. Prof. Dr. Dietrich Kirsch
78 Jahre werden
11. Dr. Andreas Hölzler, Schwaig
15. Dipl.-Phys. Ludwig Aumüller,
Freigericht
18. Dr. Karl Schulte, Köln
24. Dipl.-Ing. Herbert Krinninger,
Bergisch Gladbach
77 Jahre werden
8. Prof. Dr. Helmut Alt, Aachen
12. Dipl.-Ing. Dieter Rohde, Mannheim
16. Dr. Jürgen Baier, Höchberg
76 Jahre werden
5. Hans-Bernd Maier, Aschaffenburg
9. Dr. Egbert Brandau, Alzenau
11. Dr. Erwin Lindauer, Köln
17. Dr. Heinz-Peter Holley, Forchheim
18. Dipl.-Ing. Josef Koban, Buckenhof
28. Dipl.-Ing. Wolf-Dieter Krebs,
Bubenreuth
75 Jahre werden
3. Dipl.-Ing. Hans Lettau, Effeltrich
14. Dr. Helmut-K. Hübner, Bruchsal
20. Dipl.-Ing. Dietmar Bittermann, Fürth
22. Dr. Wolfgang Schütz, Bruchsal
23. Dipl.-Ing. Max Heller, Uttenreuth
24. Dipl.-Ing. Rudolf Weh,
Stephanskirchen
27. Dr. Kurt Fischer, Erlangen
65 Jahre werden
2. Dipl.-Ing. Marc Winter, Veitshöchheim
3. Dipl.-Ing. Karl-Heinz Wiening,
Herzogenaurach
5. Michael Klein, Großenwörden
16. Ing. grad. Eckhard Raabe, Geiselbach
21. Dipl.-Ing. (FH) Reinhold Horstmann,
Erlangen
27. Dipl.-Ing. (FH) Ulrich Hudezeck,
Nürnberg
60 Jahre wird
23. Dr. Hans-Josef Zimmer, Steinfeld
50 Jahre werden
10. Dr. Astrid Petersen, Hamburg
20. Dipl.-Ing. (FH) Jürgen Bruder,
Gundremmingen
Die KTG gratuliert ihren Mitgliedern
sehr herzlich zum Geburtstag und wünscht ihnen weiterhin alles Gute!
Top
IAEA Expands International
Cooperation on Small,
Medium Sized or Modular
Nuclear Reactors
(iaea) The International Atomic
Energy Agency (IAEA) is launching an
effort to expand international cooperation
and coordination in the design,
development and deployment of
small, medium sized or modular
reactors (SMRs), among the most promising
emerging technologies in
nuclear power.
Significant advances have been
made on SMRs, some of which will use
pre-fabricated systems and components
to shorten construction schedules
and offer greater flexibility and
affordability than traditional nuclear
power plants. With some 50 SMR concepts
at various stages of development
around the world, the IAEA is forming
a Technical Working Group (TWG) to
guide its activities on SMRs and provide
a forum for Member States to
share infor mation and knowledge,
IAEA Deputy Director General Mikhail
Chudakov said.
“Innovation is crucial for nuclear
power to play a key role in de carbonising
the energy sector,” Chudakov,
who heads the IAEA Department of
Nuclear Energy, said at a conference
on SMRs in Prague on 15 February.
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“Many Member States that are
operating, expanding, introducing or
considering nuclear power are quite
keen on the development and
deployment of SMRs.”
Global interest in SMRs is growing.
SMRs have the potential to meet the
needs of a wide range of users and to
be low carbon replacements for ageing
fossil fuel fired power plants. They
also display enhanced safety features
and are suitable for non-electric applications,
such as cooling, heating and
water desalination. In addition, SMRs
offer options for remote regions with
less developed infrastructure and for
energy systems that combine nuclear
and alternative sources, including
renewables.
The first three advanced SMRs are
expected to begin commercial operation
in Argentina, China and the
Russian Federation between 2018 and
2020. SMR development is also well
advanced in about a dozen other
countries.
The TWG, comprising some 20
IAEA Member States and international
organizations, is scheduled to
meet for the first time on 23-26 April
at the IAEA’s headquarters in Vienna.
It is part of an expanding suite of
services the IAEA offers Member
States on this emerging nuclear power
technology. These include an SMR
computer simulation programme to
help educate and train nuclear professionals;
a methodology and related
IT tool for training in assessing the
reactor technology of different SMRs;
and the SMR Regulators’ Forum.
The forum, set up in 2015, enables
discussions among Member States and
other stakeholders to share SMR
regulatory knowledge and experience.
It contributes to enhancing safety by
identifying and resolving issues that
may challenge regulatory reviews of
SMRs and by facilitating robust and
thorough regulatory decisions.
Responding to requests from
Member States in Europe, the IAEA
recently launched a project to build
regional capacities for making knowledgeable
decisions on SMRs, including
technical assessments for SMRs
that are commercially available for
near term deployment. The two-year
project seeks to contribute to meeting
growing European demand for
flexible sources of electricity that do
not release greenhouse gases. Its first
meeting will be held on 13-15 March
at the IAEA in Vienna.
An expeditious deployment of
SMRs faces challenges, including the
need to develop a robust regulatory
| | IAEA Expands International Cooperation on Small, Medium Sized or Modular Nuclear Reactors.
framework, new codes and standards,
a resilient supply chain and human
resources. And although SMRs require
less upfront capital per unit, their
electricity generating cost will
probably be higher than that of large
reactors. Their competitiveness must
be weighed against alternatives and
be pursued through economies of
scale. Detailed technical information
on SMRs under construction or design
can be found at the IAEA’s Advanced
Reactor Information System.
“Realistically, we could expect the
first commercial SMR fleet to start
between 2025 and 2030,” said Hadid
Subki, Scientific Secretary of the TWG
and a Team Leader in SMR Technology
Development at the IAEA. “We
trust this new Technical Working
Group will help further the advancement
of SMR technology and guide
the Agency in its programmes and
projects in this field.”
| | (18791436), www.iaea.org
World
Poll Shows Local Residents
Support Poland’s Plans for
First Nuclear Plant
(nucnet) A poll carried out for Poland’s
PGE EJ1, the company in charge of the
country’s first nuclear power station
project, has shown that 67% of residents
in areas around the proposed
site in northern Poland support the
potential construction of a nuclear
power station in their region.
PGE said a poll was carried out
in November and December 2017
in three municipalities, Choczewo,
Gniewino, Krokowa, all close to
Poland’s Baltic coast in the northern
province of Pomerania.
According to PGE, local residents
indicated they are in favour of the
project because of the development
and job opportunities it could bring to
their regions. The poll showed 49% of
respondents expect cheaper electricity
to be one of the benefits from a
nuclear station, while 35 % expect
local infrastructure development.
In April 2017, PGE began environmental
and site selection surveys at
two locations – Lubiatowo-Kopalino in
the municipality of Choczewo and
Żarnowiec in the municipality of
Korkowa.
The studies aim to determine the
potential impact of the project on both
the environment and local residents.
An initial round of environmental
studies has already been carried out at
both locations.
The Polish government has not
made a final decision about the
country’s nuclear programme, with
the deadline being pushed back
several times. According to latest
reports, a decision is now expected in
mid-2018.
| | pgeej1.pl
SKB, Sweden: Two Statements
on the Spent Fuel Repository
(skb) The answer was a clear yes in
SSM’s statement to the Government
on SKB’s system for final disposal of
spent nuclear fuel. The Land and
Environment Court was also positive
| | Aerial photo of the planned site of the Spent Fuel Repository (centre)
at Forsmark. The picture is a photomontage. Illustration: Phosworks.
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262
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*)
Net-based values
(Czech and Swiss
nuclear power
plants gross-based)
1)
Refueling
2)
Inspection
3)
Repair
4)
Stretch-out-operation
5)
Stretch-in-operation
6)
Hereof traction supply
7)
Incl. steam supply
8)
New nominal
capacity since
January 2016
9)
Data for the Leibstadt
(CH) NPP will
be published in a
further issue of atw
BWR: Boiling
Water Reactor
PWR: Pressurised
Water Reactor
Source: VGB
in several important respects but calls
for more documentation on the
copper canisters.
The Swedish Radiation Safety
Authority (SSM) has reviewed SKB’s
applications under the Nuclear Activities
Act and recommends the Government
to grant a licence for a final
repository for spent nuclear fuel in
Forsmark and an encapsulation plant
in Oskarshamn.
The statement from the Land and
Environment Court (MMD) is also
positive in several important respects.
The court says yes to the issues
relating to the Forsmark site, the rock,
the buffer and the environmental
impact statement. The court also
approves the encapsulation plant and
increased capacity in the interim
storage facility Clab. However, the
court wants SKB to present more
documentation on the properties of
the canister and safety in the
long term. Furthermore, it wants
an investigation of the issue of responsibility
after closure, which has also
been requested by the munici pality.
We can conclude that we have
not been able to answer the court’s
questions regarding the copper
canister fully. At the same time, the
Government’s expert authority SSM
wrote in its statement that SKB has
the potential to meet the legislative
requirements on safe final disposal,
says SKB’s managing director Eva
Halldén in a comment.
SKB will provide documentation
That the two authorities have come
to such different conclusions is in
part due to the fact that they have
tried the applications under different
legislations, SSM under the Nuclear
Activities Act and MMD under the
Environmental Code. They also have
different licensing procedures. SSM
grants a licence in several steps with
continuous updates of the safety
analysis. But the court must say yes
or no based on the currently available
documentation.
The issue now lies with the
Ministry of the Environment and
Energy for further investigation and
SKB is working to develop the documentation
on the canister required
by the court.
This is material that we have
planned to produce for the preliminary
safety analysis. The difference
now is that we will prioritise the work
Operating Results November 2017
Plant name Country Nominal
capacity
Type
gross
[MW]
net
[MW]
Operating
time
generator
[h]
Energy generated. gross
[MWh]
Month Year Since
commissioning
Time availability
[%]
Energy availability
[%] *) Energy utilisation
[%] *)
Month Year Month Year Month Year
OL1 Olkiluoto BWR FI 910 880 720 660 116 6 744 904 253 976 759 100.00 94.12 99.88 92.69 100.75 92.46
OL2 Olkiluoto BWR FI 910 880 720 662 904 5 794 145 243 611 284 100.00 79.69 99.64 78.73 101.18 79.43
KCB Borssele PWR NL 512 484 720 367 102 3 021 010 157 825 451 99.78 74.15 99.78 74.55 99.75 72.16
KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0
KKB 2 Beznau 7) PWR CH 380 365 720 276 072 2 646 900 130 879 056 100.00 87.20 100.00 86.71 100.93 86.16
KKG Gösgen 7) PWR CH 1060 1010 720 768 486 7 788 300 304 398 935 100.00 92.37 99.99 91.99 100.69 91.66
KKM Mühleberg BWR CH 390 373 720 278 340 2 838 690 124 050 935 100.00 92.24 99.85 91.61 99.12 90.80
CNT-I Trillo PWR ES 1066 1003 720 764 776 7 740 744 238 234 461 100.00 91.36 99.95 91.07 99.24 90.09
Dukovany B1 PWR CZ 500 473 720 362 651 2 456 677 108 267 051 100.00 62.82 100.00 62.46 100.74 61.29
Dukovany B2 PWR CZ 500 473 720 360 040 2 950 413 104 273 041 100.00 75.24 100.00 74.70 100.01 73.61
Dukovany B3 PWR CZ 500 473 655 314 334 2 623 607 102 248 463 90.97 75.75 86.99 65.95 87.32 65.46
Dukovany B4 PWR CZ 500 473 361 174 635 2 371 933 102 900 084 50.14 69.14 48.36 59.29 48.51 59.18
Temelin B1 PWR CZ 1080 1030 720 781 214 8 664 341 106 292 500 100.00 100.00 99.96 99.96 100.47 100.08
Temelin B2 PWR CZ 1080 1030 720 787 897 6 819 241 100 683 563 100.00 78.34 100.00 78.01 101.32 78.77
Doel 1 PWR BE 454 433 720 325 983 3 277 563 133 890 536 100.00 90.76 99.47 90.23 99.32 89.84
Doel 2 PWR BE 454 433 720 328 778 3 268 119 131 921 768 100.00 91.40 99.71 91.03 100.27 89.30
Doel 3 PWR BE 1056 1006 0 0 6 732 621 251 169 221 0 78.97 0 78.79 0 79.12
Doel 4 PWR BE 1084 1033 720 773 286 7 054 678 253 727 128 100.00 83.08 97.81 82.31 98.06 80.49
Tihange 1 PWR BE 1009 962 158 124 135 2 815 111 290 078 185 21.98 36.57 17.32 35.78 17.03 34.79
Tihange 2 PWR BE 1055 1008 720 766 981 6 637 622 248 156 690 100.00 82.44 100.00 78.63 101.62 78.83
Tihange 3 PWR BE 1089 1038 701 759 129 8 614 260 268 094 957 97.37 99.76 96.66 99.69 96.71 98.57
Operating Results January 2018
Plant name
Type
Nominal
capacity
gross
[MW]
net
[MW]
Operating
time
generator
[h]
Energy generated, gross
[MWh]
Time availability
[%]
Energy availability Energy utilisation
[%] *) [%] *)
Month Year Since Month Year Month Year Month Year
commissioning
KBR Brokdorf DWR 1480 1410 744 957 521 957 521 341 149 580 100.00 100.00 93.97 93.97 86.59 86.59
KKE Emsland 4) DWR 1406 1335 744 1 010 637 1 010 637 336 333 920 100.00 100.00 100.00 100.00 96.55 96.55
KWG Grohnde DWR 1430 1360 744 977 338 977 338 367 604 917 100.00 100.00 94.85 94.85 91.28 91.28
KRB C Gundremmingen 4) SWR 1344 1288 744 982 159 982 159 321 562 051 100.00 100.00 100.00 100.00 97.58 97.58
KKI-2 Isar DWR 1485 1410 744 1 082 908 1 082 908 342 681 231 100.00 100.00 99.98 99.98 97.73 97.72
KKP-2 Philippsburg 1,2,4) DWR 1468 1402 744 1 062 603 1 062 603 356 230 119 100.00 100.00 99.92 99.92 96.06 96.06
GKN-II Neckarwestheim DWR 1400 1310 744 1 006 200 1 006 200 321 129 334 100.00 100.00 99.40 99.39 96.80 96.80
News
atw Vol. 63 (2018) | Issue 4 ı April
differently and complete it faster
than what was planned, says Helene
Åhsberg, SKB’s project manager for
the licensing process.
No referendum
Östhammar Municipality planned to
hold a referendum on the final repository
on March 4. But at a meeting in
the municipal council in the end of
January, it was decided to cancel the
referendum.
| | (18791534), www.skb.se
Yucca Mountain:
Can the US Finally End
the $12 Billion Impasse?
(nucnet) A US federal advisory panel
recently took a step in what could be a
lengthy process to determine if a deep
geological nuclear waste repository
should finally be built at Yucca Mountain,
a project that has been on the
drawing board since the 1970s at a
cost of around $ 12 bn (€ 9.7 bn).
The panel held a meeting to receive
input on reconstructing an electronic
library for documents needed to
decide on the US Department of
Energy’s Yucca licence application.
The meeting, at the Nuclear Regulatory
Commission’s headquarters in
Maryland, came one week after
another development: the White
House pledged $120m of funding in
its 2019 federal budget proposal to
restart licensing for the Yucca site,
north of Las Vegas in Nevada, and
to establish an interim storage programme
to address the growing
stockpile of nuclear waste produced
by nuclear plants across the nation.
After decades of wrangling, could
the US finally be on course to resolve
the question of what to do with
the high-level nuclear waste from
the nation’s 99 commercial nuclear
reactors?
| | www..energy.gov
US Nuclear Industry Calls
for Advanced Reactor Fuel
Cycle Infrastructure
(nucnet) The US Nuclear Energy
Institute has warned that preparations
should begin now to develop a
national fuel cycle infrastructure to
support the operation of advanced
reactors that are expected to begin
deployment in the 2020s and 2030s.
The Washington-based nuclear
industry lobby group said interest in
the development of advanced nuclear
reactor designs has been increasing in
recent years. Many of these designs
will require uranium fuel that is
enriched to a higher degree than
in the current worldwide fleet of lightwater
reactors. Fuel for advanced
reactors, enriched in U-235 to
between 5% and 20%, is called
high-assay low-enriched uranium
(HALEU).
Some of the advanced-performance
fuels being developed for use
with the existing reactor fleet also will
require HALEU. However, there are no
US-based facilities that manufacture
HALEU on a commercial scale. While
small quantities of HALEU materials
may be obtained on an interim basis
by “blending down” existing government
stocks of surplus high-enriched
uranium (HEU), those HEU materials
are limited in supply and not readily
available, the NEI said.
“Thus, for the long-term operation
of advanced reactors, as well as for
advanced fuels in existing reactors, a
robust new infrastructure for HALEU
fuel manufacture is needed.”
An NEI white paper says establishing
such a capability will better
position the US to advance nuclear
safety and non-proliferation policies
around the world, while helping to
ensure a robust commercial industry
domestically in the decades ahead.
On the other hand, “if the United
States and its allies have to depend on
foreign, state-owned enterprises to
meet fuel needs, it will be in a much
weaker position to influence these
policies globally”, the paper says.
| | Details online:
http://bit.ly/2FnZwOF
Reactors
IAEA Sees Safety Commitment
at Spain’s Almaraz
Nuclear Power Plant
(iaea) An International Atomic Energy
Agency (IAEA) team of experts said
the operator of Spain’s Almaraz
Nuclear Power Plant demonstrated a
commitment to the long-term safety of
the plant and noted several good practices
to share with the nuclear industry
globally. The team also identified areas
for further enhancement.
The Operational Safety Review
Team (OSART) today concluded an
18-day mission to Almaraz, whose
two 1,050-MWe pressurized-water
reactors started commercial operation
in 1983 and 1984, respectively.
Centrales Nucleares Almaraz-Trillo
(CNAT) operates the plant, located
about 200 km southwest of Madrid.
OSART missions aim to improve
operational safety by objectively
assessing safety performance using
the IAEA’s safety standards and proposing
recommendations for improvement
where appropriate. Nuclear
power generates more than 21 per
cent of electricity in Spain, whose
seven operating power reactors all
began operation in the 1980s.
“The team saw notable achievements
made by Almaraz in recent
years, such as implementing a comprehensive
management system, as
well as significant equipment renewal
plans, to establish safety as the
overriding priority at the plant,” said
Team Leader Peter Tarren, Head of the
IAEA’s Operational Safety Section.
“We found that people at every
level were willing to discuss their
work and how they might learn from
this OSART mission. They want to
keep enhancing the safety and
reliability of Almaraz.”
The 14-member team comprised
experts from Brazil, Bulgaria, France,
Germany, Mexico, the Russian Federation,
Sweden, United Arab Emirates,
the United Kingdom and the United
States of America, as well as three
IAEA officials.
The review was the 200th OSART
mission conducted by the IAEA since
the service was launched in 1982. It
covered the areas of leadership and
management for safety; training
and qualification; operations; maintenance;
technical support; operating
experience; radiation protection;
chemistry; emergency preparedness
and response; accident management;
human, technology and organizational
interactions and long-term
operation.
The team identified a number of
good practices that will be shared
with the nuclear industry globally,
including:
The use of a film-forming amine
compound to significantly reduce
the transport of potential corrosive
products to the steam generators.
The use of a cross-functional
indicator to show the cumulative
effect of equipment status and
planned activities for daily operations.
The installation of a centralized
vacuum system for cleaning, decontaminating
and discharging liquid
waste into the plant´s disposal system.
The mission made a number of
recommendations to improve operational
safety, including:
The plant should implement
further actions related to management,
staff and contractors to enforce
standards and expectations related
to industrial safety.
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The plant should take measures
to reinforce and implement standards
to enhance the performance of reactivity
manipulations in a deli berate
and carefully-controlled manner.
The plant should improve the
support, training and documented
guidance for Severe Accident Management
Guideline users in order to
mitigate complex severe accident
scenarios.
The team provided a draft report of
the mission to the plant’s management.
The plant management and the
Nuclear Safety Council (CSN), which
is responsible for nuclear safety
oversight in Spain, will have the
opportunity to make factual comments
on the draft. These will be
reviewed by the IAEA and the final report
will be submitted to the
Government of Spain within three
months.
The plant management said it
would address the areas identified
for enhancement and requested a
follow-up OSART mission in about
18 months.)
| | (18791443), www.iaea.org
Tianwan-3 Passes Commissioning
Tests at 100% Power
(nucnet) The Tianwan-3 nuclear
reactor unit in Jiangsu province,
northeastern China, has successfully
operated for 100 hours at 100% of its
design power level without interruption,
Russian state nuclear corporation
Rosatom said.
Rosatom said the 990-MW VVER
V-428M unit, which started to deliver
electrical energy to the grid on
30 December 2017, has undergone a
series of tests during the 100-hour
operation period required by regulators
before giving green light for
commercial operation.
Construction of Tianwan-3 began
in December 2012. The Tianwan
| | Swiss regulator approves safety case for restart of Beznau-1 (Photo: Axpo).
nuclear station is the largest economic
cooperation project between Russia
and China, an earlier statement had
said.
Tianwan-1 and -2, also VVER
V-428M units, began commercial
operation in 2007. The Tianwan-4
VVER V-428M unit is also under construction
by Russia while Tianwan-5
and -6 will be indigenous Generation
II+ CNP-1000 units.
| | en.cnnc.com.cn
Swiss Regulator Approves
Safety Case for Restart of
Beznau-1
(nucnet) Switzerland’s Federal
Nuclear Safety Inspectorate, ENSI,
has given the go-ahead for the restart
of the Beznau-1 nuclear unit after
approving the safety case presented
by owner Axpo following the discovery
in 2015 of flaw indications in
the reactor pressure vessel (RPV).
ENSI said in a statement that
Axpo had carried out “extensive
investigations and analyses” to
demonstrate that the RPV is safe.
Materials testing has shown
that agglomerates in the RPV do not
affect its key properties and structural
integrity analysis has shown that
the RPV does not contain any flaws
that could lead to its failure. “IRSN
is satisfied that work has been done
to all appropriate national and international
standards,” the statement
said.
Axpo said the safety case for
Beznau-1, the world’s oldest commercial
nuclear plant still in operation,
corroborates earlier assessments
and investigations, and validates the
existing safety margin for the safe
operation of the plant for 60 years.
Operator KKB will now begin the
return to service process with the
plant expected to be operating at full
load by the end of March 2018.
In December 2015 Axpo submitted
a roadmap ENSI detailing plans for
further investigations of flaw indications
in the RPV. During a scheduled
outage that began in May 2015,
inspections of the RPV registered
findings at some points in the base
material of the RPV indicating
“ minimal irregularities in the fabrication
process”, Axpo said. The company
carried out further measurements
and analyses and submitted a
report to ENSI.
In July 2015, Axpo announced
that the restart of Beznau-1 had been
postponed while the flaw indications
were investigated further. Then in
August, ENSI called for additional
investigations.
Beznau-2 was not affected by the
flaw indications and was returned to
service after its scheduled outage in
2015.
| | www.bkw.ch
Kursk II Passed
Construction Milestone
(rosatom) Kursk II began reinforcing
the foundation slab for the reactor
building of Unit 1. This operation
became the year’s key event on the
construction site of the Kursk plant.
On 21 December 2017, the first
16-ton reinforced concrete block was
installed on the rebar of the lower
foundation belt. According to the
project design, the foundation comprises
105 reinforced concrete blocks
with a total weight of 1,600 tons. This
will enable the construction team
to start concreting the foundation
slab of the reactor building in the
first half of 2018.
Prior to putting the first concrete
block, a rebar coupler engraved with
the words “The future is shaped today.
The first coupling sleeve of the innovative
VVER-TOI power unit” was
ceremonially installed into the foundation
reinforcement.
VVER-TOI (which means ‘a standard
optimized and automated power
unit based on VVER technology’)
reactors meet Russian and global
safety requirements and have a longer
service life and higher installed
capacity than existing reactors of
the Kursk Nuclear Power Plant.
Alexander Mikhailov, Governor of
the Kursk Region, noted that it was
an honor for the region to build
and commission one of the world’s
first nuclear plants with advanced
VVER-TOI reactors. “Construction of
Kursk II designed to meet the latest
global standards offers our region
development prospects for the entire
News
atw Vol. 63 (2018) | Issue 4 ı April
Rosenergoatom had planned to
build two BN-1200 units at Beloyarsk
with commercial operation scheduled
by 2025. But construction depended
on the results of operating the pilot
Beloyarsk-4 BN-800 plant, which
began commercial operation in
October 2016.
There is another commercially
operational sodium-cooled FBR at
Beloyarsk, the BN-600. Both the
BN-600 and the BN-800 are smaller
versions of the BN-1200. There are
also two permanently shut-down
light-water reactors at the site.
| | www.rosatom.ru
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| | Kursk II passed construction milestone.
21st century. Just a few Russian
regions have such opportunities,” he
stressed.
Vyacheslav Fedyukin, Director
of Kursk NPP, noted it was symbolic
that the event happened on the
25 th anniversary of RosEnergoAtom
and 10 years after the foundation of
Rosatom, the companies that shaped
the newest history of Russia’s nuclear
industry. “Construction of Russia’s
first VVER-TOI-based power unit
proves that the national nuclear
power industry is always at the
cutting edge of science and engineering.
The new generation VVER-TOI
units are state-of-the-art facilities
made to the best of Russia’s nuclear
engineering knowledge,” he added.
At the moment, other operations
are also underway at the construction
site of Kursk II. Among them is excavation
of 1.2 million cub m of soil
to be completed in 2017, with over
800,000 cub m of sand, gravel
and aggregate already put in the
foun dation of Kursk II buildings
and structures. Construction of a
330/10 kV substation and preparation
of technical documents for its commissioning
are also drawing to a
close.
For reference:
Kursk II is designed to replace the
existing Kursk Nuclear Power Plant
that will be taken out of operation in
the years to come. Its first two units
with VVER-TOI, a new-type reactor,
will be commissioned simultaneously
with decommissioning of Units 1 and
2 of the existing nuclear station.
According to the master schedule of
Kursk II, Unit 1 will be commissioned
in late 2023 to be followed by Unit 2 in
late 2024.
| | (18791501),
ww.rosatom.ru
Russia Confirms Plans to
Revive BN-1200 Fast Breeder
Reactor Project
(nucnet) Russia plans to begin construction
of its first industrial-sized
sodium-cooled fast neutron reactor in
the 2020s after saying three years ago
that the project had been postponed,
the head of state nuclear corporation
Rosatom Alexei Likhachev told president
Vladimir Putin.
According to a transcript of a
meeting posted on the Kremlin’s
website, Mr Likhachev told Mr Putin
that fast breeder reactors (FBRs) have
significant advantages over existing
reactor types and Rosatom is proposing
that Russia goes ahead
with its plans for the Generation IV
BN-1200.
FBRs have been and are being
explored or constructed in Russia,
France, India, China, Japan and the
US. They allow a significant increase
in the amount of energy obtained
from natural, depleted and recycled
uranium. The technology also enables
plutonium and other actinides to be
used and recycled.
Russia operates the BN-600 and
BN-800 FBR units at Beloyarsk and
the BOR-60 fast breeder research
reactor at the Research Institute
of Atomic Reactors (RIAR) site in
Dimitrovgrad, southwest Russia.
BOR-60 is used to test fuel cycle,
sodium coolant technologies and a
range of design concepts for fast
breeder reactors.
In 2015, Rosatom said construction
of the planned BN-1200 at the
Beloyarsk nuclear power station in
central Russia had been postponed
until at least 2020, with state
nuclear operator Rosenergoatom
citing the need to improve fuel
for the reactor and questioning the
project’s economic feasibility.
Austria Begins Legal Action
Against EC Over Hungary’s
Paks Nuclear Project
(nucnet) Austria has filed a legal
complaint against the European Commission
with the European Court of
Justice in Luxembourg for allowing
Hungary to expand its Paks nuclear
power station.
Austrian minister of sustainability
and tourism Elisabeth Köstinger said
in a statement that nuclear power
“must have no place in Europe” and
Austria will not “not budge one
centimetre” from its anti-nuclear
stance.
The EC started an investigation
into state aid given to the Paks 2
project in November 2014. Last March
it approved the project to build two
new reactors, to be financed with the
help of Russia’s state atomic energy
corporation Rosatom, after regulators
said Hungarian authorities had
agreed to several measures to ensure
fair competition.
In January 2018, Austria announced
it planned to sue the EC over
the decision. “EU assistance is only
permissible when it is built on common
interest. For us, nuclear energy is
neither a sustainable form of energy
supply, nor is it an answer to climate
change”, a statement by the ministry
of sustainability said at the time.
The two planned units at Paks 2
nuclear power station are expected
to begin commercial operation in
2026 and 2027, Attila Aszódi,
the Hungarian government’s commissioner
for the Paks 2 project,
told a conference in Brussels late l
ast year.
An agreement signed in 2014
will see Russia supply two VVER-
1200 pressurised water reactors for
Paks 2 and a loan of up to €10bn
($12.3bn) to finance 80% of the
€12bn project.
| | www.bundeskanzleramt.gv.at
News
atw Vol. 63 (2018) | Issue 4 ı April
266
NEWS
Company News
Framatome Completes
Purchase of Schneider
Electric’s Instrumentation and
Control Nuclear Business
(framatome) Framatome announced
that it completed its purchase of
Schneider Electric’s nuclear instrumentation
and control offering. With
this transaction, Framatome adds to
its engineering expertise and expands
its instrumentation and control (I&C)
offerings.
I&C systems are the central nervous
system of a nuclear power plant,
allowing operators to control reactor
operations. Modernizations, upgrades
and ongoing support are vital to manage
economic, long-term operation of
nuclear power plants, which provide
reliable, low-carbon electricity.
“With the integration of Schneider
Electric’s nuclear instrumentation
and control offering, we offer truly
added value to our customers with
a global technical expertise and
market know-how on I&C solutions
for the nuclear market,” said Bernard
Fontana, Chairman of the Managing
Board and Chief Executive Officer of
Framatome. “We welcome our new
colleagues to Framatome’s worldwide
team of I&C engineers and experts.”
This acquisition adds the nuclearqualified
version of Tricon and the
SPEC 200 platform to Framatome’s
nuclear safety I&C offerings, which
include the TELEPERM XS digital
platform, and non-computerized
analog solutions and instrumentation
for nuclear power plants.
This broadens the base of plants
worldwide for which Framatome
serves as the original equipment manufacturer
for safety I&C systems. It also
expands Framatome’s project and
engineering capacities for non-safety
I&C systems in the nuclear energy
market, relying on Schneider Electric’s
commercial TRICON and Foxboro
platforms.
Framatome also becomes the exclusive
service provider to the nuclear energy
market for the SPEC 200, nuclearqualified
Tricon and Foxboro systems.
| | www.framatome.com
Framatome Continues
Ramping up Production
at Its Le Creusot Site
(framatome) On January 25, 2018,
Framatome received the green light
from the French Nuclear Safety
Authority (ASN) and EDF to resume
manufacture of forgings for the
French nuclear fleet at its Le Creusot
site. This decision allows the plant to
continue ramping up its production
with a target of 80 ingots per year.
The authorization is an outcome of
the improvement plan launched at the
beginning of 2016 on the site following
a series of quality audits. With the completion
of all the actions necessary for
the resumption of production for the
French nuclear fleet and overall progress
of 90% to date, the plan will be
fully closed out in the first half of 2018.
The actions will then be incor porated
into the site’s continuous improvement
processes. Customers in France and
abroad, as well as all the safety
authorities concerned, have been kept
regularly informed of the actions
undertaken. Numerous reviews and
inspections have been conducted in order
to observe the progress of the plan
and integrate stakeholders’ feedback.
David Emond, Senior Executive
Vice President of Framatome’s Component
Manufacturing Business Unit,
comments: “The authorization to
resume manufacture of forgings for
the French nuclear fleet is a very
good news for the site that confirms
the successful execution of its improvement
plan. The 230 employees
at the Le Creusot site are engaged
in its deployment on a day to day basis
so that we can supply our customers
with equipment meeting the most
stringent safety and quality requirements
within agreed deadlines. I
want to thank them for the substantial
work they have accomplished on
the site over the last two years.”
Maintaining and developing the
skills of the Le Creusot plant teams
is a key element of the site’s improvement
plan, with a particular focus
on strengthening the nuclear safety
culture.
Framatome already invested
7.5 million euros at the site in 2017
to make the Le Creusot site a center
of excellence for the manufacture
of forgings for the nuclear industry,
and will pursue this effort in 2018.
Major milestone reached
in review of manufacturing
records
Moreover, a major milestone has
been reached in the review of legacy
manufacturing records at the Le
Creusot site. The first stage in the
inspection process which is being
applied to all records relating to
forgings produced for the nuclear
industry, a key stage consisting in
identifying findings, is now complete.
The analysis of these findings and the
processing of deviations will continue
until the end of 2018, in coordination
with customers and safety authorities.
Of the 6,000 records identified
during the initial survey, 3,854 correspond
to forgings installed on nuclear
installations.
At Framatome’s Jeumont and
Saint-Marcel sites, the audit has been
finalized since the summer of 2017 and
no deviation impacting the safety of
components has been brought to light.
| | www.framatome.com
JNFL and MHI Become
Shareholders of
Orano 2017 Revenue
(orano) The Orano Board of Directors
noted the completion of the capital
increase reserved for Japan Nuclear
Fuel Limited (JNFL) and Mitsubishi
Heavy Industries, Ltd. (MHI) for
a total of €500 million.
Pursuant to the initial agreements
signed with JNFL and MHI in
March 2017, the funds corresponding
to their total investment in Orano
had been placed in trust on July 26,
at the same time as the completion
of the capital increase reserved for
French State 2. These funds were
released and used for the subscription
of JNFL and MHI to Orano’s second
capital increase.
This transaction follows the
completion on December 31, 2017 of
the sale of the majority control of
Framatome (formerly New NP) by
AREVA SA to EDF as well as the
fulfillment of the regulatory closing
conditions related to the addition of
an equity stake in Orano of both Japanese
investors.
Orano’s capital is now held by the
French State (45.2%), the CEA
(4.8%)3, AREVA SA (40%), JNFL
(5%) and MHI (5%).
This transaction is the last major
step in the restructuring of the French
nuclear industry, undertaken in 2015,
and marks the end of the constitution
phase of the Orano group. With a
strengthened financial structure and
sound strategic partnerships, Orano
now has the means to grow and reach
its goal of being a leading player in the
production and recycling of nuclear
materials, in waste management and
dismantling.
Appointment of a new
independent director
After completion of Orano’s second
capital increase, the Orano General
Meeting, also held on February 26,
2018, appointed Patrick Pelata as
independent director.
| | www.orano.group
News
atw Vol. 63 (2018) | Issue 4 ı April
Westinghouse Electric
Company Signs Cooperation
Agreement for Lead-cooled
Fast Reactor Development
(westinghouse) Westinghouse Electric
Company has signed a Cooperation
Agreement for lead-cooled fast
reactor (LFR) technology development
with the Italian National Agency
for New Technologies, Energy and
Sustainable Economic Development
(ENEA) and Ansaldo Nucleare. The
agreement demonstrates each party’s
commitment to collaborating toward
the development of a next-generation
nuclear plant based on LFR technology,
which is both “walk-away”
safe and economically competitive
across global energy markets.
“This agreement is an exciting
step towards the development of a
lead-cooled fast reactor for the
marketplace,” said Ken Canavan,
Westinghouse chief technology officer
and vice president, Global Technology
Office. “The LFR is game-changing
technology for clean energy industries,
and Westinghouse is pleased to
be working with such experienced
partners to bring this innovative
concept to fruition.”
Beyond baseload electricity
generation, the high-temperature
operation of the LFR will allow for
a broad range of applications such
as an effective load-following
capability enabled by an innovative
thermal energy storage system,
delivery of process heat for industrial
applications and water desalination.
ENEA is a world leader in research
and development on lead-based
systems, and currently operates
among the finest and largest experimental
facilities for LFR research in
the world.
Ansaldo Nucleare has vast experience
in nuclear power plant design,
supply, service and decommissioning,
and has played leading roles in
multiple international LFR development
programs for the past 15 years.
| | www.westinghousenuclear.com
BKW übernimmt Experten
für Strahlenschutz
(bkw) Die BKW Konzerngesellschaft
Dienstleistungen für Nukleartechnik
(DfN) übernimmt das ebenfalls
auf den kerntechnischen Bereich
spezia lisierte Unternehmen Technischer
Strahlenschutz (TSS). Dadurch
stärkt die BKW ihre Kompetenzen in
diesem Gebiet und baut sie weiter aus.
Dies vor dem Hintergrund der geplanten
Stilllegung des Kernkraftwerks
Mühleberg und zahlreicher weiterer
Kernkraftwerke in Europa.
Mit der Übernahme des Strahlenschutzunternehmens
DfN hat die BKW
bereits im letzten Jahr ihre bestehenden
und bewährten Kom petenzen im hochspezialisierten
Nukleartechnik-Bereich
erweitert. Der Eintritt der TSS in den
Unter nehmensverbund der BKW stellt
nun einen weiteren Ausbau in diesem
Gebiet dar. Die TSS ergänzt die
Strahlenschutzkompetenzen innerhalb
der BKW Gruppe und verstärkt diese
auch im Hinblick auf die Still legung
des Kernkraftwerks Mühleberg.
In Europa ist ausserdem eine Vielzahl
weiterer Stilllegungsprojekte in
Planung oder bereits im Gang. Der
Strahlenschutz spielt bereits beim
Betrieb von Kernkraftwerken eine
wichtige Rolle. Mit der Stilllegung
und den dabei ausgeführten Demontage-
und Freimessarbeiten nehmen
die Strahlenschutzarbeiten zu. Für
Strahlenschutzdienstleisterinnen wie
TSS und DfN bietet der wachsende
Stilllegungsmarkt daher ein grosses
Potenzial und die Möglichkeit, sich
weiterzuentwickeln.
Die DfN und die TSS haben bereits
verschiedentlich auf Projektbasis
zusammengearbeitet. Die erfolgreiche
Kooperation wird künftig
weiter ausgebaut, was mit einer
gegenseitigen Stärkung einhergeht.
Um eine optimale Zusammenarbeit
zu ermöglichen, wird die TSS in die
DfN integriert.
Die TSS mit Sitz in Geilenkirchen
im deutschen Bundesland Nordrhein-
Westfalen wurde 1979 gegründet
und zählt 15 Mitarbeitende. Das
Unternehmen bietet ein qualitativ
hochwertiges und breites Angebot von
Dienstleistungen im kerntechnischen
Bereich. Dazu gehören neben dem
Strahlenschutz die Dekontamination,
die Abfallentsorgung, die Dosimetrie
sowie die Abwicklung von Transporten
radioaktiver Stoffe.
| | (18791521), www.bkw.ch
Companies
China Approves $ 100 Billion
Merger of Leading
Nuclear Companies
(nucnet) China has approved the
merger of nuclear power producer
China National Nuclear Corporation
(CNNC) with nuclear plant builder
China Nuclear Engineering and
Construction Corporation (CNECC),
the state-run China Daily news agency
said.
According to the China Daily, the
combined assets of the new company
will be worth about $100bn (€80bn),
while its workforce will be about
150,000 employees.
CNNC is China’s number two
nuclear power producer and CNECC
the country’s top nuclear power plant
builder.
China Daily said the merger is in
line with efforts by China to streamline
the state-operated sector of its
economy and reduce the number of
state-owned companies administered
by central government.
Approval for the merger was
confirmed by the State-Owned Assets
Supervision and Administration
Commission (SASAC) in a one-line
statement posted on its website.
| | (18800822), en.cnnc.com.cn
267
NEWS
Research
| | BKW übernimmt Experten für Strahlenschutz © BKW.
NRG: Every Day,
30,000 Patients Benefit From
Medical Isotopes From Petten
(nrg) Medical isotopes are indispensable
for diagnosing and treating
cancer. Demand for them is set to soar
over the next 20 years, but supplies
are diminishing. To put the spotlight
News
atw Vol. 63 (2018) | Issue 4 ı April
• Separative work: 58.00–92.00
268
NEWS
2016
January to June 2016
• Uranium: 26.50–35.25
• Conversion: 6.25–6.75
• Separative work: 58.00–62.00
July to December 2016
• Uranium: 18.75–27.80
• Conversion: 5.50–6.50
• Separative work: 47.00–62.00
| | NRG: Every day, 30000 patients benefit from medical isotopes from Petten View of the pool type reactor
core. (Courtesy: JRC)
on the world of medical isotopes, the
platform 30000perdag.nl has been
launched. The aim of the platform and
the accompanying campaign is to
boost awareness that the Netherlands
must continue leading the field in
cancer treatment.
The future
Over the next 20 years, the number of
cancer diagnoses is expected to rise by
70%. Fortunately, health care is
constantly improving, partly through
the use of medical isotopes. However,
there are only 6 reactors worldwide
which can produce medical isotopes,
one of which is closing next year.
This means that whilst demand for
medical isotopes is growing worldwide,
supplies are diminishing.
30000perdag.nl
An online information park for a
wide audience has been built on
30000perdag.nl. Visitors can learn all
about medical isotopes here: from raw
materials to the reactor in Petten to
applications in the hospital. By opening
up that world, NRG in Petten
wants to show (former) cancer
patients and their families and
acquaintances what is needed to be
able to treat cancer, and request
support for medical isotopes and good
cancer treatment in the Netherlands
and abroad.
Former cancer patients play
starring role in campaign
The campaign uses 3 video interviews
with cancer survivors. The interviews
were conducted by presenter Fien Vermeulen,
herself a former lymphoma
patient. Fien drives with former
patients Anouk (26), Alexander (42)
and Manon (34) to the research
reactor in Petten, where they talk
about their remarkable experiences in
times of uncertainty. Each of them
represents one of the 30,000 people
who benefit or have benefitted from
medical isotopes every day.
Anyone can demonstrate their
support by liking the Facebook page
30.000perdag. Another very visible
form of support is available through
the T-shirts that can be ordered via
30000perdag.nl. These enable former
patients and supporters to show their
backing for the campaign.
| | (18800822), www.nrg.eu
Market data
(All information is supplied without
guarantee.)
Nuclear Fuel Supply
Market Data
Information in current (nominal)
U.S.-$. No inflation adjustment of
prices on a base year. Separative work
data for the formerly “secondary
market”. Uranium prices [US-$/lb
U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =
0.385 kg U]. Conversion prices
[US-$/kg U], Separative work
[US-$/SWU (Separative work unit)].
January to December 2013
• Uranium: 34.00–43.50
• Conversion: 9.25–11.50
• Separative work: 98.00–127.00
January to December 2014
• Uranium: 28.10–42.00
• Conversion: 7.25–11.00
• Separative work: 86.00–98.00
January to December 2015
• Uranium: 35.00–39.75
• Conversion: 6.25–9.50
2017
January 2017
• Uranium: 20.25–25.50
• Conversion: 5.50–6.75
• Separative work: 47.00–50.00
February 2017
• Uranium: 23.50–26.50
• Conversion: 5.50–6.75
• Separative work: 48.00–50.00
March 2017
• Uranium: 24.00–26.00
• Conversion: 5.50–6.75
• Separative work: 47.00–50.00
April 2017
• Uranium: 22.50–23.50
• Conversion: 5.00–5.50
• Separative work: 45.50–48.50
May 2017
• Uranium: 19.25–22.75
• Conversion: 5.00–5.50
• Separative work: 42.00–45.00
June 2017
• Uranium: 19.25–20.50
• Conversion: 5.55–5.50
• Separative work: 42.00–43.00
July 2017
• Uranium: 19.75–20.50
• Conversion: 4.75–5.25
• Separative work: 42.00–43.00
August 2017
• Uranium: 19.50–21.00
• Conversion: 4.75–5.25
• Separative work: 41.00–43.00
September 2017
• Uranium: 19.75–20.75
• Conversion: 4.60–5.10
• Separative work: 40.50–42.00
October 2017
• Uranium: 19.90–20.50
• Conversion: 4.50–5.25
• Separative work: 40.00–43.00
November 2017
• Uranium: 20.00–26.00
• Conversion: 4.75–5.25
• Separative work: 40.00–43.00
December 2017
• Uranium: 23.50–25.50
• Conversion: 5.00–6.00
• Separative work: 39.00–42.00
2018
January 2018
• Uranium: 21.75–24.00
• Conversion: 6.00–7.00
• Separative work: 38.00–42.00
News
atw Vol. 63 (2018) | Issue 4 ı April
February 2018
• Uranium: 21.25–22.50
• Conversion: 6.25–7.25
• Separative work: 37.00–40.00
| | Source: Energy Intelligence
www.energyintel.com
Cross-border Price
for Hard Coal
Cross-border price for hard coal in
[€/t TCE] and orders in [t TCE] for
use in power plants (TCE: tonnes of
coal equivalent, German border):
2012: 93.02; 27,453,635
2013: 79.12, 31,637,166
2014: 72.94, 30,591,663
2015: 67.90; 28,919,230
2016: 67.07; 29,787,178
I. quarter: 56.87; 8,627,347
II. quarter: 56.12; 5,970,240
III. quarter: 65.03, 7.257.041
IV. quarter: 88.28; 7,932,550
| | Uranium spot market prices from 1980 to 2018 and from 2008 to 2018. The price range is shown.
In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.
269
NEWS
2017:
I. quarter: 95.75; 8,385,071
II. quarter: 86.40; 5,094,233
III. quarter: 88.07; 5,504,908
| | Source: BAFA,
some data provisional
www.bafa.de
EEX Trading Results
February 2018
(eex) In February 2018, the European
Energy Exchange (EEX) achieved a
total volume of 274.3 TWh on its
power derivatives markets (February
2017: 200.8 TWh) which is a yearon-year
increase of 37 %. In doing
so, EEX was able to grow its power
derivatives volumes across all market
areas.
In total, the German and Austrian
markets (Phelix-DE, Phelix-AT and
Phelix-DE/AT) increased by 12 % to
169.7 TWh. This includes 153.4 TWh
from the benchmark product Phelix-
DE which achieved its highest volume
since launch in April 2017. Volumes
in the French market more than
doubled to 26.9 TWh (February 2017:
12.3 TWh) while Italian power
volumes grew substantially to
40.0 TWh (February 2017: 22.1 TWh).
Furthermore, on the Spanish market,
volumes increased by more than
250 % to 8.3 TWh (February 2017:
2.3 TWh).
The February volume comprised
173.5 TWh traded at EEX via Trade
Registration with subsequent clearing.
Clearing and settlement of all exchange
transactions was executed by European
Commodity Clearing (ECC).
The Settlement Price for base
load contract (Phelix Futures) with
| | Separative work and conversion market price ranges from 2008 to 2018. The price range is shown.
)1
In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.
delivery in 2019 amounted to 33.85 €/
MWh. The Settlement Price for peak
load contract (Phelix Futures) with
delivery in 2019 amounted to 42.40 €/
MWh.
On the EEX markets for emission
allowances, the total trading volume
increased by 57% to 144.2 million
tonnes of CO 2 in February (February
2017: 91.7, million tonnes of CO 2 ).
Primary market auctions contributed
75.1 million tonnes of CO 2 to the total
volume. On the spot secondary
market, volumes more than doubled
to 4.5 million tonnes of CO 2 (February
2017: 2.0 million tonnes of CO 2 ). On
the EUA Futures market, EEX was able
to increase volumes by 80% to
37.3 million tonnes of CO 2 (February
2017: 20.7 million tonnes of CO 2 ).
Furthermore, 27.4 million tonnes of
CO 2 were traded in EUA Options
which is the highest monthly volume
so far in this product.
The EUA price with delivery in
December 2018 amounted to
8.80/10.15 €/ EUA (min./max.).
| | www.eex.com
MWV Crude Oil/Product Prices
January 2017
(mwv) According to information and
calculations by the Association of the
German Petroleum Industry MWV e.V.
in January 2018 the prices for
super fuel, fuel oil and heating oil
noted inconsistent compared with
the pre vious month December 2017.
The average gas station prices for Euro
super consisted of 136.84 €Cent
( December 2017: 136.84 €Cent,
approx. +-0.0 % in brackets: each
information for pre vious month or
rather previous month comparison),
for diesel fuel of 120.48 €Cent
(119.01; +1.24 %) and for heating oil
(HEL) of 62.27 €Cent (60.65 €Cent,
+2.67 %).
The tax share for super with
a consumer price of 136.84 €Cent
(136.84 €Cent) consisted of
65.45 €Cent (47.83 %, 65.45 €Cent)
for the current constant mineral oil
tax share and 21.85 €Cent (current
rate: 19.0 % = const., 22.12 €Cent)
for the value added tax. The product
price (notation Rotterdam) consisted
of 40.17 €Cent (29.36 %, 37.18 €Cent)
and the gross margin consisted of
9.37 €Cent (6.85 %; 12.36 €Cent).
Thus the overall tax share for super
results of 66.83 % (66.83 %).
Worldwide crude oil prices
(monthly average price OPEC/Brent/
WTI, Source: U.S. EIA) were again
significantly higher, approx. +8.36 %
(+2.34 %) in January 2018 compared
to December 2017.
The market showed a stable
development with higher prices; each
in US-$/bbl: OPEC basket: 66.85
(62.06); UK-Brent: 69.08 (64.37);
West Texas Inter mediate (WTI): 63.7
(57.88).
| | www.mwv.de
News
atw Vol. 63 (2018) | Issue 4 ı April
270
Czechs and Balances and Why ‘Ugly’
Nuclear Deserves a Political Makeover
NUCLEAR TODAY
Author
John Shepherd
Shepherd
Communications
3 Brooklands
West Sussex
BN43 5FE
Links to reference
sources:
Dana Drábová
interview:
http://bit.ly/2Ik7WaJ
European Investment
Bank announcement:
http://bit.ly/2Ik7WaJ
Yonhap News
Agency report:
http://bit.ly/2FyvZkw
As if Europe does not have enough on its plate to deal with at the moment – politically and economically just for starters
– could Brussels be on a collision course with the Czech government over the country's plans to expand nuclear energy?
There is certainly friction over the issue between Prague and
the European Commission (EC), to put it mildly. But why?
The veteran head of the Czech Republic’s State Office
for Nuclear Safety, Dana Drábová, last month accused
other EU member states of “pressurising” Prague over the
early closure of its oldest nuclear reactor units.
Drábová reportedly told an energy conference in the
country: “There is immense pressure developing that the
operating life of nuclear reactors will be limited to 40 years.
That means that our political representatives, whoever they
might be, sometime around 2023 will face a battle over a
further 10-year extension for Dukovany. The current State
Energy Framework counts on the lifetime of the Dukovany
reactors ending sometime between 2030 and 2040.”
The nuclear safety chief later told Czech Radio the
pressure was coming from “the 14 countries which are not
using nuclear power and some of which regard it as
something ugly”. If the pressure continued, she predicted
there would be a concerted “willingness… to get rid of
these nuclear plants in Europe as fast as possible”.
Drábová’s comments came against a backdrop of the
Czech government saying it would appoint an expert team
to consider proposals to break up the majority state-owned
electricity firm CEZ. The move was one of several options
mooted to support financing of the construction of a new
nuclear power plant at Dukovany.
Analysts say the new nuclear plant could be built by the
traditional energy unit, which would be fully state-owned
and therefore in the best position to take on the risks of
high costs that the utility could not if it were an entity with
private owners.
Czech prime minister Andrej Babiš is backing proposals
to build the Dukovany reactor, around 50 kilometres north
of the (anti nuclear) Austrian border, to replace a Soviet-era
reactor. But this would mean persuading the EC to exempt
the project from strict EU rules on government bids.
If the Czech government fails in its quest, it could consider
doing a deal with Russia, which would undoubtedly
be very much along the lines of the nuclear construction
and financing deal Moscow signed recently with EU
member Hungary.
If, dear reader, you now have a sense of déjà vu, you
would be right. You may recall that Hungary went through
a similar nuclear battle with the EC, despite Hungary’s parliament
fully backing proposals to build two new nuclear
reactor units in that country.
Initially, the EC said in November 2015 it had started
legal action against Hungary over a contract signed with
Russia’s Rosatom to build two units at the existing Paks
plant. Brussels expressed concern about the project’s
compatibility with EU public procurement rules. However,
the EC eventually cleared the issue and a state aid investigation
into the project financing for the ‘Paks II’ project
was subsequently dropped by the EC.
There was a similar clash with the EC when the UK first
unveiled plans to invest in building the Hinkley Point C
nuclear plant.
So is the latest tussle between Prague and Brussels
really over concerns about state-aid rules or is it more a
worrying trend of interference to stop nuclear in its tracks?
And is the conflict really worth it…?
Czech PM Babiš said following an official visit to
Hungary last January, where he attended a summit of
prime ministers of the Visegrad Group countries, that he
and Hungarian counterpart Viktor Orbán discussed the
potential for “further developing” cooperation in sectors
such as the nuclear energy industry.
But far more intriguing was what Babiš claimed was the
attitude of Visegrad leaders about relations with the
institutions of the EU. According to a statement issued by
the Czech government, Babiš said the leaders agreed it was
“necessary to depoliticise Brussels and the EC”. Apparently,
the leaders believe that when it comes to EU affairs,
“ member states, prime ministers and presidents, should
have the main say”, according to Babiš.
If there is behind-the-scenes pressure to stamp out
nuclear wherever it might try to cling on or prosper in the
EU, where is that effort coming from and why? Of course,
it is no secret that Austria and Germany strongly oppose
any expansion of nuclear power in Europe. Having lived
and worked in Germany, I never understood that
politically- inspired decision – but as a guest in the country
for which I have great admiration I respect its decision.
Austria’s approach has always puzzled me more – being
willing as it is to host the International Atomic Energy
Agency (IAEA) and enjoy all the ‘fruits’ that that privilege
brings, not least in the economic benefit of having the
agency based in Vienna.
But back to the Czech project. As a possible fight with
the EC shapes up, it is not only Moscow that is set to benefit
from yet another new nuclear power order from an EU
nation.
South Korea is also reportedly circling – keen to tempt
Prague to consider its nuclear technology, according to
Seoul’s Yonhap News Agency.
Can the EU really afford such a quarrel – again – with a
member state over nuclear? And why should European
skills, knowhow and investment not be channelled into the
Czech nuclear project?
I am struck by the EC’s approach to another industrial
sector and how contrasting it is. The EC is currently working
at full tilt to develop a European battery cell industry,
with the goal of ensuring the EU is not overwhelmed by
competition from Asian battery makers for products such as
electric vehicles and energy storage devices.
The EU’s vice-president for the energy union, Maroš
Šefčovič, said in February “there are many extremely
interesting actions that we need to pursue, including (a)
simplification of approval procedures and permitting
processes in the EU”. Indeed the European Investment Bank
has already approved a loan for the construction and
operation of what it said will be a first-of-a-kind demonstration
plant in Sweden, for the manufacturing of lithium- ion
batteries.
The EC’s support for the development of such technology
across EU member states is of course admirable, but one
hears nothing of state-aid rules and complications here!
Why is it that nuclear cannot win such favourable attention
and support? Does it really have to be this way – and
hasn’t the EC learned anything from the UK’s Brexit vote
about treading carefully in issues that are seen by member
states of national importance?
Nuclear Today
Czechs and Balances and Why ‘Ugly’ Nuclear Deserves a Political Makeover ı Jubair Ahmed Shamim and Kune Yull Suh
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