Core design analysis of the supercritical water fast reactor [Elektronische Ressource] / vorgelegt von Magnus Mori
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Core design analysis of the supercritical water fast reactor [Elektronische Ressource] / vorgelegt von Magnus Mori

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CORE DESIGN ANALYSIS OF THE SUPERCRITICAL WATER FAST REACTOR Von der Fakultät für Maschinenbau der Universität Stuttgart zur Erlangung der Würde eines Doktor-Ingenieurs (Dr. -Ing.) genehmigte Abhandlung Vorgelegt von Magnus Mori aus Legnago (VR), Italien Hauptberichter: Prof. G. Lohnert, Ph.D. Mitberichter: Prof. Dr. T. Schulenberg Tag der mündlichen Prüfung 20. Januar 2005 Institut für Kernenergetik und Energiesysteme der Universität Stuttgart 2005 Tu Vuò Fa L'Americano Puorte o cazone cu 'nu stemma arreto 'na cuppulella cu 'a visiera alzata. Passe scampanianno pe' Tuleto camme a 'nu guappo pe' te fa guardà! Tu vuò fa l' americano! mmericano! mmericano siente a me, chi t' ho fa fa? tu vuoi vivere alla moda ma se bevi whisky and soda po' te sente 'e disturbà. Tu abballe 'o roccorol tu giochi al basebal ' ma 'e solde pe' Camel chi te li dà? ... La borsetta di mammà! Tu vuò fa l' americano mmericano! mmericano! ma si nato in Italy! siente a mme non ce sta' niente a ffa o kay, napolitan! Tu vuò fa l' american! m Comme te po' capì chi te vò bene si tu le parle 'mmiezzo americano? Quando se fa l 'ammore sotto 'a luna come te vene 'capa e di:"i love you!?" Tu vuò fa l' americano mmericano! mmericano siente a me, chi t'ho fa fa? tu vuoi vivere alla moda...

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Published 01 January 2005
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CORE DESIGN ANALYSIS
OF THE
SUPERCRITICAL WATER FAST REACTOR


Von der Fakultät für Maschinenbau der Universität Stuttgart
zur Erlangung der Würde eines
Doktor-Ingenieurs (Dr. -Ing.) genehmigte Abhandlung


Vorgelegt von
Magnus Mori
aus Legnago (VR), Italien



Hauptberichter: Prof. G. Lohnert, Ph.D.
Mitberichter: Prof. Dr. T. Schulenberg
Tag der mündlichen Prüfung 20. Januar 2005


Institut für Kernenergetik und Energiesysteme der Universität Stuttgart
2005 Tu Vuò Fa L'Americano

Puorte o cazone cu 'nu stemma arreto
'na cuppulella cu 'a visiera alzata.
Passe scampanianno pe' Tuleto
camme a 'nu guappo pe' te fa guardà!

Tu vuò fa l' americano!
mmericano! mmericano
siente a me, chi t' ho fa fa?
tu vuoi vivere alla moda
ma se bevi whisky and soda
po' te sente 'e disturbà.

Tu abballe 'o roccorol
tu giochi al basebal '
ma 'e solde pe' Camel
chi te li dà? ...
La borsetta di mammà!

Tu vuò fa l' americano
mmericano! mmericano!
ma si nato in Italy!
siente a mme
non ce sta' niente a ffa
o kay, napolitan!
Tu vuò fa l' american! m

Comme te po' capì chi te vò bene
si tu le parle 'mmiezzo americano?
Quando se fa l 'ammore sotto 'a luna
come te vene 'capa e di:"i love you!?"

Tu vuò fa l' americano
mmericano! mmericano
siente a me, chi t'ho fa fa?
tu vuoi vivere alla moda...
Renato Carosone



A l'alta fantasia qui mancò possa;
ma già volgeva il mio disio e 'l velle,
sì come rota ch'igualmente è mossa,
l'amor che move il sole e l'altre stelle
(Dante: Pd. XXXIII, 143-145)

al nonno Gino e allo zio Sergio


Acknowledgments
The number of people that I would like to thank for their help, assistance, support, and
encouragement is indeed very large; therefore I really hope I am not forgetting anyone.
First of all I would like to thank Dr. Werner Maschek, he was always available, helpful,
and very supportive, he’s been my scientific and personal guide for the last few years and
above all a very good friend.
I would then like to thank Prof. Günter Lohnert, Prof. Thomas Schulenberg, Dr.
Wolfgang Bernnat, and Prof. Eckart Laurien, for their guidance, for the positive
discussions and their interest in my work; Dr. Andrei Rineiski for his enthusiasm, his
assistance, the uncountable number of ideas, and the cheerful discussions; Dr. Edgar
Kiefhaber for the numerous suggestions and the valuable comments, aside from coping
with my presence across the corridor; Mr. Frank Kretzschmar for his German lessons, his
patience, and his help; JAPCO for having funded this work!
I would like to thank the SIMMER team, above all Dr. Tohru Suzuki and Prof. Koji
Morita, for their incredible kindness, tolerance and availability and Dr. Xue-Nong Chen,
who was the main developer of MXN and who never complained about my complaints.
Finally, special thanks to all my colleagues, particularly Michael, Eva, Claudia, Valentin,
Walter, who helped and supported me not only with my work, but especially with my
everyday life troubles, not to mention selling my old Maserati.
A special mention goes then to my friends in Italy, and to all the members of the Pirate
Crew, who made my survival in Germany possible and enjoyable. A very special thank
you to the First Mate who managed to put up with me, who more than once helped me
reclaim my castaway motivations and who, together with my big Italian family, was
always a safe harbor to moor in the time of the tempest and of the sun, the wind in time
of calm.


Abstract
Light Water Reactor technology is nowadays the most successful commercial
application of fission reactors for the production of electricity. However, in the next
years, nuclear industry will have to face new and demanding challenges. The need
for sustainable and cheap sources of energy, the need for public acceptance, the need
for even higher safety standards, the need to minimize waste production are only a
few examples. It is for these very reasons that a few next generation nuclear reactor
concepts were selected for extensive research and development. Super critical water
cooled reactors are one of them.
The use of a supercritical coolant would in fact allow for higher thermal efficiencies
and a more compact plant design. As a matter of fact, steam generators, or steam
separators and driers would not be needed thus, significantly reducing construction
costs. Moreover, because of the high heat capacity of supercritical water,
comparatively less coolant would be needed to refrigerate the reactor. Consequently,
a water-cooled reactor with a fast neutron spectrum could potentially be designed:
the SuperCritical water Fast Reactor.
This system presents unique features combining well-known fast and light water
reactor characteristics in one design (e.g. the tendency to a positive void reactivity
coefficient together with Loss Of Coolant Accidents, as design basis). The core is in
fact loaded with highly enriched Mixed OXide fuel (average plutonium content of
~23%), and presents a peculiar and significant geometrical and material
heterogeneity (use of radial and axial blankets, solid moderator layers, several
enrichment zones). The safety analysis of this very complex core layout, the
development of suitable tools of investigation, and the optimization of the void
reactivity effect through core design, is the main objective of this work.
7 of 150
Zusammenfassung
Bei der Leichtwasserreaktortechnologie handelt es sich um die zur Zeit erfolgreichste
kommerzielle Anwendung von Spaltreaktoren zur Erzeugung von Elektrizität. In den
kommenden Jahren steht die Nuklearindustrie jedoch vor neuen großen
Herausforderungen. Der Bedarf an nachhaltigen und preisgünstigen Energiequellen,
die öffentliche Akzeptanz, immer höhere Sicherheitsstandards und die Minimierung
der Abfallproduktion sind nur ein paar Beispiele. Vor diesem Hintergrund wurden
ein paar Reaktorkonzepte der nächsten Generation ausgewählt und umfangreicher
Forschung und Entwicklung unterzogen. Ein solches Konzept bezieht sich auf
superkritische wassergekühlte Reaktoren.
Die Verwendung eines superkritischen Kühlmittels würde tatsächlich höhere
thermische Wirkungsgrade und eine kompaktere Anlagenauslegung ermöglichen.
Dampferzeuger bzw. Dampfseparatoren und –trockner wären nicht mehr
erforderlich, wodurch die Baukosten beträchtlich reduziert würden. Darüber hinaus
wäre aufgrund der hohen Wärmekapazität von superkritischem Wasser eine
vergleichsweise geringe Kühlmittelmenge zur Kühlung des Reaktors erforderlich.
Somit könnte auch ein wassergekühlter Reaktor mit einem schnellen
Neutronenspektrum konzipiert werden: der Schnelle Superkritische Wasserreaktor.
Dieses System besitzt einzigartige Merkmale und vereint die bekannten
Eigenschaften von schnellen Reaktoren und Leichtwasserreaktoren in sich (z. B.
positiver Voidkoeffizient und Kühlmittelverluststörfälle als Auslegungsgrundlage).
Der Kern besteht aus hoch angereichertem Mischoxidbrennstoff (durchschnittlicher
Plutoniumgehalt ~ 23 %) und ist sowohl von der Geometrie als auch vom Material
her sehr heterogen (radiale und axiale Blankets, Feststoffmoderatorschichten,
mehrere Anreicherungszonen). Eine Sicherheitsanalyse dieser sehr komplexen
Kernanordnung, die Entwicklung geeigneter Analysewerkzeuge sowie die
Optimierung des Voidkoeffizienten mittels Kernauslegung sind Gegenstand der
vorliegenden Arbeit.

8 of 150 Table of Contents
Abstract .............................................................................................................................7
Zusammenfassung ............................................................................................................8
Table of Contents ..............................................................................................................9
Nomenclature..................................................................................................................11
Subscripts and Superscripts...........................................................................................13
Dimensionless numbers..................................................................................................13
Acronyms .........................................................................................................................15
INTRODUCTION ...........................................................................................................19
1.1 Motivation .........................................................................................................20
1.2 Present State of the Technology ......................................................................24
1.3 Aim of this Work...............................................................................................28
CHAPTER 2 .....................................................................................................................31
MATHEMATICAL AND PHYSICAL MODELS ...........................................................31
2.1 System description, geometry, and integration domain.................................32
2.2 Description of the models.................................................................................36
Fluid-Dynamics Model......................................................................................36
Neutronics Model ..............................................................................................45
CHAPTER 3 .....................................................................................................................55
Numerical Methods.........................................................................................................55
3.1 Neutronics56
The Monte-Carlo method..................................................................................57
Multigroup, discrete ordinates method............................................................63
3.2 Fluid-dynamics and coupling procedures........................................................68
The Newton-Raphson method ..........................................................................68
The Coupling Procedure ...................................................................................70
3.3 Verification of the models.................................................................................74
9 of 150 Table of Contents
CHAPTER 4 .....................................................................................................................77
RESULTS ........................................................................................................................77
4.1 Neutronics Analyses for the SCFR ..................................................................78
Neutron physics of the void effect ....................................................................78
Deterministic analysis of the void effect in the SCFR ....................................79
Remarks on the deterministic analyses...........................................................85
4.2 Refined neutronics analyses for the SCFR......................................................86
Geometry model, nuclear data libraries, and data processing options. .........87
Summary of the computed results and void effect uncertainties ...................90
Influence of absorbers on safety parameters...................................................97
4.3 Optimization of safety parameters for the SCFR ...........................................98
Description of the new advanced MCNP model ..............................................99 of the calculations and results....................................................101
New, improved void effect blanket.................................................................105
Void effect calculations with tight blanket geometry....................................106
Effect of the adoption of the new cross section libraries...............................107
Investigation of the adoption of new core configurations and materials .....113
4.4 Coupled calculations for the SCFR................................................................121
Analysis of the iterations................................................................................123
Void effect calculations ...................................................................................127
Effect of fuel composition on void effect.........................................................128
Effect of the improved void configurations ....................................................131
CONCLUSIONS............................................................................................................133
Recommendations for additional studies.....................................................................137
References......................................................................................................................139
Appendix A ....................................................................................................................145
Description of the nuclear data libraries used in this work................................145
Appendix B147
Analysis of core performance ................................................................................147
10 of 150