PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Foreword
The organizing and technical program committees of PHYSOR2014 cordially welcome you. This is the second
PHYSOR conference in Japan - the PHYSOR conference returns to Japan after 18 years since it was heldin Mito in 1996, hosted by the former Japan Atomic Energy Research Institute (JAERI). For PHYSOR2014,
the Japan Atomic Energy Agency (JAEA) and Kyoto University Research Reactor Institute are the official
hosts of the conference. A broad selection of people from Japanese universities and industries participate in
the organizing committee for the preparation of PHYSOR2014. The technical program committee consists ofinternational experts and has an important role, including high quality reviews for submitted papers.
We would like to host you in the spirit of “Omotenashi,” a spirit of selfless hospitality of Japan, and we would
like to provide a well-organized, technically rich, and informative meeting, continuing the traditions of theprevious successful PHYSOR meetings, e.g., Mito in 1996, Long Island in 1998, Pittsburgh in 2000, Seoul in
2002, Chicago in 2004, Vancouver in 2006, Interlaken in 2008, Pittsburgh in 2010, and Knoxville in 2012.
The main theme of PHYSOR2014 is the role of reactor physics towards a sustainable future. Japan has theexperience of a severe accident, at the Fukushima-Daiichi Nuclear Power Station due to the massive Tsunami
triggered by the earthquake, offshore of the north-east part of Japan on March 11, 2011. Many local residents
are still evacuated from the region around the Fukushima-Daiichi NPS. It is one of the significant impacts of a
nuclear accident. A visit to the Fukushima-Daiichi NPS is planned as a technical tour of the conference. You
can directly observe the current state of affairs at the Fukushima-Daiichi NPS through this tour.
We believe that we should reconfirm that safety shall be the first priority in the utilization of nuclear power. Only
the reactor physics can explain “why it is a nuclear reactor,” i.e., the fundamental mechanism of the fission
chain reaction. In this context, it is clear that nuclear reactor physics is a very important and fundamental area
for the safety of nuclear power. During this conference, we would like to reconfirm the importance of reactor
physics in the nuclear engineering discipline with you.
We think PHYSOR2014 also provides you a very good opportunity to experience Japan. Kyoto is animpressive, historic city and the conference venue is located near the center of the historic area of Kyoto.
Many beautiful shrines, temples, historic architectures, and museums are located within walking distance fromthe conference venue. You can experience not only the traditional Japanese culture, but also “cool” Japan in
Kyoto, including fine dining.
We hope that the PHYSOR2014 conference will be fruitful for you from various aspects.

Shigeaki Okajima and Ken Nakajima
General Chairs

Akio Yamamoto
Technical Program Chair

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Acknowledgement
The organizers would like to express their heartfelt appreciation to the following organizations and individuals:• Financial sponsors for their invaluable contributions to support our conference.
• Co-sponsoring technical societies worldwide who publicize our conference.
• International experts who contributed by recruiting technical papers, by reviewing conference papers, by
chairing the technical sessions, and/or by being willing to be jury for the best poster and best student paper
awards.
• The plenary, banquet, and luncheon speakers, who use their invaluable time to make their address during
our conference.
• Organizers of the special sessions and workshops, who provide an invaluable opportunity for technicalinformation exchange on specific specialized topics.
• The hosts of the technical tours: the Kyoto University Research Reactor Institute, the Monju fast reactor of
JAEA, the Kumatori Works of Nuclear Fuel Industries, Ltd. and the Tokyo Electric Power Company.
We personally would like to offer our special thanks to the members of the organizing and technical programcommittees who have been doing a lot of work to prepare PHYSOR2014 during the approximately two years
since Kyoto was selected as the conference location for PHYSOR2014 in the ANS RPD meeting at the 2012
ANS annual meeting.
Shigeaki Okajima and Ken Nakajima
General Chairs

2

Akio Yamamoto
Technical Program Chair

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Organizing Committee
Honorary chair

Toshikazu Takeda

University of Fukui

General chairs		

Shigeaki Okajima
Ken Nakajima

Japan Atomic Energy Agency
Kyoto University

Technical Program
Chair
Co-chairs
						

Akio Yamamoto
Hany Abdel-KhalikMark DeHart
Imre Pazsit
Igor Zmijarevic

Members
		
		
		
		
		
		
		
		

Yuichiro Ban
Tomohiro Endo
Akinori Giho
Yoshio Kimura
Takanori Kitada
Hideki Matsumoto
Kazuteru Sugino
Masayuki Tojo
Takuya Umano

Nagoya University
North Carolina State University
Idaho National Laboratory
Chalmers University of Technology
Commissariat à l'énergie atomique et aux énergies
alternatives
Toshiba Corporation
Nagoya University
Shikoku Electric Power Company
Chuden CTI
Osaka University
Mitsubishi Heavy Industries
Japan Atomic Energy Agency
Global Nuclear Fuel Japan
Toshiba Corporation

Steering
Chairs		
General affairs
		
		
		
		
		
		
Finance		
Publication
		
		
		
Web
		
Public relations
		
Local
		
		
		

Tokio Fukahori
Cheolho Pyeon
Kazufumi Tsujimoto
Shigeaki Aoki
Tadashi Ikehara
Tomohiko Iwasaki
Etsuro Saji
Masahiro Tatsumi
Tadashi Ushio
Teruhiko Kugo
Suzuro Sugano
Go Chiba
Tetsushi Hino
Takanori Sugawara
Wilfred van Rooijen
Kenji Yokoyama
Masato Tabuchi
Kenya Suyama
Tadafumi Sano
Kazuki Kirimura
Yasuhiro Kodama
Kenji Nishihara
Satoshi Takeda

Japan Atomic Energy Agency
Kyoto University
Japan Atomic Energy Agency
Mitsubishi Nuclear Fuel
Global Nuclear Fuel Japan
Tohoku University
Mitsubishi Heavy Industries
Nuclear Engineering Limited
Nuclear Fuel Industries
Japan Atomic Energy Agency
Japan Atomic Energy Agency
Hokkaido University
Hitachi Limited
Japan Atomic Energy Agency
University of Fukui
Japan Atomic Energy Agency
Nuclear Engineering Limited
Japan Atomic Energy Agency
Kyoto University
Mitsubishi Heavy Industries
Nuclear Fuel Industries
Japan Atomic Energy Agency
Nuclear Fuel Industries

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Technical Program Committee
Technical Program Chair
Co-chairs

Akio Yamamoto

Nagoya University

Hany Abdel-Khalik

North Carolina State University

Mark DeHart

Idaho National Laboratory

Imre Pazsit

Chalmers University of Technology

Igor Zmijarevic

Commissariat à l'énergie atomique et aux énergies alternatives

Technical Program Committee Members

4

A. Alberto

D. Ginestar

D. LeeM. Pusa

M. Tsuji

A. Alfonsi

S. Goluoglu

J. LeeC. Pyeon

K. Tsujimoto

C. Anistratov

D. Griesheimer

Y.K. Lee

C. Rabiti

P. Tsvetkov

M.N. Avramova

A. Gruel

Y.O. Lee

P. Ravetto

V. Tuomas

Y. BanA. Haghighat

J. Leppänen

F. Reitsma

T. Ueki

H. Berger

T. Hazama

L. MaoG. Rimpault

G. van den Eynde

J. Bess

K. Hesketh

G. Marleau

J. Roberts

R. van Geemert

P. Blaise

T. HinoM. Marshall

P. Romano

W.F.G. van Rooijen

B. Bromley

D. Holcomb

R. Martineau

M. Rosa

A. Vasiliev

L. CaoS. Hong

H. Maruyama

R. Sanchez

J. Vegh

M. Carta

J. Hori

E. MasielloV. Sanchez

P. Vinai

J. ChangJ. Hutchinson

H. Mathieu

T. SanoK. Wang

G. Chiba

D. IlasH. Matsumoto

S. SenT.A. Wareing

S. Chiba

M. Ishikawa

F. Mellier

D. Serghiuta

W. Wiesenack

N.Z. Cho

K. IvanovT. Mitsuyasu

W. ShenU. Wim

H. Choi

T. Ivanova

S. Monti

H.J. Shim

H. Wu

D. Cokinos

H. Iwamoto

P. Mosca

G. Sjoden

Z. Wu

R. Dagan

H.G. Joo

Y. Nagaya

R. Slaybaugh

Z. Xu

S. Degweker

T. Kamei

Y. Nakano

L. Snoj

T. Yagi

M. DeHart

T. Kameyama

H.R. Nasif

A. Stanculescu

K. Yamaji

D. Diamond

J. Katakura

Y. Nauchi

A. Stankovskiy

A. Yamamoto

S. Dulla

T. Kawano

E. Nichita

M. Sugawara

T. Yamamoto

E. Dumonteil

T.K. Kim

K. Nishihara

T. Sugawara

W.S. Yang

J. Dyrda

Y. Kimura

D. Noyelles

N. Sugimura

K. Yokoyama

R.J. Ellis

K. Kirimura

T. Obara

K. Sugino

K. Yoshioka

T. EndoT. Kitada

S. OhkiN.V. Sultanov

R. Yoshioka

S. Eugene

Y. Kitamura

Y. Ohoka

T. Sutton

J. Zerr

M. Ferrer

J.L. Kloosterman

H. OigawaK. Suyama

I. Zmijarevic

P. Ferroni

K. Kobayashi

S. Okajima

M. Tabuchi

W. Zwermann

H. Ferroukhi

Y. Kodama

K. Okumura

K. TadaC. Fiorina

H. Koike

M. Ouisloumen

T. Taiwo

G. Flanagan

E. Koonen

S. Palmtag

Y. Takahashi

B. Forget

S. Kosaka

I. Pazsit

S. Takeda

M. Fratoni

M. Kostal

S. Pelloni

N. Takemoto

M. Fukushima

T. Kozlowski

G. Perret

A. Talamo

B. Ganapol

J. Krepel

B. Petrovic

M. Tatsumi

A. GihouE. Lahoda

J. Pounders

M. Tojo

D. GillN. LeClair

T. Primm

A. TrkovPHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Track Leaders
1 Reactor Analysis Methods
C. Demazière

H. G. Joo

R. Sanchez

K. Smith

N. Z. Cho

F. Rahnema

K. Yamaji

J. Leppänen

H. J. Shim

K. Wang

M. Tojo

2 Deterministic Transport Theory
Y. Azmy3 Monte Carlo Methods
T. Kitada

4 Verification, Validation and Uncertainty Analysis
H. Abdel-Khalik

G. Chiba

W. F. G. van Rooijen

5 Nuclear Criticality Safety
J. Bess

T. EndoT. Yamamoto

6 Reactor Physics Experiments
P. Blaise

H. Unesaki

7 Reactor Concepts and Designs
J. LeeC. Liangzhi

B. Petrovic

N. Takaki

8 Reactor Operation and Safety
F. Franceschini

T. Mitsuyasu

G. Sjoden

9 Transient and Safety Analysis
Y. BanS. Dulla

K. Ivanov

R. Jacqmin

Y. O. Lee

P. Ravetto

10 Nuclear Data
S. Chiba

11 Research Reactors and Spallation Sources
G. van den Eynde

I. L. Montoya

K. Nishihara

12 Fuel Cycle and Actinide Management
T. Kim

K. Tsujimoto

A. Worrall

13 Radiation Applications and Nuclear Safeguards
Y. Kitamura

S. A. Pozzi

14 Education in Reactor Physics
B. Forget

T. Kameyama

T. Kozlowski

15 Research Related to Fukushima Accident
A. Haghighat

S. Kosaka

K. Suyama

SS1 Molten Salt Reactors

I. Pazsit

SS2 Reactor Physics and Criticality Safety Activities in OECD/NEA Working Party

M. DeHart

SS3 Hybrid Particle Transport Methods for Solving Complex Problems in Real-Time

A. Haghighat
F. Rahnema

SS4 Advanced Geometry Processing in Deterministic and Monte Carlo Methods

H. J. Shim

SS5 Multiscale, Multiphysics Approaches in Nuclear Science and Engineering Applications

M. DeHart
R. C. Martineau

SS6 Nuclear Criticality Safety of Fuel Debris

N. Takaki

SS7 Control Rod Withdrawal Tests Performed during the PHENIX End-of-Life Experiments

S. Monti

SS8 Reactor Physics of Non-Traditional LWR Fuel Design

B. Petrovic

Selection Committee Members for Best Poster Award
Cao Liangzhi (Xi'an Jiaotong Univ.)
Yonghee Kim (KAIST)

Shinya Kosaka (MHI)

Patrick Blaise (CEA)

Benoit Forget (MIT)Christophe Demazière (Chalmers Univ. of Tech.)

Selection Committee Members for Best Student Award
Hyung Jin Shim (Seoul National Univ.) Andrew Worral (ORNL)
Wilfred van Rooijen (Univ. of Fukui)

Kan Wang (Tsinghua Univ.)

Fausto Franceschini (Westinghouse) Aldo Dall'Osso (AREVA)

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Sponsors
Financial
The Kyoto University
Foundation

Non Financial

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

General Information
Registration
Registration is required for all attendees and presenters. Name plate is required for admission to all events.
Full registration includes: All technical sessions, book of abstracts, CD proceedings, welcome cocktail,hosted lunch, and conference banquet.
Student registration includes: All technical sessions, book of abstracts, CD proceedings, welcome cocktail
and hosted lunch (option: conference banquet +JPY10,000).

Conference Proceedings
Conference Proceedings are included with the program book in CD-ROM format.

Guidelines for Speakers
There will be five parallel sessions. Each presentation will last 15 minutes at the maximum, followed by 5
minutes for questions. In order to allow conference participants to attend the presentation of papers in different
sessions in a timely manner, we, as organizers, will request the chairpersons to comply with the time schedulerigorously. In view of the given time constraints, please make sure that your presentation fits within theprescribed 15 minutes limit leaving adequate time for questions from the audience.
The laptops going to be used in sessions will have the following programs installed:• Microsoft Office Word 2013
• Microsoft Office Excel 2013
• Microsoft Office PowerPoint 2013
• Acrobat Reader Version 10
The laptops will be using Microsoft Windows 7 operating system with the Service Pack 1 installed. Default
Windows font set will be available on these machines. We highly recommend that you create a PDF versionof the presentation so that you can switch to the PDF in case of a problem with the PowerPoint. A microphone
will be used for the presentation, please make sure that you keep close to the microphone during your talk. We
also request the workshop speakers to adhere to the same rules if they plan to use our computers.

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Special Events
General Events
Welcome Cocktail

Place: Mizuho_D
Date: Sunday Sep. 28 16:00-20:00

Conference Banquet

Place: Mizuho_C, D
Date: Tuesday Sep. 30 18:30-21:00
Speakers: Dr. Ronald J. Ellis, Chair, ANS Reactor
Physics Division and Prof. Toshikazu Takeda,
Honorary Chair, Univ. of Fukui

Hosted Lunches

Place: Mizuho_C, D
Date: Monday Sep. 29 11:30-13:00
Speaker: Etsuro Saji (Mitsubishi Heavy Industries)
Place: Mizuho_C, D
Date: Tuesday Sep. 30 12:00-13:30
Speaker: Sun-Doo Kim (KEPCO Nuclear Fuel)

Technical Tours
Fukushima Dai-ichi Nuclear Power Plant(35 persons. JPY47,000/person, including train andbus on 10/2 and 10/3; accommodation on 10/2 inTokyo)
10/2: Conf. Hotel (Afternoon) → Tokyo (Stay)
10/3: Tokyo → Fukushima → Fukushima Dai-ichi
NPP → Tokyo (Evening; Dismiss)Monju in Tsuruga(20 persons. JPY 6,500/person, including bus
transportation and lunch)
10/3: Conf. Hotel (Morning) → Monju → Conf. Hotel(Evening)
Monju is a 280 MWe prototype sodium cooled fast
breeder reactor (FBR) using plutonium-uranium
mixed oxide fuel. Monju achieved initial criticality in
April 1994. In August 1995, Monju became the firstJapanese FBR to generate power. It remains the
only Japanese facility that can generate electricity
using a fast breeder reactor. In December 1995,
sodium coolant leaked from a temperature sensor
in the secondary system piping. Since then, power
generating operation has been suspended. Monju
8

is still in the limelight as an international key facilityof the fast breeder reactor R&Ds and is expected todemonstrate its reliability as an operational power
plant, and establish sodium handling technology.

Kumatori Area (NFI and KURRI)(20 persons. JPY 6,500/person, including bus
transportation and lunch)
10/3: Conf. Hotel (Morning) → Kumatori → Conf.
Hotel (Evening)+ Nuclear Fuel Industries, Ltd. (NFI, Kumatori Works)
Nuclear Fuel Industries, Ltd. (NFI) was establishedin 1972 and provides nuclear fuel and fuel-related
services. It operates two fuel processing plants, a
PWR fuel plant at Kumatori, and a BWR fuel plantat Tokai. Kumatori works has supplied over 9,000
PWR fuel assemblies for electric power companies.
NFI will continue to supply reliable and high-quality
products, and work towards a safer and more
comfortable society.
+ Kyoto University Research Reactor Institute (KURRI)
The Kyoto University Research Reactor Institute(KURRI) was established in 1963 for the joint use
program among Japanese universities to promote
the research and education in the fields of nuclear
energy and radiation application. Two nuclear
reactors, the Kyoto University research Reactor(KUR) and the Kyoto University Critical Assembly(KUCA), and related research facilities have been
being used since then, and nowadays greater
expectations are being put on the research andeducation activities at our institute for the issuesof energy and environment and for the innovative
applications of radiation.
In addition, proton beams from the Fixed Field
Alternating Gradient (FFAG) accelerator complex
installed in the accelerator room is led to KUCA forthe purpose of executing feasibility study on theaccelerator-driven system (ADS).
The tour will include the KUCA facilities and the
FFAG accelerator.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
September 28 - October 3, 2014

Meeting Rooms

Registration desk

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Sunday September 28, 2014

Technical Workshops
8:00 - 12:00# 5 New Features and Capabilities in the Serpent 2 Monte Carlo Code		
Room; Mizuho_A (4F)		Organizer; J. Leppänen (VTT, Finland)
8:00 - 12:00# 3 TEPCO's Activity on the Investigation of Fukushima Daiichi Accident / 				 Current Status of Fukushima-Daiichi NPS and Decommissioning Process		Room; Mizuho_B (4F)		Organizer; S. Mizokami (Tokyo Electric Power Company, Japan), 							M. Yamamoto (Tokyo Electric Power Company, Japan)# 6 Attila4MC - CAD Integration, Automated Deterministic Variance Reduction, andGUI Setup for MCNP		
Room; Mizuho_C (4F)		Organizer; G. Failla (Varian Medical Systems, USA)
8:00 - 12:00

8:00 - 12:00# 2 Workshop on Accelerator-Driven System in PHYSOR 2014
		
Room; Hiei (2F)		Organizer; T. Sugawara (JAEA, Japan), C. Pyeon (KURRI, Japan)
13:30 - 17:35 # 1 Short Course on Uncertainty Characterization in Nuclear Calculations
		Room; Mizuho_A (4F)		Organizer; H. S. Abdel-Khalik (North Carolina State University, USA)
13:30 - 17:35 # 4 Neutron Noise Techniques for Reactor Diagnostics
		Room; Mizuho_B (4F)		Organizer; I. Pazsit (Chalmers Univ. of Technol., Sweden),			C. Demazière (Chalmers Univ. of Technol., Sweden),			V. Dykin (Chalmers Univ. of Technol., Sweden)
13:30 - 17:35 # 7 Hybrid Particle Transport Methods for Solving Complex Problems in Real-Time		
Room; Mizuho_C (4F)		Organizer; A. Haghighat (Virginia Tech., USA), F. Rahnema(Georgia Tech., USA),			B. Petrovic (Georgia Tech., USA)
13:30 - 17:35 # 8 NESTLE 3D Nodal Core Simulator: An Overview of Latest Features and Capabilities
		Room; Hiei (2F)		Organizer; G. I. Maldonado (Univ. of Tennessee, USA),			N. P. Luciano (Univ. of Tennessee, USA), S. D. Hart (Univ. of Tennessee, USA)

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014, - 8:50 AM - Mizuho_A, B

Plenary Sessions 1 & 2
8:50 AM

Kord Smith

Professor, Massachusetts Institute of Technology (USA)

Advances in Reactor Physics and Computational Science
9:25 AM

Akira Yamaguchi

Professor, Osaka University (Japan)

Status of Nuclear Power in Japan
10:20 AM

Jun Matsumoto

General Manager, Fukushima Daiichi Decontamination and Decomissioning Engineering Company,
Tokyo Electric Power Company (Japan)

Topics on Fukushima Daiichi Decontamination and Decommissioning
10:55 AM

Robin Forrest

Head of Nuclear Data Section, International Atomic Energy Agency (Austria)

Advances in Nuclear Data

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Mizuho_A

Track1-1 Reactor Analysis Methods
Session Chair: Tarmer Bahadir(Studsvik Scandpower, Inc.), Masayuki Tojo(GNF-J)
13:00 PM

Hitachi's Advanced Technologies
Guest Speaker: H.Soneda
Hitachi-GE Nuclear Energy, Japan

13:20 PM

Development of Enhanced SPH Method for Pin-by-Pin
Core Calculations
S.Takeda, T.Ushio, Y.Ohoka, Y.Kodama
Nuclear Fuel Industries, Ltd., Osaka, Japan
Enhanced SPH method is proposed for the accurate calculation in
the geometry which has large spectral interference between fuelassemblies. The SPH factor is created in order to preserve the reaction
rate of each cell in the pin-by-pin core calculation and reduce homogenization error. Since the discontinuity of the neutron flux between
fuel assemblies are not considered in the conventional SPH method,another SPH method which uses discontinuity factor has beendeveloped. Using this SPH method, the discontinuity at the outermost
part of fuel assembly can be incorporated to the SPH factor. On theother hand, this SPH method does not consider the spectrum of theouter portion of the fuel assembly. This problem leads the difficultyof accurate evaluation of spectral interference in the system such as
UO2/MOX multi-assembly geometry. In this paper, enhanced SPH
method which takes both spectrum and discontinuity at the outermost
part of fuel assembly is proposed. And the performance of enhanced
SPH method is evaluated by calculation results of 2-dimensional
single assembly geometry and multi-assembly geometry. The results
show that newly developed method can calculate the multi-assembly
geometry accurately even if the spectral interference between the fuelassemblies is large. And it is confirmed that enhanced SPH factor
can be created by the proposed method and contribute to accurate
prediction of neutronics parameters in pin-by-pin core calculation.

13:40 PM

Study on Cross Section Correction Using SPH
Method for a Whole Core Heterogeneous MOC
Calculation
A.Giho (1), E.Yoshida, K.Miyawaki (2), K.Ohori, Y.Umebara (1),
T.Takeda(3)1)Shikoku Electric Co., Inc., Kagawa, Japan, 2)Yonden Engineering Co.,
Inc., Kagawa, Japan, 3)Shikoku Instrumentation Co.,Inc., Kagawa, JapanThe applicability of the SPH method to a whole core heterogeneous
MOC calculation is investigated. The sensitivities to the calculation
error of the discretization parameters of MOC regarding the ray trace
path width, the number of azimuthal angles and the spatial meshsize are investigated in a two-dimensional whole core configuration
which consists of MOX and UO2 fuels. The cross section sets of
MOX and UO2 fuels are generated in single assembly geometry
without SPH method and with SPH method. Although the sensitivities
to the calculation error of the ray trace path width and the numberof azimuthal angles are quite large without SPH method, thesesensitivities are decreased with SPH method. This result reveals
that the SPH method is effective to reduce the discretization error
in a whole core MOC calculation. On the other hand, the maximum
error of power distribution is increased under the coarser conditionin spite of SPH correction and this maximum error is obtained near
the boundary between MOX and UO2 fuels. This result means thatspectral interference effect among fuel assemblies should be taken
into account. The sensitivity to the calculation error of the spatial meshsize is not decreased with SPH method. It is observed that the spatial
mesh size in the moderator region has quite large effect on pin power
distribution error whether the SPH method is used. These resultsmean that the spatial distributions of neutron flux and neutron source
are quite important in the moderator region near the fuel boundary.

14:00 PM

Asymptotic, Multigroup Flux Reconstruction andConsistent Discontinuity Factors
T.J.Trahan(1,2), E.W.Larsen(2)1)Los Alamos National Laboratory, New Mexico, USA, 2)University ofMichigan, Michigan, USA
Recent work has led to an asymptotically-derived expression forreconstructing the neutron flux from lattice functions and multigroupdiffusion solutions. The leading-order asymptotic term is the standard
expression for flux reconstruction, i.e., it is the product of a shapefunction obtained through a lattice calculation and the multigroupdiffusion solution. The first-order asymptotic term is a formally small
correction. The correction is negligible when the gradient of the
diffusion solution is small, but large when the gradient of the diffusion
solution is large. Inclusion of this first-order correction term cansignificantly improve the accuracy of the reconstructed flux. One may
define discontinuity factors (DFs) to make certain angular momentsof the reconstructed flux continuous across interfaces between
assemblies in 1-D. Indeed, the standard assembly discontinuity
factors used often in reactor analysis make the zeroth moment (scalarflux) of the reconstructed flux continuous. The inclusion of the firstorder correction term in the flux reconstruction gives us an additional
degree of freedom that can be used to make two angular moments of
the reconstructed flux continuous across interfaces by using current
DFs, in addition to flux DFs, in the multigroup diffusion calculation.
Numerical results demonstrate that for certain problems (those notgenerally requiring colorset calculations to obtain DFs), using fluxand current DFs together is more accurate than using only flux DFs.
Furthermore, DFs that make the first angular moment (current) andsecond angular moment of the reconstructed flux continuous acrossinterfaces are more accurate than using DFs that make the zeroth andfirst angular moments (scalar flux and current) of the reconstructed
flux continuous.

14:20 PMA Posteriori Reconstruction of the Flux Profile in
the Case of Localized Axial Heterogeneities: An
Application to the Modeling of PWR Spacer Grids
E.Girardi, A.Aktogu, H.Leroyer, C.Meriot
Electricité de France-R&D, Clamart, France
In this paper we discuss the possibility to model localized
heterogeneities (e.g. fuel assembly spacer grid), within the context of
few-group homogeneous diffusion calculation. This work is motivatedby the research of a model being able to precisely take into accountboth global (reactivity effect) and local (axial profile) effect. Based on
the assumption that the grid can be seen as a local perturbation, theaxial flux profile is then reconstructed a posteriori by incorporating
additional local informations to the diffusion core calculation, following
the basic principles of the well-known pin-power reconstruction
procedure for homogeneous assembly calculations. After describing
the main steps allowing to establish the axial reconstruction model, its
validity is tested on a simple fuel assembly test-case, and verified withrespect to the reference axial profile computed by the Monte Carlocode TRIPOLI4. Then an optimization work, aimed at reducing theaxial form factor computing cost, is undertaken. We show that the axial
form factor can be computed on a simplified cylindrical fuel cell, andwe propose a faster calculation scheme, based on the APOLLO2 SN
solver. The final computing time for the form factor is reduced at ~10s,with a very limited precision loss. One-group fission rate discrepancies
are within ±0.5% in a wide range around the central part of theassembly, and it increases up to -6% maximum in the outermost partof the assembly, in all likelihood due to the 2-groups homogeneous
diffusion hypothesis and not to the factorization approach.

14:40 PM

Effects of Advanced Radial Submeshing Methods on
Pin Power Reconstruction for an EPR Core Design
P.Mala, S.Canepa, H.Ferroukhi, A.Pautz
Paul Scherrer Institute, Switzerland
This paper presents an evaluation of the new radial submesh methodof the advanced core simulator SIMULATE-5 (S5) and the effects on
pin power reconstruction for an EPR core design. The developed S5
core model is first verified both against results from the predecessor
SIMULATE-3 (S3) code as well as to a vendor solution. The latter

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Mizuho_A

Track1-1 Reactor Analysis Methods
Session Chair: Tarmer Bahadir(Studsvik Scandpower, Inc.), Masayuki Tojo(GNF-J)
provides confidence in the developed model while the consistency
with the former constitutes an intrinsic verification of the new S5
methodology in terms of performing the critical leakage correctionof nuclear data that was previously handled by the lattice code.
Thereafter, the impact of various submeshes on the S5 assembly andpin power results is studied, starting from a coarse 1x1 assemblysize mesh typical of conventional nodal codes down to a fine 17x17
submesh structure representative of the lattice full pin-by-pin geometry.
Since focus is on the radial submeshing, only 2-D axially-integrated
pin powers are investigated and it is found that the various submesh
subdivisions give comparable results, with differences typically around±1%. It is however confirmed that submesh effects increase acrossand within fuel assemblies as function of stronger flux heterogeneities
and gradients (e.g. core/reflector interfaces and/or control rods). An
observation is that submesh “mixture” effects can occur, especially ifsubmeshes include gadolinium bearing fuel rods, and compared to afine mesh solution, this could produce even larger differences than asolution obtained with a simple coarse assembly size mesh.

15:00 PM

Study on Robust Energy Group Structure to Spectral
Interference for PWR Pin-By-Pin Core Analysis
S.Wada, T.Kitada(1), S.Takeda, T.Ushio(2)1)Osaka University, Osaka, Japan, 2)Nuclear Fuel Industries, Ltd.,
Osaka, JapanThe energy group structure which reduces the error caused byinterference effect is studied. The error in pin-power distribution
becomes large at interference area of UO2 and MOX fuel assemblies.
Although the error could be reduced by increasing the number ofcollapsed energy groups, the target of this study is to find robustenergy group structure against spectral interference without changing
the number of collapsed energy groups and without introducing any
additional parameters. Considering the characteristics of cross-section
curve and the change of neutron flux caused by interference effect,
we obtain the robust energy group structure against interference
effect and evaluate the robustness of this energy group structure incases of a multi-assembly and a single-assembly having the different
interference effect.

13

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Mizuho_B

SS2-1 Reactor Physics and Criticality Safety Activities in OECD/NEA Working
Party
Session Chair: Mark DeHart(INL), Teruhiko Kugo(JAEA)
13:00 PM

Activities of OECD/NEA on Scientific Issues ofReactor Systems and Criticality Safety - Current
Status and Future Plan
Guest Speakers: J.Gulliford(1), M.C.Brady Raap(2)1)OECD/NEA, Issy-les-Moulineaux, France, 2)Pacific Northwest National
Laboratory, Richland, USA

13:20 PM

Uncertainty Analysis of the OECD/NRC Oskarshamn-2
BWR Stability Benchmark
I.Gajev, W.Ma(1), T.Kozlowski(2)1)Royal Institute of Technology, Stockholm, Sweden, 2)University of
Illinois, Urbana-Champaign, USA
On February 25, 1999, the Oskarshamn-2 NPP experienced astability event which culminated in diverging power oscillations witha decay ratio of about 1.4. The event was successfully modeled by
the TRACE/PARCS coupled system code, and further uncertainty
analysis of the event is described in this paper. The results show very
good agreement with the plant data, capturing the entire behavior of
the transient including the onset of instability, growth of the oscillations(decay ratio) and oscillation frequency. This provides confidence inthe prediction of other parameters which are not available from theplant records. This paper shows also how an uncertainty method was
implemented for the event. Comparing the calculated uncertainty
with the measured uncertainty gives confidence in the BWR stability
prediction.

13:40 PM

Analysis of the OECD/NEA Oskarshamn-2 FeedwaterTransient and Stability Benchmark with SIMULATE-3K
A.Dokhane, H.Ferroukhi, A.Pautz
Paul Scherrer Institute, Switzerland
The OECD/NEA recently launched an international benchmark on
a combined feedwater transient and stability event that occurredat the Swedish Oskarshamn-2 (O2) nuclear power plant (NPP).
The primary benchmark objective is to assess advances in coupledneutronic/thermal-hydraulic codes for simulations of challenging
transients including the appearance of unstable power oscillations.
The Paul Scherrer Institut (PSI) is participating in this benchmark inorder to enlarge the validation basis of its advanced stability analysis
methodology currently under development for Swiss BWRs andbased on the state-of-the-art SIMULATE-3K (S3K) code. This paper
presents the development, optimization and validation of a S3K modelfor the first phase of the O2 benchmark, namely the analysis of theentire event. With the optimized model, the S3K solution is compared
to available benchmark data both for at steady-state and transientconditions. For the latter, the qualitative as well as quantitative
behavior of S3K results is compared and discussed in relation to
the experimental observations. The modeling aspects found in thiscontext to mostly affect the S3K ability to reproduce the event are alsopresented. Finally, the S3K model indicates that if the reactor scram
had not been introduced, the observed diverging oscillations would
reach a maximum amplitude before decaying back into a stable state.
However, it was also found that the core could instead have evolved
into limit-cycle oscillations if a stabilization of the feedwater flow andtemperature had occurred just before the scram signal.

14:00 PM

Data Assimilation for Kinetic Parameters Uncertainty
Analysis
E.Ivanov, T.Ivanova(1), I.Kodeli(2), V.Mastrangelo(3)1)Institut de Radioprotection et Surete Nucleaire, Fontenay-aux-Roses,
France, 2)Jožef Stefan Institute, Ljubljana, Slovenia, 3)PACS-IPN-Orsay,
Orsay, France
Several years ago, the OECD/NEA Nuclear Science Committee (NSC)
established the Expert Group on Uncertainty Analysis in Modeling(UAM-LWR) after thorough discussions of the demands from nuclear
research, industry, safety and regulation to provide the best estimate
predictions of nuclear systems parameters with their confidencebounds. UAM objectives include among others, the quantification of
uncertainties of neutronic calculations with respect to their value for14

the multi-physics analysis.
Since the kinetics parameters and their uncertainties are of particularinterest for these studies the deterministic approaches for analysis of
uncertainties in nuclear reactor kinetic parameters (neutron generation
lifetime and delayed neutron effective fraction) have been developedin frame of the UAM-LWR. The approach uses combination ofgeneralization of perturbation theory to reactivity analysis, and theGeneralized Perturbation Theory (GPT) for sensitivity computation.
It has been applied to the UAM complementary fast neutron SNEAK
test case that has unique set of experimental data for βeff. In thisexample, the covariance matrices of nuclear data have been derived
from COMMARA library by Bayesian adjustment upon the set of thefast neutron integral benchmarks with the BERING code package.
Then, the kinetic parameters uncertainties with their correlations havebeen applied to simplified model of a reactivity insertion transient
where relative uncertainty of power peak was taken as figure of
merit. The results demonstrate that the uncertainties due to nuclear
data impact significantly the energy release in a coupled transientmodeling. It was also found that such uncertainties become higher if
the correlations between uncertainties of different lumped parameters
are taken into account.

14:20 PM

Criticality and Reactor Physics Benchmark
Experiments: Influence of Nuclear Data Uncertainties
W.Zwermann, F.Weiss, M.Clemente, A.Aures, K.Velkov
GRS, Garching, Germany
A number of LWR-type criticality and reactor physics experiments,
mainly from the ICSBEP and IRPhEP Handbooks, many of whichwere already used in the past for international benchmarks in theframework of OECD/NEA working groups, are being evaluatedwith respect to uncertainties in the basic nuclear data. For this, thesampling based uncertainty and sensitivity analysis tool XSUSA alongwith the Monte Carlo code KENO-V.a as transport solver is employed.
Particular emphasis is put on experiments where differential quantities,
mainly reaction rate distributions, were measured; the uncertainties
of such quantities are not directly accessible to tools based on firstorder perturbation theory. With respect to multiplication factors andreactivity differences, all results are compared with corresponding
results obtained with TSUNAMI-3D from the SCALE 6.1 system; theagreement is very good for all assemblies under consideration. With
respect to fission rate distributions, the uncertainty analyses yield only
moderate uncertainties from nuclear data; therefore, in general the
total uncertainty is dominated by measurement uncertainties, which
include the uncertainties of technological parameters. The work is
continuously being extended; in the future, also non-LWR specificassemblies, mainly relevant for GEN-IV reactors, will be investigated.

14:40 PMThe Evaluation of the Subcritical Experiments
Performed in the IPEN/MB-01 Research Reactor
Facility for the IRPhE Project
A.D.Santos, S.M.Lee, R.Jerez, R.Diniz
IPEN-CNEN/SP, São Paulo, Brazil
This work presents the evaluation of the subcritical experiments
performed at the IPEN/MB-01 research reactor facility. The
experimental method developed at the IPEN facility showed itselfvery valuable in the resolution of the several subcritical parameters
employed in the subcritical kinetic model of Gandini and Salvatores.
The evaluated data which were acceptable for publication in
the IRPhE handbook are of very good quality and suitable for abenchmark problem. The theoretical analyses show that the MCNP5
results, together with the ENDF/B-VII.0 library, for for keff, ρgen, Σρgen,and ksub are in a very good agreement to the benchmark values. In thecase of the calculation of ζ MCNP5 does a good job when there is justone additional source, but with the source is distribute like in the caseof the intrinsic source, MCNP5 misses part of the physics because theintrinsic source is not weighted by the adjoint flux according to theirposition. In other words, the intrinsic source that it is in the center ofthe core is more effective to induce fissions than the one that it is in
the periphery. This kind of MCNP5 deficiency imposes restrictions in
the determination of ζ and consequently in the determination of φ*. The
ksub MCNP5 results were insensitive to the source treatment and only
a few cases were slightly outside of the 3σ range of the benchmark
values.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Mizuho_B

SS2-1 Reactor Physics and Criticality Safety Activities in OECD/NEA Working
Party
Session Chair: Mark DeHart(INL), Teruhiko Kugo(JAEA)
15:00 PM

Polaris: A New Two-Dimensional Lattice Physics
Analysis Capability for the SCALE Code System
M.A.Jessee, W.A.Wieselquist, T.M.Evans, S.P.Hamilton, J.J.Jarrell,
K.S.Kim, J.P.Lefebvre, R.A.Lefebvre, U.Mertyurek, A.B.Thompson,
M.L.Williams
Oak Ridge National Laboratory, Oak Ridge, USA
Polaris is a new 2-dimensional (2-D) lattice physics capability in the
SCALE code system for the analysis of light water reactor fuel designs.
In this paper, the Polaris calculational methods are summarized andresults are provided for a series of computational benchmarks. The
summary includes the implementation of the relatively new resonanceself-shielding approach called the embedded self-shielding method,
the implementation of a new 2-D method-of-characteristics neutrontransport solver, and the integration of the SCALE/ORIGEN depletionand decay solver for depleting material compositions. Polaris
calculations are compared with reference continuous energy Monte
Carlo solutions for a UO2 fuel computational benchmark. Because
Polaris is integrated into the SCALE code system, Polaris has beenutilized as part of the new SCALE/Sampler code sequence, which
provides stochastic uncertainty analysis for the impact of cross-section
uncertainties on lattice physics calculations. SCALE/Sampler results
are summarized for selected benchmark calculations.

15

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Hiei

SS1-1 Molten Salt Reactors
Session Chair: Imre Pazsit(Chalmers Univ. of Tech.), Ritsuo Yoshioka(International Thorium Molten-Salt Forum)
13:00 PM

Experimental Modelling and Numerical Analysis of a
Molten Salt Fast Reactor
B.K.Yamaji, A.Aszodi
Institute of Nuclear Techniques, Budapest University of Technology andEconomics, Budapest, Hungary
In this paper experimental and numerical investigation of the
MSFR (Molten Salt Fast Reactor) concept will be presented. This
homogeneous, single region liquid fuelled fast reactor concept uses
fluoride-based molten salts with fissile uranium and thorium andother heavy nuclei content with the purpose of applying the thorium
cycle and the burn-up of transuranic elements. Molten salt reactors
with liquid fuel have a unique safety related property that needs clear
understanding. In the core neutron flux and fission distribution isdetermined by the flow field through distribution and transport of fissile
material and delayed neutron precursors. Since the MSFR concept
has a single region homogeneous core without internal structures,
it is a difficult task to ensure stable flow field, which is also strongly
coupled to the volumetric heat generation. These considerations
suggest that experimental and numerical modelling (including theoption of coupled neutronics-thermal-hydraulics) would be needed tobetter understand the flow phenomena in such geometry.
A scaled and segmented experimental mock-up of MSFR was
designed and built at BME NTI with the purpose of investigating theflow behavior inside the core region using particle image velocimetry.
Not only the basic flow behavior inside the core region can beinvestigated but measurement data can also provide resource for thevalidation of computational fluid dynamics models, specific problems
or phenomena (for example inlet geometry, optional internal structures,
mixing) may be studied as well. Measurement results of steady state
conditions will be presented with comparison of measurement data
and results of numerical analyses.

13:20 PM

Remark on the Propagating Neutron Noise in a MSR
V.Dykin, I.Pazsit(1), R.Sanchez(2)1)Chalmers University of Technology, Gothenburg, Sweden, 2)
Commissariata l'Energie Atomique et aux Energies Alternatives, Gif-surYvette cedex, France
The neutron noise induced by propagating perturbations in a bare
1-D Molten Salt Reactor (MSR) model is calculated and analyzed
using one-group diffusion theory. The neutron noise for different noise
sources of which two have not been accounted for, corresponding to
the fluctuations of the fission and absorption cross sections as wellas to the fuel velocity is calculated and the results are qualitatively
compared. Unlike in previous work, the solution is obtained through thematrix Green’s function of the flux and precursor equations being kept
separate. It is shown that in the case when the noise is represented by
the fluctuations of the fission cross-section, the noise source attains
a complex structure which is different from that in traditional reactors.
On the other hand, in the cases investigated, despite all qualitative
differences in the noise calculation procedure as well as in the
structure of the noise source, it turns out that the noise induced by theabsorption and the fission cross sections follow a similar behaviour. In
addition, it is observed that the inclusion of the fluctuations in the fuelvelocity examined in this paper slightly suppresses the total neutron
noise for low frequency region i.e. below ~ 2 Hz but on the other
hand it enhances the latter one by one of order of magnitude for highfrequencies i.e. above ~ 2 Hz compared to the effect of other noise
sources. The results contribute to the understanding and interpretationof the neutron noise in MSRs.

13:40 PMThe Two-Group Point-Kinetic Component of Neutron
Noise in an MSR
V.Dykin, I.Pazsit
Chalmers University of Technology, Gothenburg, Sweden
The calculation of the point kinetic component of the neutron noisein two-group diffusion theory in Molten Salt Reactors (MSRs) using
different techniques is discussed. First, the point kinetic componentof the noise is calculated from the full space-frequency dependent
solution analytically by a projection to the static adjoint. Then, thepoint-kinetic solution is derived by solving the simplified point kinetic
equations. Both results are thereafter analyzed and compared

16

quantitatively. This comparison shows that the solution of the simplified
point kinetic equations significantly differs from the exact one andcannot reconstruct some important features of the true solution. The
similar discrepancies between two methods were also observed andconfirmed in earlier one-group MSR calculations.

14:00 PM

Neutronics of Fluid Fuel System with Perfect
Remixing
S.Dulla, P.Ravetto(1), A.K.Prinja(2)1)Politecnico di Torino, Torino, Italy, 2)University of New Mexico, USA
The neutronics of a homogeneous fluid-fuel nuclear system withperfect remixing implying a uniform distribution of delayed neutron
precursors inside the core is studied in this work. A one-groupdiffusion model is adopted for the neutron balance and an analyticaltreatment is used throughout thewhole analysis. The critical problemis
first solved, allowing the determination of a set of eigenfunctions
characteristic of the physico-mathematical problem considered. The
time-dependent problem is then solved by expanding the neutron flux
in terms of these eigenfunctions and the results are compared withthose obtained through a standard Helmholtz eigenfunction expansion.
The use of these eigenfunctions show to be advantageous, since theasymptotic state of the system can be represented by the fundamental
eigenfunction only.

14:20 PM

An Innovative Approach to Dynamics Modeling and
Simulation of the Molten Salt Reactor Experiment
M.Zanetti, L.Luzzi, A.Cammi(1), C.Fiorina(2)1)Politecnico di Milano, Milano, Italy, 2)Paul Scherrer Institute, Villigen,
Switzerland
The Molten Salt Reactor Experiment (MSRE) was a circulating fuelthermal reactor built and operated in the sixties. As the only Molten
Salt Reactor (MSR) testing facility for which extensive experimental
data are available, it can be considered as a reference for thedevelopment of modeling approaches for the studies related to the
Gen-IV MSR. In this work, a geometric multi-scale approach has
been adopted for the simulation of the MSRE plant. The data andthe experimental results relative to the U-233 fuelled reactor are considered. The neutronic parameters have been determined using theMonte Carlo code Serpent. The reactor core is divided into three radial
regions, each one described by a 3D channel in which Navier-Stokes
and energy conservation equations plus delayed neutron precursors(DNP) balance equations are solved. Determination of the generated
power is obtained employing a point kinetics like equation, fed with
importance weighted values of temperatures and DNP concentrations.
The remaining part of the plant, that includes the primary andsecondary cooling circuits, is modeled by means of zero-dimensional
components. The results attained with such modeling approach arecompared with experimental data both in time and frequency domain,
showing good agreement. The adopted approach, thanks to thepunctual, coupled solution of the governing equations in the core,
gives better insights into the thermal behavior of the graphite and its
effects on MSR dynamics than commonly used correlation-based
solvers.

14:40 PM

Safety Criteria and Guidelines for MSR Accident
Analysis
R.Yoshioka, K.Mitachi(1), Y.Shimazu(2), M.Kinoshita(3)1)International Thorium Molten-Salt Forum, Yokohama, Japan, 2)
University of Fukui, Fukui, Japan, 3)University of Tokyo, Tokyo, Japan
Accident analysis for Molten Salt Reactor (MSR) has been investigatedat ORNL for MSRE in 1960s. Since then, safety criteria or guidelines
have not been defined for MSR accident analysis. Regarding thesafety criteria, the authors showed one proposal in this paper. In
order to establish guidelines for MSR accident analysis, we have toinvestigate all possible accidents. In this paper, the authors describe
the philosophy for accident analysis, and show 40 possible accidents.
They are at first classified as external cause accidents and internalcause accidents. Since the former ones are generic accidents, weinvestigate only the latter ones, and categorize them to 4 types, suchas power excursion accident, flow decrease accident, fuel-salt leak

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Hiei

SS1-1 Molten Salt Reactors
Session Chair: Imre Pazsit(Chalmers Univ. of Tech.), Ritsuo Yoshioka(International Thorium Molten-Salt Forum)
accident, and other accidents mostly specific to MSR. Each accident
is described briefly, with some numerical results by the authors.

15:00 PM

Reactivity-Insertion-Transient Analysis of a Fluoride
Salt Cooled High Temperature Reactor
Y.Yang, F.Yao, Z.Yang, S.Qiang, Z.Jie(1,2)
1)Shanghai Institute of Applied Physics, Shanghai, China, 2)KeyLaboratory of Nuclear Radiation and Nuclear Energy Technology,
Chinese Academy of Sciences, Shanghai, China
The Fluoride salt cooled High temperature Reactor (FHR) is aninnovative reactor design that uses conventional TRISO hightemperature fuel with a low-pressure liquid salt coolant. The designof this reactor is currently in progress both in China and in the UnitedStates. An FHR based on ordered pebble bed core design is being
planned for construction by the Shanghai Institute of Applied Physics(SINAP). This paper provides a preliminary reactivity insertion transient
analysis of an FHR of SINAP’s pebble core design, using RELAP5/
MOD4.0 code. New models and methodologies are developed forseveral prototypical facilities that are based on SINAP’s pebble
bed concept, and different types of reactivity insertion transient are
analyzed. SINAP’s design is currently in progress; the ultimate goal of
the transient analysis is to acquire the capability of RELAP5/MOD4.0
for performing FHR core design.

17

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Atago

SS5 Multiscale, Multiphysics Approaches in Nuclear Science and Engineering
Applications
Session Chair: Richard Martineau(INL), Takeshi Mitsuyasu(Hitachi)
13:00 PM

Influence of an Sn Solver in a Fine-Mesh Neutronics/
Thermal-Hydraulics Framework
K.Jareteg, P.Vinai, C.Demaziere, S.Sasic
Division of Nuclear Engineering, Department of Applied Physics,
Chalmers University of Technology, Gothenburg, Sweden
In this paper a study on the influence of a neutron discrete
ordinates (SN) solver within a fine-mesh neutronic/thermal-hydraulicmethodology is presented. The methodology consists of coupling aneutronic solver with a single-phase fluid solver, and it is aimed atcomputing the two fields on a three-dimensional (3D) sub-pin level.
The cross-sections needed for the neutron transport equations
are pre-generated using a Monte Carlo approach. The coupling is
resolved in an iterative manner with full convergence of both fields. A
conservative transfer of the full 3D information is achieved, allowing
for a proper coupling between the neutronic and the thermal-hydraulicmeshes on the finest calculated scales. The discrete ordinates solver
is benchmarked against a Monte Carlo reference solution for a twodimensional (2D) system. The results confirm the need of a highnumber of ordinates, giving a satisfactory accuracy in keff and scalar
flux profile applying S16 for 16 energy groups. The coupled frameworkis used to compare the SN implementation and a solver based on theneutron diffusion approximation for a full 3D system of a quarter of asymmetric, 7x7 array in an infinite lattice setup. In this case, the impactof the discrete ordinates solver shows to be significant for the coupledsystem, as demonstrated in the calculations of the temperature
distributions.

13:20 PM

High-Fidelity Multi-Physics Calculations for Light
Water Reactors Using Coupled CTF/TORT-TD/
FRAPTRAN
J.W.Magedanz, M.N.Avramova
The Pennsylvania State University, Pennsylvania, USA
This paper describes the development and testing of a high-fidelity
multi-physics system consisting of the codes CTF, TORT-TD, and
FRAPTRAN. CTF is an improved version of the subchannel code
COBRA-TF, while TORT-TD is a multi-group discrete-ordinates
neutron kinetics code used for pin-wise analysis of LWR’s.
FRAPTRAN, the most recent addition to the system, is a fuelperformance code, which was incorporated in order to provide moreaccurate feedback to the thermal-hydraulic and kinetics calculations.
The high-fidelity multi-physics system has an increased level offidelity over current methods by modeling on the pin scale, taking into
account the angular dependence of the neutron flux, and taking into
account the full range of phenomena that occur inside the fuel rods. It
is intended to be used to study the circumstances in which such higher
fidelity methods provide a significant benefit. The results obtained here
demonstrate the system capability for a set of simple problems basedon fuel pin properties obtained from the OECD/NRC PWR MOX/UO2Core Transient Benchmark.

13:40 PMThe Coupling of the Neutronic Transport Application
RATTLESNAKE to the Nuclear Fuels Performance
Application BISON under the MOOSE Framework
F.N.Gleicher, J.Ortensi, B.W.Spencer, D.Gaston, Y.Wang,
S.R.Novascone, J.D.Hales, R.L.Williamson, R.C.Martineau
Idaho National Laboratory, Idaho, USA
The MOOSE based reactor physics tool MAMMOTH provides thecapability to seamlessly couple the neutron transport application
RATTLESNAKE to the fuels performance application BISON toproduce a higher fidelity tool for fuel performance simulations. The
ultimate purpose of this coupling is to provide a tool with the predictive
capabilities to gain new knowledge and help resolve fundamental
questions in the fuel performance arena, i.e. high-burnup structures,pellet-cladding interaction, missing pellet surface, etc. RATTLESNAKE
solves the self-adjoint angular flux transport equation, derived from thelinearized Boltzmann transport equation, and provides a sub-pin level
resolution of the multigroup neutron flux. BISON solves the coupled
thermomechanical equations for the fuel on a sub-millimeter scale.
The coupling within the MOOSE framework allows both applications tosolve their respective systems on aligned and unaligned unstructured
finite element meshes. MAMMOTH uses the power density

18

calculated by RATTLESNAKE to compute the local burnup evolution.
Subsequently, MAMMOTH transfers the power density and burnup
distribution to BISON with the MOOSE Multiapp transfer system.
BISON in turn is able to provide sub-pin level temperature for crosssection feed back effects. Multiple depletion cases were run with oneway and two-way data transfer in MAMMOTH for RATTLESNAKEBISON. The one-way eigenvalues obtained show good agreement withthe reference values obtained from the lattice physics code DRAGON4
while the two-way eigenvalue show expected differences. The power
distributions obtained are consistent with both DRAGON4 and the
SERPENT Monte Carlo code. The one-way and two-way calculations
produce power density results that are comparable with those of theinternal, static, Lassmann style model in BISON. Differences in the
power densities arise from the use of better neutron energy deposition
parameters obtained from the DRAGON4 tabulations, and differencesin the fuel temperature arise from a different thermal expansion
models in the fuel.

14:00 PMA Model of Two-Stage Core Calculation Method
Coupled with Subchannel Analysis for Boiling Water
Reactors
T.Mitsuyasu, K.Ishii, N.Nakadozono, M.Aoyama
Hitachi, Ltd., JapanThe two-stage core analysis method is widely used for BWR core
analysis. The purpose of this study is to develop the core analysis
model coupled with subchannel analysis within the two-stage
calculation scheme using an assembly-based thermal-hydraulics
calculation in the core analysis. This model does not change thethermal-hydraulics scheme of the core analysis. Rather, it appends thesubchannel void distribution to the previous uniform analysis in latticephysics, and couples that with the subchannel analysis which axiallycalculates full assembly and uses the flow condition that produces
the maximum void fraction in the operation core. The subchannel
void distribution of one node from the subchannel analysis is only
normalized and used for the lattice physics. The developed model was
evaluated for the heterogeneous problem with multiple enrichments.
The developed model could decrease the eigenvalue differences bymore than half of that of the uniform case and made the differencesof assembly power the same as the uniform case. Furthermore, it
could reduce the root mean square differences to more than half of
those of the uniform case in the low and high enrichment fuels. The
computation times of the lattice physics become 2.3 times longer. The
extended computing time does not prevent core analysis because
the nuclear data are prepared in advance of the core analysis. As theresult of the evaluation, the model can incorporate the subchannnel
effect to the core analysis.

14:20 PM

Subspace Methods for Multi-Physics Reduced Order
Modeling in Nuclear Engineering Applications
B.A.Khuwaileh, J.M.Hite, H.S.Abdel-Khalik
North Carolina State University, Raleigh, USA
This manuscript proposes a new extension to a reduced ordermodeling algorithm, previously introduced for single-physics models,
to multi-physics models. This manuscript focuses on looselycoupled physics models wherein the output of one physics model is
fed as input to the next physics model, and each physics model issolved separately while freezing all other physics models. The idea
is to perform three reductions at each physics-to-physics interface,
one based on the upstream physics, another for the downstream
physics, and a third for the interaction thereof. Accurately capturing
the interaction between the reduced physics models is an essential
feature of the proposed algorithm, and will be the key measure forits success. For standard model execution, this interaction is often
captured using an iterative technique that loops over the different
physics until convergence or meeting some stopping criterion. We
adopt a similar approach in which the effective dimensionality of eachphysics-to-physics interface is updated iteratively until a user-definederror tolerance is met. A quarter PWR fuel assembly depleted to 32
GWD/MTU by iteratively solving the quasi-static transport-depletion
approximation is used to exemplify the application of the proposed
algorithm. Active subspaces for the nuclear cross-sections andneutron flux are determined, and compared to the active subspaces
obtained without the physics coupling.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Atago

SS5 Multiscale, Multiphysics Approaches in Nuclear Science and Engineering
Applications
Session Chair: Richard Martineau(INL), Takeshi Mitsuyasu(Hitachi)
14:40 PM

Extension of the Entropy Viscosity Method to Flows
with Friction Forces and Source Terms
M.Delchini, J.C.Ragusa(1), R.A.Berry(2)1)Texas A&M University, College Station, USA, 2)INL, Idaho Falls, USA
In this paper, we extend the entropy viscosity method to 1-D Euler
equations with source terms present. The entropy viscosity method
has been successfully applied to hyperbolic equations such as
Burgers equation and the Euler equation system. This method
consists in adding dissipative terms to the governing equations so
as to ensure the entropy minimum principle. The dissipative terms
contain a viscosity coefficient (function) that locally modulates theamount of dissipation. This viscosity coefficient is based on theentropy production that occurs in the wiggles, discontinuities, andshocks of hyperbolic equation systems. By adding source terms to theEuler equations (friction and gravity forces to the momentum equationand heat sources/sinks in the energy equation), the entropy viscosity
method must be modified to account for the entropy production
due to these additional terms. Tests are run for a 1D channel, usingpressurized water reactor (PWR) conditions, with the RELAP-7 code
based on the MOOSE framework. The equations are discretizedwith a continuous Galerkin finite element method (FEM) using linear
polynomials along with a second-order, implicit temporal scheme(BDF2).

15:00 PM

Efficient Finite Element Field Interpolation for
Multiphysics Applications
D.Lebrun-Grandie(1), J.C.Ragusa(1), R.Sampath(2)1)Texas A&M University, College Station, USA, 2)Houston, TX, USA
We have proposed a two-stage algorithm for data field interpolation
for parallel multiphysics simulations. In such simulations, source andtarget meshes between physics components are not necessarily
coincident (for example, neutronic/thermal-hydraulic coupledcomputations) or non-penetration conditions between multiple bodies
must be enforced (for instance, pelletclad interaction in nuclear fuelperformance). Our approach employs a coarse parallel search todetermine the subset of elements where data points of interest arelocated. This stage uses an octree representation and space-filling
curve. In the second stage, a local fine search is performed by inverse
the coordinate mapping. This nonlinear system is solved by meansof a Newton technique and several initial guess options are proposed
to improve the robustness of the nonlinear solve. Weak and strong
scaling studies were performed.

19

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Cosmos

Track11-1 Research Reactors and Spallation Sources
Session Chair: Ronald J.Ellis(ORNL), Gaillot Stephan(CEA)
13:00 PMAn Updated Core Design for the Multi-Purpose
Irradiation Facility MYRRHA
G.Van den Eynde, E.Malambu, A.Stankovskiy, R.Fernandez, P.
BaetenSCK-CEN, Belgium
MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech
Applications) is a multipurpose research facility currently being
developed at SCK-CEN. It will be able to work in both critical andsubcritical modes and, cooled by lead-bismuth eutectic. It will play a
key role in the development of the Pb-alloy technology needed for the
LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will demonstrate
the ADS (Accelerator Driven System) full concept by coupling a
proton accelerator, a spallation target and a subcritical reactor at areasonable power level to allow operation feedback. MYRRHA will
also contribute to the study of partitioning and transmutation of highlevel waste. Recently, a new core design with a longer active core has
been proposed. This paper presents the first neutronic analyses forthis design improvement.

13:20 PM

Feasibility Study of Installing a Thermal to 14 MeV
Neutron Converter into a Research Nuclear Reactor
L.Snoj, A.Trkov, I.Lengar, A.Kolšek, A.Jazbec, V.Radulović,
G.Žerovnik(1), P.Sauvan, F.Ogando, J.Sanz(2)1)Jozef Stefan Institute, Ljubljana, Slovenia, 2)Universidad Nacional de
Educacion a Distancia Ingenieria Energetica, Madrid, Spain
A deuterium-tritium (DT) based thermal-to-fast neutron converter
will be installed into the TRIGA reactor at Jožef Stefan Institute. This
paper presents preliminary results of a feasibility study where different
aspects have been addressed. The thermal column is the mostappropriate irradiation position for the DT converter in the reactor.
From potential active converter materials, LiD yields the most 14 MeV
neutrons. Lithium enrichment affects the required thickness of theactive converter material, but not significantly the 14 MeV neutron
yield. For coupled neutron-tritium transport calculations, the modified
MCUNED code was used. Relevant DT neutron activation monitor
materials for spectrum unfolding have been identified and neutron flux
criteria established.

13:40 PM

Development and Validation of a New APOLLO2Based Calculation Scheme Dedicated to Ex-Core Rod
Irradiations in the OSIRIS MTR Reactor
F.Chevallier, F.Malouch, S.Santandrea
Alternative Energies and Atomic Energy Commission (CEA), Gif sur
Yvette Cedex, FranceThis paper describes a new neutron calculation scheme dedicated tolong-duration fuel-rod irradiations carried out with the GRIFFONOS
device in the periphery of the core of the OSIRIS material testing
reactor (CEA/Saclay Center). With this calculation scheme, basedon the APOLLO2 neutron transport code, the neutron flux and fission
power can now be computed taking into account most irradiation
parameters such as the core-loading map, the control rod effects
or the fuel depletion in the core assemblies. Validation has been
performed by comparisons to reference calculations using TRIPOLI-4
Monte-Carlo code and measured quantities. Obtained discrepancies
on neutron flux and fission power in the fuel rod are less than 5%.

14:00 PM

Neutronic Designs and Analyses of a New CoreModerator Assembly and Neutron Beam Ports for thePenn State Breazeale Reactor
D.Ucar, K.Unlu, B.J.Heidrich, K.N.Ivanov, M.N.Avramova
The Pennsylvania State University, Pennsylvania, USA
A new core-moderator assembly and five new neutron beam ports aremodeled and designed for the Penn State Breazeale Reactor (PSBR).
The PSBR is an open pool, light water cooled, and moderated 1-MW
research reactor with seven neutron beam ports. The existing coremoderator assembly design does not allow simultaneous utilizationof all the available beam ports; only two beam ports, namely #4 and20#7, are currently in use for research and education in the facility.
Moreover, the prompt gamma-rays produced at the back side of
the heavy water moderator tank shine into neutron beam tube #4.
Subsequently that is hampering the quality of the experimental dataat the existing beam port facilities. The proposed design eliminates
all the limitations of the existing design and provides multiple highintensity and clean neutron beams to a new and expanded beam hallutilizing various instruments and techniques. The new design featuresa crescent-shaped moderator tank, which couples the reactor core tofour thermal ports and one cold neutron beam port with three curved
guide tubes for various cold neutron beam techniques. The modelingof the new PSBR design was achieved with highly detailed neutronics
simulations using several stochastic simulation tools developed for the
PSBR. The simulation results revealed the optimal design parameters
and neutronics performance of the new beam ports, such that thethermal neutron beam intensity was significantly increased and thetotal prompt gamma dose was drastically decreased in the new beam
port facilities.

14:20 PM

Design Studies for a Multiple Application Thermal
Reactor for Irradiation Experiments (MATRIX)
M.A.Pope, H.D.Gougar(1), J.M.Ryskamp(2)1)Idaho National Laboratory, Idaho, USA, 2)Retired, USA
The Advanced Test Reactor (ATR) is a high power density test
reactor specializing in fuel and materials irradiation. For more than 45
years, the ATR has provided irradiations of materials and fuels plus
secondary missions such as the production of radioisotopes. Should
unforeseen circumstances lead to the decommissioning of ATR,
the U.S. Government would be left without a large-scale materials
irradiation capability to meet the needs of its nuclear energy andnaval reactor missions. In anticipation of this possibility, work was
performed under the Laboratory Directed Research and Development(LDRD) program to investigate test reactor concepts that could
serve the current missions of the ATR along with an expanded set of
secondary missions. A survey was conducted in order to catalogue
the anticipated needs of current and potential customers. Concepts
were then evaluated to fill the role for this reactor, dubbed the MultiApplication Thermal Reactor for Irradiation eXperiments (MATRIX).
This evaluation indicates that the baseline MATRIX design describedherein can achieve longer cycle lengths than ATR given a particular
batch scheme and using uranium enriched to less than 20%. The
volume of test space in In-Pile-Tubes (IPTs) in MATRIX is larger thanthose in ATR with comparable magnitude of neutron flux. Furthermore,
MATRIX has a greater number and volume of irradiation spaces
having high fast neutron flux than ATR. From the analyses performedin this work, it appears that a new materials test reactor, based upon
the MATRIX concept, can meet the anticipated needs of current andnew high flux irradiation programs for the foreseeable future.

14:40 PM

Simulated Irradiation of Samples in HFIR for Use as
Possible Test Materials in the MPEX (Material Plasma
Exposure eXperiment) Facility
R.Ellis, J.Rapp
Oak Ridge National Laboratory, Oak Ridge, USA
The importance of Plasma Material Interaction (PMI) is a major
concern in fusion reactor design and analysis. The Material-Plasma
Exposure eXperiment (MPEX) facility will explore PMI under fusion
reactor plasma conditions. Samples with accumulated displacements
per atom (DPA) damage produced by irradiations in the High Flux
Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL)
will be studied in the MPEX facility. The project presented in thispaper involved performing assessments of the induced radioactivity
and resulting radiation fields of a variety of potential fusion reactor
materials. The scientific code packages MCNP and SCALE wereused to simulate irradiation of the samples in HFIR; generation anddepletion of nuclides in the material and the subsequent composition,
activity levels, gamma radiation fields, and resultant dose rates as afunction of cooling time. These state-of-the-art simulation methods
were used in addressing the challenge of the MPEX project tominimize the radioactive inventory in the preparation of the samples
for inclusion in the MPEX facility.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 13:00 PM - Cosmos

Track11-1 Research Reactors and Spallation Sources
Session Chair: Ronald J.Ellis(ORNL), Gaillot Stephan(CEA)
15:00 PMMCNPX Analysis of Delayed Neutron Fraction in
Beryllium Reflected Cores
S.Kalcheva

SCK-CEN, Belgium
The behavior of the effective delayed neutron and effective delayed
photo neutron fractions in beryllium reflected cores is investigated. An
MCNPX analysis of the dependence of the effective beta fractions on
the fuel depletion is performed for different fuel types.
An alternative method for estimation of the delayed neutron fraction
during the reactor operation, avoiding double criticality simulation
on prompt neutrons only (kp), is compared to the standard MCNP
approach (1 − kp ⁄ kp+d). The methodology is based on MCNP tally
calculations of fission integrals and actual delayed neutron fraction,
B km , for each fissile isotope m (235U, 239Pu, 241Pu) in a given reactor
configuration and fuel type k. The methodology is verified on a genericinfinite lattice and after that applied for the MCNP whole core modelof the BR2 reactor at different depletion steps during the reactor
operation.
The influence of the poisoning of the beryllium reflector on the effective
beta fraction is also studied. Criticality and spectra calculations taking
into account delayed neutrons and delayed photo neutrons are
performed using MCNPX 2.7.0.

21

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Mizuho_A

Track1-2 Reactor Analysis Methods
Session Chair: Richard Sanchez(CEA), Hideki Matsumoto(MHI)
15:45 PM

Application of the Efficient Consistent Spatial
Homogenization Method in Neutron Transport Theory
to a Gas Cooled Thermal Reactor Problem
S.Yasseri, F.Rahnema
Georgia Institute of Technology, Atlanta, USA
In this paper, the accuracy and computational efficiency of theefficient consistent spatial homogenization method (ECSH) in neutrontransport theory is assessed in a 1D benchmark problem characteristicof gas cooled thermal systems that are extremely challenging forconventional homogenization methods because of their longer neutron
mean free path than water-based thermal reactors. The ECSH
method is an extension of the consistent spatial homogenization
method by using: (1) B-spline instead of Fourier series for the
expansion of the spatial domain in the auxiliary cross section term and(2) a source iteration scheme instead of local fixed-source calculationsin the re-homogenization procedure. Furthermore, the effect of theangular expansion order in the definition of the auxiliary cross section
is studied. This method can be viewed as a significant improvement inaccuracy of standard homogenization methods used for VHTR wholecore analysis in which core environment effects are pronounced. It
is shown that the ECSH method can reproduce the heterogeneoustransport solution with up to 4 times faster computational speed,
depending on the configuration of the control rods while maintaining
reasonable accuracy and robust re-homogenization procedure.

16:05 PM

Application of the Hybrid Diffusion-Transport Spatial
Homogenization Method to a High Temperature Test
Reactor Benchmark Problem
G.Kooreman, F.Rahnema, S.Yasseri
Georgia Institute of Technology, Atlanta, USA
The recently developed Hybrid Diffusion-Transport Spatial
Homogenization (DTH) Method was previously tested on a benchmark
problem typical of a boiling water reactor. In this paper, the DTH
method is tested in a 1-D benchmark problem based on the JapaneseHigh Temperature Test Reactor (HTTR). This acts as a verification of
the method for a reactor that is optically thinner than the original BWR
test benchmark.

16:25 PM

Normalization Methods for Diffusion Calculations with
Various Assembly Homogenizations
C.Brosselard, H.Leroyer, M.Fliscounakis, E.Girardi, D.Couyras
EDF R&D/Sinetics, Clamart, France
When considering diffusion core calculations, a standard transportdiffusion SPH equivalence procedure is performed to preserve
infinite lattice reaction rates. Since the corresponding equivalencecoefficients are defined up to a multiplicative constant, a normalization
is applied aiming at preserving global leakage at assemblies
boundary. Supposing a linearly anisotropic flux in the transportreference calculation, preserving partial currents is equivalent topreserving boundary fluxes in a periodic lattice. This hypothesis
may be questionned when dealing with rodded or MOX assemblies
that present strong anisotropic fluxes. In this prospect, this paper
compares a new normalization method based on boundary partial
currents conservation to more familiar normalization techniquessuch as flux-volume conservation and Selengut normalization. The
comparison has been carried out on various 2D 3x3 assembly clusters
involving several assembly types (UOX, MOX, UOX-Gadolinium and
UOX-Pyrex), control rods and burn-up gradients, for 2-group diffusion
calculations performed with the new EDF calculation scheme (Gabv2Cocagne) for different homogenization models (homogeneous,heterogeneous and pin-by-pin) with respect to Apollo-2 transportreference calculations. The new current normalization has proven to
be accurate over the whole range of configurations, all the more as
the flux is anisotropic at boundary assemblies, when associated withany homogenization model.

22

16:45 PMOn the Practical Feasibility of Continuous-EnergyMonte Carlo in Spatial Homogenization
J.Leppänen(1), R.Mattila(2)1)VTT Technical Research Centre of Finland,VTT, Finland, 2)Finnish
Radiation and Nuclear Safety Authority, Helsinki, FinlandThis paper aims to evaluate the practical feasibility of using thecontinuous-energy Monte Carlo method for producing homogenizedgroup constants for deterministic core simulators. The calculations arecarried out using the Serpent 2 Monte Carlo code and ARES nodal
diffusion fuel cycle simulator. A test case from a previous validation
study is repeated with varying number of neutron histories in groupconstant generation. The impact of statistical variation in the resultsof ARES simulations is evaluated, and the corresponding calculation
times used to provide an order-of-magnitude estimate for the overall
computational cost for generating the full set of group constants
covering all state points. It is concluded that, while computationallyexpensive, Monte Carlo based spatial homogenization involving
burnup and thousands of state points per assembly type is within therange of feasibility using modern computer clusters.

17:05 PMA Dynamic Homogenization Model for Pebble Bed
Reactors
M.Grimod(1), R.Sanchez, F.Damian(2)1)Fraz. Arpuilles-Aville, Aosta, Italy, 2)CEA de Saclay, Gif-sur-Yvette,
FranceWe have used a macro stochastic model to provide a newhomogenization paradigm for pebble bed core calculations. Our
method results in an iterative core homogenization which for eachhomogenized region accounts for interactions between different
pebble types and the neighboring regions. Hence, at convergence
pebble depletion and power computation are done using the finegroup transport fluxes. We discuss the validation of the method
and give a new and accurate estimation of the critical height for the
HTR-10 international benchmark. Finally, we introduce a generalformulation for the search for the equilibrium core with a multipass fuelmanagement where the pebble recycling strategy is defined by a limitof the actual pebble burnup in the core, as opposed to a fixed numberof passes through the core. We also extend this formulation to accountfor a radial loading distribution.

17:25 PM

Homogenization of the Step Characteristic Scheme in
Phase Space
D.Anistratov, J.Jones
NC State University, North Carolina, USA
We apply homogenization methodology to formulate a transport
discretization scheme for solving k-eigenvalue problems on coarse
spatial and angular grids in 1D slab geometry. The proposed
homogenized scheme is based on the step characteristic (SC)method. It is algebraically consistent with fine-mesh SC equations. We
analyze sensitivity of this scheme to perturbations in homogenizedcross sections and other coecients of the scheme. The presented
results show numerical stability of the developed method to small
perturbations in its parameters.

17:45 PMSpatial Rehomogenization of Cross Sections andDiscontinuity Factors for Nodal Calculations
A.Dall'Osso
AREVA NP, Paris, France
Cross sections used in nodal calculations (interpolated from multiparameterized tables) come from infinite medium flux homogenization.
For highly heterogeneous compositions this average value does not
permit to represent correctly the reaction rate when the flux shapeis different with respect to the infinite medium shape. The error
induced is not negligible, for example, for assemblies with control
rod under flux gradient. Several approaches have been proposed toreduce this kind of error. A class of them, called rehomogenization,aims to determine a homogenization correction to the cross sections
corresponding to the difference between flux shapes in actual

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Mizuho_A

Track1-2 Reactor Analysis Methods
Session Chair: Richard Sanchez(CEA), Hideki Matsumoto(MHI)
environment and in infinite medium. This kind of approach neglects
the discontinuity factors, which remain at the value computed in
infinite medium condition. The purpose of the method developed
here is to extend the rehomogenization method to the discontinuity
factors. The accuracy of the method and its limitations are shown inseveral significant configurations. Its implementation in the AREVA NP
reactor core simulation code ARTEMISTM has shown better evaluationof control rod worth, which is often overestimated if no correction isapplied.

23

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Mizuho_B

SS2-2 Reactor Physics and Criticality Safety Activities in OECD/NEA Working
Party
Session Chair: Mark DeHart(INL), Teruhiko Kugo(JAEA)
15:45 PM

16:25 PM

L.Buiron, G.Rimpault, B.Fontaine (1) , T.K.Kim, N.E.Stauff,
T.A.Taiwo(2), A.Yamaji, J.Gulliford(3), E.Fridmann(4), A.Kereszturi,
I.Pataki, A.Tota(5), K.Kugo, K.Sugino, M.M.Uematsu(6), R.Lin Tan,
T.Kolowski(7), C.Parisi(8), A.Ponomarev(9)Y.K.Lee, E.Brun, X.Alexandre

Evaluation of Large 3600MWth Sodium-Cooled Fast
Reactor OECD Neutronic Benchmarks

1)CEA, Saint Paul-lez-Durance, France, 2)ANL, Argonne, USA, 3)
OECD/NEA, Issy-les-Moulineaux, France, 4)HZDR, Dresden, Germany,5)CER, Budapest, Hungary, 6)JAEA, Ibaraki, Japan, 7)UIUC, Urbana,USA, 8)ENEA, Rome, Italy, 9)KIT, Eggenstein-Leopoldshafen, Germany
Within the activities of the Working Party on Scientific Issues ofReactor Systems (WPRS) of the OECD, an international collaborationis ongoing on the neutronic analyses of several Generation-IV
Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes
the results obtained by participants from institutions of different
countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for thelarge core numerical benchmarks. These results have been obtainedusing different calculation methods and analysis tools to estimate thecore reactivity and isotopic composition evolution, neutronic feedbacks
and power distribution. For the different core concepts analyzed, asatisfactory agreement was obtained between participants despite thedifferent calculation schemes used. A good agreement was generally
obtained when comparing compositions after burnup, the delayed
neutron fraction, the Doppler coefficient, and the sodium void worth.
However, some noticeable discrepancies between the k-effective
values were observed and are explained in this paper. These aremostly due to the different neutronic libraries employed (JEFF3.1,
ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations
methods.

16:05 PM

Evaluation of Medium 1000 MWth Sodium-Cooled
Fast Reactor OECD Neutronic Benchmarks
N.E.Stauff, T.K.Kim, T.A.Taiwo(1), L.Buiron, G.Rimpault, Y.K.Lee,
E.Brun (2), A.Yamaji, J.Gulliford (3), N.Guilliard, W.Bernnat (4),M.Uematsu, K.Sugino, T.Kugo(5), I.Pataki, A.Kereszturi, A.Tota(6),A.Ponomarev(7), N.Messaoudi(8), R.Tran, T.Kozlowski(9)1)Argonne National Laboraory, Argonne, USA, 2)French Alternativeand Atomic Energy Commission, France, 3)OECD, NEA, Issy-lesMoulineaux, France, 4)Stuttgart University (IKE), Stuttgart, Germany,5)Japan Atomic Energy Agency, Ibaraki, Japan, 6)MTA EK Centre forEnergy Research (CER), Budapes, Hungary, 7)Karlsruhe Institute ofTechnology (KIT), Eggenstein-Leopoldshafen, Germany, 8)Institute ofAdvanced Nuclear Systems, SCK-CEN , Belgium, 9)University of Illinois
at Urbana-Champaign (UIUC), Urbana, USA.
Within the activities of the Working Party on Reactor and System(WPRS), an international collaboration is ongoing on the neutronic
analysis of several Generation-IV Sodium-cooled Fast Reactor (SFR)
concepts. Eleven institutions are participating in the analysis of a setof four SFR cores. Two “large” SFR core designs that were proposedby CEA are included in the set. These large SFR cores generate 3,600
MW(th) and employ oxide and carbide fuel technologies. Two “medium”
SFR core designs proposed by ANL complete the set. These medium
SFR cores generate 1,000 MW(th) and employ oxide and metallic
fuel technologies. This paper summarizes the results obtained by
the benchmark participants for the medium cores. Nine participating
institutions (ANL, CEA/Cadarache, CEA/Saclay, CER, JAEA, KIT,SCK-CEN, UIUC, and IKE) modeled these two SFR cores while using
different calculation methods and systems to estimate the k-effective,
reactivity feedbacks, isotopic composition evolution, and power
distribution. For the different core concepts analyzed, a satisfactory
agreement between participants was obtained despite the different
calculation schemes used. A good agreement is generally obtained
when comparing the burnup composition evolution, the delayed
neutron fraction, the Doppler coefficient, and the sodium void worth.
However, some noticeable discrepancies between the k-effective
values were observed and are explained in this paper. These aremostly due to the different neutronic libraries employed (JEFF3.1,
ENDF/B-VII.0 or JENDL-4.0) and in a less extend calculations
methods.

SFR Whole Core Burnup Calculations with TRIPOLI-4
Monte Carlo Code
CEA-Saclay, Gif sur Yvette Cedex, France
Under the Working Party on Scientific Issues of Reactor Systems(WPRS) of the OECD/NEA, an international collaboration benchmark
was recently established on the neutronic analysis of four Sodiumcooled Fast Reactor (SFR) concepts of the Generation- IV nuclear
energy systems. As the whole core Monte Carlo depletion calculationis one of the essential challenges of current reactor physics studies,
the continuous-energy TRIPOLI-4 Monte Carlo transport code
was firstly used in this study to perform whole core 3D neutronic
calculations for these four SFR cores. Two medium size (1000MWt) and two large size (3600 MWt) SFR of GEN-IV systems wereanalyzed. The medium size SFR concepts are from the Advanced
Burner Reactors (ABR). The large size SFR concepts are from theself-breeding reactors. The TRIPOLI-4 depletion calculations weremade with MOX and metallic U-Pu-Zr fuels for the ABR cores andwith MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The
whole core reactor physics parameters calculations were performed
for the BOEC and EOEC (Beginning and End of Equilibrium Cycle)
conditions. This paper summarizes the TRIPOLI-4 calculation resultsof Keff, βeff, sodium void worth, Doppler constant, control rod worth,and core power distributions for the BOEC and EOEC conditions.
The one-cycle depletion calculation results of the core inventory of U
and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days
depletion irradiation for the ABR cores, 410 days for the large MOX
core, and 500 days for the large carbide core.

16:45 PM

Summary and Status of OECD/NEA UAM-LWR
Benchmark
M.N.Avramova, K.N.Ivanov(1), E.Royer(2), A.Yamaji, J.Gulliford(3)1)The Pennsylvania State University, Pennsylvania, USA, 2)INSTN- CEA Saclay, Gif sur Yvette Cedex, France, 3)OECD/NEA, Issy-lesMoulineaux, France
In recent years there has been an increasing demand from nuclear
research, industry, safety and regulation for best estimate predictions tobe provided with their confidence bounds. Consequently OECD/NEA
has initiated an international Uncertainty Analysis in Modeling (UAM)
benchmark focused on uncertainty analysis in best-estimate coupledcode calculations for design, operation, and safety analysis of LWRs.
The title of this benchmark is: “Benchmarks for uncertainty analysis inmodelling (UAM) for the design, operation and safety analysis of LWRs”,
or often referred to as “OECD/NEA UAM-LWR Benchmark”. Reference
systems and scenarios for coupled code analysis are defined to study theuncertainty effects for all stages of the system calculations. Measured datafrom plant operation are available for the chosen scenarios. The generalframe of the OECD/NEA UAM LWR benchmark consists of three phaseswith different exercises for each phase. This paper summarizes thebenchmark activity and discusses the status of each phase and exercise.
Selected results are shown to illustrate the technical approach in defining
different exercises and findings from benchmark analysis. For the firsttime this benchmarks aims to determine the uncertainty in LWR system
calculations at all stages of a coupled reactor physics/thermal-hydraulics/
fuel performance calculation. The full chain of uncertainty propagationfrom basic data, engineering uncertainties, across different scales (multiscale), and physics phenomena (multi-physics) are tested on a numberof benchmark exercises for which experimental data are available and forwhich the power plant details have been released. As part of this effort,
the development and assessment of different methods or techniques
to account for the uncertainties in the calculations are investigated andreported, which contributes to the state-of-the-art in this area.

17:05 PM

Uncertainty and Sensitivity Analysis of OECD/NEA
UAM Fuel Thermal Behaviour Benchmark Using a
Falcon/URANIE Methodology
C.Cozzo, Y.Yun, O.Leray, H.Ferroukhi
Paul Scherrer Institut, Villigen, Switzerland
Within the framework of the OECD UAM Phase 2 benchmark, an
uncertainty and sensitivity analysis of the fuel thermal behavior is
performed using the Falcon code along with the URANIE platform. In a

24

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Mizuho_B

SS2-2 Reactor Physics and Criticality Safety Activities in OECD/NEA Working
Party
Session Chair: Mark DeHart(INL), Teruhiko Kugo(JAEA)
first part, focus is on an uncertainty quantification of the predicted fueltemperature pertaining to input parameters specified as manufacturing
tolerances of the fuel rod design. It is observed that the mean value of thefuel maximum temperature slightly increases with burnup and that thestandard deviation and its variation remain relatively small. In a second
part, a global sensitivity analysis method based on Sobol indices isapplied in an attempt to estimate the relative contributions of the various
input parameters to the fuel temperature uncertainty. It is found that theuncertainty is essentially dependent on the pellet-gap behavior. At lower
burnups when the gap is open, the pellet OD and cladding ID are found as
the key parameters while after gap closure, all parameters contribute in acollaborative manner to the fuel temperature.

17:25 PM

New PSI Methodology for Manufacturing andTechnological Uncertainty Quantification
M.Pecchia, A.Vasiliev, O.Leray, H.Ferroukhi(1), A.Pautz(1,2)
1)Paul Scherrer Institut, Switzerland, 2)École Polytechnique Fédérale deLausanne, Villigen, Switzerland
A new methodology, referred to as Manufacturing and Technologicalparameters Uncertainty Quantification (MTUQ), is under development
at PSI. Based on global uncertainty and sensitivity analysis methods,
MTUQ aims at advancing state-of-the-art for the treatment ofgeometrical/material uncertainties in LWR computations in general andfor the MCNPX Monte-Carlo neutron transport code in particular. For
the latter, the development is currently focused primarily on Criticality
Safety Evaluations (CSE). In that context, the key components are a
dedicated modular interface with the MCNPX code and a userfriendly
interface to model functional relationship between system variables.
A unique feature is an automatic capability to parameterize variables
belonging to so-called ‘repeated structures’ such as to allow for
perturbations of each individual element of a given system modelledwith MCNPX. Concerning the statistical analysis capabilities, these
are currently implemented through an interface with the ROOT
platform to handle the random sampling design. This paper presents
the current status of the MTUQ methodology development as well asa first assessment on the basis of an on-going OECD/NEA benchmark
dedicated to uncertainty analyses for CSE. The presented results
illustrate the overall capabilities of MTUQ and underline its relevancein predicting more realistic results compared to a methodology
previously applied at PSI for this particular benchmark.

17:45 PM

Re-Evaluation and Continued Development of
Shielding Benchmark Database SINBAD
I.A.Kodeli(1), P.Ortego(2), A.Milocco(3), G.Žerovnik(1), R.E.Grove(4),A.Yamaji(5), E.Sartori(6)1)Jožef Stefan Institute, Ljubljana, Slovenia, 2)SEA Shielding, Madrid,
Spain, 3)Universita degli Studi di Milano, Bicocca, Italy, 4)ORNL, Oak
Ridge, USA, 5)OECD/NEA, Issy-les-Moulineaux, France, 6)retired from
OECD/NEA Data Bank, Issy-les-Moulineaux, France
The Expert Group on Radiation Transport and Shielding was started
at the OECD/NEA in June 2011 with the mandate to, among others,
monitor, steer and support the continued development of the Shielding
Integral Benchmark Archive and Database (SINBAD), in cooperation
with RSICC. The present status of this database is presented, whichgrew since its beginnings in the 1990’s to an international reference
database containing at present 100 benchmarks relevant for fission,
fusion and accelerator shielding applications. As part of its activities
a thorough revision of 45 benchmark experiments was completed
recently in order to verify in details the completeness and consistencyof the benchmark information, in particular concerning the evaluationof the experimental sources of uncertainty. This review process is
expected to provide users with a proper choice and help them makebetter use of the experimental information and is planned to beextended to other available benchmarks.

25

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Hiei

SS1-2 Molten Salt Reactors
Session Chair: Imre Pazsit(Chalmers Univ. of Tech.), Ritsuo Yoshioka(International Thorium Molten-Salt Forum)
15:45 PM

Hybrid Spectrum Molten Salt Reactor

J.Krepel, K.Mikityuk(1), B.Hombourger, A.Pautz(2), M.Zanetti,
M.Aufiero, L.Luzzi(3)1)PSI Switzerland, Villigen, Switzerland, 2)PSI Switzerland / EPFL
Lausanne, Lausanne, Switzerland, 3)POLIMI, Milano, Italy
The Molten Salt Reactor (MSR) concept has a unique feature, if
compared to the majority of other reactor designs, that its fuel is liquid.
This property creates, on one hand, several technical challenges;
on the other hand, it offers flexibility in shaping and designing of theactive core. Accordingly, single fuel fluid can, for instance, circulate
through several core zones with different moderation ratios. Thispossibility was already considered in the past; however, in relation to
the thermal MSR. In the presented study, an extreme case of hybrid
spectrum MSR is proposed and preliminarily analyzed. It is concluded
that hybrid spectrum may provide several advantages and could beapplicable especially during the initial phase of the fuel cycle or by thetransition to equilibrium cycle.

16:05 PM

Thorium Conversion Optimization in Two-Fluid
Molten-Salt Reactor
J.FrybortUJV Rez / Czech Technical University in Prague, Husinec-Rez, Czech
Republic
Molten-Salt Reactors (MSR) are an attractive reactor system forvarious purposes. They can be designed to be operated in a fastneutron spectrum for spent fuel transmutation or in a thermal
spectrum. Thermal MSRs provide an ideal platformfor conversionof thorium to 233U. Flowing salt can be continuously reprocessed tominimize neutron losses due to neutron absorption in fission products.
This study deals with a static neutronic optimization of a Two-Fluid
MSR concept. Such a reactor features two separated molten-saltstreams in the reactor core. One salt contains fissile material 233U, theother thorium. Separation of these streams improves the conversion
capabilities of MSRs. Such a design was analysed for Molten-Salt
Breeder Reactor (MSBR) development. This reactor was not realized,
but it is used as a reference for this study.
Monte-Carlo code MCNP5 was used to model a simplified MSBR coreand for calculation of the breeding capabilities of this design. Several
basic geometric parameters were selected for evaluation of their effecton characteristics of the reactor. Based on this analysis, an improved
designed was prepared with shorter fissile material doubling time. The
whole analysis was carried out for fresh fuel composition. It is possible
to expect that on-line fuel reprocessing will limit fuel compositionchanges during reactor operation. Only effect of 233Pa accumulationon the thorium conversion was studied for several fuel reprocessing
rates.

16:45 PMUse of MCDancoff Factor Correction for MultiGroup Fuel Depletion Analyses of Liquid Salt Cooled
Reacotors
M.Huang, B.Petrovic
Georgia Institute of Technology, Atlanta, USA
Liquid Salt Cooled Reactors (LSCRs) are high temperature reactors,
cooled by liquid salt, with a TRISO-particle based fuel in a solid form
(stationary fuel elements or circulating fuel pebbles); this paper isfocusing on the former. In either case, due to the double heterogeneity,
core physics analysis require different considerations with more
complex approaches than LWRs core physics calculations. Additional
challenges appear when using the multi-group approach. In this paper
we examine the use of SCALE6.1.1. Double heterogeneity may be
accounted for through the Dancoff factor, however, SCALE6.1.1 does
not provide an automated method to calculate Dancoff Factors for fuelplanks with TRISO fuel particles. Therefore, depletion with continuous
energy Monte Carlo Transport (CE depletion) in SCALE6.2 beta wasused to generate MC Dancoff factors for multi-group calculations.
MCDancoff corrected multi-group depletion agrees with the results for
CE depletion within ±100 pcm, and within ±2σ. Producing MCDancoff
factors for multi-group (MG) depletion calculations is necessary
to LSCR analysis because CE depletion runtime and memory
requirements are prohibitive for routine use. MG depletion with
MCDancoff provides significantly shorter runtime and lower memory
requirements while providing results of acceptable accuracy.

17:05 PM

Comparative Studies on Plutonium and 233U Utilizationin miniFUJI MSR
A.Waris, S.Pramuditya, K.Basar(1), I.K.Aji, R.Wirawan, F.Monado,
M.N.Subkhi(2)1)Faculty of Mathematics and Natural Sciences, Institut Teknologi
Bandung, Bandung, Indonesia, 2)Institut Teknologi Bandung, Bandung,Indonesia
Molten salt reactor (MSR) has many merits such as safetyenhancement and capability to be used for hydrogen production. A
comparative evaluation of plutonium and 233U utilization in miniFUJI
MSR has been performed. Reactor grade plutonium (RGPu), weapon
grade plutonium (WGPu), and super grade plutonium (SGPu) havebeen utilized in the present study. The reactors can obtain theircriticality condition with the 233U concentration in the Th-233U fuel,
RGPu concentration in Th-RGPu fuel, WGPu concentration in ThWGPu fuel, and SGPu concentration in Th-SGPu fuel of 0.52%, 5.76%,
2.16%, and 1.96%, respectively. The Th-233U fuel results in the soft
neutron spectra of miniFUJI reactor. The neutron spectra turn into
harder with the enlarging of plutonium concentration in loaded fuelwhere Th-RGPu fuel gives the hardest neutron spectra.

16:25 PM

17:25 PM

Y.Jeong, S.Choi, D.Lee

B.Merk, J.Konheiser

Development of Computer Code Packages for Molten
Salt Reactor Core Analysis
Ulsan National Institute of Science and Technology, Ulsan, Republic of
Korea
This paper presents the implementations of the Oak Ridge National
Laboratory (ORNL) approach for Molten Salt Reactor (MSR) core
analysis with two nuclear reactor core analysis computer code
systems. The first code system has been set up with the MCNP6Monte Carlo code, its depletion module CINDER90 and the PYTHON
script language. The second code system has been set up with
the NEWT transport calculation module and ORIGEN depletion
module connected by TRITON sequence in SCALE code, and the
PYTHON script language. The PYTHON script language is used
for implementing the online reprocessing of molten-salt fuel, andfeeding new fertile material in the computer code simulations. In
this paper, simplified nuclear reactor core models of a Molten Salt
Breeder Reactor (MSBR), designed by ORNL in the 1960’s, and
FUJI-U3 designed by Toyohashi University of Technology (TUT) in the2000’s, were analyzed by the two code systems. Using these, various
reactor design parameters of the MSRs were compared, such as themultiplication factor, breeding ratio, amount of material, total feeding,neutron flux distribution, and temperature coefficient.

26

On an Optimized Neutron Shielding for an Advanced
Molten Salt Fast Reactor Design
HZDR, Dresden, Germany
The molten salt reactor technology has gained renewed interest. In
contrast to the historic molten salt reactors, the current projects are
based on designing a molten salt fast reactor. Thus the shielding
becomes significantly more challenging than in historic concepts. Onevery interesting and innovative result of the most recent EURATOM
project on molten salt reactors – EVOL – is the fluid flow optimized
design of the inner core vessel using curved blanket walls. The
developed structure leads to a very uniform flow distribution. The
design avoids all core internal structures. On the basis of this newgeometry a model for neutron physics calculation is presented
and applied for a shielding optimization. Based on these results an
optimized shielding strategy is developed for the molten salt fast
reactor to keep the fluence in the safety related outer vessel below
expected limit values. A lifetime of 80 years can be assured, but thesize of the core/blanket system has to be significantly increased
and will finally be comparable to a sodium cooled fast reactor. The
HELIOS results are verified against Monte-Carlo calculations with verysatisfactory agreement for a deep penetration problem.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Atago

SS7 Control Rod Withdrawal Tests Performed during the PHENIX End-of-Life
Experiments
Session Chair: Stefano Monti(IAEA), Shigeo Ohki(JAEA)
15:45 PM

international forum to verify the method, computer codes and crosssection data employed at IGCAR for the physics design calculations of
sodium cooled fast reactors, by keeping the experimental data as thereference.

V.Pascal, G.Prulhiere, M.Vanier, B.Fontaine(1), F.Varaine(2)16:45 PMCEA Contribution to the Analysis of the Control Rod
Withdrawal Test Performed During PHENIX End-ofLife Experiments (IAEA Common Research Program)
1)CEA, DEN, DER, SPRC/LEDC, Saint-Paul-lez-Durance, France, 2)
CEA, DEN, DER, SPRC, Saint-Paul-lez-Durance, France
In 2007 the IAEA, within the framework of the Technical Working
Group on Fast Reactors (TWG-FR) activities, decided to launch a
Coordinated Research Project (CRP), devoted to benchmarking
analyses on “Control Rod Withdrawal Test” performed during the“PHÉNIX End-of-Life Experiments”. This test program was conductedby the CEA, EDF and AREVA before the final shutdown of the
prototype power fast reactor PHÉNIX in order to gather important data
and knowledge about several aspects of the operation and safety ofpool-type sodium-cooled fast reactors. The overall CRP objective wasto improve the participants’ analytical capabilities in various fields ofresearch and design of sodium-cooled fast reactors.
Among the accident sequences that are to be taken into account,inadvertent withdrawal of a control rod is considered. During operation
at nominal power, such a sequence induces a general power increase
and local deformations of the power shape. Afterwards, the local
fuel temperature increases can thereby lead to fuel melting and clad
failure. The quasi-static control rod withdrawal test was especiallydesigned to gather power local data on fissile sub-assemblies andto complete validation databases of neutronic codes. The maximal
deformation of the power shape reached ±12% and was obtained
when two control rods were shifted in opposite directions.
After a description of the test and the measurement methods, thispaper presents some results obtained in the course of the test withspecial emphasis on control rod efficiencies and power deformation bysubassemblies. This paper also discusses CEA results obtained in thecourse of the benchmark with the European neutronic code used forfast reactors design, ERANOS-2.2.

16:05 PM

IAEA Benchmark Calculations on Control Rod
Withdrawal Test Performed During PHENIX End-ofLife Experiments - JAEA’s Calculation Results
K.Takano, T.Mouri, Y.Kishimoto, T.Hazama
Japan Atomic Energy Agency, Fukui, Japan
This paper describes details of the IAEA/CRP benchmark calculation
by JAEA on the control rod withdrawal test in the Phenix End-of-Life
Experiments.
The power distribution deviation by the control rod insertion/
withdrawal, which is the major target of the benchmark, is wellsimulated by calculation. In addition to the CRP activities, neutron
and photon heat transport effect is evaluated in the nuclear heating
calculation of the benchmark analysis. It is confirmed that the neutron
and photon heat transport effect contributes to the improvement of theabsolute power calculation results in the breeder blanket region.

16:25 PM

Benchmark Analysis of PHENIX Control Rod
Withdrawal End-of-Life Experiments
K.Devan, A.Bachchan, M.Alagan, C.Perumal
Indira Gandhi Centre for Atomic Research, Kalpakkam, India
The post-analysis of control rod withdrawal end-of-life experiments,
performed in the PHENIX reactor during June 2009, were carried out
at IGCAR as a part of IAEA Coordinated Research Project (CRP) on“Control rod withdrawal and sodium natural circulation tests performed
during the PHENIX end-of-life experiments”. The main objective of
this CRP was to assess the prediction capability of distorted radial
power distribution due to absorber rod movement (insertion and/or
withdrawal) at nominal power. In addition, core reactivity, absorber rod
worth, S-curve, maximum neutron flux, SA-wise power and sodium
heating deviation with respect to the reference core for various criticalcore configurations were analyzed. 3-D diffusion theory calculations
using FARCOB (IGCAR) and ERANOS-2.1 (European) code systems
were made for this analysis. IGCAR prediction of radial power
distribution due to absorber rod movement is very close to the measured values. No significant deviation is observed between the resultsof FARCOB and ERANOS-2.1 and also between the results of other
CRP participants. This benchmark exercise has provided a wideCalculation of the PHENIX End-of-Life Test “Control
Rod Withdrawal” with the ERANOS CodeE.Ivanov, V.Tiberi
Institut de Radioprotection et Surete Nucleaire, Fontenay-aux-Roses,
FranceThe Institute of Radioprotection and Nuclear Safety (IRSN) being
established as technical support organization for French publicauthorities is in charge of safety assessment of both operating andunder construction reactors and nuclear facilities. It provides safety
studies of advanced and innovative projects like fast sodium cooled
reactors as well. In this context, one of the IRSN objectives is toevaluate comprehensively the accuracy of numerical tools and theirperformance on studies of safety relay items. Reactor physics studies
step in the safety assessment support from different points of view,
among which the design of core and its protection system. They are
essential in the cores behavior analysis in normal, perturbed andaccidental conditions in order to assess the integrity of the first barrier
and the exclusion of prompt criticality and re-criticality risks. The
codes capability to compute in an accurate manner the fission power
distribution in the core during the whole reactor lifetime could indicate
the codes’ accuracy for many so-called spatial dependent values
calculations. The IAEA Coordinated Research Project on the Phénix
end-of-life test “Control Rod Withdrawal” has been a good opportunity
to check the capability of calculation tools by comparison with themeasured radial power distributions on fast reactor. IRSN participatedto this benchmark with the ERANOS code package developed by CEA
for fast reactors studies. The challenge for this code package was thatin the considered core configurations the neutron fields were notably
deformed. This paper presents the results obtained in the frameworkof the benchmark activity. A relatively good agreement has been foundwith available measures considering the approximations done in themodeling. The work underlines the importance of precise knowledgeof the details of burn-up distribution as it could impact the calculationsof the power distribution.

17:05 PM

Simulation of PHENIX Control Rod Withdrawal
Experiments with SIMMER-IV
V.Kriventsev, F.Gabrielli, A.Rineiski
Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen,
Germany
The “end-of-life” tests performed in the PHÉNIX reactor in 2009, inparticular the Control Rod (CR) withdrawal experiments provide an
excellent opportunity for validation and verification of the reactor
physics computer codes and modelling approaches. SIMMER-IV, amodern three-dimensional reactor safety code, has been recently
employed at KIT for simulating these experiments in the framework ofa benchmark exercise organized under the IAEA project. In this paper,
we report and discuss main results obtained with SIMMER at KIT. The
reactor reactivity, power and neutron flux distributions calculated withSIMMER-IV are in good agreement with advanced neutronics codes,
such as ERANOS, while the CR reactivity worth is overestimated due
to neglecting heterogeneity effects. We show that SIMMER neutronics
model can be improved by employing a correction that is based on
the results of cell calculations performed with ERANOS. The study
confirms that the 3D SIMMER-IV code can accurately predict major
fast reactor neutronics parameters, provided that a special treatment
is employed for CR modelling.

17:25 PM

Study of the Effect of Heterogeneity of the Control
Rods in the PHENIX Reactor
F.Gabrielli, A.Rineiski, M.Marchetti, M.Massone, V.Kriventsev
Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy
Technologies (IKET), Eggenstein-Leopoldshafen, Germany
Neutron cross-section processing for fast reactor sub-regions
containing control rods has to take into account heterogeneity effects

27

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Atago

SS7 Control Rod Withdrawal Tests Performed during the PHENIX End-of-Life
Experiments
Session Chair: Stefano Monti(IAEA), Shigeo Ohki(JAEA)
in order to get a reliable assessment of the control rod reactivity worth.
Several numerical methods have been developed in the past with asupport of experimental campaigns and are employed nowadays toreasonably treat such effects by using sophisticated neutronics codes.
The SIMMER code is employed at KIT for severe accident analysesof metal-cooled fast reactors and other reactor systems. Neutron
cross-sections are processed in the original SIMMER version byapproximating each reactor region as a homogeneous medium. This
simplified treatment results in the overestimation of the control rod
worth. Efforts are therefore going on to extend the code in order to take
into account heterogeneity effects in the control rod subassemblies. In
this paper a technique proposed in the past was applied to take into
account these effects as basis for further SIMMER extensions. With
this aim a 3D (HEX-Z) ERANOS model of the PHÉNIX reactor has
been assessed. Further, the corresponding 3D (XYZ) PARTISN model
has been employed, the latter code being now introduced in SIMMER
as a new neutronics solver. Results show that the employed technique
improves the capability of SIMMER and ERANOS codes to predictthe control rod reactivity worths. Results also show that the HEX-Z
ERANOS and XYZ PARTISN models for PHÉNIX reasonably agree.

17:45 PM

IAEA Benchmark Calculations on Control Rod
Withdrawal Test Performed during PHENIX End-of-Life
Experiments - Benchmark Results and Comparisons
V.Pascal, G.Prulhiere, M.Vanier, B.Fontaine (1) , K.Devan,
P.Chellapandi(2), V.Kriventsev(3), S.Monti(4), K.Mikityuk(5), A.
Chenu(5), M.Semenov(6), T.Taiwo(7), K.Takano(8), V.Tiberi(9)1)CEA, DEN, DER, SPRC/LEDC, Saint-Paul-lez-Durance, France,2)Indira Gandhi Centre for Atomic Research, Tamil Nadu, India, 3)
Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany,4)International Atomic Energy Agency, Vienna, Austria, 5)PSI, Villigen,
Switzerland, 6)IPPE, Kaluzhskaya Oblast, Russian, 7)Argonne National
Laboratory, Argonne, USA, 8)Japan Atomic Energy Agency, Fukui,Japan, 9)IRSN/PSN-Exp/SNC/LNR, France
In 2007, the IAEA, within the framework of the Technical Working
Group on Fast Reactors (TWG-FR) activities, decided to launch a
Coordinated Research Project (CRP) devoted to benchmark analysesof the “Control Rod Withdrawal Test performed during the PHÉNIX
End-of-Life Experiments”. This test program was conducted by the
CEA, EDF and AREVA before the final shutdown of the French
prototype power fast reactor PHÉNIX in order to gather important data
and knowledge about several aspects of the operation and safety ofpool-type sodium-cooled fast reactors. The overall CRP objective wasto improve the participants’ analytical capabilities in various fields ofresearch and design of sodium-cooled fast reactors.
Among the accident sequences that are to be taken into account, theinadvertent withdrawal of a control rod is considered. During operation
at nominal power, such a sequence induces a general power increase
and local deformations of the power shape. Afterwards, the local
fuel temperature increases can thereby lead to fuel melting and clad
failure. The quasi-static control rod withdrawal test was specificallydesigned to gather power dataof fissile sub-assemblies and tocomplete validation databases of neutronic codes. The description of
the test results are presented in reference. The maximal deformationof the power shape reached ±12% and was obtained when two control
rods were shifted in opposite directions.
After a description of the neutronic codes used in the benchmark
and their respective assumptions, this paper presents the results
obtained in the course of the benchmark as well as the comparisonswith experimental parameters with special emphasis on reactivity
estimations, control rod efficiencies, power estimations and power
deformation by sub-assemblies.

28

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Cosmos

Track11-2 Research Reactors and Spallation Sources
Session Chair: Gert Van den Eynde(SCK/CEN), Cheolho Pyeon(KURRI)
15:45 PM

16:45 PM

S.Dulla, M.Nervo, P.Ravetto

T.Sano, Y.Fujihara, J.Zhang, Y.Takahashi, K.Nakajima

Politecnico Di Torino, Torino, Italy
For the operation of subcritical source-driven systems it is very
important to develop reliable techniques to monitor the subcriticality
level, in order to be able to establish the effective distance fromcriticality and promptly detect changes in the reactivity that may be
relevant for the safety of the plant. In this paper, a new method tointerpret flux measurements in a subcritical system is presented,
generalizing a technique that has been recently proposed for sourcefree nuclear systems. The method is based on the mathematical
relationship between the power, its derivative, the convolution integral
appearing in the delayed neutron precursor balance equations andthe stable period of the multiplying system within the point kinetic
model. The assessment of the method is carried out interpreting
flux evolutions obtained from numerical transient simulations. The
results presented prove that the method can yield accurate reactivity
predictions for various physical situations and can be of interest for
accelerator-driven system technology.

Kyoto University Research Reactor Institute, Osaka, JapanThe neutron spectrum in the in-core irradiation equipments between
LEU and HEU core are compared and irradiation characteristics of theequipments in LEU core are evaluated by using the MVP2.0 with the
JENDL-4.0. Compared with the HEU core, the thermal neutron fluxin the LEU fuel is decreased by about 14% because of the thermal
neutron capture by U-238. The thermal neutron flux in the HYD and
the SSS is decreasing about 4 to 8% compared with the HEU core
because those irradiation equipments are located at the center ofcore and near the fuel region. On the other hand, since the pneumatic
tubes and the SLANT are located at the graphite reflector or theoutside of the reflector, the differences in the thermal neutron fluxes
at the irradiation equipments between the LEU core and the HEU
core are about 1 to 5 %. The difference in the thermal flux between
LEU core and HEU core becomes smaller, if an irradiation field is
positioned at a reflector region. In addition, the experimental validation
via irradiation of MoO3 pellet is carried out. As the result, the quantityof Mo-99 produced by the (n, γ) method is depend on the neutron
spectrum in the irradiation field.

A Method for Reactivity Monitoring in Subcritical
Source-Driven Systems

16:05 PM

Evaluation of Neutron Spectrum at In-Core Irradiation
Equipments in KUR with Low Enriched Uranium Fuel

Intepretation of Experimental Measurements on the
SC-1 Configuration of the VENUS-F Core

17:05 PM

S.Dulla, M.Nervo, P.Ravetto(1), S.Argiro, S.Beole, M.Masera(2),G.Mila(3), G.Bianchini, M.Carta, V.Fabrizio, V.Peluso(4), F.Gabrielli,
A.Rineiski(5), A.Kochetkov, P.Baeten, W.Uyttenhove, G.Vittiglio, J.
Wagemans(6)M.Yamanaka(1), C.Pyeon, T.Yagi, T.Misawa(2)1)Politecnico di Torino, Torino, Italy, 2)Universita di Torino / INFN, Istituto
Nazionale di Fisica Nucleare, Torino, Italy, 3)INFN, Istituto Nazionale di
Fisica Nucleare, Torino, Italy, 4)ENEA C.R. Casaccia, Galeria (Roma),
Italy, 5)Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen,
Germany, 6)Institute of Advanced Nuclear Systems, SCK CEN , Mol,
Belgium
Within the FREYA Project, methods to interpret flux measurementsin the VENUS-F core are being studied in order to reconstruct the
subcriticality level of the facility. In this work, after the presentationof results obtained with standard techniques such as the Area
Method, we introduce an alternative approach to the experimental
determination of the reactivity. This method, whose validity has
been tested by computational exercises, makes use of generalmathematical properties of the point kinetics system of equations
and has been recently extended for subcritical system analysis.
The evaluation of spatial correction factors is also carried out using
deterministic transport evaluations (ERANOS code). The statisticaland systematic uncertainty of the results in terms of reactivity is
discussed and numerical results are presented.

16:25 PM

Neutronic Characteristics of Solid Targets in
Accelerator-Driven System at Kyoto University
Critical Assembly
C.Pyeon(1), H.Nakano, M.Yamanaka(2), T.Yagi, T.Misawa(1)1)Kyoto University, Research Reactor Institute, Osaka, Japan, 2)Kyoto
University, Graduate School of Energy Science, Kyoto, JapanAt the Kyoto University Critical Assembly, a series of the acceleratordriven system (ADS) experiments are being carried out by couplingwith the fixed-field alternating gradient (FFAG) accelerator, and thespallation neutrons generated by 100 MeV protons from the FFAG
accelerator are successfully injected into the cores. In the ADS
experiments, the neutron characteristics are investigated throughthe static and kinetic analyses, when the source neutron spectrum
is varied by the kind of solid target (W, W+Be and Pb-Bi). From
the results of the experiments, the neutron yield is found to belarge in the W target, whereas the discrepancy is observed in thecomparison between the experiments and the calculations. The staticparameters are estimated to be fairly well in the analyses of neutron
multiplication and subcritical multiplications factor, with the use of
115
In(n, γ)116mIn reaction rate distribution in the core region. From thekinetic experiments, the dependence on source neutron spectrum
is significantly observed in the prompt neutron decay constant andsubcriticality, when the kind of solid target is varied.

Accuracy of Thorium-Loaded Accelerator-Driven
System Experiments at Kyoto University Critical
Assembly1)Graduate school of Energy Science, Kyoto University, Kyoto, Japan, 2)
Nuclear Science Engineering Division, Research Reactor Institute, Kyoto
University, Osaka, Japan
Thorium plate irradiation and thorium-loaded accelerator-driven
system (ADS) experiments are carried out at the Kyoto University
Critical Assembly (KUCA) to examine the accuracy of reaction rates,
subcriticality and neutron profiles on thorium capture reactionsin the critical and subcritical states for the ADS experiments with
14 MeV neutrons and 100 MeV protons. In the previous study at
KUCA, thorium capture and fission reaction rates were measured,and showed a large discrepancy between the calculations andexperiments. From the lessons learned from previous study, thethorium foil is employed in this study because the thorium fuelplate could not be suitable for the improvement of the experimental
errors caused by the effects of self-shielding, volume source, andperturbation of neutron flux from its large dimension. The difference
between the reaction rate C/E (calculation / experiment) values, withthe use of thorium foil, is within the relative difference of 10%. Here,the numerical analysis is conducted by MCNPX-2.5.0 with ENDF/
B-VII.0 and JENDL/D-99. Subcritical measurements with 14 MeV
neutrons and 100 MeV protons indicate different prompt neutron
decay constants at the same core, when the neutron source isvaried. The neutron profiles on thorium capture reactions reveal good
agreement between the calculation and the experiments.

17:25 PM

Shutdown Transients Analysis for Reflector Devices
Power Calculations in Jules Horowitz Material Testing
Reactor (JHR)
P.Console Camprini (1), M.Sumini (2), C.Gonnier, B.Pouchin,
P.Sireta, S.Bourdon(3)1)ENEA, Italy, 2)University of Bologna,Italy, 3)CEA, France
Jules Horowitz Material Testing Reactor (JHR) is planned to be thefirst European nuclear experimental facility of next decades thanks
to its testing capacity. High flux level according to 100 MW power is
exploited through many test slots. Fast core spectrum allows highdose rates for material testing and thermal neutron flux is achieved
inside a large reflector. Here fuel samples are irradiated insideexperimental devices – namely MADISON, ADELINE and MOLFI– and each specific power is then worth to be evaluated for safety
reasons. Moreover, devices transients require particular analyses forreactor shutdown conditions, in order to evaluate power behavior. All
nuclear heating effects are concerned and related time-dependent
description is carried out in this work. First, thermal hydraulic and29

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Monday September 29, 2014 - 15:45 PM - Cosmos

Track11-2 Research Reactors and Spallation Sources
Session Chair: Gert Van den Eynde(SCK/CEN), Cheolho Pyeon(KURRI)
neutronic core model is implemented through DULCINEE code toobtain core transients. Then, detailed power calculations for reflector
devices are obtained through an enhanced multi-point kinetics modelaccounting for every device which is now thought of as a single
lumped system - coupled with reactor core as external source. Coredevice coupling coefficients to define this model are finally obtainedby means of Monte Carlo simulations with TRIPOLI 4.8 code, about
different core fuel compositions – namely Beginning of Cycle (BOC),
Xenon Saturation Point (XSP), Middle of Cycle (MOC) and End of
Cycle (EOC). Complete power deposition in devices is obtained
through TRIPOLI simulations considering promp

17:45 PM

Neutronic Analysis of the PULSTAR Reactor Using
Monte Carlo Simulations
A.I.Hawari, J.L.Wormald, V.H.GilletteNorth Carolina State University, North Carolina, USA
Neutronic analysis of the PULSTAR nuclear reactor was performed insupport of its utilization and power upgrade from 1-MWth to 2-MWth.
The PULSTAR is an open pool research reactor that is currently fueledwith UO2 enriched to 4% in U-235. Detailed models were constructedof its core using the MCNP6 Monte Carlo code and its standard
nuclear data libraries. The models covered all eight variations of thecore starting with the first critical core in 1972 to the current corethat was configured in 2011. Three dimensional heterogeneousmodels were constructed that faithfully reflected the geometry of
the core and its surroundings using the original as-built engineeringdrawings. The Monte Carlo simulations benefited extensively frommeasurements that were performed upon the loading of each coreand its subsequent operation. This includes power distribution andpeaking measurements, depletion measurements (reflecting acore’s excess reactivity), and measurements of reactivity feedback
coefficients. Furthermore, to support the PULSTAR’s fuel needs,
the simulations explored the utilization of locally existing inventory offresh UO2 fuel that is enriched to 6% in U-235. The analysis shows
reasonable agreement between the results of the MCNP6 simulations
and the available measured data. In general, most discrepancies
between simulations and measurements may be attributed to thelimited knowledge of the exact conditions of the historical measurements and the procedures used to analyze the measured data.
Nonetheless, the results indicate the ability of the constructed models
to support safety analysis and licensing action in relation to the ongoing upgrades of the PULSTAR reactor.

30

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Mizuho_A

Track1-3 Reactor Analysis Methods
Session Chair: Han Gyu Joo(SNU), Wei Shen(CCSN)
8:00 AM

Research Reactor In-Core Fuel Management
Optimisation Using the Multiobjective Cross-Entropy
MethodE.B.Schlunz, P.M.Bokov(1), J.H.Van Vuuren(2)1)South African Nuclear Energy Corporation SOC Ltd (Necsa), South
Africa, 2)Stellenbosch University, South Africa
The in-core fuel management optimisation (ICFMO) problem has
been studied for several decades. Very little research has, however,been aimed at multiobjective optimisation involving the fundamental
notion of Pareto optimality. In this paper, the recently developed
multiobjective optimisation using the cross-entropy method (MOO
CEM) algorithm is applied to a multiobjective ICFMO problem for the
first time. A derivation of the MOO CEM algorithm is presented for
ICFMO, along with a constraint handling technique. The algorithm
is applied to a biobjective test problem for the SAFARI-1 nuclear
research reactor. The Pareto set approximated by the algorithm
is compared to solutions obtained by typical operational reload
strategies. The results indicate that the MOO CEM algorithm for
multiobjective ICFMO is a robust and efficient method which is able toobtain a good spread of trade-off solutions. The method may therefore
greatly aid in the decision making of a reactor operator tasked withdesigning reload configurations.

8:20 AM

Transient Cycle Fuel Management Optimization of a
Pressurized Water Reactor
T.K.Park(1), H.G.Joo, H.J.Shim, C.H.Kim(2), J Il Yoon(3)1)FNC Technology Co. Ltd., Yongin-si, Korea, 2)Seoul National
University, Seoul, Korea, 3)KEPCO Nuclear Fuel Co., Daejeon, Korea
This paper concerns with how to optimally determine enrichmentsof fuel assembly (FA) batches of beginning-of-life (BOL) and reload
cycle cores of a pressurized water reactor (PWR) plant which runs
on a multi-batch, multi-cycle fuel management scheme. As a way todetermine the optimum FA enrichments, a multi-cycle, multi-objective
FA loading pattern (LP) optimization problem for the transient cycle
cores involving the BOL and the reload cycle cores of the PWR plant
is solved by the adaptively constrained discontinuous penalty functionbased (ACDPF-based) multi-objective simulated annealing (MOSA)
algorithm in combination with the commercial core neutronics designcode ASTRA (Advanced Static and Transient Reactor Analyzer).
The applicability and the effectiveness of the ACDPF-based MOSA
algorithm is examined in terms of its solution to the first three transient
cycle FA LP optimization problem of Yonggwang Nuclear Unit 4(YGN4) a PWR plant in Korea. The practicality and usefulness of the
ACDPF-based MOSA algorithm as an optimizer to determine optimum
enrichments of BOL and reload cycle cores are discussed.

8:40 PMA Multi-Level Parallel Computation of Reactor Cores
Using GPU for Loading Pattern Optimization
T.Okubo, T.Endo, A.Yamamoto
Nagoya University, Nagoya, Japan
Efficient and rapid computation of multiple loading patterns using GPU
is studied aiming application to loading pattern optimization of LWR.
The loading pattern has significant impacts on safety and economy
of a reactor. However, design of loading pattern is a combinatorial
optimization problem, thus it is computationally intensive task. In
order to address this issue, efficient and rapid computation method ofloading patterns using massively parallel computing capability of GPU
is studied in the present paper. Though GPU has higher computational
performance than CPU, but different computational algorithm andcoding approach are necessary to maximize the performance of GPU,due to different architecture of GPU. In the present study, a multilevel parallel computing approach is examined considering hardware
architecture of GPU, i.e., parallel computing is carried out not only inspatial mesh-wise, but also in loading pattern-wise. In other words,
multiple loading patterns are simultaneously computed and domain(mesh-wise) decomposition is applied to each loading pattern.
With the present approach, computational efficiency using GPU isapproximately four times higher than that of CPU. The present core
analysis algorithm can be used for screening of poor loading patterns
in optimization process.

9:00 AM

Exact-to-Precision Generalized Perturbation Theory
for Reactor Design Calculations
C.Wang, H.S.Abdel-Khalik
North Carolina State University, North Carolina, USA
In this manuscript we have presented the application of exact-toprecision generalized perturbation theory (E PGPT) we recently
developed to the reactor design calculations. EPGPT approach can beused to reconstruct the neutron flux and the responses of interest bymeans of a Neumann series expansion. Their application to reactor
design calculations can be performed by pre-computing a set offorward fluxes and GPT fluxes, in this work the main details about how
this is done are presented. The EPGPT approach makes use of both
GPT approach and state reduction methods, i.e. proper orthogonal
decomposition or POD of snapshots, to reduce the number of model
executions and to allow one to calculate the higher-order variationsin the responses of interest due to general parameters perturbations
without having to re-execute the forward model. The computational
cost is also inexpensive as it only involves simple inner products
operations. This provides an enabling tool to analyze the impact ofparameters variations on the responses variations as many times asneeded for reactor design calculations.

9:20 AM

Depletion GPT-Free Sensitivity Analysis of the TMI
Reactor Eigenvalue Model
C.Kennedy, H.Abdel-Khalik
North Carolina State University, North Carolina, USA
For reactor design calculations, generalized perturbation theory (GPT)
has been recognized as the most computationally efficient approach toperforming sensitivity analyses of models with many input parameters,
i.e. the majority of reactor design problems during at least one
component of the design phase. Most of these design problems are
multi-component models where one or more of the codes lack GPTfunctionality. Additionally, stochastic Monte Carlo codes may not havea well-understood GPT formulation that can be implemented into thesoftware. Without GPT support, the computationally efficient approachto performing a sensitivity analyses is unavailable or computationallyinfeasible. The sensitivity analysis is a critical step in determining apreferred solution or propagating uncertainties through a series ofdesign codes.
This paper uses a fundamental adjoint algorithm, denoted GPTFree, that emulates depletion perturbation theory (DPT) without theneed to set up or solve the GPT equations. Specifically, this algorithm
constructs a reduced order model (ROM) by developing a subspacefrom the set of response sensitivity profiles formed using perturbation
theory. This algorithm determines a ROM that is independent of
the number of responses and is instead dependent on the rank orcomplexity of the underlying physical model.
This paper demonstrates the GPT-Free algorithm for depletion reactor
calculations performed in SCALE6 using a TMI 9 assembly mini-coremodel. The reactor, which is depleted to 8.1 GWd, is used to evaluate
the algorithm’s ROM reduction error with the κ-metric. The algorithms
results are then benchmarked against forward sensitivities.

9:40 AM

The "Virtual Density" Theory of Neutronics: A
Generic Method for Geometry Distortion Reactivity
Coefficients
M.Reed(1), K.Smith, B.Forget(2)1)Terra Power, Washington, USA, 2)Massachusetts Institute ofTechnology, Massachusetts, USA
Geometry distortions (bowing, flowering, and axial swelling of fuelassemblies) provide crucial negative reactivity feedback in fast
reactors. Currently, no generic, accurate, and efficient method exists
for computing distortion reactivities. Our solution is the “virtual density”
theory of neutronics, which alters material density (isotropicallyor anisotropically) instead of explicitly changing geometry. While
geometry is discretized, material densities occupy a continuous
domain; this allows density changes to obviate any re-meshingrequired for geometry changes. Although rudimentary forms of thistheory exist in Russian literature, they are only applicable to cases
in which entire cores swell uniformly. Thus, we conceive a genericand pragmatic form of “virtual density” theory to model non-uniform

31

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Mizuho_A

Track1-3 Reactor Analysis Methods
Session Chair: Han Gyu Joo(SNU), Wei Shen(CCSN)
and localized geometry distortions. We implement this theory in finite
difference diffusion for 3-D coarse mesh models of the Fast Flux Test
Facility (FFTF) and Jōyō benchmarks. We model a panoply of nonuniform anisotropic swelling scenarios, including axial swelling ofindividual assemblies, axial swelling of each mesh cell in proportion
to its fission power, and radial core flowering with arbitrary axial
dependence. In 3-D coarse mesh Cartesian cores with explicit coolant
gaps, we model assembly axial swelling and assembly row motion
with arbitrary axial dependence. We validate “virtual density” theory in
two ways: (1) by building distorted geometries in continuous energyMonte Carlo or (2) by slightly perturbing dimensions of mesh cells
in diffusion. In all examined cases, distortion reactivity coefficientsagree with Monte Carlo well within 2σ uncertainty and with diffusionreferences precisely. We culminate with the VirDenT code, which
computes reactivity insertions for arbitrary 3-D assembly distortions inrealistic fast reactor designs. VirDenT and Monte Carlo typically agree
within a few pcm.

32

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Mizuho_B

Track5-1 Nuclear Criticality Safety
Session Chair: John Bess(INL), Toshihiro Yamamoto(KURRI)
8:00 AMin the Monte Carlo code TRIPOLI4, based on the calculation of acontinuous energy adjoint flux in the reference situation, convoluted to
the forward flux of the perturbed situation.

M.S.Hodges, C.E.Sanders

9:00 AM

Nuclear Criticality Safety in the United States: Recent
Events, Trends and a Review of the Safety Culture
University of Nevada, Las Vegas, USA
If nuclear energy is to play an important role in our economies in the
future, fissile materials must be handled safely over the whole fuelcycle. Although a wealth of information is available from more than 50
years of cumulative knowledge acquired, case-specific analyses will
be needed and will dominate criticality safety. There have been no
nuclear criticality accidents in the United States since 1978, althoughseveral significant near misses have occurred. Previously, informationon these near misses has not been compiled into a single source,which has left a missing piece of the puzzle in criticality safety. To
alleviate this, these events are reviewed and a new system of criticality
accident near miss categorization has been developed. Analysis of
the information shows the nuclear criticality accident safety culture
practiced in the United States has been successful in eliminating
serious events. In the coming years there will likely be further
clarification of potential nuclear fuel cycle strategies, each one with its
specific needs in criticality safety. Criticality safety calls for constantsupport and attention. A sound understanding and correct applicationof the principles of nuclear criticality safety are vital to the nuclear
industry. The objective is to pursue an accidentfree goal, while keeping
in mind the repercussions that an avoidable criticality excursion could
have. Diversity in national policies has existed and is likely to exist
into the near future. Therefore, it is worth sharing criticality information
beyond national policies.

Criticality Calculation of Fuel Debris in Fukushima
Daiichi Nuclear Power Station
A.Tsuchiya, T.Kondo, H.Maruyama(1), M.Yamaoka, R.Kimura, Y.
Moriki, Y.Hayashi, Y.Takeuchi(2), K.Yamaji, M.Nakano(3), K.Oyama,
A.Takagi(4)1)Hitachi-GE Nuclear Energy, Ltd., Ibaraki, Japan, 2)Toshiba
Corporation, Kanagawa, Japan, 3)Mitsubishi Heavy Industry, Ltd.,
Hyogo, Japan, 4)Tokyo Electric Power Company, Tokyo, Japan
In the Fukushima-Daiichi Nuclear Power Station units 1, 2 and 3,
we have estimated that the nuclear fuel bundles have become nonuniform fuel debris, which could not remain at the normal core locationin the reactor pressure vessel (RPV), but could have fallen to the RPV
bottom head and accumulated in the pedestal floor. At the present, thefuel debris is estimated to be sub-critical by measurements of short life
time fission product (FP) gas and water temperature. In the phase of
introducing small circulation loop or fuel debris removal, the part of fueldebris may flow into the liquid waste treatment and cooling facilities.
Thus, we have studied the criticality scenarios on each location wherefuel debris is estimated to exist. The criticality evaluation based on thecriticality scenarios has been performed with conservative conditions,because the debris properties are unidentified. The criticality
evaluation shows the range of condition for criticality.

8:20 AM

9:20 AM

F.Michel-Sendis, M.Bossant, N.Soppera(1), I.Gauld(2)A.Ranta-aho

A New OECD/NEA Database of Nuclide Compositionsof Spent Nuclear Fuel
1)OECD Nuclear Energy Agency, Issy-Les-Moulineaux, France, 2)Oak
Ridge National Laboratory, Tennessee, USA
The SFCOMPO database of nuclide compositions of spent nuclear
fuel is hosted by the OECD Nuclear Energy Agency since 2001.
Since 2011, a collaborative effort led by the OECD/NEA Data Bankand Oak Ridge National Laboratory, under the guidance of the NEA
Expert Group on Assay Data of Spent Nuclear Fuel, has resulted in
the creation of a new enhanced relational database structure and asignificant expansion of SFCOMPO, now containing experimental
assay data for a wide selection of international reactor designs. This
paper aims at describing the new SFCOMPO Database developed at
NEA in terms of functionalities, contents and foreseen developments.
This new database is expected for public release in 2014.

8:40 AMOECD EGBUC Benchmark VIII - Comparison of
Calculation Codes and Methods for the Analysis of
Small-Sample Reactivity Experiments
P.Leconte, A.Santamarina
CEA Cadarache, Saint Paul-Lez-Durance, France
Small-sample reactivity experiments are relevant to provide accurate
information on the integral cross sections of materials. One of thespecificities of these experiments is that the measured reactivity worth
generally ranges between 1 and 10 pcm, which precludes the use ofMonte Carlo for the analysis. As a consequence, several papers havebeen devoted to deterministic calculation routes, implying spatial and/or energetic discretization which could involve calculation bias. Within
the Expert Group on Burn-Up Credit of the OECD/NEA, a benchmark
was proposed to compare different calculation codes and methods forthe analysis of these experiments. In four Sub-Phases with geometries
ranging from a single cell to a full 3D core model, participants wereasked to evaluate the reactivity worth due to the addition of small
quantities of separated fission products and actinides into a UO2 fuel.
Fourteen institutes using six different codes have participated in the
Benchmark. For reactivity worth of more than a few tens of pcm, theMonte-Carlo approach based on the eigen-value difference method
appears clearly as the reference method. However, in the case ofreactivity worth as low as 1 pcm, it is concluded that the deterministic
approach based on the exact perturbation formalism is more accurate
and should be preferred. Promising results have also been reported
using the newly available exact perturbation capability, developed

Design of an Efficient Calculation Model of BWR Cold
Critical Experiments for Validation
Teollisuuden Voima Oyj, Eurajoki, Finland
The term burnup credit is used when the calculated spent fuelcomposition is credited in the criticality safety analysis as opposed to
the fresh fuel assumption. Applicable standards place requirements
for the validation of the burnup codes that are used in the analysis.
Unfortunately, there is a lack of high quality BWR radiochemical assay
data suitable for validation. In order to circumvent this difficulty, BWR
cold critical experiments could be used for the validation.
A disadvantage in the use of reactor measurements is the numberof detail that needs to be fed into the calculation model. An accurate
modelling would require thousands of assembly burnup calculations
and setting up a core model with hundreds of thousands of fuelmaterial compositions and different control rod designs present in thecore. Clearly, a simplified approach would be very valuable for themodelling of cold critical experiments with Monte Carlo codes.
A simplified way of modelling BWR cold critical experiment has been
considered in this work. In this approach, only the most relevant partof the core is described in a detailed manner and suitable boundary
conditions are applied in other parts of the core by replacing theassembly and control rod data with representative designs.
In this work BWR cold critical measurements of Olkiluoto 1 andOlkiluoto 2 units were used to demonstrate the quality of the approach.
SIMULATE-3 calculations were made in order to compare different
calculation models for 58 cold critical experiments. The results showthat the simplified core model with suitable boundary conditions isrobust, accurate and neutronically equivalent with the detailed model.
The results suggest that instead of modelling all 500 assemblies in
the core including nodal burnup calculations with a depletion code,
only 48 assemblies need to be considered. Futhermore, instead ofmodelling all control rod types in the core, considering one or two
rod designs is sufficient for validation purposes. This means thatthe computational effort can be reduced by more than an order ofmagnitude without significant effect on the final results. For the same
reason the work needed to prepare benchmarks from the experiments
is reduced significantly. The procedure can be looked upon as being a
conservative calculation model for criticality code validation or a way
to create simplified benchmarks from cold critical experiments.

33

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Mizuho_B

Track5-1 Nuclear Criticality Safety
Session Chair: John Bess(INL), Toshihiro Yamamoto(KURRI)
9:40 AMFirst Burnup Credit Application for Transport andStorage Cask Using French Experiments

M.Tardy(1), S.Kitsos, G.Grassi(2), A.Santamarina, L.S.Felice,
C.Riffard(3)1)Areva, Montigny-le-Bretonneux, France, 2)AREVA NC, Paris La
Defence, France, 3)CEA/DEN/DER/SPRC, Saint Paul-Lez-Durance,
FranceThe burnup credit (BUC) methodology for a transport and storage
cask application including actinides and fission products is
implemented at AREVA TN using the French BUC calculation routefor PWR UO2 used fuel. The methodology is based on the connectionof the French depletion code DARWIN2 and the French Criticality
Safety Package CRISTAL V1. The BUC methodology includes theexperimental validation of the computation codes dedicated to thecalculation of the used fuel inventory calculations. Indeed, the resultsof the comparison Calculation-Experiment (C-E)/E allow to determine
either a set of isotopic correction factors (ICF) for the BUC nuclides
considered in the criticality calculation or keff-penalty terms directly
used for the definition of the keff-acceptance criterion for the criticality
assessment of the transport and storage cask. These ICFs or keffpenalty terms are one of the key of the BUC method to guarantee theconservativeness of the fuel reactivity in safety-criticality calculations
using BUC approach.
A French BUC program has been developed at CEA/Cadarache in
the framework of the CEA-AREVA collaboration in order to validate
fuel inventory calculations. This program involves two kinds ofexperiments: chemical analyses and microprobe measurements of
PWR irradiated fuel pins (French P.I.E. program) on one hand; andreactivity worth measurements of the BUC nuclides in the MINERVE
reactor on the second hand.
This paper highlights, through a first industrial AREVA TN's applicationof the BUC method including fission products, that the French P.I.E.
program and reactivity worth measurements in MINERVE reactor
are suitable for the implementation of burnup credit in transportand storage cask applications loaded with PWR-UOX used fuelsassemblies.

34

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Hiei

Track2-1 Deterministic Transport Theory
Session Chair: Nam Zin Cho(KAIST), Masato Tabuchi(NEL)
8:00 AM

A Collision Probability Based Method to Compute
Cross Sections Sensitivities for the Subgroup SelfShielding Technique
M.Dion, G.Marleau
École Polytechnique de Montréal, Montréal, Canada
We present a method to evaluate self-shielded cross sections
sensitivity to light isotopes densities. Our method is based on a
subgroup self-shielding model, using physical probability tables. We
propose a simplified way to evaluate the derivatives of the collision
probabilities, and show these can be used to compute the derivativesof the weight function required to average resonant cross sections
in a multigroup scheme. We present self-shielded cross sections
sensitivity coefficients based on a variation of the hydrogen density fora relatively simple lattice, and we compare with reference values using
complete calculations.

8:20 AM

Improvement of a Convergence Technique for MOC
Calculation with Large Negative Self-Scattering Cross
Section
M.Tabuchi, M.Tatsumi(1), A.Yamamoto, T.Endo(2)1)Nuclear Engineering, Ltd., Osaka, Japan, 2)Nagoya University,
Nagoya, Japan
In the method of characteristics, convergence behavior of inner
iteration would be degraded by the transport correction. A previous
study revealed that large negative self-scattering cross section due
to the transport correction worsens convergence of inner iteration. In
order to resolve this issue, a technique to improve the convergence
behavior based on the successive over relaxation (SOR) method was
proposed in the previous study. However, considerable calculation
time is necessary to estimate the optimum acceleration factor for the
SOR method especially in large scale geometries. In order to address
this drawback, a fast estimation scheme of the acceleration factors is
proposed in this study. The present approach can provide acceleration
factors which ensure convergence of iterative calculation. In thepresent approach, the equation to obtain the appropriate acceleration
factor is derived by using Gerschgorin’s theorem. Based on the equation, the acceleration factors can be easily obtained in each energy
group with negligible calculation time. Verification calculations are
performed in various critical experiments, in which degradation of
convergence by the transport correction is significant. When thetransport correction is used without the SOR method, all iterative
calculations diverged. On the other hand, when the SOR method withthe acceleration factors based on the present approach is applied,
convergent solutions are successfully obtained in all cases. Fromthese results, validity of the present approach is confirmed.

8:40 AM

Boundary Acceleration Techniques for CMFDAccelerated 2D-MOC
S.G.Stimpson, B.S.Collins, B.M.Kochunas, T.J.Downar
University of Michigan, Michigan, USA
Coarse mesh finite difference (CMFD) has been used successfully foraccelerating the eigenvalue and scalar fluxes for transport problem
solutions. In the conventional use of CMFD to accelerate transport
solution methods, it provides homogenized scalar fluxes on the pin- or
assembly-basis and drives the eigenvalue calculation. The transport
sweeper provides currents at coarse mesh boundaries which areused to inform the CMFD current coupling coefficients. In this paper
we present three methods that use the CMFD solutions to modify the
incoming flux boundary condition of the 2D-MOC sweeps that are theprimary driver of the transport solution. The first technique (P0) uses
a ratio of the scalar surface fluxes to scale the incoming angular flux.
This is the most basic technique, since all incoming angular fluxes are
scaled isotropically. The second technique (DP0) uses a ratio of thepartial currents to scale the incoming angular flux, so the update isangularly dependent in half-space. The third technique (P1) uses theP1 approximation of the angular flux to compute the scaling factor forthe angular flux. This requires both the scalar flux and net current oneach coarse mesh surface. These techniques can be applied to both
system and parallel boundaries, which will reduce the additional outer
iterations incurred in certain spatially decomposed parallel transport
sweep algorithms. Results for various parallel configurations are alsopresented for the C5G7 application and 2D quarter core slice AMA
Problem 5.

9:00 AM

A Low Order Nonlinear Transport Acceleration
Scheme for the Method of Characteristics
L.Li, K.Smith, B.Forget(1), R.Ferrer(2)1)Massachusetts Institute of Technology, Massachusetts, USA, 2)
Studsvik Scandpower, Inc., Idaho Falls, USA
This study presents a new physics-based multi-grid nonlinear
acceleration method, henceforth referred to as the Low-Order Operator(LOO) method, which uses a coarse space-angle multi-group Methodof Characteristics (MOC) neutron transport calculation to accelerate
the fine space-angle MOC calculation. LOO is designed to capture
more angular effects than diffusion-based acceleration methods
through a transport-based low-order solver. LOO differs from the two
existing transport-based acceleration schemes in that it emphasizes
simplified coarse space-angle characteristics and preserves physics
in quadrant phase-space. The details of the method, including the
restriction step, the low-order iterative solver, and the prolongation
step are discussed in this work. The performance of LOO is comparedto the Coarse-Mesh Finite Difference method (CMFD) in benchmark
problems. LOO shows comparable convergence compared to CMFD,
and does not require under-relaxation to converge, making it a robust
acceleration scheme.

9:20 AM

p-CMFD Acceleration and Nonoverlapping Local/
Global Iterative Transport Methods with 2-D/1-D
Fusion Kernel
S.Yuk, N.Z.Cho
Korea Advanced Institute of Science and Technology, Daejeon, Korea
Conventional reactor core analysis, based on isolated single-assembly
lattice calculation and diffusion nodal core calculation, has thelimitation due to the absence of treatment of accurate inter-assemblytransport effect. We consider two approaches in this paper: 1) wholecore fine-group deterministic transport calculations are accelerated
by the partial-current based coarse-mesh (in space, angle, andenergy) finite-difference (p-CMFD) method, and 2) whole-core domain
is decomposed into nonoverlapping local problems and the local
problem transport solutions are wrapped around by the p-CMFD
methodology in a two-level iterative scheme to provide whole-coretransport solution. In both approaches, the 2-D/1-D fusion method isused as solution kernel, which employs two-dimensional method ofcharacteristics (MOC) in radial direction and one-dimensional SN-like
method in axial direction. We report results of the two approaches
applied to the three whole-core configurations of the C5G7 OECD/
NEA three-dimensional benchmark problem.

9:40 AM

Application of the SDD-CMFD Acceleration Method toParallel 3-D MOC Transport
B.Kochunas, B.W.Kelley, S.G.Stimpson, E.Larsen,
T.J.Downar
University of Michigan, Michigan, USA
In this paper the spatial domain decomposed coarse mesh finite
difference (SDD-CMFD) method is applied as an acceleration
technique to a parallel implementation of the 3-D method ofcharacteristics (MOC) for a series of problems to assess theeffectiveness of the method for practical applications. The SDD-CMFD
method assumes the problem domain is divided into independent
parallelizable sweep regions globally linked within the framework ofa CMFD-like system. Results obtained with the MPACT code are
examined for three problems. The first analysis is of multi-dimensional,
1-group, infinite homogeneous media problems that compare thenumerically-measured rate of convergence to that predicted by the
1-D Fourier analysis performed in previous work. It is observed thatthe rate of convergence of the numerical experiments has similar
behavior to that predicted by the Fourier analysis for variations of
optical thickness in the coarse cell and spatial sub-domain. However,
the rate of convergence is measured to be slightly less than that predicted by Fourier analysis. The algorithm is applied to the Takeda 3-D
neutron transport benchmark, and compared to a standard source

35

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Hiei

Track2-1 Deterministic Transport Theory
Session Chair: Nam Zin Cho(KAIST), Masato Tabuchi(NEL)
iteration. In the analysis of this problem, the method is observed to
speed up convergence, significantly reducing the number of outer
iterations by a factor of nearly 20x and reducing the overall run time bya factor of about 10x. Finally, the method is applied to a realistic PWR
assembly, which is observed to converge in 7 outer iterations, a factorof 150x less than source iteration, using the SDD-CMFD acceleration
method, and have an estimated speedup of ~34x over conventionalsource iteration.

36

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Atago

Track3-1 Monte Carlo Methods
Session Chair: Jaakko Leppänen(VTT), Takanori Kitada(Osaka Univ.)
8:00 AM

Unstructured Mesh Based Multi-Physics Interface
for CFD Code Coupling in the Serpent 2 Monte Carlo
Code
J.Leppänen, V.Valtavirta, T.Viitanen(1), M.Aufiero(2)1)VTT Technical Research Centre of Finland, Espoo, Finland, 2)
Politecnico di Milano, Milano, Italy
This paper presents an unstructured mesh based multi-physics
interface implemented in the Serpent 2 Monte Carlo code, for the
purpose of coupling the neutronics solution to componentscale thermal
hydraulics calculations, such as computational fluid dynamics (CFD).
The work continues the development of a multi-physics coupling
scheme, which relies on the separation of state-point informationfrom the geometry input, and the capability to handle temperatureand density distributions by a rejection sampling algorithm. The newinterface type is demonstrated by a simplified molten-salt reactor test
case, using a thermal hydraulics solution provided by the CFD solver
in OpenFOAM.

8:20 AM

Analysing the Statistics of Group Constants
Generated by Serpent 2 Monte Carlo CodeT.Kaltiaisenaho, J.Leppänen
VTT Technical Research Centre of Finland, Espoo, Finland
An important topic in Monte Carlo neutron transport calculations is
to verify that the statistics of the calculated estimates are correct.
Undersampling, non-converged fission source distribution and intercycle correlations may result in inaccurate results. In this paper, westudy the effect of the number of neutron histories on the distributionsof homogenized group constants and assembly discontinuity factors
generated using Serpent 2 Monte Carlo code. We apply two normality
tests and a so-called “drift-in-mean” test to the batch-wise distributionsof selected parameters generated for two assembly types taken from
the MIT BEAVRS benchmark. The results imply that in the tested
cases the batch-wise estimates of the studied group constants can be
regarded as normally distributed. We also show that undersampling is
an issue with the calculated assembly discontinuity factors when thenumber of neutron histories is small.

8:40 PM

Theoretical Prediction on Underestimation of
Statistical Uncertainty for Fission Rate Tally in Monte
Carlo Calculation
T.Endo, A.Yamamoto, K.Sakata
Nagoya University, Nagoya, JapanA theoretical model to predict underestimation of standard deviation forthe mean of fission rate in each tally region is proposed on the basisof the behavior of higher order modes in fission source distribution
and the autoregressive (AR) model. The predicted underestimationof standard deviation for the mean of fission rate is compared withthat obtained by actual Monte Carlo calculations and they agree welleach other. Dependency of underestimation of standard deviation onnumber of tally regions and spatial position is clarified by the proposed
theoretical model. The present theoretical model can be used toquantitatively predict underestimation of standard deviation of local
fission rate tally.

9:00 AM

Analysis of Tally Correlation in Large Light Water
Reactors
B.R.Herman, B.Forget, K.Smith (1), P.K.Romano, T.M.Sutton,
D.J.Kelly, III, B.N.Aviles(2)1)Massachusetts Institute of Technology, Massachusetts, USA, 2)Knolls
Atomic Power Laboratory - Bechtel Marine Propulsion Corporation, New
York, USAA study of tally correlation was performed for a 2-D radial slice of the
BEAVRS light water reactor. This reactor has a dominance ratio of
approximately 0.995. Convergence rates of fission source tallies of anassembly and pin mesh were analyzed. Results of this study indicatethat these tallies do not converge at the ideal rate of the inverse of thesquare root of number of tally realizations. Ideal convergence rates

are achieved only when tally realizations are uncorrelated. Correlation
exists because source sites of one generation of neutrons are fromfission sites produced from a previous generation. The degree ofcorrelation was investigated by computing autocorrelation coefficients.
High autocorrelation coefficients of about 0.7 for lag one wereobserved for assembly mesh tallies. Pin mesh tallies were much lesscorrelated. In addition, autocorrelation coefficients were not affectedby increasing number of neutrons in a tally batch. Three methods are
discussed to reduce correlation: increasing number of initial discard
batches, multiple generations per tally batch and CMFD feedback. The
only significant decrease in correlation was observed when applying
CMFD feedback. Lag one autocorrelation coefficients decreased fromabout 0.7 to 0.4. No significant difference in tally convergence rates
were observed when applying CMFD feedback.

9:20 AM

Higher-Mode Applications of Fission Matrix Capability
for MCNP
S.E.Carney(1), F.B.Brown, B.C.Kiedrowski(2), W.R.Martin(1)1)University of Michigan, Ann Arbor, USA, 2)Los Alamos National
Laboratory, NM, USA
The fission matrix method, implemented into the MCNP6 Monte Carlocode, can be used to provide estimates of the fundamental modefission distribution, the dominance ratio, the eigenvalue spectrum, andhigher mode forward and adjoint eigenfunctions of the fission neutron
source distribution. It can also be used to accelerate the convergenceof the power method iterations. The higher-mode fission sources canbe used in MCNP6 to determine higher mode fluxes and tallies, bothforward and adjoint. These higher mode fluxes are necessary forimportant applications such as second-order perturbation theory andquasi-static calculations. The flux calculations are theoretically justified
here, along with a justification of eigenmode expansions during source
convergence. Forward fluxes and the relative uncertainties for a 2D
PWR are shown, both of which qualitatively agree with expectations.
Adjoint-weighted flux for a 3-group slab problem is calculated andfound to agree with discrete ordinates results. Lastly, eigenmodeexpansions are performed during source convergence from two
intial distributions for the 2D PWR problem; observed decay rates ofcoefficients agree closely with expectation.

9:40 AMA Symmetric View Hiding the Ugly Truth
D.Mennerdahl
E Mennerdahl Systems, Täby, Sweden
The general recommendation to use symmetry in Monte Carlo codekeff calculations is questioned. A solution using a symmetric model
provides less information and may hide serious errors that would
be apparent when using a full model calculation. Convergenceindicators such as dominance ratio and Shannon entropy haveappropriate applications but they do not always give correct andreliable information about symmetric model calculations. Further, themethods do not see the symmetry in a full model, apparent to the userof the code. The user often accounts for symmetry in a full model byhaving the same material or geometry unit in the symmetric regions. If
average tallies are requested, they can be obtained from the full model
results. The integral values are the same but the differences between
local reactions in symmetric regions are informative.
The symmetry recommendation, together with false convergenceindicators, encourages the user to believe that the symmetric model
calculations are more reliable than full model calculations, when thesame Monte Carlo statistics are applied. Examples are given in thepaper, demonstrating the points made. The most basic example
of symmetry is a calculation model for an infinite material. Anotherexample is a cylinder with azimuthal symmetry. In both cases it isclear that more symmetry is not better. This is demonstrated for cubeswith different dimensions and mirror reflection all around. Applicationof symmetry is necessary for a simulation of such geometry due tocomputer memory storage limitations (less than an infinite value).
Better convergence is not a relevant point.

37

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Cosmos

SS8 Reactor Physics of Non-Traditional LWR Fuel Design
Session Chair: Bojan Petrovic(Georgia Tech.), Yoichiro Shimazu(Univ. of Fukui)
8:00 AM

I2S-LWR Equilibrium Cycle Core Analysis
D.Salazar, F.Franceschini(1), B.Petrovic(2)1)Westinghouse, Pennsylvania, USA, 2)Georgia Tech., Atlanta, USA
This paper presents the preliminary core design developed for the
Integral Inherently Safe LWR (I2S-LWR). The I2S-LWR is a conceptual
design of a ~1,000 MWe (2,850 MWt) integral PWR with inherent
safety features. The baseline core configuration contains 121 fuelassemblies with a 19×19 square lattice with 144-in active fuel height.
The reference fuel option is U3Si2 in advanced stainless steel cladding,which is envisioned to enhance accident tolerance compared tothe current UO2/Zr system. The viability of UO2 fuel is also underinvestigation as an alternative option for accelerated deployment. A
12-month cycle 3-batch fuel management scheme has been foundto be the economic optimum for the I2S-LWR with the current 235U
licensed enrichment. The U3Si2 and UO2 core designs implementing
this option are presented in this paper together with a comparison of
their reactor physics performance at the equilibrium cycle.

8:20 AM

Uranium Nitride Composite Fuels in a Pressurized
Water Reactor: Exploration of Multi-Batch Cycle
Length and UB4 Admixture for Reactivity Control
N.R.Brown, M.Todosow, K.J.Mcclellan
Brookhaven National Laboratory, New York, USA
In the aftermath of the Fukushima accident, uranium nitride (UN)-based
fuel composites have been presented as potential light water reactor
fuels with enhanced accident tolerance. Enhanced accident tolerance
implies improved accident performance versus the present UO2Zr fuel system as well as similar or improved nominal performance
under operational conditions. UN-based composite fuels consist ofa primary UN phase that is “shielded” from water via a secondary
phase. This allows the potential benefits of the UN fuel, including
enhanced thermal conductivity, to be realized while significantly
reducing hydrolysis of the nitride phase. We compared performanceof several nitride composites in a PWR. Performance parameters
included quantities like cycle length for three-, four-, and five-batch
fuel management schemes as well as reactivity coefficients. Althougha nitride composite fuel would most likely be paired with an advanced
cladding, our analyses assume Zr-based cladding. However, manyof the qualitative comparisons from our analyses have relevancefor proposed advanced cladding materials, especially those that are
relatively neutron transparent (e.g. silicon carbide). We studied UN/
U3Si5, UN/U3Si2, UN/UB4, UN/ZrO2, and UN/U3Si2/UB4. In all cases
UN is the primary phase, with small fractions of U3Si5, U3Si2, UB4, or
ZrO2 as a secondary phase. For the UN/UB4 cases, the boron content
is 100% enriched in 11B, but for UN/U3Si2/UB4 cases natural boron isused for reactivity control. This paper presents a summary of some of
the key results and an overview of the research philosophy. In general,
the full-core equilibrium cycle reactivity coefficients for the nitride andnitride composite fuels were within the design limits for the reference
PWR. Increasing the number of batches appears to be a viable
alternative to increasing the cycle length relative to the nominal UO2Zr fuel system. We found that the reactivity suppression of integral fuelburnable absorber rods could be effectively reproduced using small
fractions of UB4 admixture homogeneously distributed in the fuel.

8:40 AM

Impact of Coating on Nitride Fuel Performance in
PWRs
F.Heidet, A.M.Yacout
Argonne National Laboratory, Argonne, USA
In order for uranium nitride fuel to be used in light water reactors
and benefit from its high thermal conductivity and high density, it isnecessary to develop a mitigation strategy to prevent uranium nitrideoxidation with water/steam at high temperatures. One possible
strategy is the use of a protective layer between the fuel and cladding(coating fuel pellet), preventing the water/steam to contact with thefuel. Another strategy consists in using thin-layer coating applied
to uranium nitride fuel particles before the fuel sintering step. It is
important for the materials used not to have a detrimental effect on
the neutronics performance of the fuel. The impact on the neutronics
performance of using Al2O3 coatings has been assessed in this work.
The main effect is the reduction of the heavy metal mass, as compared38

to pure uranium nitride fuel, which results in slightly softer spectrum.
However, when compared with uranium oxide fuel which has a similar
heavy metal mass, the achievable cycle length and discharge burnup
are found to be nearly identical since para- sitic absorption in Al2O3 isrelatively small. Further development and application of Al2O3 coatings(and possibly other materials) can ultimately lead to the deployment of
uranium nitride fuel in light water reactors.

9:00 AM

Optimization of Fully Ceramic Micro-Encapsulated
Fuel Assembly for PWR
R.A.Shapiro, M.J.Vincenzi(1), M.Fratoni(2)1)The Pennsylvania State University, Pennsylvania, USA, 2)University ofCalifornia, Berkeley, USA
Fully ceramic micro-encapsulated (FCM) fuels consist of TRISO
particles embedded in a ceramic (SiC) matrix to form fuel pellets androds, and are expected to largely improve fuel performance in lightwater reactors during beyond design basis accident conditions. The
viability of FCM fuel is hindered by a considerable lower heavy metalload compared to conventional fuel. This deficiency is compensatedusing high-density fuel (UN), large kernels, large fuel rods, and higher
enrichment. This study evaluated the assembly design requirements
for FCM fuel to match the neutronics performance of conventional fuel,in particular to reach similar cycle length while major safety criteria—
reactivity coefficients, and control system worth, are met. It was foundthat with a traditional 1717 assembly FCM fuel must be enriched close
to 20% in order to achieve similar cycle length as conventional fuel,but an optimization of the assembly design (99) could reduce such
enrichment to about 16%.

9:20 AM

Fully Ceramic Microencapsulated Fuels:
Characteristics and Potential LWR Applications
J.J.Powers, A.Worrall, K.A.Terrani, J.C.Gehin, L.L.Snead
Oak Ridge National Laboratory, Oak Ridge, USA
This paper summarizes the characteristics of the fully ceramic
microencapsulated (FCM) fuel concept and two potential light water
reactor (LWR) applications of FCM fuels: for actinide management
and as an accident tolerant fuel (ATF). Recent progress in FCM fueldevelopment includes production of uranium mononitride kernels,
fabrication of FCM pellets and pins, and irradiation testing of matrix
samples and FCM pellets. Potential applications of FCM fuel in
LWRs appear promising based upon studies performed by several
organizations; however, further efforts are needed to investigate
various design aspects in further detail and explore promising newareas of research such as new fuel pin and assembly designs
or alternate materials of interest. Current challenges in FCM fueldevelopment and LWR applications for FCM fuels include low heavy
metal fuel loading densities and increased uncertainties in analysis
due to several different factors. Overall, LWR FCM concepts appear
feasible for both actinide management and as an ATF.

9:40 AM

Neutronic Challenges of Advanced Boiling Water
Reactor Designs
K.Shirvan, M.Kazimi
MIT, Massachusetts, USA
The advancement of Boiling Water Reactor technology has been underinvestigation at the Center for Advance Nuclear Energy Systems at
MIT. The advanced concepts under study provide economic incentives
through enabling further power uprates (i.e. increasing vessel power
density) or better fuel cycle uranium utilization. The challenges inmodeling of three advanced concepts with focus on neutronics arepresented. First, the Helical Cruciform Fuel rod has been used insome Russian reactors, and studied at MIT for uprating the power in
LWRs through increased heat transfer area per unit core volume. The
HCF design requires high fidelity 3D tools to assess its reactor physics
behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher
power density over existing BWRs, while using current licensing tools
and existing technology. Its larger assembly size requires stronger
coupling between neutronics and thermal hydraulics compared tothe current practice. Third is the reduced moderation BWRs, which

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 8:00 AM - Cosmos

SS8 Reactor Physics of Non-Traditional LWR Fuel Design
Session Chair: Bojan Petrovic(Georgia Tech.), Yoichiro Shimazu(Univ. of Fukui)
had been proposed in Japan to enable breeding and burning of fuelas an alternative to sodium fast reactors. Such technology suffersfrom stronger sensitivity of its neutronics to the void fraction than the
traditional BWRs, thus requiring exact modeling of the core conditions
such as bypass voiding, to correctly characterize its performance.

39

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 10:20 AM - Mizuho_A

Track1-4 Reactor Analysis Methods
Session Chair: Jess Gehin(ORNL), Scott Palmtag(Core Physics Inc.)
10:20 AM

11:20 AM

J.C.Gehin, A.T.Godfrey, T.M.Evans, S.P.Hamilton(1),F.Francheschini(2)M.Ryu, Y.S.Jung, H.H.Cho, H.G.Joo

Watts Bar Unit 1 Cycle 1 Zero Power Physics Tests
Analysis with VERA-CS
1)Oak Ridge National Laboratory, Oak Ridge, USA, 2)Westinghouse
Electric Corporation, Pennsylvania, USA
The Consortium for Advanced Simulation of Light Water Reactors(CASL) is developing a collection of methods and software products
known as VERA, the Virtual Environment for Reactor Applications,
including a core simulation capability called VERA-CS. A key milestone
for this endeavor is to validate VERA against measurements fromoperating nuclear power reactors. The first step in validation against
plant data is to determine the ability of VERA to accurately simulate
the initial startup physics tests for Watts Bar Nuclear Power Station,
Unit 1 (WBN1) cycle 1. VERA-CS calculations were performed with
the Insilico code developed at Oak Ridge National Laboratory using
cross section processing from the SCALE system and the transport
capabilities within the Denovo transport code using the SPN method.
The calculations were performed with ENDF/B-VII.0 cross sectionsin 252 groups (collapsed to 23 groups for the 3D transport solution).
The key results of the comparison of calculations with measurements
include initial criticality, control rod worth critical configurations, control
rod worth, differential boron worth, and isothermal temperature
reactivity coefficient (ITC). The VERA results for these parameters
show good agreement with measurements, with the exception of
the ITC, which requires additional investigation. Results are also
compared to those obtained with Monte Carlo methods and a current
industry core simulator.

10:40 AM

AP1000 Ⓡ PWR Reactor Physics Analysis with VERACS and KENO - Part I: Zero Power Physics Tests
F.Franceschini(1), A.Godfrey, J.C.Gehin(2)1)Westinghouse, Pennsylvania, USA, 2)ORNL, Tennessee, USA
Westinghouse has applied the Core Simulator of the Virtual
Environment for Reactor Applications, VERA-CS, under developmentby the Consortium for Advanced Simulation of LWRs (CASL) tothe core physics analysis of the AP1000® PWR. The AP1000 PWR
features an advanced first core with radial and axial heterogeneities,
including enrichment zoning, multiple burnable absorbers, and a
combination of light and heavy control banks to enable the MSHIMTM
advanced operational strategy. These advanced features makeapplication of VERA-CS to the AP1000 PWR first core especiallyrelevant to qualify VERA performance. A companion paper at thisconference describes the power distribution analysis of the AP1000
PWR with VERA-CS and the KENO Monte-Carlo code. This paper
describes the results obtained for the startup physics tests simulationsof the AP1000 PWR first core (critical boron, rod worth and reactivity
coefficients), supporting the excellent numerical agreement reported in
the companion paper for the power distribution.

11:00 AM

AP1000 Ⓡ PWR Reactor Physics Analysis with VERACS and KENO - Part II: Power Distribution
F.Franceschini(1), J.C.Gehin, A.T.Godfrey(2)1)Westinghouse, Pennsylvania, USA, 2)ORNL, Tennessee, USA
Westinghouse has applied the Core Simulator of the Virtual
Environment for Reactor Applications, VERA-CS, under developmentby the Consortium for Advanced Simulation of LWRs (CASL) to thecore physics analysis of the AP1000® PWR. The AP1000 PWR
features an advanced first core with radial and axial heterogeneities,
including enrichment zoning, multiple burnable absorbers, and a
combination of light and heavy control banks to enable the MSHIMTM
advanced operational strategy. These advanced features makeapplication of VERA-CS to the AP1000 PWR first core especiallyrelevant to qualify VERA performance. A companion paper at thisconference describes the results obtained with VERA-CS and the
KENO Monte-Carlo code for startup physics tests simulations of
the AP1000 PWR first core (critical boron, rod worth and reactivity
coefficients). This paper describes the results of detailed power
distribution comparisons between VERA-CS and KENO, and confirms
the excellent numerical agreement reported in the companion paper
for the startup physics tests simulations.

40

Solution of the BEAVRS Benchmark Using thenTRACER Direct Whole Core Transport CodeSeoul National University, Seoul, Korea
The BEAVRS benchmark is solved by the nTRACER direct wholecore transport code to assess its accuracy and to examine the solution
dependence on modeling parameters. A sophisticated nTRACER
core model representing the BEAVRS core is prepared after aseries of study to ensure solution accuracy. The resulting solutions
for several hot-zero-power (HZP) states are compared first with thecorresponding Monte Carlo solutions and then with the measured
data which include the control rod worths as well as the critical boron
concentrations (CBC). The core depletion calculation is performed
for the initial and second cycles with a set of approximated power
histories and the calculated CBCs are compared with the measured
data. The comparison results show that the criticality, control rod bankworths at HZP and the boron let-down curves of two cycles agree wellwith the measurements within ~200 pcm and ~20 ppm, respectively.

11:40 AM

Calculation of the Pressure Vessel Fluence in the
Hungarian VVER-440 Plants for the Lifetime Extension
G.Hegyi, G.Hordósy, A.Keresztúri, C.Maráczy, E.Temesvári
Magyar Tudományos Akadémia Energiatudományi Kutatóközpont,
Budapest, Hungary
Originally the design operational lifetime of the VVER-440 plants
was 30 years, which was based on a rather conservative estimation.
To extend its service life is an economically desirable goal, keeping
in mind safety precautions. One of the limiting factors for the plant
lifetime is the embrittlement of the reactor pressure vessel (RPV) due
to the interaction of the solutes in the vessel with the vacancies andother products created by the incident neutrons. In order to qualify
this effect the neutron fluence at the RPV must be well known. For
the proper analysis of this quantity the detailed operational history(from initial irradiation to near future) needs to be known and a large
number of shielding calculations have to be performed. In Hungary
a coupled calculation tool was employed. The outgoing current andthe scalar flux at the periphery of the core are calculated by the
KARATE code. The neutron source for calculation of RPV fluence by
the MCNP code is based on these quantities. Considerable researchhas been performed on validating this method. At Paks NPP, Hungary,a major project was launched to investigate the possibility of lifetime
extension up to 60 years while new fuel containing gadolinium pinsand corresponding reloading schemes are being introduced. The
paper details the evaluation of the fluence using the above outlined
procedure, the data used, the fluence obtained, and the uncertainty of
the result.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 10:20 AM - Mizuho_B

Track5-2 Nuclear Criticality Safety
Session Chair: Bo Feng(ANL), Tomohiro Endo(Nagoya Univ.)
10:20 AM

Uncertainty Evaluation of Reactivity in Single andMulti-Region TSUNAMI Modeling Analysis for Dry
Cask Storage
Q.T.Newell, C.E.Sanders
University of Nevada, Las Vegas, USA
Uncertainties in the reaction probabilities have the potential to affectevery major facet of uranium use in operational lifecycles. One ofthese operations includes disposal of used nuclear fuel. Evaluation of
the uncertainties in a given data library can provide assurances in thearea of dry cask storage. This study investigates single-region versus
multi-region modeling of a large cask system with the TSUNAMI3D (Tools for Sensitivity and Uncertainty Analysis Methodology
Implemented in Three Dimensions) control module (for sensitivity anduncertainty analysis) in the SCALE6 computer code. The differencein the sensitivity coefficients for three nuclides (U-235, Pu-239, andEu-153) are considered by modeling single and multi-region fuelassemblies within a single region storage cask. Overall, the study
shows that there are no distinct advantages of using a multi-region
assembly compared to a single region assembly to calculate thesensitivity coefficients of a storage cask. However, developing multiple
regions in the problem provides users a different way of modeling,
than simply running more particles.

10:40 AM

Transient Analysis in Super Critical Condition for
Several Fuel-Solution Tanks System with Different
LayoutH.Kikuchi(1), T.Obara(2)1)Department of Nuclear Engineering, Tokyo Institute of Technology,
Tokyo, Japan, 2)Research Laboratory for Nuclear Reactors, Tokyo
Institute of Technology, Tokyo, Japan
To confirm that changes of power and the total integral power incritical accidents are affected by different layouts of fuel-solution
tanks, transient analyses were performed for a single-tank systemand different layouts of a three-tank system using an integral kinetic
analysis method for a weakly coupled system. The changes of power
and the integral power were analyzed with different layouts of fuelsolution tanks. The code used for the analysis took into account theneutronic coupling between the tanks. The results show that different
layouts of fuel-solution tanks cause different transients of power inthe supercritical condition by different neutronic coupling and different
increases of integral power compared to the integral power in a singletank system. The integral power can be underestimated if neutronic
coupling between the tanks is ignored.

11:00 AM

Comparison of Gamma Dose Rate Calculations for
PWR Spent Fuel Assemblies
B.Feng, R.N.Hill(1), R.Girieud, R.Eschbach(2)1)Argonne National Laboratory, Argonne, USA, 2)CEA Cadarache,Saint-Paul-lez-Durance, France
Gamma dose rate calculations from the U.S. Department of Energy
and the French Commissariat à l'Énergie Atomique et aux Énergies
Alternatives were compared for PWR spent fuel assembly models
with UO 2 fuel at 33 MWd/kg burnup and MOX fuel at 60 MWd/kg. Both sides used their reference physics codes in a three-step
calculation procedure involving depleting fresh assemblies to obtain
the discharge compositions, obtaining the isotopic gamma source
after 30 years of simulated radioactive decay, and applying the source
to heterogeneous 3-D transport models to tally the flux at one meteraway from the axial midpoint through air. The fluxes were ultimately
converted to dose rates with different energy-dependent conversion
factors. This study was able to pinpoint the intermediate calculations
that exhibited the largest sources of discrepancy between the two
approaches. Most notably, the U.S. gamma source calculationaccounts for Bremsstrahlung and adjusts its multi-group photon
release rates to conserve energy. Despite these differences, very
similar dose rates were calculated by both approaches. For the UO2case, which was intended to benchmark a frequently-cited reference
study, both the DOE and CEA calculated 30-year dose rates between
4.6 and 5.8 Sv/h with various conversion factors, which are roughly
three times lower than the reference study’s results. For the MOX
case, the calculated dose rates ranged from 8.5 to 11 Sv/h. Given thesame conversion factor, the largest difference between DOE and CEAvalues was about 1.5 Sv/h.

11:20 AMThe KBS-3 Spent Fuel Canister Criticality Calculation
during 100,000 Years
S.Wu, B.Yang(1), H.Wu, Q.Wei, Y.Liu(2)1)East China Institute of Technology, Ministry of Education, Nanchang,
China, 2)East China Institute of Technology, Nanchang, China
KBS-3 system is a mature disposal system, the spent fuel is directlystored in canister without reprocessing. This paper studies the KBS3 spent fuel canister criticality security issue under groundwater
penetration accident from 100 to 100,000 years. The compositionof the material PWR spent fuel assemblies are calculated by using
MCBurn software, which is a coupling of MCNP and ORIGEN2.1.
The calculation results are used in MCNP5 to calculate the effective
multiplication factor keff of the assemblies in different volumes ofgroundwater penetration. The keff maximum value in initial period
is 1.0968, the error is about 0.735%. And then, four PWR spent
fuel assemblies are set in the canister, The paper study finds thatthe maximum keff of canister is 0.79609 in groundwater penetration
accident. The calculation results show that the KBS-3 spent fuelcanister is safe in criticality security issue.

11:40 AM

Favorable Features in Kinetics of Fast Reactors with
Physically Thick 208Pb-Reflector
G.G.Kulikov, A.N, Shmelev, V.A.Apse, M.Y.Ternovykh
National Research Nuclear University MEPhI, Moscow, Russia
Light materials with small atomic mass (light or heavy water, graphiteand so on) are usually used as a neutron reflector and moderator. The
present paper proposes to use a new, very heavy neutron moderator
and reflector, namely “radiogenic lead” with dominant content ofisotope 208Pb. Radiogenic lead is a stable natural lead contained in theore deposits with thorium component, usually, they are poly-metallic
ores of rare earth elements. The lead is actually a final product ofradioactive decay chains starting from uranium and thorium (seeappendix bellow). This isotope is characterized by extremely low
micro cross-section of radiative neutron capture (~0.23 mb for thermal
neutrons, smaller than graphite and deuterium cross-sections) andlarge albedo for thermal neutrons.
The following mathematical models of neutron kinetics are applied toanalyze the quick-proceeding processes initiated by reactivity jumpin the reactor cores surrounded by physically thick, weakly absorbing
neutron reflector: one-point, two-point, multi-point and continuous
models. It was revealed that the results of one-point and two-point
models differed substantially from those obtained by multi-point andcontinuous kinetic models which were developed by authors anddescribed in the paper. Also, the paper demonstrates that only multipoint and continuous models are able to follow up adequately neutron
kinetics in the fast reactor core surrounded by a physically thick andweakly absorbing neutron reflector. The summand of the “inverse-hour”
equation defining kinetic effects of the neutron reflector was analyzed
in detail.
It is evaluated that the use of radiogenic lead makes it possible toslow down the chain fission reaction on prompt neutrons in the fast
reactor. This can improve the fast reactor safety and reduce somerequirements to the technologies used to fabricate fuel for the fast
reactor.

41

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 10:20 AM - Hiei

Track2-2 Deterministic Transport Theory
Session Chair: Farzad Rahnema(Georgia Tech. Univ.), Wu Hongchun(Xi'an Jiaotong Univ.)
10:20 AM

Iterative Properties of the Integral Transport Matrix
Method for the DD Scheme in 2D Cartesian Geometry
D.Anistratov, Y.Azmy

NC State University, Raleigh, USA
We study convergence of the integral transport matrix method (ITMM)
based on block-Jacobi strategy for solving diamond-differenced
S N equations for two-dimensional transport problems. This is aspatial domain decomposition method applied in massively parallel
computations. We consider the case of one cell per subdomain. A
Fourier analysis of the equations for S2 is performed. The analysis
shows that the iteration method in this particular case loses its
effectiveness. Numerical results of finite-medium problems arepresented to demonstrate the behavior of the ITMM for the DD
scheme that was theoretically predicted.

10:40 AM

Neutron Leakage Treatment in Reactor Physics:
Consequences for Predicting Core Characteristics
G.Rimpault, J.-F.Vidal(1), W.F.G.van Rooijen(2)1)CEA, Saint Paul-lez-Durance, France, 2)University of Fukui, Fukui,Japan
New generations of simulation tools responding to the challenges
brought by the advanced features of both 3rd+ generation Pressurized
Water Reactor (PWR) cores and 4th generation sodium fast neutron
reactor (SFR) cores are taking shape. The developments of new
simulations tools are also motivated by strict requirements of nuclear
safety authorities. The new tools have the objective of setting newreference standards for neutronic prediction and will take advantage
of innovative algorithms which have been implemented in existing
CEA codes, such as ERANOS (fast reactors) and APOLLO2 (PWR);
the new codes should at the same time remove remaining calculation
errors. Although innovative algorithms have been filling the gaps which
did exist 40 years ago between tools specifically dedicated to either
thermal neutron cores or fast neutron ones, there remains a series ofalgorithms which deserve particular attention: the treatment of leakage
in cell calculations.
This paper describes methods for treating neutron leakage in selfshielding calculations with the sub-group method, and in the cellbalance calculation. Applications of the MOC method of solution totreat neutron leakage are described. The application of the MOC
can eliminate approximations at the cell interfaces while maintaining
precise neutron leakage treatment. The new APOLLO3® code,
presently under development at CEA, is candidate for hosting suchalgorithms.

11:00 AM

Revisit Boundary Conditions for the Self-Adjoint
Angular Flux Formulation
Y.Wang, F.N.Gleicher
Idaho National Laboratory, Idaho, USA
This paper revisits the boundary conditions applied to the self-adjoint
angular flux formulation in a more rigorous variation derivation. It
includes a new way of weakly imposing the reflecting boundary
condition for preserving the symmetry of the final system of equations.
Completely equivalent parity variational forms are given. Numericalresults with the combination of these boundary conditions arepresented.

11:20 AM

Accuracy Preserving Surrogate for Neutron Transport
Calculations
C.Wang, H.S.Abdel-Khalik
North Carolina State University, Raleigh, USA
Recent advances in reduced order modeling and exact-to-precision
generalized perturbation theory are combined in a novel algorithm
that constructs a surrogate model for the Boltzmann equation,commonly used in assembly calculations to functionalize the fewgroup cross-sections in terms of the various assembly types, depletion
characteristics, and thermal-hydraulics conditions. First, the algorithm
employs reduced order modeling to determine the dominant input
parameters, aggregated in the so-called active subspace, using a42

random sample of first-order derivatives calculated using an adjoint
model. Next, exact-to-precision generalized perturbation theory
identifies an active subspace for the state solution (i.e., angular flux)and constructs a surrogate model that is parameterized over theactive subspace of the input parameters. This approach is shown tosignificantly reduce computational time needed for the analysis ofa large number of model variations, while meeting the user-defined
accuracy requirements. Numerical experiments are employed todemonstrate the mechanics and application of the proposed approachto assembly calculations commonly used in reactor physics analysis.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 10:20 AM - Atago

Track13 Radiation Applications and Nuclear Safeguards
Session Chair: Alexis C. Kaplan(Michigan Univ.), Yasunori Kitamura(JAEA)
10:20 AM

Design of Long Neutron Counter for Intensified D-T
Neutron Source
L.Yanan, L.Taosheng, Z.Siwei, Q.Fupeng, S.Gang, S.Jing, FDS Team
Institute of Nuclear Energy Safety Technology, CAS, Anhui, China
A low sensitivity long neutron counter was designed as a standard
directional flow detector to monitor neutron fluence reference valuesof an accelerator-based 14 MeV D-T neutron source with yield about
1013n/s. The energy response over 6 MeV was improved using atungsten radiator, which acts as an energy converter via the (n,xn)
reaction. Different parameters were optimized to flatten the neutronenergy response over a wide energy range. A simulation of thedesigned long neutron counter using the Monte Carlo codes MCNP
was undergone. The response function is relatively flat in the energy
range of 1 keV-20 MeV. The results show the maximal relative
variation is about 7.8%.

10:40 AM

Utilizing Simulated Rossi-Alpha Distributions to
Develop New Methods of Characterizing Spent
Nuclear Fuel
A.C.Kaplan, V.Henzl, A.P.Belian, M.P.Swinhoe, H.O.Menlove(1),M.Flaska, S.A.Pozzi(2)1)Los Alamos National Laboratory, New Mexico, USA, 2)University ofMichigan, Michigan, USA
List-mode data collection for high-activity sources such as spent
nuclear fuel has been made possible recently by advances in datacollection and data-storage capabilities. While the shift-register
technique has been utilized primarily in the past, list-mode datacollection and subsequent Rossi-alpha distribution (RAD) production
offers the same analysis approach, as well as increased flexibility
post-measurement. Simulating the RADs allows us to examine
different time domains and the effect of fissile and fertile materials on
the shape of the measured RADs. New analysis methods possible
only with RADs allow us to determine assembly multiplication with amean difference of 0.7% when tested against a benchmark set of 44
different simulated spent fuel assemblies. This method also enables
us to determine total Pu mass with fewer calibration constants anda-priori knowledge than previous methods. Here we consider different
techniques for simulating and analyzing Rossi-alpha distributions forapplications in spent nuclear fuel characterization.

liquid scintillator detector response are traditionally run in fully analog
mode. Analog simulations of cross-correlation measurements withthese codes are extremely time-consuming because the probabilityof correlated detection is extremely small, approximately equal tothe product of the probabilities of a single detection in each detector.
The new “implicit correlation” method described here increases
the number of correlation event tallies thereby, reducing varianceand required computation times. The cost of the implicit correlation
method is comparable to the cost of simulating single event detectionin the lowest efficiency detector. This method is especially usefulin the nuclear nonproliferation and safeguards fields for simulatingcorrelation measurements of shielded special nuclear material.
A plutonium metal sample was measured by a fast-neutron multiplicity
counter for verification of MCNPX-PoliMi calculations and the new
method. The fast-neutron multiplicity counter was used with pulse
shape discrimination techniques to produce a neutron-neutron
cross-correlation distribution. The new method was implemented
in MCNPX-PoliMi with a Pu-240 spontaneous fission source inthe measurement geometry. The method demonstrated good
agreement with analog simulation and measurement results and a
speed-up of a factor of 500 over analog calculations.

11:40 AMA Unique Tungsten-Based Tagging Approach for
Maintaining of Continuity of Knowledge of Nuclear
Waste Copper Canisters
D.Chernikova, K.Axell, A.Nordlund
Chalmers University of Technology, Göteborg, Sweden
A new approach to the unique tagging of nuclear waste copper
canisters is suggested. In this new method a combination of two
different techniques, radiation and ultrasonic measurements, is usedin order to get the same unique identifier of the cask. The necessarycomponent of the method is a tungsten/lead insert marked witha binary or bar code and placed inside the container. The paper
discusses results of the radiation measurements performed in lab
as the first proof of the concept, as well as results of Monte-Carloevaluation of the feasibility of proposed approach. The method makes
it possible to maintain continuity of knowledge of nuclear waste
for a time scale up to a few hundred years without comprising theenvironmental safety of casks.

11:00 AM

Antineutrino Emission from Fuels with High
Proliferation Resistance
T.Shiba, M.Fallot(1), S.Cormon(2)1)SUBATECH, Nantes, France, 2)Universite de Nantes, Nantes, France
A study on the derivation of antineutrino spectra from a normal
mixed oxide fuel and fuels with high proliferation resistance has been
conducted for the future study of the feasibility of antineutrino detector
to be used for nuclear non-proliferation purpose. Due to the large
isotopic proportion of 238Pu in fuels with high proliferation resistance,
the antineutrino detection method could be used as a verification toolof the use such fuels in fast reactors. Therefore, we have performed
simulations of the time evolution of a fast breeder reactor loading a few
types of nuclear fuels and obtained the inventory of fission products. A
subsequent summation calculation based on nuclear data derives theantineutrino energy spectra from a normal mixed oxide fuel and a fuelwith high proliferation resistance. The results show a 3% difference ofthe number of emitted antineutrinos between the normal mixed oxide
fuel and the fuels with high proliferation resistance.

11:20 AM

Validation of the Implicit Correlation Method in
MCPNX-PoIiMi Using Plutonium Cross-Correlation
Measurements
M.J.Marcath, T.H.Shin, S.D.Clarke, J.L.Dolan, M.Flaska, E.Larsen,
A.C.Kaplan, S.A.Pozzi(1), P.Peerani(2), E.Padovani(3)1)University of Michigan, Michigan, USA, 2)Joint Research Centre, Ispra,
Italy, 3)Polytechnic of Milan, Milan, Italy
Monte Carlo particle transport codes used to accurately model organic

43

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 10:20 AM - Cosmos

Track9-1 Transient and Safety Analysis
Session Chair: Kostadin Ivanov(Pennsylvania Univ.),Yuichiro Ban(Toshiba)
10:20 AM

11:20 AM

Y.Zheng, K.Zhuang, L.Cao, H.Wu

L.Maas, B.Normand

Xi'an Jiaotong University, ShannXi, China
In the molten salt reactor, the delayed neutron precursors continuously
change their position both in the core and in the external loop due tothe flow of fuel, which lead to the loss of reactivity and decrease ofeffective delayed neutron fraction. Therefore, the neutron kinetics ofmolten salt reactor is significantly different from that of conventional
reactors using solid fuels. In this study, a 3D neutron kinetics modelof molten salt reactor considering the flow effects of delayed neutron
precursors was established. The analytic basis functions expansion
nodal method for arbitrary triangular-z node was performed to solve
the three dimensional neutron diffusion equations and the methodof characteristics was used to find the solution of delayed neutron
precursor equations within the whole primary circuit. To verify the code,
calculations were preformed on a homogeneous reactor model. The
results were in good agreement with the reference results. Besides,
the influence of fuel circulation on the kinetic characteristic of reactor
was investigated. The results showed some special phenomenon in
the molten salt reactor.

Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Fontenay-auxRoses, France
Heterogeneous boron dilution can occur in a PWR in case of aleakage of a primary pump exchanger. Unborated cold water canaccumulate in the U-leg during shutdown states and form a plug.
When the primary pump restarts, this plug is transported toward
the core and may lead to the core re-criticality. To evaluate theconsequences of this transient, a sequential calculation involving the
CFD code CFX (for the vessel) and the fully coupled 3D computational
chain HEMERA V1 (for the core) is used.
This paper is divided into three parts. The first one describes the codesused for the transport of the unborated water (CFX) and for the core
response (HEMERA). The second part focuses on the modeling of
the heterogeneous boron dilution transient and the third one presents
the results. In order to demonstrate the capability of this calculating
scheme, two responses of the core have been calculated. The firstone keeps the core sub-critical during whole dilution transient and thesecond one assumes a prompt critical jump in order to demonstrate
the capability of the calculation scheme.

Development of the Neutron Kinetics Code for
Thermal Molten Salt Reactor

10:40 AM

Study of Neutron Propagation in Multigroup Transportby Space Asymptotic Methods
J.C.L.Fernandes(1), S.Dulla, P.Ravetto(2), M.T.Vilhena(3)1)Politecnico di Torino / Universidade Federal do Rio Grande do Sul,
Torino, Italy, 2)Politecnico di Torino, Torino, Italy, 3)Universidade Federal
do Rio Grande do Sul, Porto Alegre, Brasil
In this work the space asymptotic theory is used for the study ofthe propagation of neutron pulses in multigroup transport. Pulsed
experiments are of interest in the study and characterization ofsource-driven subcritical systems. The model allows for an exact
analytic treatment, although it can be applied only to highly idealized
configurations. Therefore, the objective of the investigation is to obtaina good physical insight into the phenomena and to derive highly
accurate results to be used as benchmarks. The solution is obtainedby a combined application of the Fourier and Laplace transforms.
The inversion of the Laplace transform is attained by the use of theresidue theorem, which, through the study of the singularities of thetransformed solution, allows interesting physical considerations. The
solution is exact until the particles, moving with a finite velocity, reachthe external boundary of the system. Numerical results for sometypical propagation problems are presented for one-dimensional planesystems.

11:00 AM

Sensitivity Analysis and Performance of the Adiabatic,
Theta, and Multigrid Amplitude Function Kinetics
Methods in 2D MOC Neutron Transport
S.Shaner, B.Forget, K.Smith
Massachusetts Institute of Technology, MA, USA
A sensitivity study of the solution accuracy for the transport versionof the 2D LRA benchmark was conducted for the Adiabatic, Theta,
and Multigrid Amplitude Function (MAF) methods. Transient analysis
has been extensively studied with neutron diffusion theory, but only
recently has the focus changed to investigating reactor transientsusing fine grained neutron transport theory calculations due to thecomputational cost of transport methods. The Method of Characteristics(MOC) was chosen as the transport method to investigate due to thecomputational efficiency and broad use of the method in academia
and industry. The Adiabatic, Theta, and MAF methods have been used
for diffusion theory transient calculations and appear well suited for
implementation in transport theory. Understanding how these methods
scale in transport theory will be essential in developing codes to model3D problems of industrial interest on current computer architectures.
This study seeks to validate the transient analysis approach employedin the our implementation of the Adiabatic, Theta, and MAF methods
as being computationally efficient and scalable and thus deserving
further investigation in transport theory transient analysis.

44

Computations of Heterogeneous Dilution Transients
Using CFX and HEMERA V1

11:40 AM

Prompt Behavior of Generalized-Eigenvalue Point
Kinetics Models
B.C.Kiedrowski
Los Alamos National Laboratory, NM, USA
The point kinetics model is used as a first approximation for modelingtransients in nuclear systems. Point kinetics is accurate enough inmany situations, but its performance can degrade far from criticality.
The classic approach uses a k or multiplication eigenvalue as thebasis for developing the underlying model. This work generalizes thepoint kinetics equations for any multiplicative eigenvalue. The collisionand leakage eigenvalues are studied, and preliminary results showthat in some cases the collision eigenvalue provides a more accurate
representation of the prompt period.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 13:30 PM - Mizuho_A

Track1-5 Reactor Analysis Methods
Session Chair: Myung Hyun Kim(Kyung Hee University), Masahiro Tatsumi(NEL)
13:30 PM

14:30 PMT.Kajihara, M.Tsuji, G.Chiba, Y.Kawamoto, T.Narabayashi(1),Y.Ohoka, T.Ushio(2)M.Wu

Automatic Construction of a Simplified Burn-Up
Chain Model by the Singular Value Decomposition
1)Hokkaido University, Sapporo, Japan, 2)Nuclear Fuel Industries, Ltd.,
Osaka, Japan
Nuclear reactor design analysis often requires a simplified burn-up
chain model to reduce computation time. It is difficult to construct thesimplified burn-up chain because it requires engineers to have highly
skilled techniques and in-depth knowledge of burn-up processes. This
paper develops an algorithm for automatically constructing a reducedorder burn-up chain model from the detailed model using the singularvalue decomposition (SVD). In our approach, we prepare a detailed
burn-up chain matrix A, and an extraction matrix C, which extracts
important nuclides for specific purposes such as the evaluation ofneutron multiplication factor. By applying SVD to C, we can obtain thefirst candidate nuclides of the reduced-order burn-up chain model.
Then, by applying SVD to 1st information transfer matrix F12(1) , which
defines the relationship between the first candidate nuclides andremaining nuclides, we can obtain additional candidate nuclides for the
reduced-order burn-up chain model from the remaining nuclides. We
repeat this process until the norm of the information transfer matrix is
sufficiently close to zero. Finally, all candidate nuclides chosen throughthese simplification processes are adopted as a reduced-order burnup chain model. As the test case, we reduce a detailed burn-up chain
model consisting of 1421 nuclides to a model of 219 nuclides. We canuse the resulting reduced-order model to calculate the burn-up with ahigh degree of accuracy.

13:50 PM

Generation of Simplified Burnup Chain Using
Contribution Matrix of Nuclide Production
R.Katano, T.Endo, A.Yamamoto(1), Y.Kamiyama, K.Kirimura,
S.Kosaka(2)1)Nagoya University, Nagoya, Japan, 2)Mitsubishi Heavy Industries, Ltd.,
Kobe, Japan
A procedure to automatically generate a simplified burnup chain is
proposed, which preserves prediction accuracy of target nuclides
number density using calculation results of a detail burnup chain. In
the present procedure, the contributions to production of the target
nuclides are quantitatively evaluated at first using the off-diagonal
elements of a burnup matrix and nuclides number density. Then,by considering the contributions to nuclide production, which isrepresented as a “contribution matrix”, the nuclides in detail burnup
chain are selected according to the importance represented as thecontribution matrix. Finally, a simplified burnup chain is generated withthe target nuclides and the selected nuclides from the contribution
matrix. Since the present method only utilizes the burnup matrix,
which is usually used in the common lattice physics computation, its
implementation to production codes is easy. In order to examine theeffectiveness of the present method, burnup calculation in a PWR fuelpin-cell problem with some simplification on calculation conditions iscarried out. The calculation results suggest that the present method
will be a good candidate for an automated generation method for a
simplified burnup chain.

14:10 PM

Important Fission Product Nuclides Identification
Method for Simplified Burnup Chain Construction
G.Chiba, M.Tsuji, T.Narabayashi(1), Y.Ohoka, T.Ushio(2)1)Hokkaido University, Sapporo, Japan, 2)Nuclear Fuel Industries, Ltd.,
Osaka, Japan
A method of identifying important fission product nuclides whichshould be included in a simplified burnup chain is proposed. Thismethod utilizes adjoint nuclide number densities and contributionfunctions which quantify the importance of nuclide number densities
to the target nuclear characteristics: number densities of specificnuclides after burnup. This method is tested against light water reactor
fuel pin-cell problems, and it is shown that this method successfully
identifies important fission product nuclides which should be included
in a simplified burnup chain with which nuclide number densities oftarget nuclides after burnup are well reproduced.

Application of Backtracking Algorithm to Depletion
Calculations
China Institute of Atomic Energy, Beijing, China
Based on the theory of linear chain method for analytical depletioncalculations, the burn-up matrix is decoupled by the divide andconquer strategy and the linear chain with Markov characteristic
is formed. The density, activity and decay heat of every nuclide in
the chain can be calculated by analytical solutions. Every possible
reaction path of the nuclide must be considered during the linear
chain establishment process. To confirm the calculation precision andefficiency, the algorithm which can cover all the reaction paths of thenuclide and search the paths automatically according to the problem
description and precision restrictions should be sought. Throughanalysis and comparison of several kinds of searching algorithms,
the backtracking algorithm was selected to search and calculate thelinear chains using Depth First Search (DFS) method. The depletion
program can solve the depletion problem adaptively and with highfidelity. The solution space and time complexity of the program wereanalyzed. The new developed depletion program was coupled with
Monte Carlo program MCMG-II to calculate the benchmark burn-up
problem of the first core of China Experimental Fast Reactor (CEFR).
The initial verification and validation of the program was performed by
the calculation.

14:50 PM

Modeling the Cross Section of Gadolinia Pins in the
Depletion for Pin-By-Pin Core Calculations
Y.Kodama, T.Ushio, Y.Ohoka, S.Takeda
Nuclear Fuel Industries, Ltd., Osaka, JapanSCOPE2 is a next generation core calculation code based on thesemi-analytic simplified P3 nodal method in 3-dimensional pinby-pin geometry with microscopic-correction depletion capability.
SCOPE2 uses pin-wise homogenized multi-group cross-section
library generated by AEGIS. SCOPE2 can track number densities
of major nuclides explicitly in each pin and capture effects under
off-nominal condition directly with microscopic-correction model.
In addition, the special treatment at Gd-bearing UO2 pins is implemented due to its large spatial self-shielding. It is the method thatreconstructs cross-section trends in depletion under arbitrary neutron
spectrum using number density of gadolinium in spite of burnup. To
prove the effectiveness of this method, depletion calculations in singleassembly geometry are performed. The results show that this model
can reconstruct cross-sections accurately in depletion under different
spectrum from base condition such as RCC-in depletion. Thus, thehigh effectiveness of this model is proved.

15:10 PMThe Optimized Algorithm for the Microscopic
Depletion Model in the COCAGNE Core CodeA 2-Level Core Partitioning Approach
M.Guillo, D.Couyras, F.Fevotte, F.Hoareau
EDF R&D, Clamart, France
EDF/R&D is developing a new calculation scheme based on thetransport-Simplified Pn (SPn) approach. The lattice code used is thedeterministic code APOLLO2, developed at CEA with the supportof EDF and AREVA-NP. The core code is the code COCAGNE,
developed at EDF R&D. The latter can take advantage of a
microscopic depletion solver which improves the treatment of spectral
history effects. The accuracy of this model has been proven to be very
useful for industrial uses but until now, being very time consuming aswell as memory greedy, this model could not be used in production
and needed to be re-written from the ground with speed and memory
constraints in mind. This paper introduces the new methodology,
a 2-levels core partitioning paradigm. Theses partitions, based on
physics considerations, are also well suited for modern multi-core
CPU architecture, and parallelizing in shared memory shows good
speedup. Compared to the previous version, a total speedup ofaround 100 has been obtained: roughly 20 due to the algorithm and 5
due to the parallelization on a 8 cores chip. These results have beenobtained on a machine similar to those that will be used in production.

45

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 13:30 PM - Mizuho_B

Track4-1 Verification, Validation and Uncertainty Analysis
Session Chair: Hany Abdel-Khalik(NSCU), Kenji Yokoyama(JAEA)
13:30 PM

Validation and Benchmarking of Calculation Methods for
Photon and Neutron Transport at Cask Configurations
L.Schlömer, S.Tittelbach, P.-W.Phlippen(1), J.Pillath(2), W.Filbert(3)1)WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Jülich,
Germany, 2)Forschungszentrum Julich GmbH, Jülich, Germany, 3)DBE
Technology GmbH, Peine, Germany
The reliability of calculation tools to evaluate and calculate dose rates
appearing behind multi-layered shields is important with regard to
the certification of transport and storage casks. Actual benchmark
databases like SINBAD do not offer such configurations because
they were developed for reactor and accelerator purposes. Due tothis, a benchmark-suite based on own experiments that contain dose
rates measured in different distances and levels from a transport andstorage cask and on a public benchmark to validate Monte-Carlotransport-codes has been developed. The analysed and summarised
experiments include a 60Co point-source located in a cylindrical cask,a 252Cf line-source shielded by iron and polyethylene (PE) and a bare252
Cf source moderated by PE in a concrete-labyrinth with different
inserted shielding materials to quantify neutron streaming effects
on measured dose rates. In detail not only MCNPTM (version 5.1.6)
but also MAVRIC, included in the SCALE 6.1 package, have been
compared for photon and neutron transport. Aiming at low deviations
between calculation and measurement requires precise source term
specification and exact measurements of the dose rates which havebeen evaluated carefully including known uncertainties. In MAVRIC
different source-descriptions with respect to the group-structure of
the nuclear data library are analysed for the calculation of gamma
dose rates because the energy lines of 60Co can only be modelledin groups. In total the comparison shows that MCNPTM fits very wellto the measurements within up to two standard deviations and that
MAVRIC behaves similarly under the prerequisite that the sourcemodel can be optimized.

13:50 PM

Recent Advances in the V&V of the New French CEA
APOLLO3® Neutron Transport Code : Benchmarks
Analysis of the Flux Solvers
J-M.Palau, P.Archier, J.-F.Vidal, G.Rimpault, B.Roque, P.Bourdot,
Y.Peneliau, G.Truchet, C.De Saint Jean
CEA, Saint-Paul-lez-Durance, FranceThis paper presents a synthesis of the latest advances in the
Verification and Validation (V&V) process of the new French (CEA)
deterministic neutron transport code APOLLO3® developed within theframework of a common CEA, AREVA and EDF project. It focuses
more precisely on the generic V&V of the main transport flux solversof the code (namely IDT, Minaret, Pastis, TDT and Minos,) through 1D
to 3D international benchmarks (ZPR-1D, Stepanek, C5G7, Takeda).
Precise criteria have been defined to assess the quality of each solver
by comparison with TRIPOLI4® multigroup Monte-Carlo calculations
that have been performed for each configuration. We show that puretransport flux solvers (IDT, Minaret, Pastis and TDT-MOC) based on Sn, Pn and characteristics methods meet the keff target precision criteria(100 pcm) whereas SPn solver (Minos) give satisfactory results withinreasonable computation time. The complementary of the APOLLO3®
flux solvers set is globally highlighted.

14:10 PMDeCART Code Verifications by Numerical Benchmark
Calculations of HTTR
C.J.Jeong, H.C.Lee, J.M.Noh
Korea Atomic Energy Research Institute, Daejeon, Korea
DeCART code verifications have been performed through thenumerical benchmark calculations of HTTR. The reference calculations
have been carried out using the Monte Carlo McCARD code in whicha double heterogeneity model was used. Verification results showthat the DeCART code gives less negative MTC and RTC than the
McCARD code does and thus the DeCART code underestimatesthe multiplication factors at states with high moderator and reflector
temperatures. However, the DeCART code predicts more negative FTC
than McCARD code does. In the depletion calculation for the HTTR
single cell and single block, the error of the DeCART code increases
with burnup. While the DeCART code error in a 2-dimensional core
depletion calculation decreases with burnup up to around 500 FPD.

46

14:30 PM

Development and Verification of Three-Dimensional
Hex-Z Burnup Sensitivity Solver Based on
Generalized Perturbation Theory
K.Yokoyama
Japan Atomic Energy Agency, Ibaraki, JapanA burnup sensitivity analysis solver based on generalized perturbation
theory has been developed in a multi-purpose analysis framework
MARBLE. The new solver has capability to calculate sensitivity
coefficients of transmuted nuclear fuel composition after burnup,
i.e. atomic number density, with respect to nuclear data. Based on
the diffusion theory, the new solver calculates the burnup sensitivity
coefficients not only in 2-dimensional R-Z model but also in
3-dimensional Hexagonal-Z model, that is not supported in a existingcode system in JAEA. The new solver has been verified by comparing
with results of so-called direct calculation. According to the generalized
perturbation theory, it is possible to calculate a breakdown of theburnup sensitivity coefficient by using five terms which appear in theformulation. In the present paper, a new numerical verification method
named component-wise direct calculation is proposed to check theresults of each term. A numerical experiment of the new verification
method was performed, and it shows that the new verification method
is valid. In addition, a sample application result is shown in order toevaluate calculation model effects due to difference between 2- and3-dimensional models. The sample application result suggests that
3-dimensional calculation model is necessary for some heterogeneous
core configuration.

14:50 PM

Validation of HELIOS for ATR Core Follow Analyses
S.E.Bays, E.T.Swain, D.S.Crawford, D.W.Nigg
Idaho National Laboratory, Idaho Falls, USA
This work summarizes the validation analyses for the HELIOS codeto support core design and safety assurance calculations of theAdvanced Test Reactor (ATR). Past and current core safety assurance
is performed by the PDQ-7 diffusion code; a state of the art reactor
physics simulation tool from the nuclear industry’s earlier days. Overthe past twenty years, improvements in computational speed haveenabled the use of modern neutron transport methodologies to replace
the role of diffusion theory for simulation of complex systems, such as
the ATR. More exact methodologies have enabled a paradigm-shift
away from highly tuned codes that force compliance with a bounding
safety envelope, and towards codes regularly validated against
routine measurements. To validate HELIOS, the 16 ATR operational
cycles from late-2009 to present were modeled. The computed power
distribution was compared against data collected by the ATR’s on-line
power surveillance system. It was found that the ATR’s lobe-powers
could be determined with ±10% accuracy. Also, the ATR’s cold startup
shim configuration for each of these 16 cycles was estimated andcompared against the reported critical position from the reactor logbook. HELIOS successfully predicted criticality within the tolerance
set by the ATR startup procedure for 13 out of the 16 cycles. This iscompared to 12 times for PDQ (without empirical adjustment). Thesefindings, as well as other insights discussed in this report, suggest thatHELIOS is highly suited for replacing PDQ for core safety assurance of
the ATR. Furthermore, a modern verification and validation frameworkhas been established that allows reactor and fuel performance data to
be computed with a known degree of accuracy and stated uncertainty.

15:10 PMPERSENT: Need of a Deterministic Code for Sensitivity
Analysis in 3D Geometry and Transport Theory
G.Aliberti, M.A.Smith
Argonne National Laboratory, Argonne, USA
A new code PERSENT was recently developed at ANL for sensitivity
analysis of relevant integral parameters in 3D geometry and transport
theory. The goal of this paper is to demonstrate the computational
capabilities of PERSENT with respect to the deterministic codes
commonly used for sensitivity analysis and particularly to address
the impact on sensitivity coefficients when a simplified RZ model
was generally adopted in past studies for the analysis of complicatedreactor designs.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 13:30 PM - Hiei

Track2-3 Deterministic Transport Theory
Session Chair: Yousry Azmy(North Carolina State Univ.), Kazuya Yamaji(MHI)
13:30 PM

14:30 PM

S.G.Stimpson, B.S.Collins, T.J.Downar

R.Lenain, E.Masiello, F.Damian, R.Sanchez

University of Michigan, Michigan, USA
The MPACT code being developed collaboratively at the University ofMichigan (UM) and Oak Ridge National Laboratory (ORNL) provides
users with a variety of deterministic methods for solving the 2D and
3D Boltzmann transport equation. One of these methods, the 2D/1Dtechnique, decomposes 3D problems into a 1D axial stack of 2D
radial planes. In this scheme, the 2D planes are typically solved using
a method such as the Method of Characteristics (MOC) to preserve
the geometric heterogeneity in the radial direction. These planes
are incorporated into a 1D axial solver, which can use a variety ofmethods. This work demonstrates the use of the traditional nodal
methods for solving the 1D axial problem (finite difference, NEM,
SANM, SP3), but also introduces a discrete ordinates (Sn) solver
which uses up to cubic Legendre expansion spatially and can also
incorporate higher order angular distributions of the radial transverse
leakage.
Several test cases are presented to demonstrate the accuracy ofthe solvers for various axial sizes. The first three are the 3D-C5G7
extension benchmark cases. The fourth case is a single quarter
assembly benchmark problem with explicit nozzle, plenum, and coreplate modelling known as AMA Problem 3. The final case is a quartercore benchmark problem that is an extension of the quarter assembly
problem known as AMA Problem 5. In general, the diffusion-based
axial solvers perform very well, though higher-order solvers provide
some benefit in more difficult problems, particularly rodded cases.

CEA, Saclay, France
In this paper we analyze the synergy between the Domain
Decomposition Method and the Coarse-Mesh Finite Differencetechnique. In contrast to massively parallel computations, weconstruct a coarse-grained parallelism for daily run calculations on
standard SIMD workstations based on shared memory architecture.
We evaluate the effectiveness of the algorithm for several high-fidelity
calculations spanning different types of color-sets up to the full-core.
We show that CPU times for a best-estimate 2D calculation of the
EPR can be reduced from several days to few hours using a standard
workstation.

Axial Transport Solvers for the 2D/1D Scheme in
MPACT

13:50 PM

Development of Legendre Expansion of Angular Flux
Method for 3D MOC Calculation
Y.Kato, T.Endo, A.Yamamoto
Nagoya University, Nagoya, JapanA three-dimensional transport calculation method, Legendre
Expansion of Angular Flux Method (LEAF method) has been proposedfor practical MOC calculations in three-dimensional geometry. In the
LEAF method, transport calculation is carried out on “characteristics
planes(CP)” rather than conventional ray traces, whose concept issimilar to that of the ASMOC3D method. By utilizing the characteristics
plane, required memory and computational load are significantly
reduced. Each characteristics plane forms rectangular regions,thus efficient transport calculation scheme based on transmission
probability can be adopted. Consequently, the LEAF method canpractically perform MOC calculation in three-dimensional geometry.
In the LEAF method, spatial distribution of incoming, outgoing andaverage angular fluxes in a CP are expressed as a series of the
Legendre polynomials in order to accurately capture their spatial
dependence. Calculation results in a test problem indicate theaccuracy and effectiveness of the LEAF method.

14:10 PM

Benchmark on Deterministic Time-Dependent
Transport Calculations without Spatial
Homogenisation
V.F.Boyarinov, A.E.Kondrushin, P.A.Fomichenko
National Research Center Kurchatov Institute, Moscow, Russia
The space-time neutron kinetics benchmark on deterministictransport calculations without spatial homogenization C5G7-TD has
been developed and proposed for verification of codes solving thetime-dependent neutron transport equation. The well-known C5G7
benchmark has been chosen as the base for new benchmark. The
proposed benchmark has been calculated by SUHAM-TD code,
which realizes the surface harmonic method (SHM). Authors hope toattract the attention of other researchers in order to involve them toparticipate in calculations of the proposed benchmark.

Coarse-Grained Parallelism for Full-Core Transport
Calculations

14:50 PM

Parallel Performance Results for the OpenMOC
Method of Characteristics Code on Multi-Core
Platforms
W.Boyd, K.Smith, B.Forget(1), A.Siegel(2)1)MIT, Massachusetts, USA, 2)Argonne National Laboratory, Argonne,
USA
Over the past decade, multi-core processors have become a fixture ofmodern computing systems. The shift towards multi-core architectures
has ushered in a new era of shared memory parallelism for scientificapplications. This paper describes and analyzes two parallel transport
sweep implementations in the OpenMOC method of characteristics
neutron transport code for multi-core platforms using OpenMP.
Strong and weak scaling studies are performed for Intel Xeon multicore processors using the GNU and Intel C++ compilers. In addition,
variations of both implementations using exponential intrinsics andlinear interpolation techniques to evaluate the exponential in the
MOC equations are compared and contrasted. The results for OpenMOC demonstrate 100% parallel efficiency for 12 threads on 12
cores on Intel Xeon platforms. These results illustrate the potential for
hardware acceleration for MOC neutron transport on modern multicore architectures. In addition, they provide insights as to the benefits
and shortcomings of the OpenMP parallel programming model withrespect to MOC for both multi-core processors as well as future manycore platforms.

15:10 PM

Making More Precise the Surface Pseudosources
Method for RBMK Cluster Cells
N.V.Sultanov
Russian Research Center "Kurchatov institute", Moscow, Russia
At present the KLARA code is being used for the multigroupcalculations of the VVER and RBMK cluster cells and for preparationof a few group cell characters. Some calculation instability has
appeared in the 69-group calculations of the RBMK cluster cells.
Detailed analysis of these calculations showed that this instability has
been connected with divergence of a type pole in a few integrals intime of a calculation of angular moments of the sine component of theGreen’s function. We removed this peculiarity by means of addition of
the two lines of angular moments of the sine component of the neutron
distribution function, when these peculiarities compensated eachother and a calculation became stable. Difference of the Kinf. values
between the options KLARA in G31 - approximation and PIJ becameless or equal to 0.2% for the RBMK cluster cells both with a fuel andwithout.

47

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 13:30 PM - Atago

Track3-2 Monte Carlo Methods
Session Chair: Kan Wang(Tsinghua Univ.), Yasushi Nauchi(CRIEPI)
13:30 PMA Monte Carlo Method for Prompt and Delayed Alpha
Eigenvalue Calculations
A.Zoia, E.Brun, F.Malvagi
CEA/Saclay, DEN/DANS/DM2S/SERMA, Gif-sur-Yvette Cedex, France
Several technological issues, such as for instance kinetics studiesof accelerator-driven systems, reactor start-up analysis, or reactivity
measurements, demand the asymptotic time behaviour of neutrontransport to be assessed. Typically, this amounts to solving aneigenvalue equation associated to the Boltzmann operator, whose
precise nature depends on whether delayed neutrons are taken
into account. In this work, we propose a Monte Carlo method for
determining the dominant eigenvalue of the Boltzmann operator
and the associated fundamental mode for arbitrary geometries,
materials, and boundary conditions. This approach can be applied toconfigurations involving both prompt and delayed neutrons. Extensive
verification tests of the algorithm are performed.

13:50 PM

Geometry Navigation Acceleration Based on
Automatic Neighbor Search and Oriented Bounding
Box in Monte Carlo Simulation
Z.Chen, J.Song, G.Sun, B.Wu, H.Zheng, P.Long, L.Hu(1), Y.Wu,
FDS Team(2)1)University of Science and Technology of China, Hefei, Anhui, China,2)Institute of Nuclear Energy Safety Technology, Chinese Academy of
Sciences, Hefei, Anhui, China
Geometry navigation plays the most fundamental role in Monte Carloparticle transport simulation. It’s mainly responsible for locating a particle
inside which geometry volume it is and computing the distance to thevolume boundary along the certain particle trajectory during each particle
history. Geometry navigation directly affects the run-time performanceof the Monte Carlo particle transport simulation, especially for large
scale complicated systems. Two geometry acceleration algorithms, theautomatic neighbor search algorithm and the oriented bounding box
algorithm, are presented for improving geometry navigation performance.
The algorithms have been implemented in the Super Monte Carlo
Calculation Program for Nuclear and Radiation Process (SuperMC)
version 2.0. The FDS-II and ITER benchmark models have been tested
to highlight the efficiency gains that can be achieved by using theacceleration algorithms. The exact gains may be problem dependent,
but testing results showed that runtime of Monte Carlo simulation canbe considerably reduced 50%~60% with the proposed acceleration
algorithms.

14:10 PM

Continuous-Energy Monte Carlo Methods for
Calculating Generalized Response Sensitivities Using
TSUNAMI-3D
C.M.Perfetti, B.T.Rearden
Oak Ridge National Laboratory, Oak Ridge, USA
This work introduces a new approach for calculating the sensitivityof generalized neutronic responses to nuclear data uncertainties
using continuous-energy Monte Carlo methods. The GEneralized
Adjoint Responses in Monte Carlo (GEAR-MC) method has enabled
the calculation of high resolution sensitivity coefficients for multiple,
generalized neutronic responses in a single Monte Carlo calculation
with no nuclear data perturbations or knowledge of nuclear covariancedata. The theory behind the GEAR-MC method is presented here andproof of principle is demonstrated by calculating sensitivity coefficientsfor responses in several 3D, continuous-energy Monte Carloapplications.

14:30 PMEnhancements in Continuous-Energy Monte CarloCapabilities for SCALE 6.2
B.T.Rearden, L.M.Petrie, D.E.Peplow, K.B.Bekar, D.Wiarda,
C.Celik, C.M.Perfetti, M.E.Dunn(1), S.W.D.Hart(2)1)Oak Ridge National Laboratory, Oak Ridge, USA, 2)University ofTennessee, Knoxville, USA
SCALE is a widely used suite of tools for nuclear systems modelingand simulation that provides comprehensive, verified and validated,

48

user-friendly capabilities for criticality safety, reactor physics, radiation
shielding, and sensitivity and uncertainty analysis. For more than 30
years, regulators, industry, and research institutions around the world
have used SCALE for nuclear safety analysis and design. SCALE
provides a “plug-and-play” framework that includes three deterministic
and three Monte Carlo radiation transport solvers that are select- ed
based on the desired solution. SCALE includes the latest nuclear
data libraries for continuous-energy and multigroup radiation transport
as well as activation, depletion, and decay calculations. SCALE’sgraphical user interfaces assist with accurate system modeling,visualization, and convenient access to desired results. SCALE 6.2
provides several new capabilities and significant improvements inmany existing features, especially with expanded continuous-energyMonte Carlo capabilities for criticality safety, shielding, depletion, andsensitivity/uncertainty analysis, as well as improved fidelity in nuclear
data libraries. A brief overview of SCALE capabilities is provided withemphasis on new features for SCALE 6.2.

14:50 PM

Leakage-Corrected Fast Reactor Assembly
Calculation with Monte-Carlo Code TRIPOLI4® and Its
Validation Methodology
L.Cai, Y.Pénéliau, J.Tommasi, J-F Vidal(1), C.M.Diop(2)1)CEA,DEN, DER/SPRC, Saint-Paul-lez-Durance, France, 2)CEA,
DEN,DANS/DM2S/SERMA, Gif-sur-Yvette Cedex, France
A leakage model based on B1 Homogeneous Equations has been
recently implemented in continuous-energy Monte Carlo code
TRIPOLI4®. This leakage model algorithm iterates between the pointwise Monte Carlo simulation and a B1 Homogeneous Equations
solver till reaching a final critical state in Monte Carlo simulation. The
two advantages of our leakage model compared with the others are:we use critical flux spectrum to generate the multi-group constantsfor solving the B1 Homogeneous Equations; the leakage coefficients
calculated are considered in point-wise Monte Carlo simulation. This
leakage model is validated by a predesigned numerical experiment
simulated with continuous-energy TRIPOLI4® and the obtained results
are also compared with those from SERPENT leakage model anddeterministic leakage model in ECCO code. Finally, our leakagecorrected multi-group constants are used in transport theory based
core calculation and they give out consistent multiplicative factor andneutronic balance.

15:10 PM

Impact of Nearest Neighbor Distribution of Fuel
Particle on Neutronics Characteristics in Statistical
Geometry Model
T.Koide, T.Endo, A.Yamamoto(1), K.Kirimura, K.Yamaji(2)1)Nagoya University, Nagoya, Japan, 2)Mitsubishi Heavy Industries, Ltd.,
Kobe, Japan
Effect of nearest neighbor distribution of fuel particle used in thestatistical geometry model (STGM) for the neutronics characteristics
calculation is investigated. The statistical uniform distribution issometimes assumed as a nearest neighbor distribution (NND)in STGM. This assumption is considered to be valid for the hightemperature gas reactor (HTGR) fuel compact, where small fine fuelparticles are distributed in the graphite matrix. However, the validityof this assumption is not confirmed for relatively large fuel debris
distributed in light water, which is one of the plausible situations ina light water reactor core with severe accident. In order to verify thevalidity of NND of statistical uniform distribution, two different types of
NNDs are numerically calculated: One is determined by the simple
cubic lattice where a single fuel sphere is located in the center of unit
lattice; another is obtained by random arrangement of fuel spheresin three dimensional space. Based on these NNDs, mean neutrontransmission probability values in moderator are evaluated. As a result,
we find that the mean transmission probability depends on the NNDs
even if the packing fraction of fuel particle and fuel particle radius
are identical, and the difference of mean transmission probability
among NNDs becomes larger as the optical path length in moderator
increases. Therefore, when the optical path length in moderator islarge, the simple cubic lattice geometry and the statistical uniform
distribution give significantly different neutron multiplication factors,
even if the packing fractions in both NNDs are identical. Consequently,
in order to properly estimate the neutronics characteristics such as theneutron multiplication factor, the appropriate NND should be provided
in the STGM calculation for the light water moderated system.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 13:30 PM - Cosmos

Track12-1 Fuel Cycle and Actinide Management
Session Chair: Nicholas Brown(BNL), Naoyuki Takaki(Tokyo City Univ.)
13:30 PM

Simulation of Fuel Cycles with Minor Actinide
Management Using a Fast Burnup Calculation Tool
M.Szieberth, M.Halász, S.Fehér, T.Reiss

We will then discuss of the results obtained through this method apply
to PWRs loaded with homogeneous and heterogeneous assemblies
using the following criteria: resource consumption, Pu inventories in
the cycle and waste production.

Institute of Nuclear Techniques, Budapest, Hungary
The paper presents a fast and flexible burnup model for fuel cycle
simulations which is based on the description of the one-groupcross-sections as analytic functions of the isotopic composition. This
was accomplished by multi-dimensional regression based on the
results of numerous core calculations. The developed model is able
to determine the spent fuel composition in reasonable CPU time,
and was integrated into a simplified fuel cycle model containing Gas
Cooled Fast Reactors (GFR) and conventional light water reactors(LWRs). The fuel cycle simulations revealed an advantageous effectof increased minor actinide content in the GFR core on the fuelutilization parameters. In order to explore the processes that lay
behind this effect the neutronics balance of the GFR was investigatedin equlibrium cycle conditions.

14:30 PM

13:50 PM

14:50 PM

N.E.Stauff, T.K.Kim, T.A.Taiwo

K.L.I.Björk, P.Fredriksson(1,2)

Variations in Activity, Toxicity and Decay Heat of
Nuclear Waste of Various Fuel Cycles

Evaluation Method of Equivalence Factors for MOX
Fuel and Non-Linear "Equivalent Pu-239" Formula
M.Tokashiki, S.Okui
Nuclear Fuel Industries, Ltd., JapanThe systematic method with least squares method to evaluate highaccurate equivalence factors, and non-linear “Equivalent Pu-239”
formula which is applicable to broad range of Pu isotopic composition,
have been developed. And the method to evaluate each MOX fuelrod type’s equivalence factors with high accurate equivalence for local
power peaking has been developed. The applicability of the methods
was confirmed for BWR MOX fuel.

Development of a Fuel Performance Code forThorium-Plutonium Fuel

Argonne National Laboratory, Argonne, USA
The variations in the values of activity, toxicity and decay heat ofnuclear waste, particularly spent nuclear fuel (SNF) and high-level
waste (HLW), have been assessed for 40 fuel-cycle examples at 10,100, and 100,000 years after reactor discharge. A correlation analysis
was performed to identify the parameters that have common trends
for the purpose of reducing the number of evaluation metrics requiredfor comparing the fuel cycles. Such an analysis in this paper allowed
focusing on the activity and inhalation toxicity at 10 years and 100,000
years after discharge. The variation in the 10 years activity is primarily
due to the variation in the quantity and specific activity of the fission
products. The specific activity of the fission products is consistently
higher for thorium fuel cycles and smaller when the fuel residence
time is long. The variation in the activity 100,000 years after discharge
is primarily explained by the quantity of U-233 and Pu-239 sent tonuclear waste, which is directly linked to the type of fuel used and to
the reprocessing scheme employed for these elements. The difference
between the inhalation toxicity and the activity is directly explained by
the variations in the effective dose coefficients. The heavy actinides
such as Pu, Am and Cm have a predominant effect, by 3 to 4 ordersof magnitude, on the inhalation toxicity. As a consequence, thorium
fuel-cycles always display low values of inhalation toxicity 10 years
after discharge. Uranium-plutonium fuel cycles display low 10 years
inhalation toxicity values only when their TRU content is being
reprocessed. Thorium fuel cycles that send U-233 into nuclear waste
are responsible for some of the largest values in the inhalation toxicity
at 100,000 years after discharge. However, the largest values in theinhalation toxicity 100,000 years after discharge are dominated byuranium fuel-cycles that send large amount of Pu-239 into the waste.

1)Thor Energy, Oslo, Norway, 2)Chalmers University of Technology,
Göteborg, Sweden
Thorium-plutonium Mixed OXide fuel (Th-MOX) is considered
for use as light water reactor fuel. Both neutronic and material
properties show some clear benefits over those of uranium-oxide anduranium-plutonium mixed oxide fuel, but for a new fuel type to belicensed for use in commercial reactors, its behaviour must be possible
to predict. For the thermomechanical behaviour, this is normally doneusing a well validated fuel performance code, but given the scarce
operation experience with Th-MOX fuel, no such code is available
today.
In this paper we present the ongoing work with developing a fuelperformance code for prediction of the thermomechanical behaviour of
Th-MOX for light water reactors. The wellestablished fuel performancecode FRAPCON is modified by incorporation of new correlations
for the material properties of the thorium-plutonium mixed oxide,and by develoment of a new subroutine for prediction of the radial
power profiles within the fuel pellets. This paper lists the correlations
chosen for the fuel material properties, describes the methodology
for modifying the power profile calculations and shows the resultsof fuel temperature calculations with the code in its current state ofdevelopment. The code will ultimately be validated using data from a
Th-MOX test irradiation campaign which is currently ongoing in the
Halden research reactor.

14:10 PM

Argonne National Laboratory, Argonne, USA
As part of ongoing efforts to assess nuclear fuel cycle options, four
fuel cycle options based on the same two reactor technologies havebeen studied. All four options are composed of two stages, one
which contains pressurized-water reactors (PWRs), and the other,fast spectrum accelerator-driven systems (ADS). The performance
characteristics and material mass flows have been determined forthe fuel cycle options considered, and compared. The three majordifficulties encountered when modeling and analyzing these fuel cycle
options have been to maintain the PWR fuel temperature reactivity
coefficient negative when multi-recycling MOX fuel, to design the
ADS core to be a breeder, and to achieve a high enough keff in the
ADS to avoid the accelerator power consumption to be larger thanthe power generated by the ADS core. The differences observed in
the performance characteristics and mass flows between the four
fuel cycle options analyzed are discussed in this paper. Overall it isfound that despite the four fuel cycle options being based on the same
reactor technologies and seemingly similar at first sight, they performdifferently and offer different features: resource utilization, need for
uranium enrichment, required reprocessing capacity, and material type
to be stored.

Effect of Heterogeneity in Plutonium Recycling in
Steady State PWR
M.Ernoult, S.David, X.Doligez(1), A.Nuttin, N.Capellan, O.Meplan(2),B.Leniau(3)1)Institute de Physique Nucléaire d'Orsay, Orsay, France, 2)LPSC,
Grenoble, France, 3)SUBATECH, Nantes, France
The possible delay of decades for the deployment of fourth generation
reactors brings up new issues. Countries like France which are storing
Plutonium for years in prediction of starting FBR could have to adapt
their plutonium management strategy. We have compared different
strategies for plutonium recycling in PWR reactors. We have chosen
to limit the scope of this study to PWRs considering the standard
uranium cycle, and we investigated the influence of the heterogeneity
in the assembly that would contain the recycled plutonium. The
comparison of different strategies is made at steady-state. This paper
present a specific method developed to allow complete and detailed
studies of equilibrium scenarios and how it has been implemented andintegrated inside the opensource MURE package.

15:10 PM

Two-Stage Fuel Cycles with Accelerator-Driven
Systems
F.Heidet, T.K.Kim, T.A.Taiwo

49

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 15:55 PM - Mizuho_A

Track1-6 Reactor Analysis Methods
Session Chair: Mohamed Elsawi(Khalifa University), Yunzhao Li(Xi'an Jiaotong University)
15:55 PM

Modeling of Shutdown Cooling Reactivity Effects with
SIMULATE
T.BahadirStudsvik Scandpower, Inc., Waltham, USA
During plant outages and subsequent return to power operations,the fuel isotopic inventory, relative to that of steady-state operation,changes as nuclides go through radioactive decay. Accurate
predictions of essential core neutronic parameters require that thereactivity impact of these isotopic changes be taken into account.
Extended outages, as experienced in Japan following the Fukushima
accident, as well as consecutive outages, which are close in time suchthat the residual effects from the first one influence the reacitivity of thesubsequent one, may challenge the reactivity models implemented
in three-dimensional (3D) core simulators. This paper studies themodeling issues with the core simulators for analyzing nominal andextended refueling shutdown outages as well as combination of suchwith mid-cycle outages. The shutdown cooling (SDC) reactivity models
that are implemented based on macroscopic (lumped) depletionand explicit isotope tracking are described for the Studsvik 3D coresimulators, SIMULATE-3 and SIMULATE5. Using relevant numericaltest cases, it is demonstrated that the explicit isotope-tracking model
offers robust and accurate prediction capabilities for computing
shutdown-cooling reactivity, regardless of the outage time.

16:15 PM

Implementation and Verification of the SDM in the
TITAN 3-D Sn Transport Code
N.J.Roskoff, W.Walters, A.Haghighat(1), C.Yi, G.Sjoden(2)1)Virginia Tech, Arlington, USA, 2)Georgia Tech, Atlanta, USA
The subgroup decomposition method (SDM) has recently beendeveloped as an improvement over the consistent generalizedenergy condensation theory for treatment of the energy variable indeterministic particle transport problems. By explicitly preserving
reaction rates of the fine-group energy structure, the SDM directly
couples a consistent coarse-group transport calculation with a setof fixed-source “decomposition sweeps” to provide a fine-group flux
spectrum. This paper will outline the implementation of the SDM into athree-dimensional, SN deterministic transport code such as TITAN. The
new version of TITAN, TITAN-SDM, is tested using a 1-D benchmark
problem based on the High Temperature Engineering Test Reactor(HTTR) in Japan. In addition to accuracy, this study examines theefficiency of the SDM algorithm in a 3-D SN transport code.

16:35 PM

Transport Core Solver Validation for the ASTRID
Conceptual Design Study with APOLLO3®
J-F.Vidal, C.Bay, P.Archier, J-M.Palau, G.Rimpault, B.Roque,J.Tommasi(1), A.Hébert(2)1)CEA Cadarache, Saint-Paul-lez-Durance, France, 2)EPM, Montreal,
CanadaIn this study, we are questioning the capability of 3D deterministictransport solvers available at CEA and EPM to accurately predict thevoid effect of the sodium plenum of the CFV core (GEN IV Sodium
Fast Reactor concept selected for the ASTRID project at CEA).
Discrete Ordinates, Spherical Harmonics and Simplified Spherical
Harmonics transport solvers have been tested against Monte Carlo
method on Takeda Benchmark Model 4 and on a new variant with asodium plenum. Sn solvers offer the best accuracy on multiplication
factor, void effect and control rod worth, with values very close toreference Monte Carlo ones. However, even using acceleration
techniques, computing time are quite consequent. This difficulty sould
be solved by the parallelization of the MINARET Sn solver in the
APOLLO3® platform. SPn solvers are very fast, but also quite distant from the reference, especially for the calculation of the void effectwhere they prove largely inadequate. A good compromise between
accuracy and computational time is obtained with the Pn solver
VARIANT. The corresponding nodal method is being implemented in
APOLLO3®.

50

16:55 PM

Methodology Assessment for the Evaluation of the
Coolant Void Worth in Sodium Fast Reactors with a
Low Void Effect Core Design
S.Bortot, S.Pelloni, K.Mikityuk
PSI, Villigen, Switzerland
Due to the recent renewed interest in Sodium Fast Reactor
cores featuring a low coolant void worth, an attempt was made toperform a methodology assessment for the evaluation of such nonconventional designs’ performance characteristics, which might not
be accurately predicted by traditional reactor physics deterministic
methods. The Monte Carlo code Serpent was employed to provide
the reference solutions, thanks to both its capability to model complex
3D geometries and its continuous energy representation of crosssections. Serpent results were then compared with ERANOS multigroup, 2D and 3D, transport and diffusion theory predictions. A
reasonably good agreement was found among the calculated nominal
reactor parameters (e.g., reactivity, effective delayed neutron fraction,
prompt neutron generation time) and coolant void worth involving
perturbations of the active core regions. Major discrepancies occur
when voiding both fuel zones and the large sodium plenum above
the core, though. In such cases, further investigations were carriedout concerning the impact of computational options at a crosssection generation level, for instance relative to spectrum and leakage
approximations for sub-critical regions. It may be concluded that theformer (i.e., 33 vs. 175 energy groups condensation) have an impactof the order of 200 pcm on absolute reactivity values, but does notaffect reactivity variations, whereas the latter (i.e., imposed bucklingvalues) play a substantial role, inducing differences between Serpent
and ERANOS of up to 35 %. Finally, the limits of variational nodal
methods in treating very low density zones were confirmed, being
particularly critical when employing fine energy group structures.

17:15 PM

Results of Verification of Computer Codes Used for
Analysis of BN-1200 Reactor Core Neutronics
S.Belov, M.Farakshin, A.Kiselyov, E.Marova (1), P.Alekseev,
V.Boyarinov, P.Fomichenko, V.Nevinitsa, A.Timoshinov, M.Zizin(2),I.Malysheva, A.Peregudov, K.Raskach, M.Semenov, V.Stogov,
A.Tsibulya(3)1)OKBM, Nizhny Novgorod, Russia, 2)National Research Centre“Kurchatov Institute”, Moscow, Russia, 3)Institute for Physics and PowerEngineering, Kaluga Region, Russia
The paper presents the results of neutronic calculations of a test
model of the BN-1200 reactor core with the use of engineering codesand codes based on the Monte Carlo method.

17:35 PM

APOLLO3® Based Method for 3D Warped Cores
Calculations – Application to Flowering Tests of
PhenixC.Patricot, D.Broc, E.Hourcade, K.Ammar
CEA, Gif-sur-Yvette, France
In the frame of multiphysics coupling tools development, amethodology for deformed cores calculations with APOLLO3®, aneutronic code in development at CEA Saclay, was developed. Thismethodology, based on three dimensional pixels, was applied to
Phénix flowering tests performed in 2009 and 2012, during which
reactivity and displacement were measured.
This paper first presents these flowering tests and the results ofprevious study, using RZ symmetry and ERANOS, which only
reproduces roughly the measurements. Then, our methodology is
explained, assessed with different pixel sizes and validated by acomparison with TRIPOLI-4® results. Some academic deformations
are also calculated in order to analyze the impacts of heterogeneities,
and we confirm, in particular, that the bending of assemblies can beneglected. Afterwards, mechanical calculations of flowering tests, donewith CAST3M, are very briefly described. Finally our results are given,for central and peripheral flowerings, and with different hypotheses ongaps between assemblies.
The paper concludes that real core compaction is probably driven byirradiation deformations of assemblies, and can be approximated bygaps between assemblies.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 15:55 PM - Mizuho_B

Track4-2 Verification, Validation and Uncertainty Analysis
Session Chair: Bassam Khuwaileh(NCSU), Kensuke Kojima(JAEA)
15:55 PM

Benchmark Calculation with MOSRA-SRAC for
Burnup of a BWR Fuel Assembly
K.Kojima, K.Okumura
Japan Atomic Energy Agency, Ibaraki, JapanThe Japan Atomic Energy Agency has developed the Modular
Reactor Analysis Code System MOSRA to improve the applicabilityof neutronic characteristics modeling. The cell calculation module
MOSRA-SRAC is based on the collision probability method and is
one of the core modules of the MOSRA system. To test the module
on a real-world problem, it was combined with the benchmark
program “Burnup Credit Criticality Benchmark Phase IIIC.” In
this program participants are requested to submit the neutronic
characteristics of burnup calculations for a BWR fuel assembly
containing fuel rods poisoned with gadolinium (Gd2O3), which issimilar to the fuel assembly at TEPCO’s Fukushima Daiichi Nuclear
Power Station. Because of certain restrictions of the MOSRA-SRAC
burnup calculations part of the geometry model was homogenized.
In order to verify the validity of MOSRA-SRAC, including the effectsof the homogenization, the calculated burnup dependent infinite
multiplication factor and the nuclide compositions were compared withthose obtained with the burnup calculation code MVP-BURN which
had already been validated for many benchmark problems. As a resultof the comparisons, the applicability of MOSRA-SRAC module for the
BWR assembly has been verified. Furthermore, it can be shown thatthe effects of the homogenization are smaller than the effects due tothe calculation method for both multiplication factor and compositions.

16:15 PM

Verification of the COCAGNE Core Code Using
Cluster Depletion Calculations
F.Hoareau, E.Girardi, C.Brosselard, M.Fliscounakis
EDF R&D, Clamart, France
EDF/R&D is developing a new calculation scheme based on thetransport-Simplified Pn (SPn) approach. The lattice code used is thedeterministic code APOLLO2, developed at CEA with the supportof EDF and AREVA-NP. The core code is the code COCAGNE,
developed at EDF R&D. The latter can take advantage of a
microscopic depletion solver which improves the treatment of spectral
history effects. This solver can resort to a specific correction basedon the use of the Pu239 concentration as a spectral indicator. In
order to evaluate the improvements brought by this Pu239 correction
model, one uses (3x3 assemblies) cluster depletion calculations as
test-cases. UOX and UOX/MOX clusters are both considered. As areference, APOLLO2 depletion calculations of these clusters, usinga critical boron (CB) search scheme at each calculation step, areperformed. This choice of methodology (using CB search instead ofa fixed average CB) enables to highlight historical spectral effects
related to the boron concentration. This methodology is also moreconsistent with the depletion calculation of real cores. Pin by pin
COCAGNE calculations are performed and compared with the
APOLLO2 results. The analysis of the results obtained shows that theboron concentration computed by COCAGNE gets more consistent
with APOLLO2 when the Pu239 corrector is used, especially for UOX/
MOX clusters. As for pin power distribution, the use of the Pu239
model also enables to reduce slightly the gap between APOLLO2 and
COCAGNE. This work will be extended to clusters with gadoliniumpoisoned fuel assemblies and reflector regions.

16:35 PMLWR Fuel Reactivity Depletion Verification Using 2D
Full-Core MOC and Flux Map Data
G.A.Gunow, K.S.Smith, B.Forget
Massachusetts Institute of Technology, Massachusetts, USA
Experimental quantification of PWR fuel reactivity (k-infinity) burnup
decrement biases and uncertainties using in-core flux map data fromoperating power reactors has previously been conducted employing
analytical methods to systematically determine experimental fuelreactivities that best match measured fission rate distributions. This
optimal core reactivity distribution that best matches the measured
fission rate distribution is assumed to be associated with the true fuelreactivity distribution. Some parties have questioned whether fortuitouscancellation of errors between various approximations inherent in the
3D nodal diffusion core analysis models might have caused reactivity

decrement biases and uncertainties to be unrealistically small. In
this study, the BEAVRS benchmark is modeled with both 2D, fullcore, multi-group transport calculations and 2-group nodal diffusion
calculations. The calculated in-core detector U-235 fission rates arecompared with measured fission rates supplied in the benchmark.
The same analytical methods previously mentioned are again used
to obtain fuel reactivity biases and uncertainties. Results demonstratethat fuel batch reactivities inferred from flux map data using full-coretransport calculations are nearly identical to those inferred using nodal
diffusion calculations. Consequently, nodal methods do not contributesignificantly to reactivity decrement biases in the BEAVRS cores. It
is recommended that fuel reactivity biases and uncertainties inferredfrom 3D nodal diffusion calculations remain valid.

16:55 PMCASMO-4E and CASMO-5 Analysis of the Isotopic
Compositions of the LWR-PROTEUS Phase II Burnt
PWR UO2 Fuel Samples
P.Grimm, G.Perret, H.Ferroukhi
Paul Scherrer Institute, Villigen, Switzerland
The isotopic compositions of 7 UO2 samples irradiated in a Swiss PWR
power plant, which were investigated in the LWR-PROTEUS Phase
II programme, were calculated using the CASMO-4E and CASMO-5
assembly codes and cross-section libraries based on ENDF/B-VI
and ENDF/B-VII release 1, respectively. The burnups of the samples
span a very wide range from ~40 to ~120 MWd/kg. The results for alarge number of actinide and fission product nuclides were comparedto those of chemical analyses performed using a combination ofchromatographic separation and mass spectrometry. A satisfactory
overall agreement of calculated and measured nuclide concentrations
has been found for both codes. Significant improvements over
CASMO-4E can be seen in the CASMO-5 results for the main Pu
isotopes and for the lanthanides Sm, Eu and Gd. Considering thevery high burnups of some of the samples, it is important to note thatthe differences between calculated and measured concentrations do
generally not show a trend to increase for higher burnups.

17:15 PM

Experimental Validation of Decay Heat Calculations
with VESTA 2.1
W.Haeck, R.Ichou, F.Ecrabet, J.Dannet
IRSN, Fontenay-aux-Roses, France
The removal of decay heat is a significant safety concern in nuclear
engineering for the operation of a nuclear reactor both in normal
and accidental conditions and for intermediate and long term waste
storage facilities. The correct evaluation of the decay heat produced
by an irradiated material requires first of all the calculation of thecomposition of the irradiated material by depletion codes such as
VESTA 2.1, currently under development at IRSN in France. The
use of this software for decay heat applications requires a consistentexperimental validation effort. For this purpose, PWR assembly
decay heat measurements have been selected from the CLAB, GE
and HEDL data as well as fission experiments. Using the JEFF 3.1
and ENDF/BVII. 0 nuclear data libraries, VESTA 2.1 calculates theassembly decay heat for almost all cases within 4 % of the measured
decay heat. On average, the JEFF 3.1 calculated decay heat value
appear to give a systematic underestimation of about 2 %. Whenusing ENDF/B-VII.0, this is limited to an average underestimation of 0.5%.

17:35 PM

Development and Validation of Ad Hoc ORIGEN-ARP
Libraries for Very High Burnup UO2 PWR Fuel with
SCALE/TRITON
S.Caruso(1), A.-L.Panadero(2)1)National Cooperative for the Disposal of Radioactive Waste (NAGRA),
Wettingen, Switzerland, 2)Ecole Polytechnique Fédérale de Lausanne(EPFL), Lausanne, Switzerland
The estimation of the nuclide inventory of the spent fuel to bedisposed of in the high-level waste geological repository in Switzerland
is fundamental for the realization of the repository itself. The fuelcharacterization is relevant not only for safety-relevant issues
concerning the repository, but also for all the issues related to the51PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 15:55 PM - Mizuho_B

Track4-2 Verification, Validation and Uncertainty Analysis
Session Chair: Bassam Khuwaileh(NCSU), Kensuke Kojima(JAEA)
previous phase, consisting of fuel handling and encapsulation. In
order to keep the uncertainties sufficiently low, a fuel characterization
methodology needs to be defined and validated. The large
heterogeneity of the fuel used in the five Swiss reactors makes thistask highly complex.
UO 2 PWR fuels have been simulated using SCALE/TRITON
employing the latest ENDF/B-VII cross-section data, aiming to produce
ORIGEN-ARP libraries suitable for very high burnup fuels. The results
obtained from TRITON and ORIGEN-ARP calculations are comparedto the results of previous experimental studies. The experimental
results considered here were selected from chemical isotopic analysis
and passive neutron emission performed earlier at the Paul Scherer
Institute in Switzerland. A set of fuel rod samples with different burnups
was used for the analysis. Very high burnup samples have been
investigated, particularly for the neutron output. The results show thatthe developed ORIGEN-ARP libraries provide a valid alternative tovery long and detailed fuel simulations without loss of accuracy in thedepletion capabilities.

52

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 15:55 PM - Hiei

Track2-4 Deterministic Transport Theory
Session Chair: Ricardo Barros(Univ. do Estado do Rio de Janeiro ), Tetsuo Matsumura(CRIEPI)
15:55 PM

Phase Space Bases for Response Matrix Methods
J.A.Roberts, R.L.Reed(1), B.Forget(2)1)Kansas State University, Kansas, USA, 2)Massachusetts Institute ofTechnology, Massachusetts, USA
New basis functions in space, angle, and energy are developed foruse in the eigenvalue response matrix method (ERMM) . ERMM
solves reactor eigenvalue problems by spatially decomposing a modeland linking the resulting independent nodes through approximateboundary conditions. The conditions are defined via a set of nodal
transport calculations for which the forcing function is a tensor productof basis functions in all phase space variables on one nodal surface(with vacuum elsewhere). Traditional implementations of ERMM haveused the standard bases of mathematical physics, including Legendre
polynomials and their discrete analogs. Alternatively, new bases thatincorporate some physics a priori are shown to yield satisfactory
results with far lower expansion orders than is possible with more
traditional bases. In particular, a very simple spatial basis incorporating
pin-dependent spatial variation reduces partial current errors by nearly
an order of magnitude over a standard basis. In angle, conservative
bases with appropriate angular quadratures outperform simpler bases.
In energy, incorporating representative spectral information frominfinite medium calculations results in a basis that can achieve sub-1%
pin fission rate errors with as low as a 5th order expansion in energy
for a 44- group problem. The results indicate that highly-accurate, loworder response matrix solutions should be feasible for reactor physics
analysis.

16:15 PMThe Drift Diffusion Limit of Thermal Neutrons:
Theoretical and Numerical Results
P.A.Vaquer, R.G.McClarren, M.L.Adams
Texas A&M University, TX, USA
This study presents a new approach to modeling neutron
thermalization which can greatly simplify nuclear reactor simulations.
A drift-diffusion equation is derived through an asymptotic analysisof the 1-D energy-dependent transport equation. Two test problems
were used to provide numerical comparisons of this drift-diffusion
approximation to MCNP6 calculations. Both of these test problems
include a large 1-D graphite moderator with a steady-state temperature
gradient. The results demonstrated that the drift-diffusion is able tocapture energydependent effects and can successfully be used formodeling neutron thermalization in a large moderating material.

16:35 PM

Adequacies of Different Convergence Accuracy
Measures in Full-Core Nodal Flux Computations
R.van Geemert
Neutronics Department Erlangen, AREVA GmbH, Erlangen, Germany
This article presents a computational analysis that emphasizes theimportance of achieving convergence rate invariance in deriving truly
dependable error estimates for use in convergence monitoring in
iterative nodal flux solution processes. It is argued and shown that
this property is not fulfilled by every known formula in common use
for estimation of the convergence accuracy. Based on a rigorous
theoretical convergence analysis context, a generalized convergenceerror measure concept is derived that avoids remaining computational
uncertainties that may otherwise arise due to convergence rate
influences combined with the use of a deficient error measure. A
pursued in-depth convergence analysis has confirmed that the use ofthis derived convergence error measure concept enables an optimum
convergence monitoring process. With this generalized convergencecontrol concept in place, the delivered nodal flux solution can be
expected to truly and dependably adhere to any imposed convergenceaccuracy requirement. Simultaneously, the computational effort(i.e. numbers of iterations) required for establishing that accuracy
is automatically avoided to evolve beyond minimal necessity.
This concept is in standard use in AREVA’s core simulation code
ARTEMISTM in which it enforces a significantly enhanced convergenceaccuracy.

16:55 PM

Flexible Semi-Analytical Calculation Method of
Escape Probability
T.Matsumura
CRIEPI, Tokyo, JapanA new semi-analytical calculation method of escape probability isdeveloped. Application for an infinite cylinder is introduced in present
paper. Present calculation method has advantage in flexibility of
application for various shapes of calculation cells. By approximating
the integration result of the direction of z by simple quadratic formula
divided into ranges, it became possible to perform integration to y
direction analytically. Numerical result of escape probability showshigh precision (less than 0.1%) by present semi-analytical calculation
method. Present calculation method has flexible applicability forvarious two-dimensional shapes. Also it is thought that possibility of ananalytical solution for the rector physics calculation is still large.

17:15 PM

Corrected Diamond Difference Method for Couplingfrom the Method of Characteristics to Discrete
Ordinates
M.T.H.Young, B.Collins, W.R.Martin
University of Michigan, Michigan, USA
A new SN formulation has been developed which is effective atpreserving the behavior of a coupled fine-mesh transport solver on acoarse, pin homogenized Cartesian mesh. The method uses angleandenergy-dependent correction factors and a modified formulation
for the diamond difference equations, called corrected diamonddifference (CDD). An extension of the CDD method from previous
work includes a new cross section correction step. It is shown thatthe new CDD approach is equivalent in two-dimensional problems intheory and practice. The CDD method is then applied to several threedimensional benchmark cases using one-way 2-D/3-D coupling, and isshown to provide more accurate results as compared to the previous
CDD formulation and diamond difference alone.

17:35 PM

Energy Multigroup Spectral Green's Function
Constant Nodal Method for Fixed-Source SN Problems
in X,Y-Geometry
W.A.Menezes, H.A.Filho, R.C.Barros
Universidade do Estado do Rio de Janeiro, Nova Friburgo, Brazil
Presented here is a spectral nodal method for the numerical solutionof energy multigroup, fixed-source, discrete ordinates (SN) problems
in two-dimensional rectangular geometry. This analytical coarse-meshmethod is referred to as the multigroup spectral Green ́s functionconstant nodal (SGF-CN) method as it uses the multigroup SGF
method for numerically solving the one-dimensional transverseintegrated SN nodal equations with constant approximations for thetransverse leakage terms. The only approximations in the present
X,Y-geometry nodal method occur in these transverse leakage terms,
as the energy-group transfer scattering source terms are treatedanalytically within the offered method. Numerical results to a typicalmodel problem are given to illustrate the method ś accuracy.

53

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 15:55 PM - Atago

Track3-3 Monte Carlo Methods
Session Chair: Daniel J. Kelly(Knolls Atomic Power Laboratory), Zeguang Li(Tsinghua University)
15:55 PM

Large-Scale Monte Carlo Calculations with ThermalHydraulic Feedback
A.Ivanov, V.Sanchez, R.Stieglitz(1), K.Ivanov(2)1)Karlsruhe Institute of Technology, Institute of Neutron Physics andReactor Technology, Eggenstein-Leopoldshafen, Germany, 2)The
Pennsylvania State University, Department of Mechanical and Nuclear
Engineering, Pennsylvania, USA
Monte Carlo based codes provide the most accurate solution of theparticle transport problem. Individual particle trajectories are followed,
and the interaction physics is simulated using detailed modelingof the physical reaction. The calculations are usually done using
uniform temperature and density distributions. This is a significant
approximation and leads to significantly distorted solution when applied
to hot full power conditions. In this paper a method for introducing
the thermal-hydraulic feedback by dynamic material distributions is
introduced. The global variance reduction technique has been used
to optimize the power tallying. The fission source convergence wasaccelerated by applying the Wielandt’s acceleration method. Since
the aim of this work is to solve coupled neutronic/thermal-hydraulicproblems a convergence acceleration strategy based on stochastic
approximation was proposed. The coupled system was applied to aquarter PWR core at pin and sub-channel level resolution.

16:15 PM

Sodium Void Reactivity Effect Analysis Using the
Newly Developed Exact Perturbation Theory in MonteCarlo Code TRIPOLI-4®
G.Truchet, P.Leconte, J.M.Palau, P.Archier, J.Tommasi,
A.Santamarina, Y.Peneliau(1), A.Zoia, E.Brun(2)1)CEA, DEN, DER/SPRC, Cadarache Saint Paul Lez Durance, France,
2)CEA, DEN, DM2S/SERMA, Saclay Gif-sur-Yvette-Cedex, France
The analysis of void reactivity effect is prominent interest for Sodiumcooled Fast Reactor (SFR) safety. Indeed, in case of sodium leakageof the primary circuit, void reactivity represents the main passive
negative feedback to ensure reactivity control. The core can be
designed to maximize neutron leakage and lower the average neutron
multiplication factor in the event of sodium disappearing from withinassemblies. Thus, the nuclear chain reaction is stopped. The mostpromising solution is to place a sodium region above the fuel in orderfor neutrons to be reflected when the region is filled and escape when
the region is empty. In terms of simulation, this configuration is achallenge for usual calculation schemes:
1. Deterministic codes are typically limited in their ability tohomogenize a sub-critical medium as the sodium plenum.
2. Monte Carlo codes are typically not able to split the total
reactivity effect on different components, which prevents to achievestraightforward uncertainty analysis. Furthermore, since experimental
values can sometimes be small, Monte Carlo codes may not converge
within a reasonable computation time.
A new feature recently available in the Monte Carlo TRIPOLI-4®
based on the Exact Perturbation Theory allows very small
reactivity perturbations to be computed accurately as well asreactivity effect to be estimated on distinct isotopes cross-sections.
In the first part of this paper, this new feature of the code is
described and then applied in the second part to a core configuration
composed of several layers of fuel and fertile zones below a sodium
plenum. Reactivity and its contributions from specific reactions andenergy groups are calculated and compared with the results of thedeterministic code ERANOS. The aim of this work is twofold: (1)Achieve a numerical validation of the new TRIPOLI-4® features and(2) Identify where deterministic codes might be less accurate and why– even when using them at full capacity (S16 transport, 1968 energy
groups or more).

16:35 PMMonte Carlo Perturbation Analysis on Isothermal
Temperature Reactivity Coefficient of Light-Water
Moderated and Reflected Critical Assembly
B.K.Jeon, H.J.Shim(1), C.H.Pyeon(2)1)Seoul National University, Seoul, Korea, 2)Kyoto University, Kumatori,
Japan
Experiments have been carried out on the isothermal temperature54

reactivity coefficient (ITRC) for the light-water moderated core at the
Kyoto University Critical Assembly. The temperature effect on reactivity
is analyzed by the Seoul National University Monte Carlo (MC) code,
McCARD, which well reproduce experimental data. The contributionsof the each isotope by the density changes of the core and reflector
regions and the microscopic cross section changes to the ITRCs arequantified by sensitivity analyses based on the MC adjoint-weighted
perturbation methods.

16:55 PMMonte Carlo and Thermal-Hydraulic Coupling via
PVMEXEC
D.F.Gill, D.L.Aumiller, D.P.Griesheimer
Bechtel Marine Propulsion Corporation, Pennsylvania, USA
Successful high-fidelity coupling between a Monte Carlo neutrontransport solver and a subchannel thermal-hydraulics solver has
been achieved using PVMEXEC, a coupling framework developed foranalysis of transient phenomenon in nuclear reactors. The PVMEXEC
framework provides a generic program interface for exchanging
data between solver kernels for different physical processes, suchas radiation transport, heat conduction, and fluid flow. In this study,
PVMEXEC was used to couple the in-house Monte Carlo radiationtransport code, MC21, with a locally modified version of COBRA-TF.
In this coupling scheme, MC21 is responsible for calculating threedimensional power distributions and COBRA-TF for calculating local
fluid temperatures and densities, as well as fuel temperatures. The
coupled system was used to analyze 3D single-pin and assemblymodels based on the Calvert Cliffs commercial PWR. Convergenceproperties of the coupled simulations are examined and results arecompared to simulations conducted using the existing integrated
thermal feedback kernel in MC21.

17:15 PM

Perturbation Based Monte Carlo Criticality Search in
Density, Enrichment and Concentration
Z.Li, K.Wang, J.Deng
Tsinghua University, Beijing, China
Criticality search is a very important aspect in reactor physics analysis.
Due to the advantages of Monte Carlo method and the developmentof computer technologies, Monte Carlo criticality search is becoming
more and more necessary and feasible. Existing Monte Carlo criticality
search methods need large amount of individual criticality runs andmay have unstable results because of the uncertainties of criticality
results. In this paper, a new perturbation based Monte Carlo criticality
search method is proposed and discussed. This method only needs
one individual criticality calculation with perturbation tallies to estimate
keff changing function using initial keff and differential coefficientsresults, and solves polynomial equations to get the criticality search
results. The new perturbation based Monte Carlo criticality searchmethod is implemented in Monte Carlo code RMC, and criticality
search problems in density, enrichment and concentration are takenout. Results show that this method is quite inspiring in accuracy andefficiency, and has advantages compared with other criticality searchmethods.

17:35 PMMonte Carlo Perturbation Method for Geometrical
Uncertainty Analysis Using McCARD
H.J.Park, J.Y.Cho, J.S.Song(1), H.J.Shim(2)1)Korea Atomic Energy Research Institute, Daejeon, Korea, 2)Seoul
National University, Seoul, Korea
Monte Carlo geometrical uncertainty analysis method based on theMonte Carlo number density perturbation method is introduced. In
the method, all the regions of a given system are divided into theperturbed region and non-perturbed regions. In the perturbed region,
the MC number density perturbation calculations are performed. Inorder to examine the performance of the MC geometrical uncertainty
analysis, the accuracy of σ(keff) estimated by the MC number density
perturbation techniques are validated in Godiva, PWR fuel pin, and
HMF-030 ICSBEP benchmark problem.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Tuesday September 30, 2014 - 15:55 PM - Cosmos

Track12-2 Fuel Cycle and Actinide Management
Session Chair: Alberto Talamo(ANL), Kazufumi Tsujimoto(JAEA)
15:55 PMbeen studied for the (Th,Pu)O2 case.

F.Heidet, T.K.Kim, T.A.Taiwo

16:55 PM

Thorium-Fueled Breed-and-Burn Fuel Cycle
Argonne National Laboratory, Argonne, USA
The feasibility to sustain a breed-and-burn mode of operation
with thorium fuel in a sodium-cooled fast reactor is assessed. It isnecessary to use fissile support to sustain the breed-and-burn modewith thorium based fuel due to the poor neutron economy of thorium.
A thorium-fueled B&B core concept with low-enriched uranium(LEU) support is proposed and the performance characteristics ofthis core and the associated fuel cycle are assessed. The core isfueled with natural thorium and LEU. The thorium fuel is irradiatedfor seven cycles and is discharged with a burnup of ~38%. The LEU
fuel is irradiated for only one cycle and discharged with a burnup
of ~13%. About 72% of the energy is produced in the thorium fueland the remaining 28% in the LEU fuel. The resource utilization is
improved compared to light-water reactors but is not as good as thatof conventional once-through sodium-cooled fast reactors. Other
benefits typically observed in thorium-based fuel cycles, such as morefavorable reactivity coefficients, have not been assessed as part ofthis study.

16:15 PM

Fuel Cycle Analysis of Self-Sustaining Water Cooled
Reactors with 232Th/233U Fuel and Impact of 233U (n, γ)
Cross Section Evaluations
N.R.Brown, G.Raitses, M.Todosow
Brookhaven National Laboratory, New York, USA
This paper presents the analysis of a nuclear fuel cycle based on
an intermediate spectrum water-cooled reactor with 232Th/233U oxidefuel. The objective of the analyzed fuel cycle is to attain equilibrium
operation with only fertile 232Th resource feed while using water
reactor technology. The analysis is performed in the context of a large
multi-laboratory effort to develop a compendium of fuel cycle
performance data. Several potential reactor systems are identified in
the literature and evaluated for deployment in the proposed fuel cycle.
The investigation considers concepts where more than 50% of fissions
occur in the intermediate energy regime, defined for the purposesof the study as 1 eV to 0.1 MeV. One configuration, a reduced
moderation BWR proposed by UC Berkeley, meets the minimum
criteria for discharge fissile inventory ratio when accounting for losses(>1.012). The natural resource required is 165.5 t of fertile feed per
100 GWe-yr, more than two orders of magnitude lower than thenatural resource requirement for a typical light water reactor in a once
through cycle. The Berkeley RBWR-Th concept has been identified ashaving potential issues due to lack of acceptable shutdown margin, so
we also investigated an alternative D2O-cooled PWR-like concept andshowed that it has very similar fuel cycle performance. We also foundthat the neutronic performance of intermediate spectrum 232Th/233U
fuelled reactors are highly sensitive to the 233U radiative capture crosssection in the intermediate energy range. This behavior had beenpreviously observed in the literature for molten salt fast reactors. The
discrepancy between different cross sections evaluations coupled withthe lack of available measured data in this energy regime indicatesthat measurements of the 233U radiative capture cross section would
be desirable. Integral criticality experiments are also warranted for
232
Th/233U systems with intermediate spectra.

16:35 PM

Production of 232U from Irradiation of Standard andThorium-Based Fuels in PWR Reactors
B.Leniau, M.Ernoult, X.Doligez, J.N.Wilson
Institut de physique nucléaire d'Orsay, Orsay, France
The production of small quantities of 232U can induce radiation
protection issues in the back end of the fuel cycle, particularly forthorium-based fuels. This is due to its relatively short half life (69years) and the emission of a high energy gamma ray of 2.6 MeV at theend of its decay chain. With the depletion code MURE, we determine
the different reactions pathways, and their proportions, leading to thesynthesis of 232U in UO2 and (Th,Pu)O2 fuels irradiated in a PWR.
Moreover, the impact, on the 232U production, of cycle times such as
time separating the fabrication of the fuel and its irradiation as well as
influence of the fissile content has been investigated for UO2 fuel. The
impact of the thorium ore provenance and of the plutonium quality has

Fuel Cycle Scheme Design and Evaluation forThorium-Uranium Breeding Recycle in CANDU
Reactors
B.Yang, J.Shi, G.Bi, C.Tang
Shanghai Nuclear Engineering Research and Design Institute, Shanghai,
China
This work is aimed to develop a fuel cycle scheme for thorium-uranium
breeding recycle in CANDU reactors. According to the neutronics
theory on the breeding of fissile nuclides in nuclear fuel system, theprerequisite of 232Th-233U breeding is revealed in CANDU fuel system
and a roadmap of thorium-uranium breeding recycle in CANDU
reactors is promoted. Following the roadmap, the thorium-based fuelcores were designed and the neutronics performance was evaluated.
A CANDU-6 core with rated thermal power of 2061 MW was taken as
the reference core. Based on the fuel management strategy, the spentfuel characteristics for employed fuel types were assessed, including
radiotoxicity, decay heat and gamma rays. Finally, the thorium-uranium
breeding recycle schemes were established and evaluated in contrast
with natural uranium once-through cycle, widely applied in operational
CANDU reactors. The design and evaluation results indicated that
the Th-U breeding recycles is technically feasible in current CANDU
reactors.

17:15 PM

An Inventory Analysis of Thermal-Spectrum
Thorium-Fueled Molten Salt Reactor Concepts
J.J.Powers, J.C.Gehin, A.Worrall, T.J.Harrison, E.E.Sunny
Oak Ridge National Laboratory, Oak Ridge, USA
Inventory analyses of thermal-spectrum, thorium-fueled molten salt
reactors (MSRs) have been performed to support US Departmentof Energy fuel cycle screening and evaluation activities within the. A
single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept witha fuel-bearing molten fluoride salt moderated by graphite was usedas the basis for this work. Depletion calculations were performed
using SCALE 6.1.1 with ENDF/B-VII.0 nuclear data. Equilibrium
conditions were evaluated for several design parameter sets using amethodology developed at Oak Ridge National Laboratory (ORNL)
that enables MSR analysis by performing multiple SCALE/TRITON
depletion calculations with material flow modeling calculations
between time steps. Adequate modeling approximations wereidentified by comparing results obtained from calculations that useddifferent modeling choices and levels of fidelity. Parametric analyses
examined the performance sensitivity of a thorium MSR to different
separations approaches and elemental removal efficiencies. Finally,an inventory analysis for a thorium-fueled MSR with full recycling
demonstrated how these insights can be applied and showed thatsuch a system appears feasible from a mass flow and reactivity basis.

17:35 PM

Agent-Based Dynamic Resource Exchange in
CYCLUS
M.J.Gidden, R.W.Carlsen, A.Opotowsky, O.Rakhimov,
A.M.Scopatz, P.P.H.Wilson
University of Wisconsin - Madison, Wisconsin, USA
A novel methodology for modeling the dynamic exchange of resourcesamong actors in the CYCLUS fuel cycle simulator is described.
The methodology is comprised of an information gathering step, abipartite-graph-based supply-demand matching step, and a trade
execution step. Its implementation in CYCLUS allows the simulator
to model generic fuel cycles, i.e., those in which the types offacilities and possible resource flows is not known a prior. The flowof resources at each time step is algorithmically determined basedon facilities’ resource flow preferences. The dynamicsim of both thepreference-based flows and quantity and quality constraints on thoseflows are demonstrated via two simple scenarios. The first scenario
exemplifies situations in which actors’ preferences change over time
and the second typifies situations in which quality-based constraints
limit resource flow. This generic capability provides a sufficient
framework and basis for the CYCLUS simulation engine to model
complex, advanced fuel cycles.

55

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Mizuho_A

Track1-7 Reactor Analysis Methods
Session Chair: Eleodor Nichita(University of Ontario Institute of Technology), Chang Joon Jeong(KAERI)
8:00 AMHigh Order Source Approximaion for the EFEN
MethodY.Li, L.Cao, X.Yuan
Xi'an Jiaotong University, Shaanxi, China
The flat source approximation in one dimensional Exponential
Function Expansion Nodal (EFEN) method is extended to a highorder polynomial approximation while maintaining the simplicity of
the nodal response matrix. By applying the new method to a one
dimensional PWR pin-by-pin problem, it has been observed that
quadratic source approximation is good enough for PWR pin-by-pin
calculation, while the flat source approximation causes about 5% ofrelative error to the thermal flux. By applying the new method to a one
dimensional assembly homogenized problem, it has been found that
the EFEN method with cubic source approximation can be employed
to handle PWR core diffusion problems. Numerical results suggest the
optimization of source approximation order for different energy groups
and different spacial locations to achieve more accurate results withless computing effort.

8:20 AM

Extension of Linear Source MOC Methodology toAnisotropic Scattering in CASMO5
R.M.Ferrer, J.D.Rhodes
Studsvik Scandpower, Inc., Idaho Falls, USA
The Linear Source (LS) spatial approximation, previously implemented
in the Method of Characteristic (MOC) transport solver under
the assumption of isotropic scattering, is extended to account foranisotropic scattering. The generalized formulation results in the
representation of the angular moments of the flux as a linearlyvarying function within each source region and is fully compatible with
Coarse-Mesh Finite Difference (CMFD) acceleration. The main set of
equations for this anisotropic-source LS MOC method are derived inthis work and numerical results are presented for critical assembly test
cases. The results show improvement in the accuracy of the transport
solution, but also an increase in run time and storage relative to the LS
isotropic scattering implementation.

8:40 AM

Finite Difference Equations for Neutron Flux andImportance Distribution in a Heterogeneous Reactor
without Homogenization and Diffusion Approximation
A.V.Elshin
Alexandrov Research Institute of Technology, Russia
This paper describes an application of the surface harmonics method
to derivation of fewgroup finite difference equations for neutron flux
distribution in a 3D square-lattice reactor model. The Boltzmann
neutron transport equation is used as the original equation. No
homogenization is used because the method first solves a systemof finite difference equations (with a spatial step corresponding to alattice pitch) and then constructs a detailed neutron flux distribution
in the reactor cells with assumed approximations. Non-diffusion
approximations apply to calculation of a whole reactor core if weincrease the number of trial functions for describing the neutron
flux distribution in each cell and the size of the matrices of the
few-group coefficients for finite difference equations. Few-groupfinite difference equations are derived, which describe the neutron
importance distribution (the multiplication factor in the homogeneous
eigenvalue problem) in the reactor core. The derived finite difference
equations remain adjoint to each other like the original equation ofneutron transport and its adjoint equation. Difficulties and possibility ofcalculating the neutron flux distribution (trial functions) in cells of thereactor core model are discussed.

9:00 AM

Efficient Subspace Construction for Reduced Order
Modeling in Reactor Analysis
B.A.Khuwaileh, C.Wang, Y.Bang, H.S.Abdel-Khalik
North Carolina State University, Raleigh, USA
Subspace-based reduced order modeling (ROM) is a powerful
technique for reducing the computational burden required for the56

analysis of complex engineering models, such as nuclear reactor corecalculations. It identifies a relatively small subspace that represents thevariations for the quantities of interest with quantifiable user-defined
accuracy. Focusing on neutron transport calculations employed inreactor analysis, the subspace is defined by a set of basis vectorsthat are extracted from the converged flux solution associated with aset of randomized forward model executions. In each execution, theinput cross-sections are randomly perturbed, and the corresponding
converged flux solution, referred to as snapshot, is recorded. This
work mathematically proves and numerically demonstrates that the
non-converged flux iterates can be used to approximate the reduction
subspace without compromising the user-defined accuracy. It isshown that although this is expected to be bigger than (and inclusiveof) the one based on converged flux snapshots, the computationalsavings resulting from the early termination of the iterative solution are
significant even for situations where the convergence is slow, such as
for systems with high dominance ratio (e.g., BWR). A quarter BWR
fuel assembly was modeled and the variations of multiplication factor
and neutron multi-group flux distribution were used to assess theadequacy of the proposed approach.

9:20 AM

Variational Acceleration of Fission Source Iteration for
Subcritical Source-Driven Systems
B.Ozgener, H.A.Ozgener
Istanbul Technical University, Istanbul, Turkey
Numerical evaluation of the neutron flux distribution in slightly
subcritical source-driven systems by means of multigroup diffusion
theory is considered. It has been shown that the classical fission
source iteration results in inordinately large number of iterations insuch cases. To overcome the severe convergence problems met, avariational method, coarse mesh rebalance technique, is formulated,
implemented and tested in a number of problems characterizingslightly subcritical systems. The numerical studies of such problems
show that it is possible to accomplish superior convergence rates byutilizing the proposed variational technique. The method also has thepotential for being applicable in space-time kinetics problems.

9:40 AMAn Incident Flux Coupling Calculation Study for Nodal
Method and Monte Carlo Method
X.Wang, H.Yu, L.Xu, Y.Hu, X.Yang
China Institute of Atomic Energy, China
Conventional nodal method can not calculate fine distribution of introassembly neutron flux , especially for the assemblies with complex
geometry and materials. Fine distribution of neutron flux and the otherrelated parameters, however, are very useful for reactor core design
and neutronic simulation in some special conditions, for example,
simulation of neutron irradiation in test assemblies. A method named
Incident Flux Nodal and Monte-carlo Coupling method (IFNMC), is
studied and proposed to couple nodal transportation calculation with
Monte-carlo calculation by using incident flux on the coupling interface.
The IFNMC method, first use nodal transport method to do whole coreneutronic calculation and get neutron flux distribution in full-core sacle
and the incident interface flux for the specially identified assembly,
and then use Monte-carlo method to simulate neutron behavior andget fine neutron flux distribuion within the special assembly with finegeometry description. By this way, fine distribution of neutron flux andother related information can be calculated quickly. According to thebasic algorithm and the calculation flow of IFNMC, a computer code
has been developed to achieve the coupling calculation. Two cases
have been calculated and used to verify the method, one is fuel pin
power distribution of China Experimental Fast Reactor (CEFR) and theother is radial distribution of 197Au capture reaction rate in CEFR whichis actually measured in CEFR physical startup test. The calculate
results are consistent quit well with reference data, which preliminarily
proves the prosed IFNMC method is effective and accurate, and isparticularly suitable for fine distribution of neutron flux calculation in
specially identified assembly.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Mizuho_B

Track4-3 Verification, Validation and Uncertainty Analysis
Session Chair: Nuria Garcia Herranz(Univ. Politecnica de Madrid), Go Chiba(Hokkaido Univ.)
8:00 AM

Confidence Interval Estimation by Bootstrap Method
for Uncertainty Quantification Using Random
Sampling Method
T.Endo, T.Watanabe, A.Yamamoto
Nagoya University, Nagoya, JapanThe random sampling method is simple and practical technique toquantify uncertainties of target core parameters calculated by a core
analysis code. It is noted that statistical errors are inevitably involvedin the estimated uncertainties, since the random sampling method is
a probabilistic method using pseudorandom numbers. In the present
study, the applicability of bootstrap method is investigated in order toestimate the confidence interval of the estimated uncertainty using therandom sampling method. The bootstrap method enables to evaluate
the histogram, variance, and confidence interval of sample estimates(e.g., mean, variance, median) by the resampling technique, i.e.,a series of resampling are carried out using random sampling withreplacement from an original data. In other words, limited numberof samples can be used to estimate the confidence interval of
uncertainty. Through a lattice burnup calculation for a simplified BWR
fuel assembly, it is verified that the bootstrap confidence interval isa reasonable estimator to evaluate appropriate number of samples
from the statistical point of view. If the distribution of target parameteris well approximated by a normal distribution, the confidence intervalof uncertainty can be estimated by the conventional method based on
the chi-squared distribution. The merit using the bootstrap method is
to simply estimate the confidence interval without the assumption of
normal distribution.

8:20 AM

Uncertainty Quantification of Neutronics
Characteristics Using Latin Hypercube Sampling
MethodK.Kinoshita, T.Endo, A.Yamamoto(1), Y.Kodama, Y.Ohoka, T.Ushio,
H.Nagano(2)1)Nagoya University, Nagoya, Japan, 2)Nuclear Fuel Industries, Ltd,
Osaka, JapanThe Latin Hypercube Sampling (LHS) method is applied to theuncertainty quantification of neutronics characteristics based on
the Monte-Carlo based sampling method. The Monte-Carlo based
sampling method is one of the uncertainty quantification methods ofneutronics characteristics (e.g. neutron multiplication factor) due touncertainty cross section. The sampling method has the advantagethat burnup and thermal-hydraulics feedback effects are easily
considered in complicated light water reactor core analysis. Contrary,
the sampling method has the disadvantage that statistical errors
are involved in estimated uncertainties of neutronics characteristics
since the sampling method is a probabilistic approach. Although thestatistical errors can be reduced by increasing the number of samples(e.g. number of perturbed cross section libraries), computational costalso increases. Therefore development of an efficient uncertainty
quantification method, which provides smaller statistical error of
estimated uncertainty while reducing number of samples, is highly
desirable. In the present study, adequacy and performance of the LHS
method is investigated as an efficient Monte-Carlo based sampling
method. Uncertainty quantification of multiplication factor for a BWR
fuel assembly is carried out with the LHS method. The results indicatethat uncertainty quantification using the LHS method is more efficientthan that using the conventional sampling method, which utilizes
simple random sampling of cross sections.

8:40 PM

Uncertainty Quantification of BWR Core
Characteristics Using Latin Hypercube Sampling
Methodexplicitly taken into account using a licensing-grade core simulator.
Uncertainties for various core characteristics and correlations among
them are evaluated by the statistical processing of core calculation
results based on the LHS method. The calculation results indicate that
uncertainty of critical eigenvalue (i.e., core reactivity) in the BWR core
is comparable to that of typical PWR. On the other hand, uncertainties
of assembly relative power distribution and maximum assembly
burnup in the present BWR core are much smaller than those of
typical PWR. The strong thermal hydraulics feedback effect of BWR
would significantly contribute the difference of uncertainties in BWR
and PWR.

9:00 AM

Applicability of the Cross Section Adjustment Method
Based on Random Sampling Technique for Burnup
Calculation
T.Watanabe, T.Endo, A.Yamamoto(1), Y.Kodama, Y.Ohoka, T.
Ushio(2)1)Nagoya University, Nagoya, Japan, 2)Nuclear Fuel Industries, Ltd.,
Osaka, Japan
Applicability of the cross section adjustment method based on random
sampling (RS) technique to burnup calculations is investigated. The
cross section adjustment method is a technique for reduction ofprediction uncertainties in reactor core analysis and has been widely
applied to fast reactors. As a practical method, the cross section
adjustment method based on RS technique is newly developed for
application to light water reactors (LWRs). In this method, covarianceamong cross sections and neutronics parameters are statisticallyestimated by the RS technique and cross sections are adjusted
without calculation of sensitivity coefficients of neutronics parameters,
which are necessary in the conventional cross section adjustment
method. Since sensitivity coefficients are not used, the RS-based
method is expected to be practically applied to LWR core analysis, inwhich considerable computational costs are required for estimation ofsensitivity coefficients. Through a simple pin-cell burnup calculation,
applicability of the present method to burnup calculations isinvestigated. The calculation results indicate that the present method
can adequately adjust cross sections including burnup characteristics.

9:20 AM

NUSS-RF: Stochastic Sampling-Based Tool
for Nuclear Data Sensitivity and Uncertainty
Quantification
T.Zhu, A.Vasiliev, H.Ferroukhi, A.Pautz(1), S.Tarantola(2)1)Paul Scherrer Institut, Villigen Switzerland, 2)European Commission
Joint Research Council, Ispra (VA), Italy
The “blackbox” approach of stochastic sampling (SS) methods for
simultaneous nuclear data uncertainty quantification is powerful except
it reveals little of the individual uncertainty contributions. In this work,
the SS-based tool “NUSS” developed at PSI is updated to “NUSSRF” which can estimate how much individual nuclear data uncertainty
contributes to the total output uncertainty. The new capability is basedon the Random Balance Design and FAST methods, both belonging
to the so-called Global Sensitivity Analysis. First, the implementation
of NUSS-RF is tested using a mathematical function, followed by thesensitivity and uncertainty analysis for 239Pu(n,f), 235U(n,f) and 238U(n,γ)
cross sections in Jezebel, Godiva and BWR pincell benchmarks
respectively. The results are compared to the deterministic Sensitivity/
Uncertainty “Sandwich Rule” approach (which is local). For
uncorrelated inputs, both methods have the equivalent interpretationof the input uncertainty contribution (in terms of variance fraction andSensitivity Index), hence producing good agreement in the results. For
correlated inputs, the discrepancy between the two methods broadenswith the extent of the correlations which currently limits the validation
of NUSS-RF to inputs with weak correlations.

A.Yamamoto
Nagoya University, Nagoya, Japan
Uncertainties of neutronics characteristics in a commercial BWR
core due to cross section covariance are evaluated by the Latin
Hypercube Sampling (LHS) method, which is a Monte-Carlo based
efficient sampling algorithms. The Peach Bottom Unit 2, which is a
BWR core, is a target for the uncertainty quantification in the present
study. Thermal hydraulics feedback and burnup effects are fully and57

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Mizuho_B

Track4-3 Verification, Validation and Uncertainty Analysis
Session Chair: Nuria Garcia Herranz(Univ. Politecnica de Madrid), Go Chiba(Hokkaido Univ.)
9:40 AMMOCABA: A General Monte Carlo-Bayes Procedure
for Improved Predictions of Integral Functions of
Nuclear Data
A.Hoefer, O.Buss, M.Hennebach, M.Schmid(1), D.Porsch(2)1)AREVA GmbH Offenbach, Offenbach, Germany, 2)AREVA GmbH
Erlangen, Erlangen, Germany
MOCABA is a combination of Monte Carlo sampling and Bayesian
updating algorithms for the prediction of integral functions of nuclear
data, such as reactor power distributions or neutron multiplication
factors. Similarly to the established Generalized Linear Least
Squares (GLLS) methodology, MOCABA offers the capability to utilize
integral experimental data to reduce the prior uncertainty of integral
observables. The MOCABA approach, however, does not involve any
series expansions and, therefore, does not suffer from the breakdownof first-order perturbation theory for large nuclear data uncertainties.
This is related to the fact that, in contrast to the GLLS method, theupdating mechanism within MOCABA is applied directly to the integral
observables without having to “adjust” any nuclear data. A central
part of MOCABA is the nuclear data Monte Carlo program NUDUNA,
which performs random sampling of nuclear data evaluations
according to their covariance information and converts them into
libraries for transport code systems like MCNP or SCALE. What is
special about MOCABA is that it can be applied to any integral functionof nuclear data, and any integral measurement can be taken into
account to improve the prediction of an integral observable of interest.
In this paper we present two example applications of the MOCABA
framework: the prediction of the neutron multiplication factor of awater-moderated PWR fuel assembly based on 21 criticality safety
benchmark experiments and the prediction of the power distribution
within a toy model reactor containing 100 fuel assemblies.

58

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Hiei

SS3 Hybrid Particle Transport Methods for Solving Complex Problems
in Real-Time
Session Chair: Alireza Haghighat(Virginia Tech Univ.), Kazuya Yamaji(MHI)
8:00 AM

Discretized Mesh Tools and Related Treatment
for Hybrid Transport Application with 3D Discrete
Ordinates and Monte Carlo
K.L.Manalo, G.E.Sjoden, C.A.Edgar
Georgia Institute of Technology, Atlanta, USA
Hybrid methods of neutron transport have increased greatly in use, forexample, in applications of using both Monte Carlo and deterministictransport methods to calculate quantities of interest, such as the fluxand eigenvalue in a nuclear reactor. Many 3d parallel Sn codes apply
a Cartesian mesh, and thus for nuclear reactors the representation of
curved fuels (cylinder, sphere, etc.) are impacted in the representationof proper fuel inventory, resulting in both a deviation of mass andexact geometry in the computer model representation. In addition,
we discuss auto-conversion techniques with our 3d Cartesian meshgeneration tools to allowfor full generation of MCNP5 inputs (Cartesianmesh and Multigroup XS) from a basis PENTRAN Sn model.
For a PWR assembly eigenvalue problem, we explore the errors
associated with this Cartesian discrete mesh representation, andperform an analysis to calculate a slope parameter that relates the
pcm to the percent areal/volumetric deviation (areal → 2d problems,volumetric → 3d problems). This paper analysis demonstrates a linear
relationship between pcm change and areal/volumetric deviation
using Multigroup MCNP on a PWR assembly compared to a reference
exact combinatorial MCNP geometry calculation. For the same MCNP
multigroup problems, we also characterize this linear relationship in
discrete ordinates (3d PENTRAN). Finally, for 3D Sn models, we show
an application of corner fractioning, a volume-weighted recovery ofunderrepresented target fuel mass that reduced pcm error to < 100,
compared to reference Monte Carlo, in the application to a PWR
assembly.

8:20 AMA Novel Hybrid Weighting Scheme for Multi-Group
Cross Section Collapsing
C.Yi, G.E.Sjoden, C.A.Edgar
Georgia Institute of Technology, Atlanta, USA
Multi-group cross section library generation plays an important rolein deterministic transport simulations. In this paper, a new fine-groupto broad-group cross section collapsing method is introduced. Rather
than a traditional flux weighting, the new method uses a hybrid
weighing scheme to collapse the scattering cross section matrix.
Based upon a matrix analysis approach, we generalize different
weighting schemes and derive the new hybrid weighting scheme,
which mathematically shows that it is rational for the scattering crosssection to be weighted by the (i) forward fluxes of the incoming/inbound neutron groups and (ii) the adjoint functions of the outgoing/
out-bound neutron energy groups. This approach also makes physicalsense, since it conserves the “importance flow” of particles throughscattering while collapsing cross sections. To conserve the reaction
rates at the same time, we re-normalize the hybrid weighted scatteringcross section to the original library total scattering reaction rate. We
demonstrate that the hybrid weighting scheme is more accurate,
especially for the detector response simulation problem in a DualRange Coincidence Counter (DRCC) 3-D SN transport model.

8:40 PM

Development of an Iterative Lattice-Core Coupling
Method Based on MICROX-2 Cross Section Libraries
J.Hou(1), H.Choi(2), K.Ivanov(1)1)The Pennsylvania State University, PA, USA, 2)General Atomics, CA,
USAThis paper describes an innovative online cross section generation
method, developed based on Iterative Diffusion-Transport (IDT)
calculation to minimize the inconsistency and inaccuracy in determining
physics parameters by feeding actual reactor core conditions into thecross section generation process. A 2-dimensional (2-D) pin-by-pin
lattice program, NEMA, was also developed to generate assembly
lattice parameters using the embedded MICROX-2 cross section
libraries and Nodal Expansion Method (NEM). The proposed methods
were validated against a 2-D miniature core benchmark problem forboth NEMA itself and its coupling to a reactor code by the IDT method(NEMA- DIF3D). The computational benchmark calculations haveshown that the IDT improves the eigenvalue and power distribution

predictions when compared with the conventional offline method.

9:00 AM

Solution of a Stylized European Pressurized Reactor(EPR) Benchmark Problem Using the Coarse Mesh
Radiation Transport Method (COMET)
D.Lago, F.Rahnema, D.Zhang
Georgia Institute of Technology, Atlanta, USA
In this paper, as additional verification of its accuracy and efficiency,thecoarse mesh radiation transport code COMET is compared toMonte Carlo solutions in a stylized benchmark problembased on theEuropean Pressurized Reactor (EPR). The core specifications weretaken directly from the Final Safety Analysis Report (FSAR) submitted
to the Nuclear Regulatory Commission (NRC) and the reactor wasmodeled in a stylized manner while maintaining full heterogeneity
at the pin and assembly level. Detailed results including assembly
eigenvalues, core eigenvalues, and pin fission densities using a
2-group cross section library are presented. COMET results are inexcellent agreement with MCNP with eigenvalue relative differenceson the order of 10 pcm and average pin fission density relative
differences on the order of 1-5%. Some maximum errors were on theorder of 10% due to poor statistics on the periphery of the core in thereference results.

9:20 AM

Computational Efficiency and Accuracy of the Fission
Collision Separation Method in 3D HTTR Benchmark
Problems
D.Zhang, F.Rahnema
Georgia Institute of Technology, Georgia, USAA fission collision separation method has been recently developed
to significantly improve the computational efficiency of the COMET
response coefficient generator. In this work, the accuracy andefficiency of the new response coefficient generation method is tested
in 3D HTTR benchmark problems at both lattice and core levels.
In lattice calculations, the surface-to-surface and fission density
response coefficients computed by the new method are comparedwith those directly calculated by the Monte Carlo method. In wholecore calculations, the eigenvalues and bundle/pin fission densities
predicated by COMET based on the response coefficient libraries
generated by the fission collision separation method are comparedwith those based on the interpolation method as well as the Monte
Carlo reference solutions. These comparisons have shown that the
new response coefficient generation method is significantly (about 3
times) faster than the interpolations method while its accuracy is close
to that of the interpolation method.

9:40 AMUse of the Fission Matrix Method for Solution of the
Eigenvalue Problem in a Spent Fuel Pool
W.J.Walters, N.Roskoff, K.K.Royston, A.Haghighat
Virginia Tech, Arlington, USA
In this paper, we examine the use of the fission-matrix method to
calculate the eigenvalue for a spent fuel pool. The fission matrix
coefficients are calculated using MCNP5. In order to make the method
as efficient as possible for real-time calculations, these coefficientsare pre-calculated. Certain simplifying assumptions are made basedon geometric considerations, greatly reducing the amount of precomputation, and allowing the coefficients to be used on a variety ofproblems beyond that for which the coefficients were calculated. Thismethodology is applied to a reference pool that is being designed forthe I2S-LWR project. Typically, the eigenvalue calculation in Monte
Carlo is very difficult for loosely coupled problems such as a spent fuelpool due to source convergence issues, which are not present using
the fission matrix. The fission matrix method has shown accuracy
very close to that of MCNP5, while giving results in several orders ofmagnitude less time. Total pre-calculation time was less than a single
eigenvalue calculation and can be used across many different poolconfigurations.

59

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Atago

Track3-4 Monte Carlo Methods
Session Chair: Ho Jin Park(KAERI), Toshihiro Yamamoto(KURRI)
8:00 AMXSBench - The Development and Verification of a
Performance Abstraction for Monte Carlo Reactor
Analysis
J.R.Tramm, A.R.Siegel(1), T.Islam, M.Schulz(2)1)Argonne National Laboratory, Argonne, USA, 2)Lawrence Livermore
National Laboratory, Livermore, USA
We isolate the most computationally expensive steps of a robust
nuclear reactor core Monte Carlo particle transport simulation. The
hot kernel is then abstracted into a simplified proxy application,
designed to mimic the key performance characteristics of the full
application. A series of performance verification tests and analyses
are carried out to investigate the low-level performance parametersof both the simplified kernel and the full application. The kernel’s
performance profile is found to closely match that of the application,
making it a convenient test bed for performance analyses on cuttingedge platforms and experimental next-generation high performancecomputing architectures.

8:20 AM

Development of Neutron Current Connection Method
for Whole Core Analysis Based on Monte Carlo
MethodN.Nakadozono, K.Ishii, M.Aoyama, T.Mitsuyasu, T.Hino
Hitachi Research Laboratory, Hitachi, Ltd., Ibaraki, JapanThe whole core analysis with a Monte Carlo neutron transport method
gives more accurate neutronics analysis of a heterogeneous corethan the conventional two-stage core analysis method with latticephysics and the core analysis. The whole core Monte Carlo analysis,
however, needs vast amounts of computational time and memory,
so generally parallelization is necessary. In this study, the neutron
current connection method is developed for the whole core Monte
Carlo analysis to improve its parallelization efficiency. When the wholecore analysis is spatially decomposed into many calculation nodes,
this method connects nodes by the neutron current instead of using
the information of each neutron. While this can reduce the amountof communication data between nodes, there is a discretization error
when compressing the data into macro neutron currents and iterationof the calculation between the node analysis and update of boundary
conditions is required. The validation of the method feasibility in viewof this error is evaluated. This method is installed into the multi-group
Monte Carlo code. Heterogeneous problems are evaluated and theirsolutions are compared with reference case solutions which calculates
the whole problem by direct Monte Carlo calculation. By setting
adequate division, after a sufficient number of current iterations,the calculated neutron infinite multiplication factors are agreed withreference cases within 0.1%dk and the root-mean-square differencesof pin power distribution of each fuel assembly with reference cases
are found to be within 1%.

8:40 PM

Analysis of Select BEAVRS PWR Benchmark Cycle 1
Results Using MC21 and OpenMC
D.J.Kelly, B.N.Aviles, P.K.Romano(1), B.R.Herman, N.E.Horelik, B.
Forget(2)1)Bechtel Marine Propulsion Corporation, New York, USA, 2)
Massachusetts Institute of Technology, Massachusetts, USA
MC21 and OpenMC Monte Carlo results have been comparedwith hot zero power measurements from an operating pressurizedwater reactor (PWR), as specified in the recently released BEAVRS(Benchmark for Evaluation And Validation of Reactor Simulations) flux
map data. Included in this comparison are axially-integrated full coredetector measurements and axial detector profiles.
In addition, MC21 was used in the 3-D analysis of cycle 1 using
integrated reactor feedback capabilities. A quarter core model waschosen for this study because of difficulties converging a full coremodel. This depletion encompassed quarter assembly temperature
feedback with more than 200 fission products in more than 379,000
regions. An approximate power history was utilized to model the actual
ragged power history present in cycle 1. Detailed analysis of in-coredetector data was performed with MC21 at three different exposurelevels. These also included comparisons with SIMULATE-3 in orderto further validate the MC21 code. The MC21 results agreed well with60

experiment and were comparable to SIMULATE-3 indicating that these
initial analysis methods in MC21 provide a highly accurate solution to
the analysis of a 3-D PWR.

9:00 AM

Monte Carlo Neutronics Analysis of Sodium-Cooled
Fast Reactor Benchmark with OTF Temperature andBurnup Treatment
N.S.Guilliard, W.Bernnat, J.Starflinger (1) , W.Zwermann, I.
Pasichnyk(2)1)Universität Stuttgart - IKE, Stuttgart, Germany, 2)GRS - Garching,
Germany
The sodium-cooled fast reactor design is one of six reactor
technologies proposed by the Generation-IV International Forum for
deployment during the next years. The concept combines advancesin sustainability, safety and proliferation-resistance. Therefore fouradvanced fast reactor cores had been proposed by the Working
Party on Reactor and System (WPRS). These cores are the base forneutronic analysis in a sodium-cooled fast reactor OECD Benchmark.
Eleven international organizations investigate two large coreswith 3600 MWth (oxide or carbide fuel) as well as two small cores
generating 1000 MWth with either metallic or oxide fuel. Besides
the k-effective value, reactivity feedbacks, isotopic composition at
the end of equilibrium cycle (EOC) and power distributions were
calculated. The benchmark partners used different code systems
and different data libraries. The latter had been identified as main
source for the discrepancies in the results. This paper shows selectedparts of IKE results for the large oxide fueled core performed with
MCNP6. The obtained IKE MCNP6 results are compared with thebenchmark results. The large oxide core was analyzed with both the
JEFF-3.1.2 and the ENDF/B-7.1 data set. Additionally, the k-effective
was calculated with varying treatment of the Doppler broadening.
The methods used were interpolation between two temperaturedependent data sets, a polynomial representation (Fit-OTF) and the
direct generation of data sets for the problem temperatures by the
MAKXSF code. Detailed burnup calculations were done with pinwise assignment of burnup dependent nuclide compositions in the
hexagonal core lattice. The benchmark calculations were performed
for constant coolant and fuel temperatures. Additionally, by external
coupling of MCNP6 with the thermal-hydraulic code ATHLET realistic
coolant and fuel temperatures were calculated to show the influenceof thermal hydraulic coupling.

9:20 AM

Verification of Coupled 3D Fuel Cycle Analysis with
Stable Monte Carlo Based Code, BGCore, against theNodal Diffusion DYN3D Code
D.Kotlyar, E.Shwageraus(1), M.Margulis(2), Y.Bilodid, E.Fridman(3)1)University of Cambridge, Cambridge, United Kingdom, 2)Ban Gurion
University of the Negev, Be'er Sheva, Israel, 3)Helmholtz-Zentrum
Dresden-Rossendorf, Dresden, Germany
Previous studies suggested that different schemes for coupling Monte
Carlo (MC) neutron transport with burnup and thermal hydraulicfeedbacks may potentially be numerically unstable. This issue can be
resolved by application of new Stochastic Implicit Mid-Point (SIMP)
methods. In order to assure numerical stability, the new methods
do require additional computational effort. The instability issue
however, is problem-dependent and does not necessarily occur inall cases. Therefore, blind application of the unconditionally stablecoupling schemes, and thus incurring extra computational costs,
may not always be necessary. In this paper, we attempt to develop
an intelligent diagnostic mechanism, which will monitor numericalstability of the calculations and, if necessary, switch from simple andfast explicit coupling scheme to more computationally expensive but
unconditionally stable one. To illustrate this diagnostic mechanism,
we performed a coupled burnup and TH analysis of a single BWR
fuel assembly. The reference solution was obtained by the state of
the art nodal diffusion code – DYN3D. Very good agreement wasobserved in all neutronic and TH parameters. The results indicate thatthe developed algorithm can be easily implemented in any MC based
code for monitoring of numerical instabilities. The proposed monitoring
method has negligible impact on the calculation time even for realistic
3D multi-region full core calculations.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Atago

Track3-4 Monte Carlo Methods
Session Chair: Ho Jin Park(KAERI), Toshihiro Yamamoto(KURRI)
9:40 AM

Domain Decomposition and Terabyte Tallies with the
OpenMC Monte Carlo Neutron Transport Code
N.E.Horelik, B.Forget, K.Smith(1), A.Siegel(2)1)Massachusetts Institute of Technology, Massachusetts, USA, 2)
Argonne National Laboratory, Argonne, USA
Memory limitations are a key obstacle to applying Monte Carloneutron transport methods to high-fidelity full-core reactor analysis.
Billions of unique regions are needed to carry out full-core depletionand fuel performance analyses, equating to terabytes of memory for
isotopic abundances and tally scores - far more than can fit on a single
computational node in modern architectures. This work introduces animplementation of domain decomposition that addresses this problem,
demonstrating excellent scaling up to a 2.39TB mesh-tally distributed
across 512 compute nodes running a full-core reactor benchmark on
the Mira Blue Gene/Q supercomputer at Argonne National Laboratory.

61

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 8:00 AM - Cosmos

Track9-2 Transient and Safety Analysis
Session Chair: Miriam Daeubler(KIT), Shigeaki Aoki(MNF)
8:00 AM

Recriticality Risk in PWR Spent Fuel Pools
G.Grandjean(1), W.F.G.van Rooijen, Y.Shimazu, H.Mochizuki(2)1)Institut National des Sciences et Techniques NuclÃľaires, Mines deNancy, France, 2)University of Fukui, Tsuruga, Japan
In this paper we investigated the situation in a PWR Spent Fuel Pool
(SFP) following a long-term loss of power / loss of cooling accident. In
the SFP there is a large amount of water with soluble boron between
the fuel assemblies. There may be a problem from the point of viewof criticality safety if the water of the SFP starts to boil and evaporate.
A thermal-hydraulic analysis was performed using a simplified modelof the SFP. The thermal-hydraulic analysis shows that in all cases achaotic boiling phenomenon develops. This indicates that even if there
is an issue of (near-)criticality, it will have a very intermittent nature.
The multiplication factor of the SFP was evaluated with a Monte
Carlo calculation. The neutronic analysis was performed for several
representative cooling situations. In all cases, the system remains(deeply) subcritical.

8:20 AM

Demonstration of Fully Coupled Simplified Extended
Station Black-Out Accident Simulation with RELAP-7
H.Zhao, H.Zhang, L.Zou, D.Andrs, R.Martineau
Idaho National Laboratory, ID, USA
The RELAP-7 code is the next generation nuclear reactor systemsafety analysis code being developed at the Idaho National Laboratory(INL). A number of physical components with simplified two phaseflow capability have been developed to support the simplified boilingwater reactor (BWR) extended station blackout (SBO) analyses. The
demonstration case includes the major components for the primarysystem of a BWR, as well as the safety system components for thesafety relief valve (SRV), the reactor core isolation cooling (RCIC)
system, and the wet well. Three scenarios for the SBO simulations
have been considered. Since RELAP-7 is not a severe accident
analysis code, the simulation stops when fuel clad temperaturereaches damage point. Scenario I represents an extreme station
blackout accident without any external cooling and cooling water
injection. The system pressure is controlled by automatically releasing
steam through SRVs. Scenario II includes the RCIC system but
without SRV. The RCIC system is fully coupled with the reactor
primary system and all the major components are dynamicallysimulated. The third scenario includes both the RCIC system and the
SRV to provide a more realistic simulation. This paper will describe
the major models and discuss the results for the three scenarios. The
RELAP-7 simulations for the three simplified SBO scenarios show theimportance of dynamically simulating the SRVs, the RCIC system,and the wet well system to the reactor safety during extended SBO
accidents.

8:40 AM

Code Scaling Applicability to a Cold Leg SBLOCA
Scenario in a Nuclear Power Plant
A.Querol, S.Gallardo, G.Verdú
Universitat Politècnica de València, Valencia, Spain
The knowledge of thermalhydraulic phenomena occurring in Nuclear
Power Plants (NPPs) during an accident is crucial in the assessmentof nuclear safety. Several experimental facilities have been built toreproduce some accidental scenarios obtaining measured data to
be compared with simulation results and to test the capability of thethermalhydraulic codes. This paper presents the scaling method used
to obtain a NPP scaled-up TRACE5 input from a Large Scale Test
Facility (LSTF) model to reproduce a cold leg Small Break Loss-ofCoolant Accident (SBLOCA) transient. The scaled-up TRACE5 model
has been developed conserving the power-to-volume scaling ratios
of LSTF components, initial and boundary conditions. A comparisonof the simulation results between LSTF and NPP scaled-up TRACE5
models is provided throughout different graphs, which represent themain thermalhydraulic variables. Results show that the main physicalphenomena produced during a cold leg SBLOCA are qualitatively
reproduced with the NPP scaled TRACE5 model.

62

9:00 AM

Transient Simulation of Gas Bubble in a Medium
Sized Lead Cooled Fast Reactor
C.F.Hellesen, P.Wolniewicz, P.Jansson, A.Håkansson, S.J.Svärd,
M.Österlund
Uppsala University, Uppsala, Sweden
A common problem for many liquid metal cooled fast reactor designs
is the positive void worth of the coolant. In this context, an advantage
of lead cooled fast reactors is the high temperature of coolant boiling.
In contrast to sodium cooled fast reactors this, in practice, precludes
coolant boiling. However, partial voiding of the core could result
from e.g. gas bubbles entering the core from below. This would
introduce a positive reactivity, and if the bubble is large enough.
In this paper we model this type of event using a point kinetics code
coupled to a heat transport code. The reactivity parameters are
obtained from a Monte Carlo code. The 300 MWth reactor design
Alfred is used as a test case. We show that in general the reactor
design studied is robust in such events, and we conclude that small
bubbles a measureable Power oscillation would occur. For very large
bubbles there exist a possibility of core damage. The cladding is themost sensitive part.

9:20 AMPower Ramp Transient in a Sodium-Cooled Fast
Reactor Used for Minor Actinides Transmutation
S.P.Martin, A.Ponomarev, R.Krüßmann, W.PfrangKarlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany
In this work the relevance of MOX fuel bearing minor actinides (MA)on the behavior of a SFR reactor under an unprotected transient
overpower (UTOP) event has been studied. An updated versionof the SAS-SFR code able to analyze the initial phase of accident
scenarios for sodium-cooled fast reactors fuelled with innovative oxide
mixtures (U,Pu,Np,Am,Cm)O2 has been used. We have considered
the effect of MA on reactor kinetics (i.e. reactivity coefficients, power,
and kinetic parameters), as well as on fuel thermal and mechanicalproperties (i.e. thermal conductivity, heat capacity, thermal expansion
and melting temperature). The core design used in this work is the socalled Reference Oxide core used in the CP-ESFR project at End Of
Equilibrium Cycle.

9:40 AM

Step towards Integral Validation of Energetic
Re-Criticality Prediction for Sodium Cooled Fast Reactor
T.Ivanova, E.Ivanov
Institut de Radioprotection et de Sûreté Nucléaire, Fontenay-aux-Roses,
FranceA specific feature of sodium cooled fast reactors is the energetic
re-criticality risk resulting from relocation of core materials duringa hypothetical core disruptive accident. It may lead to a severe
nuclear power excursion. The risk of mechanical energy releaseduring the accident is determined by whether the prompt criticality
is reached or not, or in other words, whether the re-criticality is
predicted with accuracy under delayed neutron fraction (ßeff) value
or not. The released energy depends on reactivity excess above ßeffvalue. Accordingly, for safety studies, the keff uncertainty should bewell established through accident evolution. A large portion of thisuncertainty originates from nuclear data. This paper presents work on
evaluation of nuclear data uncertainties for the selected snapshotsof unprotected loss of flow accident modeled for a large sodium fast
reactor and propagation of this uncertainty through the transient
process.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 10:20 AM - Mizuho_A

Track1-8 Reactor Analysis Methods
Session Chair: Youqi Zheng(Xi'an Jiaotong University), Tatsuya Iwamoto(GNF-J)
10:20 AM

Modernization Enhancements in SCALE 6.2
B.T.Rearden, R.A.Lefebvre, J.P.Lefebvre, K.T.Clarno,
M.A.Williams, L.M.Petrie, U.Mertyurek
Oak Ridge National Laboratory, Oak Ridge, USA
SCALE is a widely used suite of tools for nuclear systems modelingand simulation that provides comprehensive, verified and validated,user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For morethan 30 years, regulators, industry, and research institutions aroundthe world have used SCALE for nuclear safety analysis and design.
However, the underlying architecture of SCALE is based on a 40-yearold design with dozens of independent functional modules andcontrol programs, primarily implemented in the Fortran programming
language, with extensive use of customized intermediate files tocontrol the logical flow of the analysis. Data are passed between
individual computational codes using custom binary files that areread from and written to the hard disk. The SCALE modernization
plan provides a progression towards SCALE 7, which will provide an
object-oriented, parallel-enabled software infrastructure with state-ofthe-art methods implemented as reusable components. This paper
provides a brief overview of the goals of SCALE modernization anddetails some modernized features available with SCALE 6.2.

10:40 AMSALOME-CORE Platform: Uses for EDF R&D
Neutronic Studies
H.Leroyer, G.Boulant, A.Calloo, Y.Pora, M.Fliscounakis, T.Clerc,
R.Barate
EDF R&D, Clamart, France
EDF R&D is currently developing a new, state of the art, calculation
chain, ANDROMEDE, which relies on APOLLO-2 for assembly
calculation and on the in-house core code COCAGNE. Engineers
using these new neutronic tools have generic needs in terms ofnumerical simulation: pre-processing, post-processing, physics
coupling, uncertainty propagation, data assimilation, calculation
supervision and distribution,... These needs, shared with other generalfields (solid mechanics, fluid dynamics, thermalhydraulics) are notcovered by the perimeter of APOLLO-2 nor COCAGNE. In orderto mutualize the development effort in those generic modules, andto provide a common framework for all their numerical simulationengineers, EDF and the CEA co-developed the open-source SALOME
platform, a toolbox with generic modules for numerical simulation.
This paper presents the SALOME-CORE platform, which packages
the core code COCAGNE and the SALOME platform, and also usesof this platform for reactor physics studies: meshing of assemblies
for APOLLO-2 calculation, data assimilation for the computationof an appropriate reflector model, calculation supervision for the
optimization of energy meshes for SPn or Sn multigroup calculations,calculation distribution on supercomputers, efficient postprocessing
and visualization of the results.

11:20 AMGood Practice in Development of Advanced
Assembly/Core Calculation Methods and
Implementations of AEGIS/SCOPE2
M.Tatsumi, M.Tabuchi, H.Tagawa (1), Y.Kodama, Y.Ohoka, T.
Ushio(2)1)Nuclear Engineering, Ltd., Osaka, Japan, 2)Nuclear Fuel Industries,
Ltd., Osaka, Japan
This paper reviews the history of development of AEGIS/SCOPE2, anadvanced in-core fuel management code for PWRs. The initial project,
development of a proto-type code, was started in 1996 as a feasibility
study of the advanced calculation method/algorithm for advanced
computation environments such as distributed parallel computers
like PC-clusters which are commonly used nowadays. With successof development of the prototype code, a production-level advanced
core calculation code, SCOPE2, was developed followed by AEGIS,an advanced assembly calculation code. These codes have beendeveloped on the basis of the object-oriented programming approachand the agile software development. The authors extracted the key
factors for success of the project as good practices from the viewpointof code design, implementation, project management and verificationand validation. Those practices are universal and may be applicable toany projects in the future.

11:40 AM

Validation of LANCR01/AETNA01 BWR Code Package
against FUBILA MOX Experiments and Fukushima
Daiichi Nuclear Power Plant Unit 3 MOX Core
T.Iwamoto, T.Ikehara, M.Ono, T.Yamana, H.Suzuki
Global Nuclear Fuel-Japan, Yokosuka, JapanLANCR01 assembly code and AETNA01 core simulator are theadvanced BWR package developed by Global Nuclear Fuel (GNF).
In order to establish the applicability of the package to the plutonium
containing mixed oxide (MOX) fuel, validation tests have been
conducted against BASALA, FUBILA MOX critical experiments andthe operational data from Fukushima Daiichi Nuclear Power Plant Unit
3 (1F3) with MOX assemblies loaded at cycle 25. For BASALA andFUBILA, critical eigenvalue and the pin-by-pin fission rate distribution
by LANCR01 were compared with the experimental data. For 1F3,
AETNA01 predictions with LANCR01 assembly cross sections werecompared with the measured control blade worth and the moderator
temperature coefficient in the reactor physics tests, as well as thecold/hot eigenvalues and the in-core instrument readings during theoperation. It is concluded that LANCR01/AETNA01 system has acomparable accuracy for the MOX cores with that for the uranium
cores.

11:00 AMA Steady-State Core Analysis Code for the Modelingof Accelerator-Driven Subcritical Reactors
S.Zhou, H.Wu, L.Cao, Y.Zheng
Xi'an Jiaotong University, Xi’an, China
In order to analyze and evaluate Accelerator Driven Subcritical
Reactors (ADSR) efficiently, a neutronics/thermal-hydraulics coupling
analysis code named LAVENDER has been developed. In theneutronics calculation, the three dimensional deterministic neutrontransport method is adopted. The nuclides transmutation analysis is
implemented by the micro-depletion method with the Transmutation
Trajectory Analysis algorithm (TTA). In the thermal-hydraulics
calculation, a heat transfer model is established to consider thermal
feedback and examine thermal-hydraulics design. The validations are
performed based on several benchmarks. Numerical results indicatethat LAVENDER is reliable and efficient to be applied for the designand steady-state analysis of ADSR.

63

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 10:20 AM - Mizuho_B

Track4-4 Verification, Validation and Uncertainty Analysis
Session Chair: Oscar Cabellos(Univ. Politecnica de Madrid), Hiroki Iwamoto(JAEA)
10:20 AM

Pinpower Uncertainty Quantification of LWRPROTEUS Phase III Experiments
M.Hursin, M.Scriven, G.Perret, P.Grimm, A.Pautz
Paul Scherrer Institut, Villigen, Switzerland
CASMO-5MX is a set of Perl-based tools built around the latticecode CASMO-5 and developed at the Paul Scherrer Institut, used toperform sensitivity analysis (SA) and uncertainty quantification (UQ).
The paper describes the recent developments to the CASMO-5MX
methodology to perform pinpower UQ with the Stochastic Sampling
methodology while perturbing nuclear data and technologicalparameters.
The CASMO-5MX SS pinpower UQ is applied to the re-analysis of
the LWR-PROTEUS Phase III campaign which featured SVEA-96
Optima2 fresh BWR fuel assemblies in an attempt to explain somediscrepancy between the fission rate measurements (E) and theirassociated predictions with CASMO-5M (C).
The pinpower UQ results have shown that the pinpower uncertainty
due to nuclear data and technological parameters is in the range of
0.4% to 0.8%. This additional source of uncertainty, added to theexisting experimental uncertainty, do not fully explain the remaining
few C/E discrepancies in the pinpower map. The magnitude ofnuclear data and technological parameters uncertainty is either
underestimated or an additional source of uncertainty has not beentaken into account.

10:40 AM

Quantification of Code, Library and Cross-Section
Uncertainty Effects on the Void Reactivity Coefficientof a BWR UO2 Assembly
O.Leray, P.Grimm, H.Ferroukhi, A.Pautz
Paul Scherrer Institut, Villigen, Switzerland
A study of BWR Void Reactivity Coefficient (VRC) calculations withstate-of-the-art lattice physics codes is presented in this paper.
Although the study is performed with the CASMO code, since used at
PSI for analyses of the Swiss reactors, a first objective is to quantify
the VRC range of variation that can be expected from code/librarymodifications. A second objective is to start assessing the VRC
uncertainty related to nuclear data in modern lattice physics codes. To
these aims, a benchmark case is presented based on a representative
BWR lattice design and described in details in order to be readily
available for further assessments and comparisons with other latticecodes and/or libraries. The VRC behavior in the 0%-100% void
range and at low/high burnups is then studied as function of CASMO
code/library updates in general and of a transition from CASMO-4
to it advanced successor CASMO-5 in particular. Through this, aquantitative assessment of the VRC range of variation resulting among
other things from a much finer group structure and from updates to themost recent neutron data evaluations is made. Qualitatively, distinct
trends are observed but mainly in the high-to-very-high void range
and especially with CASMO-4 combined with the oldest libraries. To
understand this, a reactivity breakdown of the VRC is performed,
indicating that these trends are driven primarily by a non-linear 238U
capture reaction rate behavior in the epithermal to fast range. At
higher burnups, compensating effects from other nuclide/reactions
are pointed out. Finally, the methodology under development at PSI
for nuclear data uncertainty propagation with CASMO-5 is applied.
Although only cross-section uncertainties based on the SCALE-6
44-group variance-covariance matrices are at this stage considered,
the VRC uncertainty is in this manner estimated and compared to therange of variation due to code/library updates.

11:00 AM

Sensitivity and Uncertainty Analysis of Burnup
Reactivity for an Accelerator-Driven System
H.Iwamoto, M.Maier, K.Nishihara
Japan Atomic Energy Agency, Ibaraki, JapanA burnup calculation is carried out for an accelerator-driven system(ADS) proposed by the Japan Atomic Energy Agency (JAEA) withthe fourth version of JENDL, JENDL-4.0 and the previous one,JENDL-3.3. Considerable differences are seen in burnup reactivity
between the nuclear data libraries for an initial phase (first burnup
cycle) and an equilibrium phase (tenth burnup cycle). The differencesin these values are investigated using two methods: a method by64

replacing a nuclear data library by nuclide and a sensitivity analysis
technique. Among many contributors to them for the both phases,
we identify major ones; (i) the initial phase: fission cross section andfission neutron multiplicity of 238Pu, capture cross section of 241Am, and(ii) the equilibrium phase: capture cross section of 244Cm and 241Am,
and inelastic scattering cross section of 206;207Pb. The uncertainty
analysis shows that uncertainties in the bunrup reactivity deduced
from the JENDL-4.0 covariance data are comparable in magnitude to
the differences between the nuclear data libraries, and are dominatedby nuclear data parameters of 238Pu. Finally, we show the necessity ofuncertainty evaluation of the branching ratio of 241Am capture reaction.

11:20 AM

Sensitivity/Uncertainty Analysis for BWR
Configurations of Exercise I-2 of UAM Benchmark
N.Garcia-Herranz, O.Cabellos
Universidad Politècnica de Madrid, Madrid, Spain
In order to evaluate the uncertainties in prediction of lattice-averaged
parameters, input data of core neutronics codes, Exercise I-2 of theOECD benchmark for uncertainty analysis in modeling (UAM) was
proposed. This work aims to perform a sensitivity/uncertainty analysisof the BWR configurations defined in the benchmark for the purposeof Exercise I-2. Criticality calculations are done for a 7x7 BWR fresh
fuel assembly at HFP in four configurations: single unrodded fuelassembly, rodded fuel assembly, assembly/reflector and assemblyin a color-set. The SCALE6.1 code package is used to propagate
cross section covariance data through lattice physics calculations
to both k-effective and two-group assembly-homogenized crosssections uncertainties. Computed sensitivities and uncertainties for
all configurations are analyzed and compared. It was found that
uncertainties are very similar for the four test-problems, showing thatthe influence of the assembly environment on uncertainty prediction isvery small.

11:40 AM

Impact of the Fission Yield Nuclear Data Uncertainties
in the Pin-Cell Burn-Up OECD/NEA UAM Benchmark
O.Cabellos, D.Piedra, C.Diez(1), L.Fiorito(2)1)Universidad Politécnicade Madrid, Madrid, Spain, 2)SCK.CEN, Mol,
Belgium and Université Bruxelles Libre, Brussels, Belgium
The prediction of fission products and the impact of their uncertainties
to different safety-related spent fuel applications (burn-up credit,
decay heat generation, radiological safety, waste management, burnup prediction) are required for the evaluation of spent fuel systemdesigns and safety analysis options. One of the nuclear data needs
to this prediction is the independent fission yields. The mostly usedgeneral-purpose evaluated nuclear data libraries provide thesedata including their uncertainties as standard deviation, with nocorrelation between fission yields. However, new developments inthe theory and measurements of fission product yields are expected
to result in new evaluated files in the next coming years. These files
will include considerably more accurate yields including neutronenergy dependence combined with new covariance information
allowing realistic uncertainty estimates. In this paper, we focusedon the effect of fission yield covariance information on criticality anddepletion calculations. A LWR pin-cell burn-up benchmark, proposedin the general framework of the OECD/UAM Benchmark is analyzed
to address the impact of independent fission yield uncertainties.
Calculations were performed with the SCALE6 system and the ENDF/
B-VII.1 fission yield data library, adding covariance data obtained from
including covariance info of mass yields. Results are compared withthose obtained with the uncertainty data currently provided by ENDF/
B-VII.1. The uncertainty quantification is performed with a Monte Carlosampling and then compared with linear perturbation.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 10:20 AM - Hiei

Track8-1 Reactor Operation and Safety
Session Chair: Gerald Rimpault(CEA), Hiroshi Akie(JAEA)
10:20 AM

Effects of Cross Sections Libraries Parameters on the
OECD/NEA Oskarshamn-2 Benchmark Solution
P.Balestra, C.Parisi, E.Negrenti(1), F.Giannetti, G.Caruso(2)1)ENEA, “Casaccia” Research Center, Rome, Italy, 2)“Sapienza”
University, Rome, Italy
The OECD/NEA proposes a new international benchmark basedon the data collected from an instability transient occurred at the
Oskarshamn-2 NPP with the aim to test the coupled 3D Neutron
Kinetic/Thermal Hydraulic codes on challenging situations. The
ENEA “Casaccia” Research Center is participating to this benchmark,developing a computational model using RELAP5-3D© code. The
3DNK model was developed starting from the cross sections datasets
calculated by OKG, the Oskarshamn-2 licensee, using the CASMO
lattice code. Integration of neutron cross sections database in
RELAP5-3D© required data fitting by a n-dimensional polynomials,
calculations of the various polynomial coefficients and of the basecross sections values. An ad-hoc tool named PROMETHEUS has
been developed for automatically generate the RELAP5-3D ©compatible cross sections libraries. Thanks to this software it has
been easily possible to visualize the complex structure of theneutronic data sets and to derive different cross sections libraries forevaluating the effects of some neutronic parameters on the predictionof the reactor instability. Thus, the effects of the fuel temperatureand control rod history, of the discontinuity factors (averaged/not
averaged), and of the neutron poisons has been assessed. A ranking
table has been produced, demonstrating the relevance of the notaveraged discontinuity factors and of the on-transient neutron poisonscalculations for the correct prediction of the Oskarshamn-2 event.

10:40 AM

Evaluation of Operational Experiences and Reactor
Physics Tests of MOX Loaded BWR Cores
Y.Kuroda, Y.Miki, K.Nozaki, A.Fukao (1) , N.Yanagisawa (2) ,T.Makigami(3)1)TEPCO SYSTEMS CORPORATION, Tokyo, Japan, 2)Electric Power
Development Co.,Ltd., Tokyo, Japan, 3)Tokyo Electric Power Company,
Tokyo, Japan
Boiling water reactor (BWR) cores with various MOX fuel loading
ratios (i.e., ratios of the number of MOX fuel assemblies to the total
number of fuel assemblies in the core) were operated for many cycles
at two commercial BWRs in Germany. The evaluation of operational
experiences and reactor physics tests of those two plants is useful forenhancing the reliability of in-core fuel management for MOX loaded
cores. The evaluation of those experiences of MOX loaded cores has
been performed using CASMO-4 and SIMULATE-3 in order to confirm
that the evaluation results with various MOX fuel loading ratios areequivalent to those of UO2 loaded cores from the point of view of incore fuel management. This paper introduces the evaluation resultsof the critical eigenvalue in hot condition (i.e., rated and partial power
condition), the traversing in-core probe (TIP) response, the cold criticaleigenvalue, the differential control rod worth, and the moderator
temperature coefficient. The MOX fuel loading ratio has no significant
effect to those parameters in the evaluation results. Hence, it has been
confirmed that the in-core fuel management for MOX loaded corescan be performed equally with that for UO2 loaded cores.

This is conventionally referred to as a ‘power tilt’. The amplification
effect of the reflector on the power tilt has been addressed in two
steps. Firstly, we have compared the power tilt, computed with thedeterministic code CRONOS2 (the neutronics diffusion module of the
HEMERA chain) and with the probabilistic transport code MCNP, tovalidate the CRONOS2 response against a Monte-Carlo reference
and obtain a quantification of the amplification effect at zero power.
Then the feedbacks have been taken into account in the diffusion
calculations to evaluate the sensitivity of the tilt to the power level.
The explanation of the difference in the core behavior in the presenceof either a conventional or a heavy reflector has been investigatedthrough the examination of the power distributions and the eigenvalue
splitting of the two systems. The results explain the higher sensitivity
to a peripheral perturbation of a heavy-reflector-core compared to aconventionally- reflected one.

11:20 AM

Improve the Accuracy of the Power Distribution
Reconstruction Using Power Distributions of Different
Status as the Fundamental Harmonic
F.Kai, L.Fu, Z.Xuhua, G.Jiong
Tsinghua University, Beijing, China
Modular high temperature gas cooled reactor HTR-PM demonstration
plant, designed by INET, Tsinghua University, is being built in
Shidaowan, Shandong province, China. HTR-PM adopts pebble
bed concept. Like other HTR, to measure or monitor the core
power distribution is very important but challenging for HTR-PM,
as there is no in-core neutron detectors. There are some proposals
to reconstruct the power distribution from the readings of ex-coreneutron detectors. One method is harmonic synthesis method. In thispaper, improved harmonic synthesis method is proposed and tested
to increase the accuracy of the power distribution reconstruction,by using better fundamental harmonic according to different reactor
operation conditions. The numerical result shows the improvementin the reconstruction accuracy, contrasting to harmonics’ coefficientspolynomial expansion method.

11:00 AM

Coupling Effects in Large Reactor Cores: The Impact
of Heavy and Conventional Reflectors on PowerDistribution Perturbations
A.Sargeni(1), K.W.Burn(2), G.B.Bruna(1)1)IRSN, Fontenay-aux-Roses, France, 2)ENEA, Bologna, Italy
The paper is intended to provide a simple but general overview of the
problem of large-size nuclear reactor coupling, through an applicationthat demonstrates the sensitivity of the system perturbations to theeigenvalue dominance ratio. As an example, we have investigated
the sensitivity of the core radial power distribution to the nature of thereflector (conventional vs. a “heavy” stainless-steel one), by applying
a perturbation to some peripheral assemblies of an EPR-like reactor
system (as defined in the UAM GEN-III benchmark, a fresh fuel core).
The perturbation was selected to simulate possible assembly bowing
effects, engendering a variation of the inter-assembly water gap,
which can generate an azimuthal asymmetry in the power distribution.

65

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 10:20 AM - Atago

Track6-1 Reactor Physics Experiments
Session Chair: Patrick Blaise(CEA),Takuya Umano(Toshiba)
10:20 AM

Static Modal Analysis of the Current-to-Flux
Subcriticality Monitor for Accelerator-Driven Systems

W.Uyttenhove, P.Baeten, A.Kochetkov, G.Van den Eynde,
G.Vittiglio, J.Wagemans(1), D.Lathouwers, J.-L.Kloosterman,
T.H.J.J.van der Hagen(2), A.Billebaud, S.Chabod(3), F.Mellier(4),J.-L.Lecouey, F.-R.Lecolley, G.Lehaut, N.Marie(5), X.Doligez(6),M.Carta(7), V.Bécares, D.Villamarin(8)1)SCK-CEN, Belgian Nuclear Research Centre, Boeretang, Belgium, 2)
Delft University of Technology, The Netherlands, 3)CNRS-IN2P3/UJF/
INPG, France, 4)CEA/DEN/DER/SPEX/LPE Cadarache, France, 5)
ENSICAEN/Université de Caen, CNRS-IN2P3, France, 6)CNRS-IN2P3/
Université Paris Sud, Orsay, France, 7)ENEA, Rome, Italy, 8)CIEMAT,
Madrid, Spain
With Accelerator-Driven Systems (ADS), the incineration of large
quantities of minor actinides is envisaged by coupling a particle
accelerator to a subcritical core. Despite the subcritical operation of
the core, a monitoring tool is indispensable in an ADS for both safety
and operational purposes. Several experimental techniques are underinvestigation today within the European FP7 project FREYA (Fast
Reactor Experiments for hYbrid Applications) at the VENUS facility at
SCK-CEN, where the lead fast zero-power VENUS-F core is coupled
to the GENEPI-3C deuteron accelerator. The proposed monitoring
concept combines an initial reactivity determination, e.g. by the pulsed
neutron source technique, a long-term relative reactivity monitor, by
the current-to-flux (CTF) technique, and an interim cross-checking,by the source jerk technique. In this paper the CTF technique ispresented as a candidate for continuous reactivity monitoring. Static
modal analysis is applied to evaluate the point kinetics relationshipbetween accelerator source neutrons and the neutron flux. As a firsttest case, the zero-power VENUS-F SC1 core is considered. Spatial
correction on the CTF monitor is evaluated by measurement results forthe control rods worth at different detector positions. Good agreement
between the modal and experimental spatial correction factor (SCF) is
found with 9 modes. Limited SCF are found in the VENUS-F reflector,
far away from the perturbation. Modal analysis is envisaged for themonitoring of burn-up in demonstrator ADS like MYRRHA.

10:40 AM

12 Years of Franco-Japanese International Programsin EOLE for the Validation of 100%MOX Recycling in
LWRs
P.Blaise, P.Alexandre, K.Blandin, JF.Ledoux, JM.Girard,
H.Philibert, A.Roche, P.Fougeras, S.Cathalau, O.Litaize,
A.Santamarina, C.Vaglio, JF.Vidal, N.Thiollay, N.Huot, JP.Chauvin,
T.Pont, V.Laval, S.Testanière, B.Vincent
CEA Cadarache, Saint Paul-lez-Durance, France
In France, recycling was decided in 1985 and the first mixed oxide(MOX) reload was introduced in the Saint Laurent B1 PWR French
EdF Nuclear Power Plant in 1986. In order to validate the current
calculation schemes and to reduce the uncertainties, exhaustive
experimental programs were launched in the French EOLE zero
power critical facility. The first programs were focused on physicalproperties of 30%MOX recycling in French PWRs. Later, specificstudies progressed in Japan and in France in the mid 90’s to verify
the possibility of introducing high moderation 100%MOX reloads inLight Water Reactors. These studies were performed in the frameworkof development of Advanced LWR concepts (ALWRs) which include
BWRs and PWRs. They relied on current calculation schemes that
require additional validation in order to achieve similar accuracies forhigh moderation 100%MOX cores as it had already been obtained for
UO2 and 30%MOX recycling in P- and BWRs. The paper details the 3experimental programs that have been conducted in the French EOLE
critical facility of the Cadarache Research Centre between 1995 and2007, and the associated experimental techniques applied to reachtarget accuracies. These programs are :MISTRAL:(MOX Investigation of Systems which are Technically
Relevant of Advanced Light water reactors) for 100%MOX
investigation in overmoderated homogeneous and 17x17
PWR mock-up lattices
BASALA:(Boiling water reactor Advanced core physics Study Aimedat mox fuel LAttice) for 100%MOX investigation in 9x9
Advanced BWR mock-up lattices of various moderation
ratios66

FUBILA: (FUll mox core physics experiments of Bwr Initiated for
Lattice Analysis method verification and improvement) for
100%MOX investigation in High Burn 9x9 and 10x10 BWR
mock-up lattices with void increase.
After a description of those main experimental programs performedin EOLE, an overview of their feedback on experimental uncertainty
improvement and code V&V is given.

11:00 AM

Application of the Best Representativity Method to a
Future PWR Fuel Assembly Calculation Using Four
Critical Experiments of Different Facilities
T.Umano, K.Yoshioka, M.Yamaoka(1), T.Obara(2)1)Toshiba Corporation Power Systems Company, Kawasaki, Japan, 2)Tokyo Institute of Technology, Tokyo, Japan
Since the inception of the nuclear industry, for the qualification andthe validation of the nuclear calculation codes, precise measurement
data from the critical experiments are both necessary and profitable.
To judge the experiment similarities to the objective of the actual
reactor conditions or actual reactor equipment, the concept of a“representativity factor” has recently been adopted in the criticalexperiment field, particularly for fast breeder reactors (FBRs) andfuture reactor studies. Based on this concept, a new numericalevaluation method has been developed to improve the representativityof critical experiments for the target reactor conditions. With thismethod, it is also possible to evaluate more reliable value of a physicalproperty of the target system under combining two or more criticalexperiments. As an industrial application, this new method was utilized
to correct the infinite neutron multiplication factor (k-infinity) of a futurepressurized water reactor (PWR) fuel assembly. This PWR 17×17
type fuel assembly has 6.0 wt% of 235U enrichment and 24 UO2Gd2O3 rods. Based on the method, four kinds of critical experiments
were combined. Two of them were the critical experiments for PWR
simulations at the Toshiba NCA facility. Other two were due to
ICSBEP 2010 data disk. The SCALE 6.1.2 system was utilized for
performing calculations. After combining four critical experiments, therepresentativity factor became larger and the similarity was improved.
Also k-infinity of the PWR fuel assembly was corrected. In this study,
the calculation procedures are explained. After that, calculation results
are shown with physical explanations and discussions.

11:20 AM

Characterization of Irradiation Fields for Fuel andMaterial Irradiation in the Experimental Fast Reactor
Joyo
S.Maeda, H.Saito, H.Naito, C.Ito, T.Sekine
Japan Atomic Energy Agency, Ibaraki, JapanThe Joyo MK-III core is a worldwide fast neutron irradiation field not
only for FBR development but also for use in other fields such as lightwater reactor (LWR) and fusion reactor studies, and in the non-nuclear
industry. The characterization of these neutron and gamma ray fields
is most important to utilize for irradiation tests. This paper describes
the details of distributions of neutron flux, reaction rate and gamma
heating in the MK-III core. The calculation accuracy of the coremanagement codes HESTIA, TORT and MCNP, was also evaluatedby the measured data. The calculated results in neutron calculationagreed well with the measured one. The calculation method wasvalidated and correction factors were identified. In case of gamma
heating evaluation, the calculated result is underestimated withrespect to the experimental value especially in the upper and lower SS
reflector region. Further investigations in gamma heating evaluation
are needed.

11:40 AM

Reaction Rate, Fission Product Yield, and RossiAlpha Measurements Using a HEU Metal, Copper
Reflected Critical Assembly
R.G.Sanchez, J.A.Bounds, T.A.Bredeweg, J.M.Goda, D.K.Hayes,
K.R.Jackman, G.E.Mckenzie, W.L.Myers, T.J.Grove
Los Alamos National Laboratory, Los Alamos, USAA critical experiment was performed on the Comet assembly toprovide nuclear data in a non-thermal neutron spectrum and to reestablish experimental measurement capabilities relevant to thePHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 10:20 AM - Atago

Track6-1 Reactor Physics Experiments
Session Chair: Patrick Blaise(CEA),Takuya Umano(Toshiba)
United States Department of Energy’s general purpose nuclear
criticality experiments capability and to the Technical Nuclear Forensic
program. Activation and fission foils were placed at specific locationsin the Zeus all-oralloy core, copper reflected critical experiment to infer
spectral indices data and obtain reaction rate data. After the irradiation,
passive gamma ray measurements were performed on all the foils
and several of them were packaged and shipped to Los Alamos
National Laboratory for further radiochemical analysis. The resultsfrom the non-destructive and radiochemical analyses are presented.
Finally, Rossi-α measurements were performed on a slightly
modified configuration from the configuration used for the activation
measurements. The Rossi-α results are presented and compared topast measurements performed using other critical assemblies.

67

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 10:20 AM - Cosmos

Track9-3 Transient and Safety Analysis
Session Chair: Silva Kalcheva(SCK/CEN), Mikio Tokashiki(NFI)
10:20 AMSIMMER-III Modeling of Gas Cooled Fast Reactor
X.N.Chen, L.Andriolo, A.Rineiski, W.Maschek
Karlsruhe Institute of Technology (KIT), Karlsruhe, Germany
This paper deals with extension and application of the SIMMERIII code for safety studies of a gas cooled fast reactor. The equationof state of the helium gas and its thermal physical properties havebeen prepared and implemented in the code. The geometric, thermal
hydraulic and neutronic models have been set up for the ALLEGRO
reactor. The code and the associated model are verified by comparing
steady state and unprotected loss of flow with 20% remained flowrate (ULOF-20%) results with those done by other project partners.
Reasonable or good agreements have been achieved for major
physical variables.
The unprotected loss of coolant accident (ULOCA) case is a severetransient case with core melting and degradation that was simulated
only by SIMMER in the project. In the initiating phase the clad
becomes molten; this triggers the first power excursion. Then the fuelbecomes more mobile and further power excursions take place, whichlead to core melting and degradation. The fuel is ejected by power
excursion and then moves relatively slowly to the lower part of vessel.
Finally there are only a few kilograms of fuel escaping to the vessel
outside (into reactor container) and the released thermal energy is
about 6 GJ within a period of one minute. The final power stays below
one MW and the reactor is in a deep sub-criticality state, since 1/2 fuelbecomes noneffective.

10:40 AM

Validation of the Subchannel Code CTF against theBenchmark Data of the OECD/NEA PSBT
A.Abarca, R.Miró, P.Hidalga, T.Barrachina, G.Verdú
Universitat Politècnica de València (UPV), Valencia, Spain
Nuclear reactor safety analysis has been increased rapidly in thelast decades. Due to this, advanced methods for core phenomena
predictions are developed, specially related to thermal-hydraulics using
subchannel codes. CTF is a version of the subchannel code COBRATF improved by the ISIRYM. This code must be validated against fullscale high-quality experimental data, which allow comparing results
for several case types. The benchmark of the OECD/NEA PWRSubchannel and Bundle Test (PSBT) has been carried out in order tovalidate the features of CTF during steady state and transient casesin Light Water Reactors (LWR). The objective of the benchmark is
to analyze the void fraction predicted by CTF in 4 different cases forsteady state and 4 transient cases. Benchmark cases are performed
with 3 different bundle types.

11:00 AM

Validation of CASMO5 / SIMULATE-3K Using theSpecial Power Excursion Test Reactor III E-Core: Cold
Start-Up, Hot Start-Up, Hot Standby and Full Power
Conditions
G.Grandi
Studsvik Scandpower, Inc., ID, USA
The Special Power Excursion Reactor Test (SPERT) III E-core wasa pressurized-water, nuclear research facility constructed to analyze
the reactor’s kinetic behavior under initial conditions similar to thoseof commercial LWRs. The SPERT III E-core, except for its size,
resembles a PWR. The initial test conditions were representative of awide range of initial and transient conditions: cold start-up, hot startup, hot standby, and full power. This paper is a sequel to a paper
presented in PHYSOR 2012. In the previous work, only the cold startup tests were analyzed. The aim of the present work is to extend theprevious work and validate the code system CASMO5 / SIMULATE3K for the full range of initial and transient conditions expected inreactivity initiated accidents.
The CASMO5 / SIMULATE-3K model was developed using datafrom the original SPERT experiments and the feedback models
recommended by Studsvik for LWR applications. The CASMO5 /SIMULATE-3K results are summarized in terms of: peak power,
reactivity compensation at the time of peak power and the energy
release to time of peak power. The CASMO5 / SIMULATE-3K transient
simulations are in good agreement with the experimental data. The
main uncertainty in the transient calculations is the initial position of
the control rods. Differences in the control rod positions, and hence in68

the initial reactivity insertion, can explain the differences between thecalculated and experimental results.

11:20 AM

Validation of the Nodal Kinetics Code System
GALAXY/COSMO-K Using the SPERT-III E-Core
Experiments
K.Yamaji, Y.Takemoto, K.Kirimura, S.Kosaka, H.Matsumoto
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
This paper describes validation of the GALAXY/COSMO-K system
to simulate reactivity insertion accidents (RIA) in the Special Power
Excursion Reactor Test III (SPERT-III) E-core tests. Mitsubishi Heavy
Industries, Ltd. (MHI) has developed the new neutronics and thermalhydraulic (T/H) coupling code system SPARKLE-2. SPARKLE-2
consists of the core neutronic, the T/H and the plant system calculation
codes. Under the present circumstances, the SPARKLE-2 code
system has been applied to the Anticipated Transient without Scram(ATWS) in Design Extension Condition (DEC) analyses, which arerequired by new Japanese regulation. GALAXY/COSMO-K is used as
the neutronic code in SPARKLE-2. In the previous study, GALAXY/COSMO-K has been verified and validated using several well-known
benchmark problems. In this study, GALAXY/COSMO-K simulates the
RIA analyses in cold zero power, hot startup, hot stand-by and hotoperation conditions of the SPERT-III. It is found that the GALAXY/COSMO-K system represents the experiment results well against
any initial conditions and insertion reactivity cases in the SPERT-III
experiments. In addition, by using the SPERT-III experiment analyses,
we performed sensitivity analyses about Doppler reactivity.

11:40 AM

Propagation of Nuclear Data Uncertainty for a Control
Rod Ejection Accident Using the Total Monte-Carlo
MethodD.F.D.Cruz, D.Rochman, A.J.Koning
Nuclear Research and Consultancy Group NRG, Petten, The
Netherlands
The Total Monte-Carlo method was applied to a full 3-D core of atypical pressurized water reactor (PWR) for the uncertainty analysis
of key reactor parameters as result of uncertainties in nuclear dataof the isotopes 235,238U, 239Pu, and thermal scattering data for H in
H2O. Both steady-state and transient reactor states were considered.
Uncertainty evaluation of key physical quantities is of importance inthe safety assessment during licensing of new nuclear power plants(NPP), and upgrading of operating conditions of NPP already inoperation. This work focus on the uncertainties in important reactor
parameters followed during the evolution of a control rod ejection
accident, where the control rod with the highest worth is supposed tobe ejected from the core within 0.1 seconds. Among other parametersstudied, the clad-to-coolant heat flux (and the associated uncertainty)
was considered, which is an important parameter in relation with thedeparture of nucleate boiling (DNBR) safety criteria used for this typeof reactivity insertion accident scenarios.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Mizuho_A

Track1-9 Reactor Analysis Methods
Session Chair: Richard Sanchez(CEA), Baocheng Zhang(WH)
13:30 PM

Resonant Upscattering Effects on 238U Absorption
Rates
C.Mounier, P.Mosca

Commissariat a l'energie atomique et aux energies alternatives, Gif-surYvette Cedex, France
The new requirements of accuracy in reactor physics calculations
justify a review of the current models, like that one adopted in the
cross-section processing, for a better evaluation of the keff and theresonant absorption rates. In this context, the aim of this paper is toinvestigate the effects of the free gas kernel developed by Sanchez on
the 238U absorption rates. Homogeneous medium tests point out theincrease of the absorption rates in the left wing of the resonances dueto the upscattering produced by the new kernel. Heterogeneous tests
show that the absorption in the left wing of the resonances is mostly
affected by the scattering anisotropy in the laboratory system.

13:50 PMThe Up-Scattering Treatment in the Fine-Structure
Self-Shielding Method in APOLLO3®
L.Lei-Mao, I.Zmijarevic
CEA, Gif-sur-Yvette Cedex, France
The use of the exact elastic scattering in resonance domain introduces
the neutron upscattering which must be taken into account in thedeterministic transport code. We present the newly implemented upscattering treatment in the fine-structure self-shielding method of
APOLLO3 ® . Two pin cell calculations have been carried out in orderto evaluate the impact of the up-scattering treatment. The results arecompared to those obtained by the Monte Carlo code TRIPOLI-4 ®
with its newly implemented DBRC model. The comparison of k-eff
values on the examples of single cell calculations shows a very good
agreement between the APOLLO3 ® up-scattering treatment and theTRIPOLI-4 ® DBRC model, which is less than 30 pcm for UOX fuel andless than 110 pcm for MOX. Also, the differential effects of asymptotic
versus exact kernel produced by APOLLO3 ® compared to TRIPOLI-4 ®,do not exceed 20 pcm for the UOX cell and 40 pcm for the MOX cell.
A detailed comparison of the U238 absorption rates shows clearly the
influences of the first four big resonances of U238 to the calculation
results.

14:10 PM

Target Motion Sampling Temperature Treatment
Technique with Track-Length Estimators in OpenMC- Preliminary Results
T.Viitanen, J.Leppanen(1), B.Forget(2)1)VTT Technical Research Centre of Finland, VTT, Finland, 2)
Massachusetts Institute of Technology, MA, USA
This paper examines the applicability of the Target Motion Sampling(TMS) temperature treatment method together with track-length
estimators. Several track-length estimator based quantities arecalculated in four test cases using a preliminary implementationof the method in OpenMC. The results are compared to an NJOYbased reference. The study reveals a statistically significant bias in
the estimator results, but the errors were found to only affect limited
energy regions, and their magnitude is relatively small as long as
energy-integrated estimators are considered.

Doppler broadened before transport starts, and the impact on crosssection loading is negligible for all but a few isotopes. Results arecompared with those obtained by using multigroup libraries. Current
results compare favorably with these expected results.

14:50 PM

Verification of Doubly-Heterogeneous Self-Shielding
Method Based on Equivalance Theory
S.Choi, D.Lee(1), M.L.Williams(2)1)UNIST, Ulsan, Korea, 2)ORNL, Oak Ridge, USA
A new methodology has been developed to treat resonance selfshielding in double heterogeneity systems such as VHTR. The new
method adopts equivalence theory in both micro- and macro- level
heterogeneities and uses the resonance integral table with a modified
interpolation parameter, i.e., background cross section. This paper
presents the implementation of the new method and verification resultsfor various pin cell problems. The verification results show acceptableaccuracy compared to the results of Monte Carlo code. A formulationof correction factor for Dancoff factor has also been introduced to
improve the accuracy.

15:10 PM

Quantification of Resonance Interference Effectfor Multi-Group Effective Cross-Section in Lattice
Physics Calculation
H.Koike, K.Yamaji, K.Kirimura, S.Kosaka, H.Matsumoto
Mitsubishi Heavy Industries, Ltd., Kobe, Japan
This paper presents the quantification of resonance interference effectfor multi-group effective cross-section in lattice physics calculation.
In the resonance self-shielding method based on the equivalence
theory, the resonance interference effect among multiple nuclides
cannot be treated directly to the multi-group effective cross-section.
The continuous energy or the ultra-fine-group treatment can directly
consider the effect, but the application to the fuel assembly geometry
is not realistic with practical computation time. In the present study,
the resonance interference effect to the multi-group effective crosssection is simply quantified by the resonance interference factor (RIF)
in order to confirm the benefit for considering the effect. The RIF is
generated for the typical pin-cell geometry of water moderated system.
The multi-group effective cross-sections with and without RIFs arecompared with the continuous energy Monte-Carlo result. As a result,the significant impact for considering the resonance interference
effect is confirmed to the limited nuclide, reaction type and energy
group. Fortunately, these have small effect on k-infinity because
the resonance interference effect is mainly induced by the wide
resonances of 238U to the other minor nuclides (e.g., 235U, 239Pu) in thelimited resonance energy ranges. The results also show that the effectis small to the absorption cross-section of 238U, which is the dominantresonance nuclide in the fuel. The quantification results in the present
study indicate a useful material to investigate the more advanced
resonance treatment for the next generation lattice physics code.

14:30 PM

Problem-Dependent Doppler Broadening ofContinuous-Energy Cross Sections in the KENO
Monte Carlo Computer Code
S.W.D.Hart, G.I.Maldonado(1), C.Celik, L.C.Leal(2)1)The University of Tennessee, Knoxville, USA, 2)Oak Ridge National
Laboratory, Oak Ridge, USA
This paper discusses recent progress in the SCALE KENO Monte
Carlo code to generate problem-dependent, Doppler-broadenedcross sections. The approach investigated in this study uses a finitedifference method to calculate the temperature-dependent crosssections for the 1D data, and a simple linear-logarithmic interpolationin the square root of temperature for the probability tables. With thecurrent approach, the temperature-dependent cross sections are

69

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Mizuho_B

Track4-5 Verification, Validation and Uncertainty Analysis
Session Chair: Maria Avramova(Penn State University), Willem F.G.van Rooijen(Univ. of Fukui)
13:30 PM

Deterministic Approach of the Decay Heat Uncertainty
Due to JEFF-3.1.1 Nuclear Data Uncertainties with the
CYRUS Tool and the DARWIN2.3 Depletion Code
V.Vallet, S.Lahaye, A.Tsilanizara, L.San-Felice, R.Eschbach
CEA/DEN, Gif sur Yvette Cedex, France
The knowledge of the decay heat value and its associated uncertainty
is of paramount importance for safety concerns in nuclear power
plants, storage and transport facilities. The CYRUS tool has beendeveloped in order to perform sensitivity analysis and uncertainty
propagation on fuel inventories and decay heat due to the propagationof the variance and covariance data of all the nuclear data involvedin the decay heat calculation (cross-sections, branching ratios,
radioactive decay constants, fission yields, mean decay energies),
for all kind of fuels and reactors. This tool works together with thereference depletion code DARWIN2.3 and is based on a deterministic
method. In the meantime, the CEA/DEN also developed a probabilistic
approach based on the URANIE uncertainty platform and the
MENDEL depletion code.
The purpose of this paper is to set up the bases of the deterministic
method in CYRUS, to discuss the choice of the different physicalcorrelations between the parameters, and to validate CYRUS by anumerical comparison with the probabilistic approach implemented in
URANIE/MENDEL. Two preliminary applications of elementary fission
bursts have been investigated: the thermal fission of 235U and thefast fission of 239Pu. All the nuclear data come from the JEFF-3.1.1
evaluation. Eventually, the comparison of the deterministic approachof CYRUS and the probabilistic approach reveals good agreement
between the two methods.

13:50 PM

Bias and Uncertainty Assessment of Pressurized
Water Reactor Fuel Isotopics Using SCALE
R.N.Bratton, K.N.Ivanov(1), W.A.Wieselquist, M.A.Jessee(2)1)The Pennsylvania State University, Pennsylvania, USA, 2)Oak Ridge
National Laboratory, Tennessee, USA
The purpose of this study is to investigate bias and uncertainty infuel isotopic calculations for a well-defined radiochemical assay
benchmark with Sampler, the new sampling-based uncertainty
quantification tool in the SCALE code system. Isotopic predictions arecompared to measurements of fuel rod MKP109 of assembly D047
from the Calvert Cliffs Unit 1 core at three axial locations, representing
a range of discharged fuel burnups. A methodology is developed
which quantifies the significance of input parameter uncertainties andmodeling decisions on isotopic prediction by comparing to isotopic
measurement uncertainties. The SCALE Sampler model of the D047
assembly incorporates input parameter uncertainties for key input datasuch as multigroup cross sections, decay constants, fission productyields, the cladding thickness, and the power history for fuel rod
MKP109. The effects of each set of input parameter uncertainty on theuncertainty of isotopic predictions have been quantified. In this work,
isotopic prediction biases are identified and an investigation into theirsources is proposed; namely, biases have been identified for certain
plutonium, europium, and gadolinium isotopes for all three axial
locations. Moreover, isotopic prediction uncertainty resulting from only
nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160.

14:10 PM

Inventory Uncertainty Quantification and Propagation
Using TENDL Covariance Data in FISPACT-II
J.C.Sublet(1), J.W.Eastwood, J.G.Morgan(2)1)United Kingdom Atomic Energy Authority, Abingdon, UK, 2)Culham
Electromagnetics Ltd., Abingdon, UK
The new inventory code FISPACT-II provides predictions of inventory,
radiological quantities and their uncertainties using nuclear data
covariance information. Central to the method is a novel fast pathways
search algorithm using directed graphs. The pathways output provides(1) an aid to identifying important reactions, (2) fast estimates of
uncertainties, (3) reduced models that retain important nuclides
and reactions for use in the code’s Monte-Carlo sensitivity analysis
module. Described are the methods that are being implemented for
improving uncertainty predictions, quantification and propagation using
the covariance data that the recent nuclear data libraries contain. In
the TENDL library, above the upper energy of the resolved resonance70

range, a Monte Carlo method in which the covariance data comefrom uncertainties of the nuclear model calculations is used. The
nuclear data files are read directly by FISPACT-II without any further
intermediate processing. Variance and covariance data are processed
and used by FISPACT-II to compute uncertainties in collapsed
cross-sections, and these are in turn used to predict uncertainties ininventories and all derived radiological data.

14:30 PM

Validation of CASMO5 Spent Fuel Isotopics with
Decay and Fission Yield Uncertainties
J.Hykes, J.Rhodes
Studsvik Scandpower, Inc., Idaho Falls, USA
We examine the effects of the uncertainty in decay and fission yielddata on CASMO5- computed nuclear fuel isotopics, comparing theresults to measurements. The uncertainties of radioactive decay
constants, decay and (n,γ) capture branching fractions, and fission
yields are extracted from the ENDF/B-VII.1 data files or estimated
where they are lacking. These nuclear data uncertainties arepropagated through the CASMO5 lattice depletions via Monte Carlosampling. The distribution of computed isotopic compositions arecompared to the Fukushima Daini-2 measurements of Nakahara. The
uncertainties in the calculated values are able to explain some of thecalculation-to-measurement discrepancies, particularly for Am-242m
and the samarium isotopes.

14:50 PM

Uncertainty Propagation and Sensitivity Analysis in
the ALEPH Monte Carlo Burnup Code: Applications to
Fission Pulse Decay Heat Calculations
L.Fiorito(1), G.van den Eynde, A.Stankovskiy(2), O.Cabellos, C.
Diez(3)
1)SCKCEN, Mol, Belgium / ULB, Brussels, Belgium, 2)SCKCEN, Mol,
Belgium, 3)UPM, Madrid, Spain
Uncertainty propagation and sensitivity analysis for burnup codes
are key topics in the nuclear field because of their huge involvementin safety-related problems. Consequently, a correct evaluation of agiven response variation after propagating the system parameter
uncertainties is a primary feature for any burnup code.
In this paper we described the options of linear sensitivity analysis,
either forward or adjoint, recently implemented in the ALEPH Monte
Carlo burnup code. The testing phase was carried out against
the fission pulse decay heat problem for 239Pu thermal fission
and compared to total Monte Carlo results, also produced by thecode through its dedicated routines. The different techniques givecomparable results thus supporting the consistency of the methods. To
conclude, we also calculated and listed the parameters that contributethe most to decay heat perturbations.

15:10 PM

Uncertainty Quantification of Spent Fuel Nuclide
Compositions Due to Cross Sections, Decay
Constants and Fission Yields
O.Leray, P.Grimm, M.Hursin, H.Ferroukhi, A.Pautz
Paul Scherrer Institut, Villigen, Switzerland
Uncertainty Quantification (UQ) helps to understand discrepancies
between predictions and experiments by providing confidence intervals
in the results. A tool, referred to as SharkX, is under development at
PSI for UQ studies in relation to the CASMO-5 lattice physics code.
In this paper, calculation-to-experimental ratios (C/Es) for the nuclide
compositions of a UO2 spent fuel sample from the LWR-PROTEUS
Phase II program are presented with both experimental and nuclear
data uncertainties. The uncertainty propagation of the nuclear datais performed using a stochastic sampling method for cross sections
and decay constants but also for fission yields. For the latter, a new
perturbation methodology taking into account physical constraints
is developed and detailed in this paper. This methodology helps toquantify the uncertainty on kinf and the nuclides compositions withrespect to the fission yields uncertainties. Its results are comparedwith that of straightforward perturbations and basic normalization
methodologies without consideration of correlations. These results
also highlight the very strong impact of the uncertainty of the I-133
yield due to U-235 fissions (σ = 64% in ENDF/B-VII.1) on both kinf

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Mizuho_B

Track4-5 Verification, Validation and Uncertainty Analysis
Session Chair: Maria Avramova(Penn State University), Willem F.G.van Rooijen(Univ. of Fukui)
and caesium composition uncertainties. An alternative source of
uncertainty on this fission yield has also been considered and shows
more reasonable results.

71

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Hiei

Track7-1 Reactor Concepts and Designs
Session Chair: Bojan Petrovic(Georgia Tech. Univ.), Akiyuki Tsuchiya(Hitachi GE)
13:30 PM

Conceptual Study of a Long-Life Prototype Gen-IV
Sodium-Cooled Fast Reactor (PGSFR)
D.Hartanto, Y.Kim
KAIST, Daejeon, Korea
A long-life Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR)
has been investigated from the neutronic perspective. This reactor
will serve as an alternative plan while waiting for the availability of
reprocessing technology in Korea. The objective of this study is toextend the lifetime of the PGSFR to more than 20 EFPYs and have an
average core discharge burnup that is greater than 100 GWd/MTHM.
The objective has been achieved by significantly increasing the fuelvolume fraction, the use of a better reflector, and the utilization of anaxial blanket while minimizing modifications from the current PGSFR
design. After fulfilling the design target, the long-life PGSFR has beencharacterized. Several parameters such as the conversion ratio,
coolant void reactivity, power distribution, and fast neutron fluence
have been calculated. The safety concern due to the positive reactivity
insertion due to fuel compaction of the long-life PGSFR core has also
been investigated. All neutronics calculations have been performed
using the Monte Carlo code McCARD.

13:50 PMCore Design Studies on the Fast Reactor with Flexible
Breeding Ratio
Y.Xiao, H.Wu, Y.Zheng
Xi'an Jiaotong University, Shaanxi, China
This paper presents the conceptual design of a sodium cooled fast
reactor with flexible breeding ratio. It can achieve variable breeding
ratios without significant changes in core design. This concept enables
fast reactor effectively responds to the industrial needs. The U-Pu-Zr
alloy with better breeding and thermal performance is chosen as fuel.
Blankets contain PWR spent fuels are axially and radially arranged.
The number and layout of assemblies are adjusted to achieve thetarget breeding ratios. The achieved breeding ratios varied from 1.0 to
1.4. All the reactor cores use the same fuel design. Three enrichment
are appropriate selected to ensure all reactors maintain critical during
the operation. The assembly number and their layout are adjusted toreduce the maximum linear power. Two control systems are designed
and evaluated to guarantee that the capability of either system fulfilled
its control function. Concerned reactivity coefficients are calculated
at the equilibrium cycle. The temperature distribution is obtained by
the single channel heat transfer calculation and the preliminary safety
analysis is evaluated by the quasi-static methodology to indicate theinherent safety features.

14:10 PMEnhanced Feedback Effects in Sodium Cooled Fast
Reactors Using Moderating Material - The Effect of
the Plutonium Content in the Fuel
B.MerkHelmholtz-Zentrum Dresden-Rossendorf, Dresden, Germany
The use of fine distributed moderating material to enhance thenegative feedback effects and to reduce the sodium void affecting
sodium cooled fast reactor cores is described. The influence of themoderating material on the neutron spectrum is given and evaluatedthrough impact on the capture cross sections of major materials (U238, Pu-239, and Pu- 240). The influence of the variation of the Pu
content on the efficiency of the enhancement of the Doppler effectand on the reduction of the positive coolant and sodium void effectin a representative SFR fuel assembly configuration is analyzed.
Additionally the influence of the moderating material combined withthe variation of the Pu content on the infinite multiplication factor isstudied.

72

14:30 PMAdvanced Sodium Cooled Reactor Cores Having
Thorium Blankets for Effective Burning of Transuranic
NuclidesW.S.You, S.G.Hong
Kyung Hee University, Gyeonggi-do, Korea
In this paper, a design concept of 400 MWe sodium cooled fast
reactor (SFR) cores having thorium blankets for effective burning of
TRU (Transuranics) from LWR spent fuel is described. Specifically,
we considered two recycling options of thorium blankets : 1) norecycling and 2) fully recycling. The thorium blankets are loaded in theaxially central regions of the core regions and their axial heights areadjusted so as to increase TRU burning rate and to reduce burnup
reactivity swing. Also, we analyzed the performances of the coreshaving different fuel management batch sizes and different recycling
options for the searched core configuration. The results show that theaxial thorium blankets with no recycling option can be effectively used
to increase TRU burning rate with a significant reduction of burnup
reactivity swing in comparison with typical SFR burner cores having noblankets while the recycling of thorium blanket degrades TRU burning
rate and burnup reactivity swing but it leads to a reduction of sodium
void worth and more negative Doppler coefficient.

14:50 PMPHISICS Improvements and Comparative Study withERANOS 2.2 on the Gen-IV Lead-Cooled Fast Reactor
Concept, ALFRED
F.Lodi, M.Sumini(1), G.Grasso(2), C.Rabiti(3)1)University of Bologna, Bologna, Italy, 2)ENEA, Bologna, Italy, 3)Idaho
National Laboratory, Idaho Falls, USA
Recent improvements in the PHISICS code are listed and discussed.
Also, the assessment of the PHISICS code for fast reactor design
applications is considered. For this purpose a comparison withthe deterministic code ERANOS2.2 and the stochastic MCNPX is
performed on the lead-cooled concept ALFRED. The results indicate
good general agreement between the codes on the integral quantities
usually of interest in reactor design. Also the analysis highlighted
PHISICS's possibility of keeping low the computational time evenwhen using higher order approximations.

15:10 PM

Velocity Characteristic and Stability of Wave Solutions
for a CANDLE Reactor with Thermal Feedback
V.M.Khotyayintsev, A.V.Aksonov (1) , O.M.Khotyayintseva,
V.M.Pavlovych(2)1)T. Shevchenko National University of Kyiv, Kviv, Ukraine 2)Institute for
Nuclear Research of NAS of Ukraine, Kyiv, Ukraine
We study nuclear burning wave (NBW) of a steady shape in the U-Pu
CANDLE reactor analytically and numerically taking into accountnuclear density mechanisms which are responsible for the wave
velocity formation and thermal feedback to power. One dimensionalmodel of the infinite cylindrical reactor with the absorber which does
not burn out includes equations for nuclear number densities and one
group diffusion equation for neutron flux which includes quadratic influx term responsible for instant thermal feedback in the simplest form.
We show that thermal feedback and mechanism related to finite lifetime of the short-living 239Np additively contribute to velocity formationand compete together with the 241Pu mechanism which dominates atlower velocities and leads to instability of the corresponding stationary
wave solutions. Negative feedback decreases the wave velocity andlowers possible absorber concentrations. Under realistic conditions
both neptunium and plutonium nuclear density mechanisms andthermal feedback may be important for the wave velocity formation in
CANDLE reactors.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Atago

Track3-5 Monte Carlo Methods
Session Chair: Bradley Rearden(ORNL), Marco Pecchia(PSI)
13:30 PM

Effects of Stochastic Noise on a Three-Dimensional
Monte Carlo Depletion Analysis of the H.B. Robinson
Reactor
S.J.Spychala, D.P.Griesheimer

Bettis Atomic Power Laboratory, Pennsylvania, USA
Monte Carlo depletion calculations for nuclear reactors are affected by
the presence of stochastic noise in the local flux estimates produced
during the calculation. The effects of this random noise and its
propagation between timesteps during long depletion simulations are
not well understood. To improve this understanding, a series of Monte
Carlo depletion simulations have been conducted for a 3-D, eighthcore model of the H.B. Robinson PWR. The studies were performedby using the in-line depletion capability of the MC21 Monte Carlo code
to produce multiple independent depletion simulations. Global andlocal results from each simulation are compared in order to determine
the variance among the different depletion realizations. These
comparisons indicate that global quantities, such as eigenvalue (keff),
do not tend to diverge among the independent depletion calculations.
However, local quantities, such as fuel concentration, can deviate
wildly between independent depletion realizations, especially athigh burnup levels. Analysis and discussion of the results from thestudy are provided, along with several new observations regarding
the propagation of random noise during Monte Carlo depletioncalculations.

13:50 PM

Efficient Estimation of Adjoint-Weighted Kinetics
Parameters in the Monte Carlo Wielandt Calculations
S.H.Choi(1), H.J.Shim(2)1)Korea Electrical Engineering & Science Research Institute, Seoul,
Korea, 2)Seoul National University, Korea
The effective delayed neutron fraction, βeff, and the prompt neutron
generation time, Λ, in the point kinetics equation are weighted by theadjoint flux to improve the accuracy of the reactivity estimate. Recently
the Monte Carlo (MC) kinetics parameter estimation methods by using
the adjoint flux calculated in the MC forward simulations have beendeveloped and successfully applied for reactor analyses. Howeverthese adjoint estimation methods based on the cycle-by-cycle
genealogical table require a huge memory size to store the pedigree
hierarchy. In this paper, we present a new adjoint estimation method
in which the pedigree of a single history is utilized by applying the MC
Wielandt method. The algorithm of the new method is derived and its
effectiveness is demonstrated in the kinetics parameter estimations forinfinite homogeneous two-group problems and critical facilities.

14:10 PM

Higher-Order Chebyshev Rational Approximation
Method (CRAM)
M.PusaVTT Technical Research Centre of Finland, Finland
The burnup equations can in principle be solved by computing
the exponential of the burnup matrix. However, due to the difficult
numerical characteristics of burnup matrices, the problem is extremely
stiff, and the matrix exponential solution was long considered infeasible
for an entire burnup system containing over a thousand nuclides. Afterdiscovering that the eigenvalues of burnup matrices are generally
confined to a region near the negative real axis, the Chebyshev
rational approximation method (CRAM) was introduced as a novel
method to solve the burnup equations. It can be characterized as thebest rational function on the negative real axis and it has been shown
to be capable of simultaneously solving an entire burnup systemboth accurately and efficiently. The main difficulty in using CRAM forcomputing the matrix exponential is determining the coefficients of therational function for a given approximation order. Some polynomial
CRAM coefficients have been published in 1984, and based on these
literature values, CRAM approximations up to the order 16 have beenthus far applied in burnup calculations. The topic of this paper is thecomputation of CRAM approximations and their application to burnup
equations. A Remez-type method utilizing the equioscillation property
of best approximations is used to construct the CRAM approximants
for approximation orders 1, ..., 50. Numerical results are presented
for a large burnup system and for a decay system. It is demonstrated
that higher-order CRAM can be used to accurately solve the burnup

equations even with time steps of the order of millions of years.

14:30 PM

Efficiency and Accuracy Evaulation of the Windowed
Multipole Direct Doppler Broadening Method
C.Josey, B.Forget, K.Smith
Massachusetts Institute of Technology, Massachusetts, USA
It was recently proposed to use the multipole formalism for direct
Doppler broadening by reducing the number of poles that needbroadening and approximating the contributions of distant poles.
The multipole representation takes the Reich-Moore formalism andrestructures it into a pole and residue form that can be analytically
Doppler broadened using the Solbrig kernel. The windowed multipole
method reduces the number of calculations required by approximating
the contribution of far away poles with a fitting function. Additionally,
approximate Faddeeva function evaluations can be used to reduce
computational costs. A detailed study was performed on 238U todetermine the tradeoff between efficiency, memory requirements andaccuracy. It was determined that accuracy could be gained at little tono additional cost by fitting the nonfluctuating poles in windows thatwere equally spaced in momentum instead of energy as had been
done prior. It was also determined that outer window sizes can be
tailored to a specific accuracy level. Two sets of promising parameters
for accuracy and memory requirements were selected and tested at
two different temperatures on a slowing down problem in 238U andhydrogen over the resolved resonance range. Various approximationsof the Faddeeva function evaluations were also compared todetermine the tradeoff between accuracy and efficiency. The results
indicated that a performance comparable to table lookup is achievable.
Table lookup with temperature interpolation required 0.68 μs per 238U
cross-section evaluation, while windowed multipole ranged from 0.35
to 2.55 μs depending mostly on the Faddeeva approximation used.
The windowed multipole required roughly 300 kB of storage for the
resolved resonance range covering all temperatures of interest andthree cross-sections of interest, while a typical single temperature
piece-wise linear library in the resonance range required roughly 5 MB
for the same data.

14:50 PM

Diffusion Monte Carlo Method with Transport
Corrections
A.Srivastava, Y.S.Rana, S.B.Degweker
Bhabha Atomic Research Centre, Mumbai, India
Analytical Kernel Based Diffusion Monte Carlo method has beendeveloped for simulating proposed reactor noise experiments formeasuring the degree of sub-criticality in accelerator driven systems.
Although this approach has several advantages such as speed,
elegance and exactitude, but it was applicable to a rather restrictedclass of problems, such as the bare homogeneous reactor. Thisapproach has been further developed to demonstrate its utility in a
wider class of problems involving heterogeneous media and forms the
subject of this paper. We also investigate whether and to what extent
the diffusion based Monte Carlo can be improved to give results closer
to transport theory, particularly in situations wherein diffusion theory
methods are otherwise inapplicable.

15:10 PM

Generation of One-Group Self Shielded Cross
Sections with Multi-Group Approach for Monte Carlo
Burnup Codes
D.Kotlyar(1), E.Shwageraus(2), E.Fridman(3)1)Ben Gurion University, Be’er Sheva, Israel, 2)University of Cambridge,
Cambridge, UK, 3)Helmholtz-Zentrum Dresden-Rossendorf, Bautzner
Landstraße, Germany
Allowing Monte Carlo (MC) codes to perform fuel cycle calculations
requires coupling to a point depletion solver. In order to perform
depletion calculations, one-group (1-g) cross sections must be
provided in advance. This paper focuses on generating accurate
1-g cross section values that are necessary for evaluation of nuclide
densities as a function of burnup. The proposed method is analternative to the conventional direct reaction rate tally approach,which requires extensive computational efforts. The method presented
here is based on the multi-group (MG) approach, in which pre-73

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Atago

Track3-5 Monte Carlo Methods
Session Chair: Bradley Rearden(ORNL), Marco Pecchia(PSI)
generated MG sets are collapsed with MC calculated flux. In ourprevious studies, we showed that generating accurate 1-g crosssections requires their tabulation against the background cross-section(σ0) to account for the self-shielding effect. However, in previous
studies, the model that was used to calculate σ0 was simplified by
fixing Bell and Dancoff factors. This work demonstrates that 1-g values
calculated under the previous simplified model may not agree with thetallied values. Therefore, the original background cross section model
was extended by implicitly accounting for the Dancoff and Bell factors.
The method developed here reconstructs the correct value of σ0 by
utilizing statistical data generated within the MC transport calculation
by default. The proposed method was implemented into BGCore
code system. The 1-g cross section values generated by BGCore
were compared with those tallied directly from the MCNP code. Very
good agreement (<0.05%) in the 1-g cross values was observed. The
method does not carry any additional computational burden and it isuniversally applicable to the analysis of thermal as well as fast reactor
systems.

74

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 13:30 PM - Cosmos

Track15 Research Related to Fukushima Accident
Session Chair: Alireza Haghighat(Virginia Tech Univ.), Akio Yamamoto(Nagoya Univ.)
13:30 PMAREVA Dismantling and Decommisioning Experienceand Fuel Debris Removal Approach for Fukushima
Dai-ichiGuest Speaker: K.Schauer
AREVA, France

13:50 PM

Re-Criticality Potential at Fukushimi Dai Ichi Unit 4
A.H.Wells(1), A.J.Machiels, A.G.Sowder(2)1)Consultant, Atlanta, USA, 2)EPRI, Palo Alto, USA
The damage to Fukushima Daiichi Unit 4 reactor buildings discoveredon March 15, 2011, led to concerns that the Unit 4 spent fuel pool had
leaked and might have gone dry. The possibility of a re-criticality event
occurring following reflooding of a fully drained Unit 4 spent fuel pool
was a concern. Uncertainties regarding the status of the fuel and thepool made it difficult to apply conventional criticality safety methods.
However, EPRI was able to apply an alternative criticality safety
methodology, developed for the analysis of spent fuel criticality duringtransportation accidents, to show that the most likely configuration ofdamaged fuel and pool would be a decrease in reactivity. Thus it wassafe to reflood the pool.

14:10 PMWhy a Criticality Excursion was Possible in the
Fukushima Spent Fuel Pool
A.Sargeni, G.Caplin
IRSN, Fontenay-aux-Roses, France
During the Fukushima event, IRSN performed some calculations
to assess whether a criticality excursion was likely to occur in caseof a loss of coolant of the spent fuel pools (dry-out). The resultsof these calculations show that the k-eff of the spent fuel pool canincrease when a water-air mixture is modeled within the storage (withsubsequent water density decrease), compared to the k-eff value
with a full water density within the pool. This k-eff increase reaches
more than 5% for an optimal range of water-air mixture, which is morethan usual safety margins kept to demonstrate the safety of the spent
fuel pools. This initial study showed that the pitch between the fuelassemblies, the materials (steel with more or less boron) composing
the storage rack and the water densities are the main parameters
driving the magnitude of the k-eff increase.
All the above calculations were performed with the standard routeof the CRISTAL package and, in order to interpret and understandthe here-above results, investigations were performed in parallelusing pure Monte-Carlo pointwise calculations and pure deterministic
calculations so as to first confirm, then to explain the observed results
using a reactivity splitting via the four factors formula. This paper
presents these complementary studies showing how, when water
density decreases and inter-assembly gap increases, the balance
between increasing fast fissions and water absorption may justify areactivity increase.

14:30 PM

Critical Experiments for Fuel Debris Using Modified
STACY
K.Izawa, K.Tonoike, H.Sonov, Y.Miyoshi

the severe accident of the Fukushima Daiichi Nuclear Power Station.
This report introduces the plan of criticality safety research in Japan
Atomic Energy Agency following the accident, and describes the roleof the modified STACY in the retrieval work of fuel debris from thedamaged reactor.

14:50 PM

Concept of Capture Credit Based on Neutron Induced
Gamma Ray Spectroscopy
Y.Nauchi, H.Ohta(1), H.Unesaki, T.Sano, T.Yagi(2)1)Central Research Institute of Electric Power Industry, Tokyo, Japan, 2)
Research Reactor Institute, Kyoto University, Kumatori, Japan
Capture credit (CapC) based on neutron induced gamma ray
spectroscopy (NIGS) is proposed to estimate the sub-criticality marginof fuel debris in which nuclear fuel and structural materials are comelted or mixed. By NIGS, rates of some capture reactions can bemeasured in relation to fission reactions. By the ratio, we can credit
the sub-criticality margin of the fuel debris caused by the capture
reactions. The theory of CapC is described. In order to demonstrate
how CapC is beneficial, numerical simulations are performed for ahypothetical array of canisters in which the fuel debris is stored. An
application procedure of the CapC is also proposed, which consistsof several technologies: 1) NIGS, 2) simulations of a response andan efficiency of the γ ray detection, 3) unfolding of the γ ray spectrum
to obtain reaction rates. Experimental studies of NIGS have beenlaunched in Kyoto university critical assembly facility. NIGS is firstly
studied for simulated fuel debris of a few kinds of mixture of stainlesssteel and uranium. The measured γ ray spectra and preliminary
analysis give promising results for CapC based on NIGS.

15:10 PMA Methodology for Fast and Accurate Decay Heat
Calculations for In-Pool Used Fuel Assemblies
Developed at AREVA La Hague Reprocessing Facility
A.Launay, G.Grassi, E.Lecampion, J.Lelandais(1), R.Eschbach, L.
San Felice(2)1)AREVA NC Recycling, Beaumont Hague, France, 2)CEA Cadarache,
St Paul-lez-Durance, France
The evaluation of the decay heat released from used fuels stored
in an interim storage pool is a crucial issue for safety and operation.
This paper presents a methodology developed at the AREVA NC La
Hague reprocessing facility in order to perform a fast and accurate
assessment of the decay heat for a pool containing a large variety
of fuel assemblies with different features (kind of reactor, initial
composition, number of irradiation cycles, operating power, outage
length) and cooling times ranging from 6 months to 60 years. The
key aspect of the methodology relies on the fact that, for coolingtimes higher than 6 months, only a very short number of isotopes
significantly contribute to the overall decay heat. The methodology
uses results from calculations performed by the industrial depletion
code CESAR and takes benefit from the experimental validation of
CEA reference depletion package DARWIN. A full CESAR calculation
at a given cooling time for each assembly allows building up areference database: for any cooling time higher than the reference
one, a simplified on-line calculation involving less than 30 isotopes isperformed, and all approximations made are covered by the mean of
biases relying on pre-evaluated coefficients. Results for an application
to a 10,000-used-fuel-assembly interim storage pool are given.

Japan Atomic Energy Agency, Ibaraki, Japan
Critical assemblies of the thermal neutron system are decreasing in
number in spite of their important roles in the reactor physics research,the design of nuclear facilities and the human resource training. On
the other hand, the utilization term extension of the light water reactors
brings new research themes requiring critical experiments of thethermal neutron system; e.g., the new fuel design with higher burnup, the introduction of burn up credit into the criticality safety controlof spent fuel, the criticality safety control of fuel debris generated ina severe accident of a reactor, etc. Japan Atomic Energy Agency ismodifying the Static Experiment Critical Facility (STACY) to revive thecritical experiments. The modified STACY will be an infrastructure, amultipurpose critical assembly, for the experimental research of thereactor physics on thermal neutron system. The primary mission of themodified STACY at present is the critical experiments for fuel debris tocontribute to the criticality safety control of the fuel debris generated by75

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
Track1: Reactor Analysis Methods
PAPER 1089476

A Parametric Study and Comparison of BWR Fuel
Depletion Calculations Using CASMO-4, MCNPX, and
SCALE/TRITON
C.Lin, M.Chang, R.Sheu, Y.Liu

8

8

8

Institute of Nuclear Engineering and Science, National Tsinghua
University, Hsinchu, Taiwan
CASMO-4 is a multigroup two-dimensional transport code for LWR
lattice physics calculations. MCNPX and TRITON/T6-Depl are twogeneral-purpose transport codes with depletion capability for various
fuel designs. MCNPX can use continuous-energy cross sections while
TRITON currently only supports multigroup depletion calculations.
This study presented a systematic comparison of these three codesfor depletion calculations of a typical BWR fuel assembly. Keyparameters for sensitivity studies were neutron cross-section libraries,
burnup steps, modeling of poison rods, inclusion of additional nuclides
for depletion, thermal expansion, pin-by-pin depletion, and Dancoff
factors. The CASMO-4 results were arbitrarily taken as a reference
base on which the differences of MCNPX or TRITON calculations
were evaluated. Useful observations from the comparisons wereas follows: The ENDF/B-VII cross-section library gave the mostconsistent result with CASMO-4. At least five radially subdivided
zoning of a Gd-bearing rod was necessary for depletion calculations.
MCNPX calculations were more sensitive to choices of burnup steps
and numbers of nuclides being traced in fuel inventory than TRITON
did. Applying the same thermal expansion corrections in TRITON
reduced its differences with CASMO-4 in the middle of cycle. Pin-bypin depletion is necessary but only slightly changed k profiles in thiscase compared with average depletion. Using more accurate Dancoff
factors in TRITON resulted in an excellent agreement of k values
with CASMO-4 at the early stage of burnup, but they still graduallydeviated at later burnups. Overall, both MCNPX and TRITON
predicted k profiles in this problem were within 500 pcm agreementwith CASMO-4 in the entire burnup period.

PAPER 1092007

Neutron Noise Induced by Fluctuations of the Boric
Acid Content in Pressurized Water Reactors
H.Malmir, N.Vosoughi
Department of Energy Engineering, Sharif University of Technology,
Tehran, Iran
Propagation noise induced by fluctuations of the boron content of
the coolant, in pressurized water reactors (PWRs) is addressed inthis paper. The spatial discretization of the neutron noise equations
is based on the Box-Scheme Finite Difference Method (BSFDM)
for rectangular-z, triangular-z and hexagonal-z geometries. Using
the derived equations, a 3-D 2-group neutron noise simulator isdeveloped, by which the discrete form of both the forward and adjoint
reactor dynamic transfer functions (in the frequency domain) can becalculated. In addition, both types of noise sources, namely point-likeand traveling perturbations, can be modeled by the developed neutron
noise simulator. The results and benchmarking are then reported
for the case study of a VVER-type pressurized water reactor core.
Considering the boric acid fluctuations as traveling perturbations of themacroscopic thermal absorption cross sections, the induced neutron
noise is calculated throughout the reactor core. Space- and frequencydependence of the propagation noise are also investigated in thiswork.

PAPER 1093007

Application of Westinghouse NEXUS/ANC9 CrossSection Model for PWR Accident Analyses
B.Zhang, H.Huria, I.Boyan
Westinghouse Electric Company, Pennsylvania, USA
NEXUS/ANC9 is the latest licensed PWR core design code systemdeveloped by Westinghouse. This system has demonstrated
capabilities of modeling advanced core designs with improved
accuracy in core reactivity and power distribution predictions. NEXUS/ANC9 system is being rolled out to replace the current APA system(ALPHA/PHOENIX-P/ANC) for routine core calculations. In addition
to the standard core design calculations, investigations are underway
to explore the possibility to expand the NEXUS/ANC9 application forsafety analysis, especially at accident conditions. The main focus of76

the investigation is the evaluation of the NEXUS/ANC9 cross-section
representation model conditions like high void and significant changeof core pressure. Comparisons of the predicted parameters among
ANC9, PARAGON lattice code and MCNP calculations are presented.
The results show that NEXUS/ANC9 is able to model the crosssection behavior and accurately reproduce lattice code results at all
simulated conditions.

PAPER 1102611

Numerical Dispersion and Dissipation Analysis of
Nodal Expansion Method
X.Zhou, F.Li
Institute of Nuclear and New Energy Technology, Tsinghua University/ the Key Laboratory of Advanced Reactor Engineering and Safety,
Ministry of Education, Beijing, China
The numerical property of nodal expansion method (NEM) is studied in
the paper from the perspective of numerical dispersion and dissipation,
which is brand new for nodal methods and no one else has ever tried
before. Besides, the more complicated transient convection diffusion
equation is chosen to be the research target so as to be as generaland comprehensive as possible. First, the nature and connotationof dispersion and dissipation is presented. Then, the numericaldispersion and dissipation analysis for NEM is developed with thehelp of Fourier analysis and solution for complex matrix generalized
eigenvalue problem. Through analyzing the numerical dispersion anddissipation of NEM with different order N basis functions as well ascomparing it with the central difference (CD) and first order upwind
scheme (FUS), and with numerical verification, the conclusion isdrawn finally: the numerical dispersion and dissipation of NEM is ofan advance in that it can simulate the rather difficult problems suchas steep gradients, the high frequency analytical solution, convection
dominated problems even in the coarse mesh.

PAPER 1102622

The Integration of Control Rod Calculation and VSOPJ.Guo, F.Li, C.Hao, H.Zhang, X.Zhou, K.Fan, L.Wang, J.Lu
Institute of Nuclear and New Energy Technology, Tsinghua University/ the Key Laboratory of Advanced Reactor Engineering and Safety,
Ministry of Education, Beijing, China
Accurate calculation of the control rod outside the active zone inpebble bed HTR requires special treatment, integration of thisdetailed control rod calculation with whole core calculation like VSOP
package requires more efforts, which is not realized before. Accurate
calculation of control rod is proposed to use the discontinuity factorcorrected diffusion method accompanied with transport calculationin advance. Appropriate coupling between the control rod calculation
and whole core calculation must be developed to take into account of
the buckling feedback, spectrum change, data transform and so on. In
this paper, internal spectrum calculation module inside VSOP is used
to model the control rod region, and the neutron leakage obtained
from the whole core calculation is used as the boundary condition of
the control rod region. The numerical calculation demonstrated thefeasibility and accuracy of this coupling and integration method.

PAPER 1103763

Reflector Modelling with Multi-Group Nodal
Equivalence Theory for the SAFARI-1 Research
Reactor
S.A.Groenewald(1), F.Reitsma(2)1)Necsa, Pretoria, South Africa, 2)IAEA, Vienna, Austria
Normalised Generalised Equivalence Theory is used to model the excore reflector region of the SAFARI-1 research reactor. This method isa one-dimensional homogenisation technique based on Generalised
Equivalence Theory, but with only one discontinuity factor defined pernode, and divided into the nodal parameters. The SAFARI-1 reactor
is modelled with the deterministic code system OSCAR-4. Crosssections for the reflector model is generated with NEWT (part of
the SCALE 6.1 package) and EQUIVA-1 (part of OSCAR-4), whichcalculates the NGET parameters. A period of three years in theoperational history of the SAFARI-1 research reactor is modelled.
Two models are used, one with traditional flux-volume weighted
and the other with equivalent ex-core reflector cross-sections.
The performance of the two models over the three year period iscompared. Reactor parameters such as reactivity and fuel burnup areinvestigated. Comparisons to experimental data, in particular control
rod calibrations, are also made. The model with equivalent reflector

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
parameters shows improved accuracy for control rod calibrations,a power tilt of about 10 % across the core, no noticeable change inreactivity or burnup, and significant improvement in calculational time
(reduced by over 40 %) due to a reduction in the size of the coremodel.

PAPER 1104477

Relationship Between Computed ANSI/ANS-5.1 and
ORIGEN-S Decay Heat Powers for BWR LOCA Safety
Analysis
R.Chiang
Energy Engineering Service, San Jose, USA
The decay heat power fraction computed using ANSI/ANS-5.11979 with CASMO-4 decay heat parameters is compared with the
decay heat power fraction computed using ANSI/ANS-5.1-1979 with
ORIGEN-S decay heat parameters. The comparison indicates that
the ORIGEN-based ANS-5.1 total decay power fraction appears very
close to the CASMO-based ANS-5.1 total decay power fraction due
to compensating effect between fission-product decay heat power
fraction and U-239&Np-239 decay heat power fraction, although
the CASMO-4 fission fractions and U-238 neutron capture ratio are
considered more accurate than the ORIGEN-S fission fractions
and U-238 neutron capture ratio. Therefore, it seems acceptableto calculate the total decay heat fraction using ANSI/ANS-5.1-1979
with ORIGEN-S decay heat parameters. This result is useful, since
ORIGEN-S /SCALE 5.1 are easier to run than CASMO-4.
The decay heat power fraction computed using ANSI/ANS-5.1-1979
is also compared with the decay heat power fraction computed using
ORIGEN-S directly. The comparison indicates that the ORIGEN-S total
decay heat power fraction is much smaller than the corresponding
ANS 5.1 total decay heat power fraction, which is due to the fact that
the ORIGEN-S fission-product decay heat power fraction is muchsmaller than the corresponding ANS 5.1 fission-product decay heat
power fraction. This demonstrates that the total decay heat power
fraction calculated using ORIGEN-S directly is not conservative andthat ANSI/ANS-5.1 must be used to calculate the total decay heat
power fraction for LOCA safety analysis.

PAPER 1104519

A New Method to Measuring the α Eigenvalue of a
Subcritical Reactor System
Y.Bai, B.Yang
Institute of Applied Physics and Computational Mathematics, Beijing,
China
The pulsed source method is one of the methods to confirm thetime multiplication constant (α eigenvalue) for a subcritical reactor.
Practically its applicability is determined by the characteristic of theneutron source. The analytic and numerical methods are both used
to analyze the affect of the temporal characteristic of DPF source inthe pulsed neutron method. The result indicates that the requirementof the temporal characteristic of DPF source in the pulsed neutron
method is ultra fast and ultra short.

PAPER 1105008

Progress towards an Accurate LatticeHomogenization Technique for Pressure-Tube
Supercritical Water Cooled Reactor Neutronic
Calculations
E.Nichita, P.Schwanke
University of Ontario Institute of Technology, Ontario, Canada
The Pressure-Tube Supercritical-Water Cooled Reactor (PT-SCWR)
is one of the reactor concepts investigated by the Generation-IV
International Forum as a reactor concept capable of achieving an
overall nuclear power plant efficiency in the vicinity of 50%, higher
than the approximately 35% that characterize current nuclear power
plants. The development process includes studying the neutronics
methods that can provide an adequately-accurate modelling of thecore. The PT-SCWR evolved from the Pressurized Heavy-Water
Reactor but presents a more heterogeneous core because of themuch higher discharge burnup (~40 MWd/kg vs. ~7.5MWd/kg) andgreater variation of coolant density. This work presents progress
towards addressing the nodehomogenization needs of the PT-SCWR
core by using macroscopic cross sections and discontinuity factors
calculated using quasi-exact node boundary conditions determinedusing global-local iterations between core and lattice calculations .

Results show that cross sections and discontinuity factors obtained
via the global-local iterative process are very effective at improving
the accuracy of full-core diffusion calculations in the fuel region of thecore and that some challenges remain in applying leakage corrections
to high-leakage configurations, especially those comprising reflector
nodes.

PAPER 1105869

Proposal of Subcritical PWR Core Benchmark
Problems
T.Kitada
Osaka University, Osaka, Japan
Subcritical PWR core benchmark problems are proposed in orderto clarify the error in neutron flux distribution and the magnitude ofneutron flux evaluated in fixed-source mode. Proposed benchmark
problems consist of 2-stages named Stage-I and Stage-II, respectively.
The difference caused by the discrepancy in theory used in simulation
code will be discussed for simplified 2-D and 3-D PWR core models
with fixed-source by using of same group constants in Stage-I. Then inStage-II, we are planning to discuss how to evaluate group constantsfor several types of UO2 fuel assemblies used in subcritical systemanalysis. This paper shows proposed benchmark problems withseveral results performed by Monte Carlo code as a reference.

PAPER 1105924

A New Monte Carlo-Deterministic Two-Step Method
for Fast Reactor Diffusion Analysis
W.Heo, Y.Kim
KAIST, Daejeon, Korea
This paper investigates a hybrid Monte Carlo-Deterministic hybrid
method in which the cross sections are generated with the Monte
Carlo method and they are used in 3-D diffusion reactor analysis. In
both the new hybrid and conventional methods, the cross sections are
based on a simple RZ homogeneous core model. As a computationalmodel, a 300MWe SFR (sodium-cooled fast reactor) TRU burner corehas been introduced. The generated multi-group cross sections areused by DIF3D to calculate the keff and assembly
power distribution. A whole core calculation of the heterogeneous
core model was performed using MCNP5 as the reference solution
for the evaluation of the analysis method. For an in-depth verificationof the hybrid, it was also applied to heavily rodded core. In the caseof rodded core, a new core modeling named RRZ was proposed fora better modeling of the self-shielding effect of the control assembly.
Additionally, the nodal equivalence theory was successfully applied to
fast reactor analysis for the first time in this work. To apply the nodal
equivalence theory, a simple 1-D spectral geometry was developed todetermine the flux discontinuity factors of a control assembly region.
The generated group-DF values were used to correct the crosssections of the control assembly region. Moreover, the sensitivity of
the DF to the 1-D spectral geometry model was also evaluated in thiswork. It is concluded that the hybrid method works well for the analysisof the SFR core. Particularly, the special application of the nodal
equivalence theory greatly improves the accuracy of the new hybrid
method for fast reactor analysis.

PAPER 1106049

Explicit Transverse Leakage Treatment Using an
Analytic Basis Function Expansion
S.A.Thompson, K.N.Ivanov
The Pennsylvania State University, Pennsylvania, USA
An explicit method for calculating the transverse leakage is presented
in this paper. The method is based upon the use of analytic basis
functions, which represent individual eigenfunctions of the neutron
diffusion equation. The intranodal flux solution is expressed as an
eigenspace, and can be solved by using the already calculated surface
currents and flux moments as boundary conditions. The salient featureof the method, therefore, is that no ad hoc presumptions are madewith regard to the leakage shape. The individual eigenfunctions arecalculated based upon already calculated parameters from the flux
solution and response matrix solution, and therefore no additionalparameters are introduced into the problem, which could lead to an
unwanted increase in computation time. The new transverse leakage
method is implemented in PSU’s NEM code and is tested against
the OECD/NEA 3D C5G7 rodded MOX benchmark and the C3
benchmark.

77

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
PAPER 1106127

Method for Calculation Capturing Reactions
Contribution to Total Energy Release in Nuclear
Reactors
R.B.Bahdanovich(1), G.V.Tikhomirov, I.S.Saldikov,
E.F.Kryuchkov(2)1)The Belarusian State University, Minsk, Belarus, 2)National Nuclear
Research University “MEPhI”, Moscow, Russia
The growth of computers capabilities gives us the possibility to
conduct computational experiments and obtain reasonable results.
Computer modeling allows investigating models which are impossible
for experimental investigation. This is why we can reexamine somephysical phenomena at the deeper level. The energy release innuclear reactors has been analyzed since the beginning of nuclear
industry, but there are still some features that need to be investigated.
In this paper, basic features of energy release in nuclear reactors
and its components has been examined. The method for calculation
fraction of reactions with neutron disappearance (reaction channels(n,γ), (n,α), (n,p), etc.) in total energy release has been developed.
Calculations and comparative analysis of this fraction for three modelsof Water-Water Power Reactor-1000 (WWER-1000) have been made.
The fraction of capturing energy in total energy release for WWER1000 is about 3.5%. It is shown that depending on fuel assembly’s
type the total energy release could vary by 0.5%.

PAPER 1106156

A Preliminary Analysis of the Accuracy of
Homogenized 2D Cross Section in 3D Nodal
Calculations for BWRs
A.M.Ward, A.C.Hall, T.J.Downar
University of Michigan, Michigan, USA
Current methods in reactor analysis employ a 3D nodal diffusion code
which utilizes homogenized cross sections generated from a 2D latticephysics calculation. This approach has been used successfully for awide range of LWR core design and safety analysis. With the adventof more advanced axial heterogeneous fuel assemblies, particularly
for BWRs, it is worthwhile to assess the accuracy of the 2D crosssection generation method for advanced fuel designs. The goal of thework reported here was to provide an estimate of the accuracy of 2D
cross section in this situation, specifically for control blade insertion.
Results indicated that while the power shape can be estimatedproperly in some cases, significant differences in k-eff were observed.
Work continues in the areas of void distribution, histories, and burnup.

PAPER 1106178

Deterministic Lattice Code Development at UNIST
C.Kong, S.Choi, D.Lee
Ulsan National Institute of Science and Technology, Ulsan, Korea
A new method of characteristics code, STREAM, has been developedat Ulsan National Institute of Science and Technology (UNIST). With
regard to a resonance treatment in STREAM, the gray resonancetreatment method and doubly heterogeneous self-shielding method
were implemented based on the equivalence theory. STREAM cancalculate eigenvalue and pin power distribution of a two-dimensional
multi-assembly with CMFD acceleration. In this paper, designed pin
cell and two-dimensional C5G7 benchmark problems were solved
to verify the STREAM library and solver, respectively. As a practicalproblem, the VENUS-2 benchmark from research reactor experiments
was calculated up to assembly level by the STREAM code. To checkapplicability of STREAM to new fuel analysis for a boron free reactor,
innovative fuel problems were designed and calculated. STREAM
solutions of these problems were compared with the Monte Carloreference solutions, and showed a good agreement with them on asimilar level.

PAPER 1106220

Implications of Mesh Refinement in Lattice Physics
on BWR Core Analysis and Nuclear Design
P.Forslund Guimarães, D.Bourgin, P.Olivius, J.Aurén
Westinghouse Electric Sweden AB, Västerås, Sweden
A BWR axial power tilt and reactivity bias with increasing coolant
void content has been observed due to the application of a spatiallyflat source approximation in lattice transport calculations. In order toaddress such a reactivity bias using the CCCP method of HELIOS

78

one needs to employ a sufficiently fine spatial and angular mesh in
the lattice calculations to limit any undesired axial power oscillation
to occur in real core operation and to reduce the uncertainties inimportant safety parameters such as the SDM. The challenge inthis regard is to find a sensible balance between the use of current
coupling and collision probabilities without a too large penalty incomputational cost rendering such lattice calculations impractical.
Based on both 2D lattice transport and 3D nodal core analyses, it wasrecognized that the most important mesh improvement component
was to add azimuthal coolant regions in the fuel pin cells coming witha rather low computational cost. The use of collision probabilities overlarger sub-domains of non-boiling moderator regions was consideredappropriate mainly based on its theoretical merits as neither significant
differences in results nor mesh convergence was obtained in fulllattice geometry for this modeling choice.

PAPER 1106272

Generating Multigroup Data Stochastically for aHighly Heterogeneous VHTR Problem
K.J.Connolly, F.Rahnema
Georgia Institute of Technology, Atlanta, USA
A necessary step in any reactor physics solution process is thecompilation of nuclear data. While the use of continuous energy crosssections in a whole core Monte Carlo calculation will always yieldaccurate results if enough particle histories are run, such an approach
is so computationally expensive as to be considered not feasible forpractical core eigenvalue calculations. It is therefore necessary to
produce accurate multigroup cross sections for homogenized regionsof the core, and let a faster method determine the core solution. A
particular challenge, due to its more complicated spectrum and higher
heterogeneity, is the generation of data for the VHTR, a gas-cooled
reactor whose fuel is in particle form. In this paper, the stochastic
cross section generation method MoCS-Gen is used to generate 26
group data for the VHTR. Problems in a pin cell and a fuel block aresolved, and compared to the original fully heterogeneous problem
in continuous energy and also to the same problems solved incontinuous energy but utilizing a simple volume homogenization
strategy. Results are presented which clearly show the benefit of
MoCS-Gen as an accurate multigroup cross section generation
method.

PAPER 1106273

Reference Solution for Core Physics
R.Vocka, F.Havluj
UJV Rez, a.s., Czech RepublicCore calculations of nuclear reactors are usually performed bycore physics codes (e.g. with NEM or FDM solvers) in diffusion or
SP3 approximation of the transport equation. For each fuel typeparameterized data libraries are prepared by means of a lattice code.
The data libraries are burnup dependent, and the parameterization
covers the hyperspace of admissible values of all operationalparameters (fuel temperature, moderator density, boron concentration
etc.)
This approach has two weak spots. The first is, that it is difficult
to make perfect parameterization of the data library because of
relatively broad range of the parameter values and the fact that theparameters’ effect on the macroscopic cross-sections are not mutuallyindependent. The second is that even for perfect parameterizations
with precise approximations of the data changes with respect to the
feedback parameters the so-called history effects are neglected. It
is generally difficult to assess the cumulative errors arising due tothe approximative parameterization of the data libraries and dueto the history effects. It is as well difficult to assess the efficiency oftechniques developed in order to incorporate the history effect inthe data library (such as time integration). In this paper we present
a tool for reference core calculations in which the above statedapproximations are eliminated.
This paper presents the solution method, its implementation, as wellas the results of a demonstration calculation showing the improvementof the calculation results over the traditional approach, assessing themagnitude of history and parameterization effects importance.
The most important feature of the presented method is that it provides
the perfect parameterization of macroscopic data, allowing thecore physics code developers to understand sources of modelinguncertainties by completely removing the parameterization error(including, unlike other approaches, a complete representation of thehistory effects). It is therefore a highly important tool for improving theperformance of core physics calculations with to an extent which was
not previously available.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
PAPER 1106738

Development of a Generalized Cross Section Library
Applicable to Various Reactor Types
C.Lee
Argonne National Laboratory, Argonne, USA
A method for generating a cross section library applicable to various
reactor types is presented. In this method, an ultrafine group (2158groups) cross section library is first prepared using NJOY and MC23, which includes absorption, nu-fission, and scattering resonancecross section tables as a function of the background cross section andtemperature. Subsequently, for a specific reactor or reactor type ofinterest, this base ultrafine group library is condensed to broad-groupcross section libraries using the group condensation optimization
algorithm that minimizes the change of cross section and eigenvalue
over different compositions and geometry. Based on equivalence
theory, the escape cross sections representing the local heterogeneity
effect are calculated by iteratively solving the fixed source problems
in which resonance cross sections are updated during the iteration.
Preliminary verification tests indicate that the base ultrafine groupcross section library is able to accurately estimate eigenvalues andcross sections of various compositions from different reactor types
including LWR, HTR, and SFR. The broad group cross section libraries
condensed from the base ultrafine group cross section library for a
specific reactor or reactor type show good agreements in eigenvalue
with corresponding Monte Carlo solutions.

PAPER 1119837

Monte Carlo Analysis of Doppler Reactivity Coefficientfor UO2 Pin Cell Geometry
Y.NagayaJapan Atomic Energy Agency, Ibaraki, Japan
Monte Carlo analysis has been performed to investigate the impactof the exact resonance elastic scattering model on the Doppler
reactivity coefficient for the UO2 pin cell geometry with the parabolic
temperature profile. As a result, the exact scattering model affects thecoefficient similarly for both the flat and parabolic temperature profiles;
it increases the contribution of uranium-238 resonance capture in theenergy region from 〜 16 eV to 〜 150 eV and does uniformly in theradial direction. Then the following conclusions hold for both the exact
and asymptotic resonance scattering models. The Doppler reactivity
coefficient is well reproduced with the definition of the effective fueltemperature (equivalent flat temperature) proposed by Grandi et al.
In addition, the effective fuel temperature volume-averaged over theentire fuel region negatively overestimates the reference Doppler
reactivity coefficient but the calculated one can be significantly
improved by dividing the fuel region into a few equi-volumes.

PAPER 1125254

Assessment of the Depletion Capability in MPACT
A.Zhu, B.Collins, B.Kochunas, T.Downar
University of Michigan, Ann Arbor, USA
The objective of the paper is to develop and demonstrate the depletion
capability with pin resolved transport using the MPACT code. The
first section of the paper provides a description of the depletion
methodology and the algorithm used to solve the depletion equations
in MPACT. A separate depletion library for MPACT is used basedon the ORIGEN-S library to provide the basic decay constants andfission yields, as well as the 3-group cross-sections which are used
for the isotopes not contained in the MPACT multi-group library. The
cross sections for the depletion transmutation matrix were collapsed
using the transport flux solution in MPACT based on either the 47
group HELIOS library based on ENDF-VI or a 56 group ORNL library
based on ENDF-VII. The second section of this paper then describes
the numerical verification of the depletion algorithm using two sets ofbenchmarks. The first is the JAERI LWR lattice benchmark which hadparticipants from most of the lattice depletion codes currently usedin the international nuclear community and the second benchmark isbased on data from spent fuel of the Takahama-3 reactor. The results
show that MPACT is generally in good agreement with the results of
the other benchmark participants as well as the experimental data.
Finally, a full core 2D model of CASL AMA benchmark was depleted
based on the central plane of the Watts Bar reactor core which
demonstrates the whole core depletion capability of MPACT.

PAPER 1125817

Monte Carlo Modelling of VR-1 Reactor Core
T.Bily, O.Huml, J.Frybort, J.Rataj
Czech Technical University in Prague, Czech Republic
The possibilities of reactor core analysis by precise Monte Carlocodes are gradually increasing along with the accessibility ofcomputing power. In the case of zero power research reactors,
where temperature and burn-up effects remain negligible, model canapproximate the reality to a very high degree. In such a case, mostof calculation uncertainty can be caused by uncertainties in technicalspecifications of fuel and reactor internals. Thus performance of themodelling and its predictive power can be significantly improved
via comparison with a large set of experimental data that can beacquired during reactor operation and via subtle tuning and improving
the calculation model. The paper describes the case for neutronics
calculations of VR-1 zero power reactor core.

PAPER 1126966

Assessment of the 2D/1D Implementation in MPACT
B.S.Collins, S.G.Stimpson(1), A.T.Godfrey(2), B.M.Kochunas,
T.J.Downar(1)1)University of Michigan, Ann Arbor, USA, 2)Oak Ridge National
Laboratory, Tennessee, USA
The 2D/1D method is used in the MPACT code to obtain 3D solutionsof the Boltzmann transport equation for practical reactor geometries.
The OECD C5G7 transport benchmark problem is used first to assess
the accuracy of the method with a fixed set of cross-sections. The
VERA Core Physics Progression Problems are then used to compare
the accuracy of the transport solver using a 56-group library based on
ENDFB-VII.0. Single assembly PWR designs are simulated, and theeigenvalue and pin powers are compared to continuous-energy Monte
Carlo results. A 3×3 assembly cluster with a control rod inserted into
the center assembly is then compared to Monte Carlo to assess theability of MPACT to predict a control rod worth curve. Finally, MPACT
is used to simulate the initial critical states of a full 3D initial core of a
PWR at zero power conditions.

PAPER 1127524

Coupled Neutronics and Thermal-Hydraulic Solutionof a Full-Core PWR Using VERA-CS
S.Palmtag (1) , K.Clarno, G.Davidson, T.Evans, J.Turner (2) ,K.Belcourt, R.Hooper, R.Schmidt(3)1)Core Physics Inc., North Carolina, USA, 2)Oak Ridge National
Laboratory, Tennessee, USA, 3)Sandia National Laboratories, New
Mexico, USA
The Consortium for Advanced Simulation of Light Water Reactors(CASL) is developing a core simulator called VERA-CS to model
operating pressurized water reactors (PWRs) with high resolution. This
paper describes how the development of VERA-CS is being driven
by a set of progression benchmark problems that specify the deliveryof useful capability in discrete steps. As part of this development, thispaper will describe the current capability of VERA-CS to perform a
multiphysics simulation of an operating PWR at Hot Full Power (HFP)
conditions using a set of existing computer codes coupled together ina novel method. Results for several single-assembly cases are shownthat demonstrate coupling for different boron concentrations andpower levels. Finally, high-resolution results are shown for a full-core
PWR reactor modeled in quarter-symmetry.

PAPER 1127710

Assessment of the WIMS9A/PARCS/TRACE Code
System for Power Density Calculations of theWestinghouse AP1000TM Reactor
M.A.Elsawi, A.S.Bin Hraiz
Khalifa University of Science, Technology and Research(KUSTAR), Abu
Dhabi, United Arab Emirates
The objective of this paper is to assess the accuracy of the WIMS9A/
PARCS/TRACE code system for power density calculations of theWestinghouse AP1000TM nuclear reactor, as a representative ofmodern pressurized water reactors (Gen III+). The cross section
libraries were generated using the lattice physics code WIMS9A (thecommercial version of the legacy lattice code WIMSD). Nine different
fuel assembly types were analyzed in WIMS9A to generate the twogroup cross sections required by the PARCS core simulator. The nine79

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
assemblies were identified based on the distribution of the discreteburnable absorbers (Borosilicate glass) and the integral fuel burnable
absorbers (IFBA) in each fuel assembly. The generated cross sections
were passed to the coupled core simulator PARCS/TRACE which
performed 3-D, full-core diffusion calculations from within the US NRC
Symbolic Nuclear Analysis Package (SNAP) interface. The results
which included: assembly power distribution, effective multiplication
factor (keff), radial and axial power density, and whole core depletion
were compared to reference Monte Carlo results and to a publishedreactor data available in the AP1000 Design Control Document (DCD).
The results of the study show acceptable accuracy of the WIMS9A/
PARCS/TRACE code in predicting the power density of the AP1000
core and, hence, establish its adequacy in the evaluation of theneutronics parameters of modern PWR designs. The work reported
here is new in that it uses, for the first time, the combination of
WIMS9A/PARCS/TRACE codes to perform neutronics calculation for
the Westinghouse’ AP1000TM reactor, as a representative of modern
PWRs, with its challenging core configuration.

PAPER 1127773

Feasibility of Nodal Equivalence Theory Using
Functionalized Discontinuity Factors
W.Kim, Y.Kim
KAIST, Daejeon, Korea
The conventional Simplified Equivalence Theory (SET) is based on
the single assembly lattice calculation with net-zero current boundary
condition. However, SET has a problem that it can’t reflect the actual
node interface current on the Discontinuity Factor (DF) so that it shows
discrepancy with the reference value. For more accurate reactor
core analysis, the conventional SET was modified by introducing
Functionalized Discontinuity Factors (FDFs). DFs are functionalizedusing lattice calculation result from several boundary conditions. The
FDFs are updated during iterative whole-core calculation withoutcompromising the computing time. The test calculation for onedimensional slab geometry core showed that FDFs provided muchaccurate keff value that error reduced by up to 63%.

PAPER 1127819

The Multigroup Neutronics Model of NuStar’s 3D Core
Code EGRETS.Zhang, D.Lu, T.Wang
Shanghai NuStar Nuclear Power Technology Co., Ltd., Shanghai, China
As a key component of NuStar’s core analysis system for PWR
application, EGRET is designed to perform steady-state coupledneutronic/hydraulic analysis of PWRs. This paper presents EGRET’s
unique 3D nodal diffusion model and 2D pin power reconstruction(PPR) model. Unlike the practice in most of today’s production codesthat iteratively solves the global 3D coarse-mesh problem and the
local axially 1D fine-mesh problem to handle the axial heterogeneity
within a node caused by fuel grid and partially-inserted control rod,
EGRET resolves the issue by inventing a new nodal technology andintroducing the adaptive meshing technique to follow the movementof control rod tip. The new nodal method employs fine-meshheterogeneous calculation with coarse-mesh transverse coupling suchthat the axial heterogeneous nodes can be explicitly modeled in exact
geometry and directly incorporated into the scheme of transversely
coupled coarse-mesh nodal methods. Each axial channel can haveits own fine-mesh division without the need of dividing the whole core
into radially coupled fine-meshes. There is no need to do 1D finemesh and 3D coarse-mesh iteration either. While for the PPR model,
EGRET adopts a group-decoupled direct fitting method, which avoids
both the complication of constructing 2D analytic multigroup flux
solution and any group-coupled iteration. Another unique feature of the
PPR model is that it fully utilizes all the information available from 3D
core calculation into the downstream PPR process. Particularly, for the
first time, the 1D profiles of transversely-integrated fluxes are utilized
as the additional conditions to reconstruct pin power. Numerical resultsof series of benchmark problems verify the good performance of
EGRET’s unique multi-group neutronics model.

PAPER 1128004

Simulation of Watts Bar Initial Startup Tests with
Continuous Energy Monte Carlo Methods
A.T.Godfrey, J.C.Gehin, K.B.Bekar, C.Celik
Oak Ridge National Laboratory, Tennessee, USA
The Consortium for Advanced Simulation of Light Water Reactors

80

is developing a collection of methods and software products known
as VERA, the Virtual Environment for Reactor Applications. One
component of the testing and validation plan for VERA is comparisonof neutronics results to a set of continuous energy Monte Carlosolutions for a range of pressurized water reactor geometries using
the SCALE component KENO-VI developed by Oak Ridge National
Laboratory. Recent improvements in data, methods, and parallelism
have enabled KENO, previously utilized predominately as a criticality
safety code, to demonstrate excellent capability and performance forreactor physics applications. The highly detailed and rigorous KENO
solutions provide a reliable numeric reference for VERA neutronics
and also demonstrate the most accurate predictions achievable bymodeling and simulations tools for comparison to operating plantdata. This paper demonstrates the performance of KENO-VI for theWatts Bar Unit 1 Cycle 1 zero power physics tests, including reactor
criticality, control rod worths, and isothermal temperature coefficients.Track2: Deterministic Transport Theory
PAPER 1039082

The Application and Performance of ACMFD
Acceleration in 2D/3D Full Core MOC Transport Fuse
MethodZ.Li, Y.Zhang
China Nuclear Power Technology Research Institute Shanghai Branch,
Shanghai, China
It has been shown that the Analytic Coarse Mesh Finite Difference(ACMFD) method is very robust in nodal diffusion acceleration
because of its rigorous derivation, and also should be applicablein core transport methods. In the past decade, the Method of
Characteristics (MOC) was widely studied either in 2D or 2D/1D fusecases with Coarse Mesh Finite Difference (CMFD). In this paper, theapplication of ACMFD in 2D or 2D/1D fuse MOC transport theory as
an acceleration method is presented. Numerical result indicates thatthe performance of ACMFD is similar to CMFD.

PAPER 1083797

A Coupling Method of Subgroup and Wavelet
Expansion for the Resonance Parameter Calculation
H.Wu, L.He, Y.Li, T.Zu
Xi'an Jiaotong University, Shaanxi, China
Owing to their geometric flexibility, subgroup method and wavelet
expansion method have become attractive approaches to obtaineffective self-shielding microscopic cross sections within resonanceenergy groups for geometrically complex problems. However, thesubgroup method is good in the dense resonance range, while thewavelet expansion method is good in the sparse resonance range. In
order to get the resonance parameter in the whole resonance energy
range more accurately and effectively, this paper developed a newcoupling resonance calculation model based on subgroup method
and wavelet expansion method. In this coupling model, the subgroupmethod is employed to handle the higher resonance energy groups,
and the wavelet expansion method is employed to handle the lower
resonance energy groups. At the coupling interface, they are coupledby transferring scattering source. In order to verify the coupling model,
a series of benchmark problems are calculated in this paper. It isdemonstrated that compared with subgroup method and wavelet
expansion method respectively, this coupling resonance model has
the ability to provide more exactly self-shielding microscopic crosssections in the whole resonance energy range while keeping enoughefficiency.

PAPER 1104084

Verification of Ray Effect Elimination Module in theTransport Code ARES
M.Chen, B.Zhang, Y.Chen
North China Electric Power University, Beijing, China
ARES is a multi-group of anisotropic scattering transport shieldingcode based on discrete ordinates method. The 3D radiation transportbenchmark problems proposed by Kobayashi were calculated by
ARES with sub-module ARES_RayEffect which using first collision
method for ray effect mitigation. ARES_RayEffect calculates uncollided

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
flux and first collision source moments for ARES. The uncollided flux
is obtained by a ray tracing calculation between a source point anda target mesh center. In addition, ARES_RayEffect has a modifying
factor function to improve the quality of uncollided flux calculation.
For verification, the results of MCNP code are used as reference
solution and the results of TORT with FNSUNCL3 are compared.
ARES_RayEffect introduced the modifying factor to reduce the relative
errors of meshes near the source region. For example, at the position(15,15,15) of Problem 1 case i, the relative difference of the result of
ARES with ARES_RayEffect is -2.37%, while relative difference ofthe result of TORT with FNSUNCL3 is -11.95%. The calculated total
neutron fluxes show good agreement with the MCNP solutions. For
the pure absorber cases, the maximum differences are less than 3%.
For the half scattering cases, the maximum differences are less than
11%. Numerical results demonstrate that ray effect can be effectively
mitigated.

Track3: Monte Carlo Methods
PAPER 1084078

3-D Monte Carlo Neutron-Photon Transport CodeJMCT and Its Algorithms
L.Deng, T.Ye, G.Li, B.Zhang, D.Shangguan
Institute of Applied Physics and Computational Mathematics, Beijing,
China
JMCT Monte Carlo neutron and photon transport code has beendeveloped which is based on the JCOGIN toolbox. JCOGIN includes
the geometry operation, tally, the domain decomposition and theparallel computation about particle (MPI) and spatial domain (OpenMP)
etc. The viewdata of CAD is equipped in JMCT preprocessor. The
full-core pin-mode, which is from Chinese Qinshan-II nuclear power
station, is design and simulated by JMCT. The detail pin-power
distribution and keff results are shown in this paper.

PAPER 1099429

Monte Carlo Calculation of Neutron Generation Timein Critical Reactor and Subcritical Reactor with anExternal Source
L.Zhang, X.Jiang, X.Zhang, P.Hu, T.Ma, L.Chen,W.Chen
Northwest Institute of Nuclear Technology, Xi’an, China
The neutron generation time Λ plays an important role in the reactor
kinetics. However, it is not straightforward nor standard in mostcontinuous energy Monte Carlo codes which are able to calculate theprompt neutron lifetime lρ directly. The difference between Λ and lρ aresometimes very apparent. As very few delayed neutrons are producedin the reactor, they have little influence on Λ. Thus on the assumptionthat no delayed neutrons are produced in the system, the prompt
kinetics equations for critical system and subcritical system with an
external source are proposed. And then the equations are applied to
calculating Λ with pulsed neutron technique using Monte Carlo. Only
one fission neutron source is simulated with Monte Carlo in criticalsystem while two neutron sources, including a fission source and an
external source, are simulated for subcritical system. Calculations are
performed on both critical benchmarks and subcritical system withan external source and the results are consistent with the reference
values.

PAPER 1099694

Neutron Channels Shield Design Analyses of KIPT
Neutron Source Facility
Z.Zhong, Y.Gohar
Argonne National Laboratory, Argonne, USA
Argonne National Laboratory of the United States and the Kharkov
Institute of Physics and Technology in Ukraine have been collaborating
on design and construction of an experimental neutron source
facility. It is an electron accelerator driven system (ADS) that utilizes
a subcritical assembly driven by an electron accelerator. This facility
will be used to conduct basic and applied nuclear research, produce
medical isotopes, and train young nuclear specialists. The shieldis designed to reduce the biological dose to < 0.5 mrem/hr outside
its boundary during normal operation. Monte Carlo computer code
MCNPX was utilized as the main design tool. In the shield design

analyses, the radial heavy concrete shield was deemed to have acylindrically symmetric geometry, to reduce the computation time of
MCNPX. The required thickness of heavy concrete was determined
to be 140 cm for this cylindrical shield geometry. Recently theradial shield design was finalized, with built-in neutron channels,
beam shutters and plug-in shield, and the outer shield boundary
is not cylindrical. MCNPX calculation were carried out to verify theperformance of this finalized radial shield, with the neutron and photon
doses calculated at selected locations outside the shield boundary.
The previously developed neutron source file procedure was utilized
for the neutron dose calculation, while the photon dose calculations
started with the electron source particles. Due to the asymmetricgeometry of the finalized radial shield, three dimensional weightwindows were generated and utilized for both neutron and photon
dose calculations. The calculated results show that the total radiationdose at the selected locations outside the radial shield is < 0.5 mrem/
hr.

PAPER 1101864

Development of a New Convergence Criterion forMonte Carlo Simulation with Thermal-Hydraulics
Feedback
X.Wu, T.Kozlowski
University of Illinois at Urbana-Champaign, Urbana, USA
Coupled multi-physics approach plays an important role in improving
computational accuracy. Compared with deterministic neutronics
codes, Monte Carlo codes have the advantage of higher resolution
level. In the present paper, a three-dimensional continuous-energyMonte Carlo reactor physics burnup calculation code, Serpent, iscoupled with thermal-hydraulics safety analysis code, RELAP5. A newconvergence criterion for the coupled simulation is developed basedon the statistical uncertainty in power distribution in Monte Carlo code,
rather than an arbitrarily chosen criterion in previous research.
The coupled simulation is based on the OECD-NEA/NRC PWR
MOX-UO2 Core Transient Benchmark. The convergence criterion of
normalized axial power distribution is tested on both UO2 and MOX
single assembly models. Compared with previously implemented
convergence criteria based on temperature, eigenvalue or flux (orpower), it takes into account both the local and global convergence. It
does not use a pre-set tolerance limit and is decided by the statisticalaccuracy of the Monte Carlo code itself. This new convergencecriterion is shown to be stable, more stringent and direct, equallyconvenient to use but may need a few more steps to converge.

PAPER 1102407

JCOGIN: A Parallel Programming Infrastructure forMonte Carlo Particle Transport
B.Zhang, G.Li, L.Deng, Y.Ma, D.Shangguan, A.Zhang, X.Cao,
Z.Mo
Institute of Applied Physics and Computational Mathematics, Beijing,
China
The advantages of the Monte Carlo method for reactor analysis
are well known, but the full-core reactor analysis challenges thecomputational time and computer memory. Meanwhile, the exponential
growth of computer power in the last 10 years is now creating agreat opportunity for large scale parallel computing on the Monte
Carlo full-core reactor analysis. In this paper, a parallel programming
infrastructure is introduced for Monte Carlo particle transport, named
JCOGIN, which aims at accelerating the development of Monte
Carlo codes for the large scale parallelism simulations of the full-corereactor. Now, JCOGIN implements the hybrid parallelism of the spatial
decomposition and the traditional particle parallelism on MPI and
OpenMP. Finally, JMCT code is developed on JCOGIN, which reaches
the parallel efficiency of 70% on 20480 cores for fixed source problem.
By the hybrid parallelism, the full-core pin-by-pin simulation of theDayawan reactor was implemented, with the number of the cells up to
10 million and the tallies of the fluxes utilizing over 40GB of memory.

PAPER 1104378

Void Transit Time Calculations by Neutron Noise of
Propagating Perturbation Using Complex-Valued
Weight Monte Carlo
T.Yamamoto
Kyoto University, Kumatori, Japan
This paper focuses on the propagation of neutron noise induced

81

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
by void formation in coolant/moderator moving upward in a reactor
core within the limit of the first order perturbation. A new Monte Carlo
method to solve the frequency domain transport equation of neutron
noise observed by in-core neutron detectors has been proposed byadopting the complex-valued weight Monte Carlo technique. The
technique has already been established by the author of the present
paper to implement the B1 approximation method into the Monte Carlo
method. The newly-developed Monte Carlo method is comparedwith the conventional method based on the diffusion theory forneutron noise analyses. The Monte Carlo method makes a significant
difference in neutron noise distribution as compared to the diffusion
theory, suggesting that the transport theory should be introduced for
accurate estimation of neutron noise in a reactor core. A numerical test
is conducted to simulate the measurement of void transit time or void
velocity in a reactor core by calculating a cross power spectral density
between two in-core detectors.

PAPER 1105897

Simulating Fast Transients with Fuel Behavior
Feedback Using the Serpent 2 Monte Carlo Code
V.Valtavirta, T.Ikonen, T.Viitanen, J.Leppänen
VTT Technical Research Centre of Finland, VTT, Finland
Simulating transients with reactivity feedback effects using Monte
Carlo neutron transport codes can be used for validating deterministictransient codes or estimating for example the total deposited energy
in a fuel rod following a known reactivity insertion in the system.
Recent increases in computational power as well as developments in
calculation methodology makes it possible to obtain a coupled solution
for several aspects of the multi-physics problem in a single calculation.
This paper describes the different methods implemented in Serpent
2 Monte Carlo code that enable it to model fast transients with fuelbehavior feedback. The capability is demonstrated in a prompt
critical pin-cell case, where the transient is shut down by the negative
reactivity from rising fuel temperature.

PAPER 1105995

Status of Monte Carlo Code Development at UNISTH.Lee, C.Kong, D.Lee
UNIST, Ulsan, Korea
This paper presents the status of Monte Carlo (MC) code development
at Ulsan National Institute of Science and Technology (UNIST).
UNIST Monte Carlo code named “MCS” has 3D whole core modelingcapability. ENDF/B-VII.0 and ENDF/B-VII.1 nuclear cross section
libraries are used as continuous energy cross section data. MCS
can estimate not only the fast spectrum reactor but also the thermal
spectrum reactor which requires thermal scattering capability. Twotypes of calculations are possible: criticality calculation and real time
fixed source calculation. The criticality calculation ability is verified
with selected cases of ICBEP benchmark problem, C5G7 benchmark,
VENUS-2 benchmark, and Hoogenboom benchmark problems. The
real time fixed source calculation ability is tested with noise analysis.
Furthermore, Hybrid form of neutron transport solver and modified
power iteration method are implemented in the MCS.

PAPER 1106262

Criticality Benchmarking of ANET Monte Carlo Code
T.Xenofontos, P.Savva, M.Varvayanni, N.Chrysanthopoulou,
N.Catsaros(1), J.Maillard(2), J.Silva(3), G.-K.Delipei, A.Clouvas(4)1)National Centre for Scientific Research Demokritos, Greece,2)CNRS, France, 3)Université Pierre et Marie Curie, France,4)Aristotle University of Thessaloniki, Greece
In this work the new Monte Carlo code ANET is tested on criticality
calculations. ANET is developed based on the high energy physics
code GEANT of CERN and aims at progressively satisfying several
requirements regarding both simulations of GEN II/III reactors, aswell as of innovative nuclear reactor designs such as the Accelerator
Driven Systems (ADSs). Here ANET is applied on three different
nuclear configurations, including a subcritical assembly, a Material
Testing Reactor and the conceptual configuration of an ADS. In thefirst case, calculations of the effective multiplication factor (Keff) are
performed for the Training Nuclear Reactor of the Aristotle Universityof Thessaloniki, while in the second case Keff is computed for the
fresh fueled core of the Portuguese research reactor (RPI) just after
its conversion to Low Enriched Uranium, considering the control rods
at the position that renders the reactor critical. In both cases ANET
computations are compared with corresponding results obtained by82

three different well established codes, including both deterministic(XSDRNPM/CITATION) and Monte Carlo (TRIPOLI, MCNP). In the
RPI case, Keff computations are also compared with observations
during the reactor core commissioning since the control rods are
considered at criticality position. The above verification studies show
ANET to produce reasonable results since they are satisfactorily
compared with other models as well as with observations. For thethird case (ADS), preliminary ANET computations of Keff for various
intensities of the proton beam are presented, showing also a
reasonable code performance concerning both the order of magnitude
and the relative variation of the computed parameter.

PAPER 1123217

Effective Diffusion Homogenization of Cross Sections
with the Monte Carlo Method
D.Ćalić(1), M.Kromar(2), A.Trkov(3)1)ZEL-EN d.o.o. / Institute Jozef Stefan, Ljubljana, Slovenia, 2)Institute
Jozef Stefan, Ljubljana, Slovenia, 3)Institute Jozef Stefan / International
Atomic Energy Agency, Vienna, Austria
Homogenizations of heterogeneous regions using Monte Carlo
methods is one of the possible intermediate steps to transfer fromthe deterministic codes to a full scale Monte Carlo calculations of alarge complex reactor systems. With the use of capable computer
the generation of neutron homogenized multigroup cross sections
using Monte Carlo method could be performed, which are later
used in deterministic codes to provide neutron solution on a coarse
mesh. In the Monte Carlo code Serpent two methods are used toobtain the homogenized cross sections. First method is the simplest
homogenization method based on a flux and volume weighting (FVH)
of cross sections. The second method is based on the B1 method.
While the first method suffers in the presence of the strong absorbers
the second method is applicable only for the cases with fissile
materials. The implementation of an effective diffusion homogenization(EDH) method is proposed in this paper. With the use of the Serpent
code and the EDH method coarse group cross sections for a sample
problem are calculated. Obtained cross sections are tested against
the determinist results obtained with the WIMSD code and against theB1 method within the Serpent code. Accuracy of the calculated crosssections is tested by calculating multiplication factor using diffusion
approximation in GNOMER code and comparing the results to thereference values obtained with the MCNP code.

PAPER 1228305

Estimating Local In- and Ex-Core Responses within
Monte Carlo Source Iteration Eigenvalue Calculations
K.W.Burn
ENEA, Bologna, Italy
Local in- and ex-core responses are calculated by employing variancereduction within the Monte Carlo source-iteration scheme. This is doneby employing the Direct Statistical Approach to search for an optimum
trade-off between sampling the local response and sampling thefundamental mode. Superhistories are employed to improve the tradeoff point. Realistic test problems are run that show a good agreement
between the predicted and actual calculational figure-of-merits. For the
sample problems treated, gains in efficiency over analog (i.e. without
variance reduction) range from 1 - 2 orders of magnitude for in-core
responses to many orders of magnitude for ex-core responses. An
alternative way of finding the trade-off point using the classic adjoint
flux formalism showed substantial differences for one of the problems.

Track4: Verification, Validation and Uncertainty Analysis
PAPER 1067489

Review of Neutronic Assessments Applied to Small
Reactor Core Physics
L.Chabert
AREVA TA, Aix en Provance, France
In its design division for material test reactors and research reactors,
AREVA TA has to characterize these manufactured cores. This stepis sequential with neutronics benchmarks associated with validation(standard Verification & Validation approach). The previous twopoints are embedded in core projects and can be run separately

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
especially when experimental tests are foreseen for validation
database enrichment. Methodological standard is given in order tomatch validation and benchmark process illustrated alongside with
two specific items on critical research reactors (AZUR – JHR) andsubcritical mock up (AZUR).

uncertainties. In some fast neutron systems, uncertainties of Λ arelarger than uncertainties of βeff . We conclude that uncertainties of Λ
should be properly considered when one discusses uncertainties of
βeff/Λ, which is one of measurable parameters during reactor physics
experiments.

PAPER 1084893

PAPER 1103703

A New Neutronics Analysis Code System for Fast
Reactors and Validation
T.Takeda, Y.Shimazu
Research Institute of Nuclear Engineering, University of Fukui, Fukui,Japan
A new neutronics analysis code system has been developed for
detailed analysis of fast reactor cores. The code system is composedof a calculation code of effective cross sections, an assembly
calculation code based on the method of characteristics, and a fullcore transport/diffusion calculation code. The validity of the code
system is investigated by applying it to the prototype fast reactor
Monju, and by comparing the calculation results with measured ones.

PAPER 1100864

Uncertainty Analysis of Delayed Neutron Fissile
Material Assay Using a Genetic Algorithm
R.P.Kelley, L.M.Rolison, D.Raetz, K.A.Jordan
University of Florida, FL, USA
An uncertainty analysis of the delayed neutron non-destructive
assay method was conducted to explicitly define the accuracy withwhich plutonium content in an uncharacterized sample can beassessed. Perturbing various parameters allowed for an investigationof the sensitivity of this method to various nuclear data, and it wasdetermined that the relative delayed neutron group abundances had
the largest effect on the genetic algorithm. Specifically, for a sample
containing 235U, 238U, and 239Pu, irradiations in the thermal spectrum
were shown to be more sensitive to 235U and 239Pu data, while
irradiations in a fast spectrum were shown to be more sensitive to the238
U data. The overall uncertainties of the mass estimates were 15%,5%, and 30% for 235U, 238U, and 239Pu, respectively. Finally, reducing
the first delayed neutron group abundances by a factor of three assuggested by recent research reduced the overall uncertainties to10%, 3%, and 20%.

PAPER 1102809

IR Approximation for Calculating Sensitivity andUncertainty of PWR Cells by Taking Account of SelfShielding Effect
B.Foad(1,2), T.Takeda(1)1)University of Fukui, Fukui, Japan, 2)Egypt Nuclear and Radiological
Regulatory Authority, Nassr, Egypt
A new improved method has been developed for calculating sensitivity
coefficients of neutronics parameters in pressurized water reactor
cells relative to infinite dilution cross-sections by taking account ofresonance self-shielding effect. In our paper, the IR approximation
is used in order to get accurate results in both high and low energy
groups. This method is applied to UO2 and MOX fueled PWR cells
to calculate sensitivity coefficients and uncertainties of eigenvalue
responses. We have verified the improved method by comparing
the sensitivities with MCNP code and good agreement is found.
For uncertainty, the improved results are compared with TSUNAMI1D, and demonstrate that the differences are caused by the use ofdifferent covariance matrix.

PAPER 1102810

Uncertainty Quantification of Reactor Kinetics
Parameters Using JENDL-4.0 Covariance Data
G.Chiba, M.Tsuji, T.Narabayashi
Hokkaido University, Sapporo, Japan
Nuclear data-induced uncertainties of the neutron generation time Λ,
which is one of the important reactor kinetics parameters, and related
quantities βeff and βeff /Λ are quantified with deterministic calculations. A
simple calculation method, which is based on the ordinary perturbation
theory, for Λ sensitivities is proposed. The covariance data given in
the JENDL-4.0 library is used. Numerical calculations reveal thatthe uncertainties of Λ are not negligible in comparison with the βeff

Survey on Effect of Crystal Texture of Beryllium on
Total Cross-Section to Improve Neutronic Evaluationin JMTR
N.Takemoto, T.Imaizumi, N.Kimura, K.Tsuchiya(1), J.Hori, T.Sano,
K.Nakajima(2)1)Japan Atomic Energy Agency, Ibaraki, Japan, 2)Research Reactor
Institute, Kyoto University, Kumatori, Japan
Neutronic evaluations in JMTR have been performed for irradiation
tests by Monte Carlo method with thermal neutron scattering law,
S(α, β), data for beryllium metal, etc, and the calculation accuracy offast and thermal neutron fluxes are ± 10% and ± 30%, respectively.
Analytical and experimental investigations to achieve higher
calculation accuracy, especially for the thermal neutron flux up to thefast neutron flux level, have been performed to offer higher value
data technically to the JMTR users. In order to investigate an effect of
fabrication method of beryllium material on the calculation accuracy,
total cross-sections of beryllium specimens were measured using
KURRI-LINAC, and it was found that the total cross-section was
different from the evaluated one, and depended on the crystal texture,etc. The S(α, β) data for beryllium metal was adjusted based on themeasured data, and the applicability to the neutronic evaluation in the
JMTR was verified.

PAPER 1103924

Uncertainty and Sensitivity Analysis for an OECD/
NEA HTGR Benchmark with XSUSA
A.Aures, K.Velkov, W.Zwermann (1), P.Rouxelin, K.Ivanov (2),J.Lapins, W.Bernnat(3)1)Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH,
Garching, Germany, 2)The Pennsylvania State University, PA, USA,
3)Institut für Kernenergetik und Energiesysteme (IKE), University of
Stuttgart, Stuttgart, Germany
Sensitivity and uncertainty (S/U) analyses with respect to uncertainties
in microscopic neutron cross sections and fission yields are performedfor HTGR fuel, of both pebble bed and prismatic types, with UO2 as
initial fuel. This is done with the sampling-based XSUSA program,with criticality and depletion calculation sequences from the SCALE
6.1 code system. One specific goal is to provide a tool for systematic
HTGR S/U analysis with respect to core physics in the framework of
the “IAEA Coordinated Research Program on the High-Temperature
Gas cooled Reactor Uncertainty Analysis in Modelling”. As a basis, an
HTGR fuel depletion benchmark initiated by U.S. NRC and OECD/
NEA is chosen. Criticality and depletion calculations are performed
for pebble bed and prismatic fuel arrangements in infinite latticegeometry. As sample size, a sufficiently high number is chosen to beable to determine not only uncertainties, but also the main contributors
to these uncertainties. Uncertainties and importance indicators are
determined for relevant quantities, such as multiplication factors,
isotopic inventories, and few-group cross sections. For most of theoutput quantities, a very similar behavior is observed for pebble bed
and prismatic fuel concerning uncertainties and sensitivities. Also,in many respects similarity to results for LWR fuel is obtained. The
few-group cross sections obtained from the calculations with varied
nuclear data can be used later on in steady-state and transient HTGR
core simulations, as planned within the IAEA Coordinated Research
Program.

PAPER 1104067

First Verification and Validation Steps of MENDEL
Release 1.0 Cycle Code System
S.Lahaye, P.Bellier, H.Mao, A.Tsilanizara(1), Y.Kawamoto(2)1)CEA/DEN/DANS/DM2S/SERMA/LLPR, Gif sur Yvette Cedex, France,2)Hokkaido University, Sapporo, JapanFor each new code system, verification and validation process is aneed to prove the efficiency and accuracy of the calculated physicalquantities.
MENDEL is the new CEA depletion code system, whose first release
was done at the end of 2013. It offers iso-capacity with the already83

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
well-established DARWIN. MENDEL is the successor of DARWIN,
and can be used as a stand-alone code system for reactor cycle
studies to compute interest output quantities. MENDEL also provides
its depletion solvers to both Monte Carlo TRIPOLI-4® and deterministic
APOLLO3® transport code systems.
The purpose of this paper is to present the first contributions to
MENDEL release 1.0 verification and validation process. This firstrelease has been used with nuclear data coming from both JEFF3.1.1 and ENDF/B-VII.1 nuclear data evaluations, and its results arecompared either with experimental data, either with DARWIN results.

PAPER 1104316

Application of the GRS Method for Estimation ofUncertainties of LMFBR Type Reactor Physics
Parameters with Taking into Account Macroscopic
Experiments
A.Peregudov, M.Semenov, G.Manturov, V.Koscheev, K.Raskach,
A.Tsibulya
Institute for Physics and Power Engineering, Obninsk, Russia
A number of recent studies have been devoted to estimation of errorsof fast reactor nuclear-physics parameters by the GRS (Generation
Random Sampled) method. The method is based on direct sampling
input data resulting in formation of random sets of input parameterswhich are used for multiple neutronics calculations. Once thesecalculations are performed, a statistical processing of calculation
results is carried out to determine the mean value and the variance ofeach calculation parameter of interest.
In this study the GRS full scale method was used for estimation oferrors of calculation of nuclear-physics parameters (k-eff, power
density, dose rate)for a perspective LMFBR type sodium-cooled
fast reactor with taking into account different sets of integral andmacroscopic experiments. The multiple neutronics calculations are
performed by the nodal diffusion three-dimensional code TRIGEX andCONSYST/ABBN nuclear data system. The sampling of input data,nuclide nuclear cross-sections, is based on the ABBN-93 covariancematrices.

PAPER 1104670

Benchmarking of Photon and Coupled Neutron andPhoton Process of SuperMC 2.0
J.Song, G.Sun, H.Zheng, L.Hao, P.Long, L.Hu, Y.Wu
Institute of Nuclear Energy Safety Technology, Chinese Academy of
Sciences, Anhui, China
Super Monte Carlo Calculation Program for Nuclear and Radiation
Process (SuperMC), developed by FDS Team in China, is a multifunctional simulation program mainly based on Monte Carlo (MC)method and advanced computer technology. This paper focuseson the benchmarking of physical process of photon and coupledneutron-photon of SuperMC2.0. Integral leakage rate of photon in
the spherical and hemispherical shell experiment was tested to verify
the physical process of photon and coupled neutron and photontransport. Vanadium assembly experiment and ADS benchmark
were given as comprehensive benchmarks. The correctness waspreliminarily verified by comparing calculation results of SuperMC withexperimental results and MCNP calculation results.

PAPER 1104787

In Depth Uncertainty Estimation of the Neutron
Computational Tools
G.Rimpault, Y.Peneliau, J.F.Vidal(1), S.Mirotta, A.Gandini(2)1)CEA Cadarache, Saint Paul-lez-Durance, France, 2)University La
Sapienza di Roma, Roma, Italy
Verification & validation processes are required to demonstrate theability of simulation tools to accomplish the task they have been
designed for. The values of the characteristics of both 3rd+ generation
PWR cores and 4th generation fast neutron reactor cores should be
associated with a validation dossier.
The new APOLLO3 ® code, under development at CEA, has toundergo such validation process in particular when treating reactor
cores with advanced features. In order to prepare this task, the waythat validation has been done in the past for a code like ECCO isbeing revisited. The aim of this work is to illustrate the change inparadigm between what was done 25 years ago and what is currently
done.
Two ways to quantify the bias introduced in deterministic calculations

84

exist:
1. Performing deterministic calculations in which someapproximations are eliminated or reduced.
2. Comparing deterministic calculations with stochastic ones
(TRIPOLI4).
The first approach was the one being used for ECCO when
programming it since it was accessible with computers available at thetime but is not sufficiently independent from the tool being tested. The
method could identify the origin of discrepancies with the use of theperturbation method so as to upgrade the method and/or track somepossible compensating errors.
The second approach is often done by comparing integral results suchas reactivity or sodium void worth obtained with deterministic codeswith the one of Monte Carlo codes using continuous energy libraries.
This approach is somehow lacking in depth understanding of theorigin of discrepancies, hence is not very instructive for finding a way
to improve the computational scheme. The use of the perturbation
theory is applied in this paper between ECCO/ERANOS deterministic
results and TRIPOLI4 Monte Carlo results for reactivity or sodium void
worth of Sodium Fast Reactors (SFR) so as to illustrate this innovative
approach.

PAPER 1105709

Updated Validation of the PSI Criticality Safety
Evaluation Methodology Using MCNPX2.7 and
ENDF/B-VII.1
M.Pecchia, A.Vasiliev, H.Ferroukhi(1), A.Pautz(1,2)
1)Paul Scherrer Institute,Villigen PSI, Switzerland, 2)École PolytechniqueFédérale de Lausanne, Lausanne, Switzerland
A validation of the PSI Criticality Safety Evaluation (CSE) methodology
using the most recent evaluated nuclear data library ENDF/B-VII.1
is presented in this paper. The basis for the methodology is a welldefined benchmark suite of 149 low-enriched thermal compounduranium critical experiments contained in the International Handbookof Evaluated Criticality Safety Benchmark Experiments. The CSE
validation of code/libraries is then focused on the evaluation of thelower tolerance bound along with trend analyses for both design
parameters as well as nuclear spectrum related parameters. Here,the overall performance of the new ENDF/B-VII.1 library for theseselected thermal systems is evaluated with MCNPX-2.7.0 and is foundto be very similar to that obtained with ENDF/B-VII.0, confirming theexisting literature. Moreover, no specific trends as function of design
based parameters and/or spectral parameters are observed with the
new library. However, due to some specific library changes, somedistinct differences between the libraries are shown for certain groupsof experiments the specific library changes. For completeness, thechanges compared to MCNPX-2.5.0 are also assessed in order toaddress code effects and including in that framework, an evaluation of
the impact from the upstream NJOY processing. Finally, a comparisonof ENDF/B-VII.1 results against those obtained with other libraries,
including JEFF-3.1.1, is also presented.

PAPER 1105825

Validation of Two Monte Carlo Codes for LWR Burnup
Calculations
D.Wicaksono(1,2), P.Grimm, A.Vasiliev, M.Hursin, G.Perret, H.
Ferroukhi(2)1)École Polytechnique Fédérale de Lausanne, Lausanne, Switzerland, 2)Paul Scherrer Institute,Villigen PSI, Switzerland
This paper presents an assessment conducted at the Paul Scherrer
Institut (PSI) of two Monte Carlo-based codes, namely Monteburns
and Serpent, for LWR depletion calculations. First, a validation ofpredicted spent fuel compositions is presented on the basis of a
selected Swiss spent fuel sample from the LWR-PROTEUS Phase
II program. Second, an additional assessment is performed for an
OECD/NEA burnup credit benchmark which is also related to isotopic
predictions. In both cases, equivalent pin-cell models are applied andthe results are presented in terms of relative isotopic concentrations
between measured/benchmark and predicted values. Additional
comparisons are also made to previously published PSI results using
the MCNPX/CINDER and CASMO-4E codes. The results show that
all codes provide an overall good agreement with experimental data
although a few specific trends are observed for some nuclides which
are pointed out and briefly discussed.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
PAPER 1106101

PAPER 1126078

K.H.Lee, J.Y.Cho, H.J.Park, J.S.Song

X.Shang, Z.Li, K.Wang, J.Liang

Korea Atomic Energy Research Institute, Daejeon, Korea
The benchmark calculation against various critical experiments has
been performed to validate the 2-dimensional whole core transport
code DeCART2D and its multi-group cross section library. In addition,
several multi-group libraries provided by our own library processing
system are employed to investigate the dependency on both theversion of the nuclear data library and the number of neutron energy
groups. Although these experiments are 3-dimensional, DeCART2D
is able to simulate these experiments directly with a consideration of
the axial leakage, which can be defined as the multiplication of the
diffusion coefficient with the measured geometrical buckling. The
calculated solutions for a critical condition are compared with themeasured data, which include the pin-by-pin fission rate distribution
in the central assembly as well as the eigenvalue. The comparison
results show that a discrepancy in the eigenvalue is highly different
from the multi-group library, whereas there is little difference in thefission rate. In conclusion, it can be seen that the ENDF/B-VII.1 based
multi-group library gives the most reasonable solutions.

Department of Engineering Physics, Tsinghua University, Beijing, China
Several types of V&V work are being carried out for the Reactor
Monte Carlo code RMC, including the heterogeneous wholecore configurations. In this paper, a whole PWR MOX/UO2 corebenchmark which contains both UO2 and MOX assemblies withdifferent enrichments and various burn-up points is chosen to verify
RMC’s criticality calculation capability, and the results of RMC andother codes are discussed and compared, such as eigenvalues,
assembly power distributions, pin power distributions and so on. The
discrepancies in eigenvalues and power distributions are satisfactory,
which proves the accuracy of RMC’s criticality calculation. Also, theinfluences of different cross-section libraries are discussed upon
the results of RMC. Besides these results, the detailed comparisons
between RMC and MCNP with the same ENDF/B-VII.0 cross-section
library are carried out in this paper, including the comparisons ofcontrol rod worths calculated by both RMC and MCNP. According to
the results, RMC and MCNP agree quite well in eigenvalues, power
distributions and other results. The discrepancies of eigenvalues
and control rod worth are fairly small and the relative differences ofassembly and pin power distributions are acceptable. All these results
contribute to the conclusion that the criticality calculation performanceof RMC is accurate and excellent.

Benchmarking of DeCART2D against Critical
Experiments

PAPER 1106250

Automated Reactor Records Evaluation Framework
J.Hejzlar, F.Havluj
ÚJV Řež, a. s., Plzeň, Czech Republic
The only truly reliable method for core physics code validation iscomparison against experimental data – and for power nuclear
reactors, the only reasonably acquirable kind of experimental data are
the reactor records. However, the amount of the data coming fromthe reactor operation is often so vast that it can be discouraging forthe code developers to use it properly. Thus, the validation package isfurther reduced because the data is hard to use.
This paper presents an elaborate, fully automated framework, which
was designed and implemented in our institute, for reactor records
processing and its use for core physics code validation. The workflow,
implemented as a Web 2.0 application, provides a practical andpainless solution for use of reactor records data for code development
and validation.

PAPER 1109432

Monte Carlo Based Equilibrium Cycle Analysisof One-Dimensional Breed and Burn Benchmark
Problem
Z.Xu
TerraPower LLC, Washington, USA
As part of initial neutronics codes verification and validation (V&V)
efforts for Traveling Wave Reactors (TWR) at TerraPower, theequilibrium cycle analysis of the one-dimensional breed and burnbenchmark problem was performed using Monte Carlo codes (MCNP5.1.60 and MCNP-XT), the depletion code, ORIGEN-2.2, and a setof driver Python scripts. Several goals of this work are: 1) to establish
the reference solution for the benchmark problem by series ofsensitivity calculations verifying both the equilibrium cycle modelingmethodology and base modeling input parameters; 2) to verify the
MCNP-XT transport solver against MCNP-5 for depletion problems; 3)
to demonstrate the reactor shutdown length effect. The computational
benchmark problem consists of 100 one-dimensional fuel slabs of 5
cm thickness each. Two fuel shuffling sequences examined are theinward convergent shuffling and the convergent divergent shuffling.
The equilibrium cycle modeling methodology was developed by
iterating flux calculations and depletion calculations. Investigatedsensitivities include the number of depletion slices in a fuel slab, thenumber of depletion steps per cycle, and the depletion step setting, i.e.,
fluxes and reaction rates at beginning of step (BOS), end of step (EOS),
or average of BOS and EOS. Furthermore, the reactivity effect of thereactor shutdown length is found to be about +100 pcm affecting only
beginning of equilibrium cycle (BOEC), which is caused by the holdup
of Np239 inventory in the core.

Verification of the Monte Carlo Code RMC with a
Whole PWR MOX/UO2 Core Benchmark

PAPER 1126219

An Improved Method for Inverse Uncertainty
Quantification for Nuclear Data Assessment
B.A.Khuwaileh, H.S.Abdel-khalik
Department of Nuclear Engineering, North Carolina State University,
Raleigh, USA
Safety analysis and design calculations depend on the precise
prediction of various reactor attributes, e.g., multiplication factor andpin powers. Focusing on neutronics calculations, where nuclear data
are known to constitute the major source of calculational uncertainty,
the attributes prediction can be improved by reducing the uncertainty
associated with the nuclear data. An inverse problem in tandem with acost function that accounts for the cost of nuclear data measurements
can be employed to measure and prioritize the contribution of thevarious nuclear data to the total propagated attributes uncertainties.
The results of this study can help guide the experimental program
focused on obtaining improved measurements of nuclear data.
This approach has already been proposed by others and appeared in
the literature under the name of ‘target accuracy assessment’; it has
been primarily applied in the context of critical benchmark experiments
used in support of fast reactor technology. The overarching objectiveof our work is to generalize the application of this approach to corewide reactor models, expected to be high dimensional. As a first step
towards this objective, this manuscript proposes an efficient inversion
technique for solving the associated optimization problem, expected
to be infeasible with the current direct approach due to the explosionin the dimensionality of nuclear data. In particular, we employ asubspace-based reduced order modeling algorithm for solving the
optimization problem which constrains the search to a small subspacein order to render the search computationally feasible. A quarter PWR
fuel assembly is employed to assess the performance of the proposed
algorithm, and compare it to conventional optimization techniques thatsearch the full space of nuclear data.

85

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
PAPER 1127444

PAPER 1086633

I.A.Kodeli

T.Shiozawa, T.Endo, A.Yamamoto(1), C.H.Pyeon, T.Yagi(2)Jožef Stefan Institute, Ljubljana, Slovenia
In the mathematical expression of the nuclear data sensitivity function
as derived from the first order perturbation theory both the angular
fluxes and their expansion in the Legendre flux moments are required.
Both terms are calculated by the traditional discrete ordinates transport
codes and stored in the files. This leaves the developers of the nuclear
data sensitivity and uncertainty codes several possible choices of themethod to account for the neutron flux anisotropy. Three possibilities
available in the SUSD3D code were studied here and comparedusing the FLATTOP-Pu benchmark exercise. This benchmark wasparticularly suitable due to its high flux anisotropy.

1)Nagoya University, Nagoya, Japan, 2)Kyoto University, Kumatori,Japan
The subcriticality measurement techniques using inherent neutron
source are investigated in this study. In facilities where nuclear fuel istreated, it is considered to install a real-time subcriticality monitoring
system. However, it is often unfavorable or impractical to bring an
external neutron source into a nuclear facility to carry out subcriticality
measurement. Thus, we focus on the inherent neutron source innuclear fuel such as spontaneous fission and (α, n) reaction. Byutilizing the inherent neutron source, the subcriticality measurement
technique, which does not introduce an external neutron source, canbe realized. In this study, the Feynman-α method and the neutron
source multiplication (NSM) method are chosen as techniques which
can be carried out only with the inherent neutron source, and haveaffinities with a real-time monitoring system. In the Kyoto University
Critical Assembly (KUCA), subcriticality measurement experimentsof these two techniques were carried out. From the experimental
result, it is confirmed that both techniques can be carried out onlywith the inherent neutron source. It is noted that, however, measured
values of subcriticality by both methods show different trends. While
the subcriticality measured by the Feynman-α method is fairly
constant among detectors, the measured value does not agree withthat obtained by the pulsed neutron source (PNS) method, in deep
subcritical systems. On the other hand, the subcriticality measured by
the NSM method show significant dependence on detectors. However,
the measurement values by the NSM method using detectors at
appropriate position agree with values by the PNS method even indeep subcritical systems.

On How Sensitive the Cross-Section Sensitivity
Calculations Are to PN Order Approximations

PAPER 1137713

Validation of NuStar’s PWR Core Analysis System
S.Zhang, T.Wang, X.Jiang, D.Lu, G.Chen, Y.Huang(1), Z.Liao,
Z.Pan, D.Kong(2)1)Shanghai NuStar Nuclear Power Technology Co., Ltd., Shanghai,
China, 2)CNNC Nuclear Power Operations Management Co., Ltd.,
Zhejiang, China
A new PWR core analysis system has recently been developed in
NuStar. It adopts not only conventional theoretical framework thatwidely used in today’s production code, but also latest progressthat achieved internationally and within the company. To validate
the system, a total of 44 operation cycles of Qinshan Phase 1 and2 reactors are evaluated and results are compared against themeasurement data. Statistics analyses of prediction errors for both
low-power startup physics test states and high-power normal operation
conditions demonstrate that the development of the system is quite
a success. It not only has a generic applicability for reactors withdifferent design, but also has a very good accuracy for parametersthat are measurable at the site. Among all the validated cases, only
for very rare cases that the acceptance tolerance is violated, andthe error distributions for all the validated parameters present to begood ones that are close to the normal distribution. Based on thevalidation results, the system is considered to be qualified for practicalapplications.

Track6: Reactor Physics Experiments
PAPER 1085573

Determination of the 58Ni (n,p) 58Co Reaction Cross
Section for both Ground and Isomeric States
Z.Lahodová, L.Viererbl, V.Klupák
Research Centre Řež Ltd, Czech Republic
A nickel (Ni) activation detector is used to determine fast neutron
fluence in nuclear reactors via the 58Ni (n, p) 58Co reaction. Proper
evaluation of 58Co-induced activity requires the use of correction
factors, especially the burn-up of product atoms during irradiation.
The irradiation of 58Ni with fast neutrons produces ground and excited
isomeric states of 58Co. This article discusses how to determine the
k-ratio between the cross section of ground and exited states. The
58m
Co isomer has a very high thermal neutron absorption cross section
and a short half-life compared to the ground state. The theoreticalsolution of differential equations for changes in the concentrationof a particular nuclide and experiments performed at the LVR-15
research reactor served to determine the production of 58Co in its
ground and excited states. The experiments were designed so that
the Ni activation detectors were irradiated without shielding and under
the same conditions as with cadmium (Cd) shielding. Cd shielding
captures thermal neutrons and thus prevents the burn-up of 58mCo
and 58Co products. The correct k-ratio between the cross section ofground and exited states will provide the burn-up correction factor for
58m
Co and 58Co products. The burn-up factor allows more accurate
determination of fast neutron fluence.

86

Investigation on Subcriticality Measurement Using
Inherent Neutron Source in Nuclear Fuel

PAPER 1094641

Benchmark Calculations of Sodium Fast Critical
Experiments
E.Mitenkova, N.Novikov
Nuclear Safety Institute of Russian Academy of Sciences, Moscow,
RussiaThe high expectations from fast critical experiments impose theadditional requirements on reliability of final reconstructed values,
obtained in experiments at critical facility. Benchmark calculations ofcritical experiments are characterized by impossibility of complete
experiment reconstruction, the large amounts of input data (dependentand independent) with very different reliability. It should also take
into account different sensitivity of the measured and appropriate
calculated characteristics to the identical changes of geometry
parameters, temperature, and isotopic composition of individualmaterials. The calculations of critical facility experiments are produced
for the benchmark models, generated by the specific reconstructing
codes with its features when adjusting model parameters, and using
the nuclear data library. The generated benchmark model, providing
the agreed calculated and experimental values for one or moreneutronic characteristics can lead to considerable differences for otherkey characteristics. The sensitivity of key neutronic characteristics
to the extra steel allocation in the core, and ENDF/B nuclear data
sources is performed using a few calculated models of BFS-62-3А
and BFS1-97 critical assemblies. The comparative analysis of thecalculated effective multiplication factor, spectral indices, sodium void
reactivity, and radial fission-rate distributions leads to quite different
models, providing the best agreement the calculated and experimental
neutronic characteristics. This fact should be considered during therefinement of computational models and code-verification purpose.

PAPER 1104329

Critical Experiments for BWR Fuel Assemblies with
Cluster of Gadolinia Rods
K.Yoshioka, T.Umano, M.Yamaoka(1), Y.Karino(2), J.J.Casal(3),M.Ouisloumen(4)1)TOSHIBA, Kawasaki, Japan, 2)TOSHIBA, Yokohama, Japan, 3)Westinghouse Electric Sweden, Västerås, Sweden, 4)Westinghouse
Electric Co., Pittsburgh, USA
Gadolinia-bearing fuel rods are needed for high-burnup fuels. Strong
neutron absorption of gadolinia makes an assembly heterogeneousfrom the viewpoint of reactor physics. The cluster of gadoliniabearing fuel rods is useful for higher-burnup fuels than current fuels.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
Few critical experiments have been reported for fuel assemblies
with the cluster of gadolinia-bearing fuel rods. We conducted
critical experiments for BWR fuel assemblies with the cluster ofgadolinia-bearing fuel rods in the Toshiba Nuclear Critical Assembly(NCA). Critical water level and power distribution were measured.
Measurements were compared with analyses by a continuous-energyMonte Carlo code, MCNP, with the JENDL-3.3 nuclear data library.

PAPER 1104817

The Calculation and Measurement of Fast Neutron
Reflection in the VVER-1000 Mock-Up Model Placedin the LR-0 Reactor
M.Košťál, B.Jánský, E.Novák, J.Milčák, E.Losa, J.Rejchrt,
Z.Lahodová(1), F.Cvachovec(2), M.Veskrna(3)1)Research Center Rez Ltd., Husinec-Rez, Czech Republic, 2)Universityof Defence, Brno, Czech Republic, 3)Masaryk University, Brno, Czech
Republic
The paper attempts to address the issue of the effect of biologicalshielding simulator on neutron transport in its vicinity. Neutronreflections from the biological shielding model may worsen theradiation conditions in the end part of the pressure vessel. A similar
problem emerges in connection with in the design of transport
experiments, where the reflections may distort the effect of transmitted
material on transport.
The experiments were conducted in the VVER-1000 mock-up placedin the LR-0 reactor whose modular character allows extraction of its
central part. The spectra were observed in conditions of full biologicalshielding as well as of its modification (in which case the central partof the biological shielding on was removed). The neutron spectra were
measured with the use of proton recoil method. In neutron energy
range below 1MeV the hydrogen spherical proportional counters(diameter of 40 mm) with various pressures of hydrogen were used,while in the case of over 1MeV scintillation detector with 10 x 10
mm cylindrical stilbene detector was employed. The results of theexperiment were compared with calculations. The calculations were
performed with ENDF/B-VII.0, JEFF-3.1. JENDL-3.3, JENDL-4.0,
ROSFOND 2009, and CENDL-3.1 nuclear data libraries.
On the basis of the resulting calculations and experiments, we may
conclude that neutron reflections from the biological shielding increase
the neutron flux density in the vessel. Generally, the calculation
overvalues the effect of reflections. In the case of the back concretereflector addition, the increase in neutron flux density in 0.1-0.2 MeV
was measured as 18 % while calculational result is 29 %.

PAPER 1105052

Development of Reactivity Meter with Novelty
Neutron Source Intensity Evaluation Model for BWR
Application
S.Tashiro, T.Kageyama, M.Tojo (1) , M.Nakai, K.Kagawa (2) ,Y.Iwata(3)1)Global Nuclear Fuel-Japan Co., Ltd., Kanagawa, Japan, 2)ElectricPower Development Co., Ltd., Tokyo, Japan, 3)Hitachi-GE Nuclear
Energy, Ltd., Ibaraki, JapanThe Ohma nuclear power plant has a plan to be the world’s first fully
MOX-operated plant. Therefore, it is important to evaluate some key
parameters representing core characteristics (e.g. control rod worth(CRW), moderator temperature coefficient (MTC), etc.) through somenuclear reactor physics experiments from the view point of keeping
nuclear design reliability. With respect of BWRs (Boiling Water
Reactors), Positive Period Method which forces operators to takecomplicated manipulation for core operation and to take long time has
been widely used for the evaluation of above mentioned parameters.
On the other hand, a reactivity meter which can evaluate thoseparameters in short time without complicated manipulation has a fewer
experiences in comparison with PWRs (Pressurized Water Reactors)
due to the difficulty of input parameters setting (e.g. neutron source
intensity). In this paper, a new concept of reactivity meter equippinga novelty neutron source intensity evaluation model is proposed.
Applying proposed reactivity meters to BWRs helps operators to
obtain high accuracy of core key parameters with much simpler test
steps and data analyses.

PAPER 1106265

Boron Carbide Neutron Screen for GRR-1 Neutron
Spectrum Tailoring
N.Chrysanthopoulou(1,2), P.Savva, M.Varvayanni, N.Catsaros(2)1)National Technical University of Athens, Greece, 2)NCSR Demokritos,
GreeceThe presence of fast neutron spectra in new reactor concepts(such as Gas Cooled Fast Reactor, new generation Sodium Cooled
Fast Reactor, Lead Fast Reactor, Accelerator Driven System andnuclear Fusion Reactors) is expected to induce a strong impact on
the contained materials, including structural materials (e.g. steels),
nuclear fuels, neutron reflecting materials (e.g. beryllium) and tritium
breeding materials (for fusion reactors). Therefore, effective operationof these reactors will require extensive testing of their components,
which must be performed under neutronic conditions representative of
those expected to prevail inside the reactor cores when in operation.
Depending on the material, the requirements of a test irradiation canvary. In this work preliminary studies were performed to observe thebehavior of the neutron spectrum within a boron carbide neutron
screen inserted in a hypothetical reflector test hole of the Greek
Research Reactor. Four different screen configurations were simulatedwith Monte Carlo code TRIPOLI-4. The obtained data showed that theinsertion of boron carbide caused not only elimination of the thermal (E< 1eV) component of the neutron energy spectrum but also absorptionof a considerable proportion of the intermediate energy neutrons (1·10-6
MeV < E < 1 MeV).

PAPER 1106850

Research and Development Activities forTransmutation Physics Experimental Facility in
J-PARC
T.Sugawara, H.Iwamoto, K.Nishihara, K.Tsujimoto, T.Sasa,
H.Oigawa
JAEA, Ibaraki, JapanThe Japan Atomic Energy Agency (JAEA) has the plan to construct
Transmutation Physic Experimental Facility (TEF-P) under aframework of J-PARC (Japan Proton Accelerator Research Complex)
project. TEF-P is a critical assembly which can load Minor Actinide(MA) fuels to perform reactor physics experiments for transmutation
systems such as Accelerator-Driven System (ADS) or Fast Reactor(FR). The facility can also use proton beam from the J-PARC
accelerator to investigate the controllability of ADS. Current status andactivities for TEF-P are described.

PAPER 1108237

Analysis of TCA Criticality, βeff and βeff/l Using
CASMO-4 and CASMO-5
S.Aoki(1), A.Michael, G.Ralph(2)1)MNF, Tokyo, Japan, 2)AREVA, Washington, USA
MNF and AREVA have been developing the BWR core calculation
system CASMO-4/MICROBURN-B2. This code system has been
modified to treat the ATRIUM 10XM bundle characteristics and satisfy
Japanese regulatory requirements. In this system, the cross section
data and the kinetics parameter are supplied by CASMO-4 code. The
accuracy of the treatments for nuclear criticality and transient eventsare very important for the reload core design and the verificationand the validation on these items are requested by the regulatory
agencies. MNF and AREVA calculated the neutron multiplication andthe βeff transient parameter using CASMO-4 and CASMO-5. MNF
compares the difference between CASMO-4 and CASMO-5 for bothof these parameters. The results of these calculations demonstrate
excellent agreement with differences of 0.13% in CASMO-4 and
0.19% in CASMO-5. Using the importance I value of the CASMO-5
calculation for the β eff measurement geometry, MNF confirmed thatthe βeff and βeff/l which were estimated by CASMO-4 β value and
CASMO-5 importance I were in good agreement.

PAPER 1126731

Measurement of Subcriticality Using Delayed Neutron
Source Combined with Pulsed Neutron Accelerator
T.Misawa, T.Yagi, C.H.Pyeon
Kyoto University Research Reactor Institute, Osaka, JapanA new experimental method for subcriticality measurement wasdeveloped by using delayed neutron source which is produced by87

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
external pulsed neutron source to increase accuracy of measured
results by overcoming the space dependency problem which means
difference of measured results in different detector position and often
appeared in almost all other subcriticality measurement techniques.
Experiments were performed at Kyoto University Critical Assembly(KUCA) combined with a DT accelerator to produce pulsed neutronin outside of the core repeatedly. In this method, neutron detection
counts in the prompt neutron time region which are appeared just after
injection of pulsed neutron are omitted, whereas neutron counts in the
delayed neutron time region which are appeared after disappearanceof exponential decay of the prompt neutron are adopted in analysis
based on neutron source multiplication method or neutron noise
analysis method; the variance to mean ratio method. In the delayed
neutron time region, neutron sources to initiate fission chain reactions
in subcritical state are delayed neutrons from delayed neutron
precursors which are mainly produced by fission chain reactions in
the prompt neutron time region, and delayed neutron precursors
exist only in the fuel region, which makes possible to decrease thespace dependency problem. The obtained results were comparedwith conventional pulsed neutron method, and it was found that the
space dependency problem in subcriticality measurement can be
fairly decreased by using the present new method compared withconventional one.

PAPER 1127355

Rossi-αParameter Measurement of Dalat Nuclear
Reactor by Analysis of Cross Power Spectral Density
Obtained from 2 Ion Chambers
T.M.Nguyen, C.S.Trang
Nuclear Research Institute, Dalat, Vietnam
Measurement and analysis of reactor power fluctuations from thecompensated ion chambers (CIC) placed in the reactor are a powerful
method in experimental study of nuclear reactor. This method allows
us to determine the important kinetic parameters such as Rossi-α
parameter (prompt neutron decay constant), reactivity, square moduleof reactor transfer function, effective delayed neutron fractions andneutron generation time, etc.
In this report, the authors present basis on theory, system ofequipment used to measure power spectral density (PSD) and crosspower spectral density (CPSD) of signals obtained from 2 KHK-56
ion chambers placed in the Dalat nuclear reactor. By fitting of thetheoretical curves with the measured curves, Rossi-α parameter wasdetermined.

PAPER 1127688

Analysis of Integral Experiment for Thorium Fuel
Cycle at Kyoto University Critical Assembly
Y.Takahashi, T.Sano, J.Hori, H.Unesaki, K.Nakajima
Research Reactor Institute, Kyoto University, Osaka, JapanTo measure integral neutronics characteristics of thorium loaded
core, critical experiments had been carried out at Kyoto University
Critical Assembly (KUCA). The critical experiments were performed
with various neutron spectra and thorium inventories. The thorium
loaded core has two regions which are a test zone and a driver fuelzone. The test zone consists of thorium plates and graphite plates.
In order to change the neutron spectrum of the experimental neutron
field systematically, the graphite/Th-232 ratio at the test zone had
been systematically varied by changing the combination of thethorium plates and the graphite plates in a unit cell. In this study,
the criticalities of thorium loaded core were analyzed by MVP2.0
with JENDL-4.0, JENDL-3.3. In addition, sensitivity analyses were
performed by SAGEP code and uncertainties of the numerical results
were evaluated by using cross section covariance matrix.

PAPER 1127907

Studies of Potential for Conversion of the Idaho
National Laboratory TREAT Transient Test Reactor to
Low-Enrichment Fuel
D.C.Kontogeorgakos, K.L.Derstine, G.Yesilyurt, H.M.Connaway,
A.E.Wright(1), M.D.Dehart, S.R.Morrell(2)1)Argonne National Laboratory, Argonne, USA, 2)Idaho National
Laboratory, Idaho Falls, USA
In light of a potential restart of the Transient Reactor Test Facility at
the Idaho National Laboratory, the conversion of the existing highenrichment uranium (HEU) fuel to a low-enrichment uranium (LEU)

88equivalent is being studied. This paper describes the development andvalidation of the methods to simulate the performance o f the existing
HEU core. Experimental data from critical configurations and transient
measurements have been used to develop both steady state andtransient models and corresponding simulation approaches. In thisprocess, the uncertainties of the core parameters have been identified
and potential biases established. Modeling of the HEU core haslargely been completed and is now being applied to the assessment ofcandidate LEU fuel and core designs.

Track7: Reactor Concepts and Designs
PAPER 1068764

Axially Heterogeneous Thorium Fuel Designs forTransuranic Burning in Reduced-Moderation BWRs
B.A.Lindley, G.T.Parks
University of Cambridge, Cambridge, UK
Reduced-moderation Boiling Water Reactors (RBWRs) can allow
sustained burning of transuranics (TRU), such that full actinide recycle
can be achieved. However, the void coefficient (VC) tends to become
positive with reduced moderation and high TRU loading, which canseverely limit the design feasibility and performance. This motivatesuse of thorium (Th) as an alternative feed to uranium (U), as this tends
to result in a more negative VC, leading to much improved neutronic
performance. While axially homogeneous fuel design is preferable forease of fuel fabrication, it is valuable to evaluate axially heterogeneous
fuel designs to see if improved neutronic performance can be realised,
which could lead to a reduction in fuel reprocessing and fabrication
throughput. Multi-recycled Th-TRU fuel contains Th, U, Pu and MAs,
leading to a wide range of possible fuel designs. Axially heterogeneous
designs are considered using 3D pincell calculations using the Monte
Carlo code Serpent. Radially heterogeneous assembly designs
are considered in a companion paper. Spatial separation of ThTRU and Th-U3 into regions of the order of a few thermal neutron
diffusion lengths greatly improves neutronic performance. This canbe accomplished radially or axially, but radial separation results insignificantly easier fuel fabrication. Axial seed-blanket heterogeneity
improves neutron economy at the expense of high power peaking,such that the radially heterogeneous assembly design is preferred.
Separation of Th-TRU and Th-U3 into larger regions is not effective,
due to increased power density in the Th-TRU region with voiding andwith burn-up, leading to a more positive VC. It is therefore concluded
that there is no motivation to pursue axially heterogeneous Th-TRU
RBWR burner designs.

PAPER 1090085

Method Development and Reactor Physics Data
Evaluation for Improving Prediction Accuracy of Fast
Reactors’ Minor Actinides Transmutation Performance
T.Takeda(1), T.Hazama(2), K.Fujimura, S.Sawada(3)1)University of Fukui, Fukui, Japan, 2)JAEA, Fukui, Japan, 3)Hitachi-GE
Nuclear Energy, Ibaraki, JapanThe Ministry of Education, Culture, Sports, Science and Technology
in Japan has launched a national project entitled “technology
development for the environmental burden reduction” in 2013. The
present study is one of the studies adopted as the national project.
The objective of the study is the efficient and safe transmutationand volume reduction of MAs with long-lived radioactivity and highdecay heat contained in HLW in sodium cooled fast reactors. We areaiming to develop MA transmutation core concepts harmonizing MA
transmutation performance with core safety and to improve design
accuracy related to MA transmutation performance. To validate andimprove design accuracy of the high safety and high MA transmutation
performance of SFR cores, we develop methods for calculating thetransmutation rate of individual MA nuclides and estimating uncertaintyof MA transmutation by using burnup sensitivity. Preliminary results
are shown for the methods. Also we show various measured reactor
physics data to be used to reduce the uncertainty of MA transmutation
calculations. In the present study the overall consistency of themeasured data is investigated by evaluating the usefulness of
conventional static data as well as those related to MA transmutation
obtained from various facilities like Monju, Joyo, FCA, BFS and PFR.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
PAPER 1094719

A Long Life Sodium Cooled Fast Reactor Concept
with Radial Shuffling
Z.Li, Y.Zheng, L.Cao, H.Wu
Xi'an Jiaotong University, Shaanxi, China
This paper proposes an idea of designing a long-life sodium cooled
fast reactor. The reactor aims to be with commercial scale of
1000MWe and without too much challenge on the materials andoperational conditions. The long-life is realized based on the breeding
and burning strategy, in which the periodical radial shuffling is adopted.
The one-through fuel cycle is adopted for nuclear non-proliferation,
therefore, only the uranium fuel is loaded as the seed fuel and noplutonium is recycled.The design of seed assemblies is similar to thecurrent sodium cooled fast reactor using metallic fuel. The enricheduranium-zirconium alloy with enrichment of 11.5% is used as the fuel.
The depleted uranium from the tails in enriching processes is used in
the blanket. An inward-convergent shuffling scheme is proposed. In
the scheme, the assemblies in the outer blanket region are first moved
to the inner blanket region and then shuffled to the burning region
while the assemblies in the burning region are shuffled inward ring by
ring and then skip to the outer blanket region. Preliminary calculations
had proved that the shuffling scheme is feasible and the discharged
blanket assemblies can be used as driver assemblies in anotherreactor with similar core design. The core life can reach over 38 years
with a shuffling period of 2 years. The key parameters of the core arewithin the acceptable range in current fast reactor designs.

PAPER 1099566

Critical Boron Concentration Reduction Method in aCore Design
C.J.Hah, W.S.W.A.Kahar(1), D.I.Chang, K.B.Seong(2)1)KINGS, Ulsan, Korea, 2)Korea Nuclear Fuel Company, Daejeon,
Korea
A systematic core design method is developed to determine Gdbearing fuel assembly having two types of Gd rods, low wt% of Gd
rod and high wt% of Gd rod. The purpose of the method is to lower
the critical boron concentration (CBC) of the preliminary core loading
pattern (PLP), and consequently to achieve more negative or lesspositive moderator temperature coefficient (MTC). In this method,both the ratio of the number of low-Gd rod to the number of high-Gd
rod (r) and assembly average Gd wt% (w) are the decision variables.
The target function is the amount of soluble boron concentration
reduction, which can be converted to ∆kTARGET. ∆k-tables parameterizes
k-infinite difference (∆k) between assembly in PLP and new designed
assembly with the decision variables. The ∆k-tables relating (r,w) to∆k is generated prior to the determination of Gd-bearing fuel assembly
pattern using the least square method. The constraints required todetermine a set of ∆k are physically realizable pair, (r,w), and the sumof ∆k of new designed assemblies as close to ∆kTARGET as possible. A
set of new designed assembly ∆k is determined to satisfy the targetfunction expressed as ∆kTARGET. New Gd-bearing assemblies selectedbased on valid pairs of (r,w) are replaced with existing assemblies ina PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1
used as a reference model. Shin-Kori Unit 3 is APR 1400. CASMO-3
is used to formulate ∆k-tables and MASTER for depletion calculation.
CASMO-3/MASTER calculations with new designed assemblies
produce lower CBC than the expected CBC, proving that the proposed
method works successful.

PAPER 1104423

BigT - A New Burnable Absorber Concept for PWR
M.Yahya, H.Yu, Y.Kim
KAIST, Daejeon, Korea
A new burnable absorber concept for Pressurized Water Reactor(PWR) is presented in this paper. The proposed concept, named“Burnable absorber-Integrated control rod Guide Thimble” (BigT),
requires minor modifications on standard control rod guide thimble
in existing PWR fuel assemblies. The BigT offers various design
flexibilities such that any loading pattern and core management
objective may potentially be met. Few technical challenges are
expected with the BigT, but none is a show-stopper. Preliminary lattice
analyses of the BigT absorbers installed in Westinghouse 17x17
and Combustion Engineering 16x16 fuel assemblies show that the
proposed concepts perform as well as the commercial technologies interm of lattice reactivity and power peaking management. In addition,
sufficiently high worth of control rods are obtained with the BigT inplace. Since only slight modifications are required to the existing PWR

lattice, the BigT can easily and readily be retrofitted into commercial
fuel assemblies. The capability of replacing the BigT during refuelingenables installation of fresh burnable absorbers in any assembly at
beginning-of-cycle. This in turn offers a promising solution to realizing
a low critical boron concentration core, and potentially even a solubleboron-free PWR. All neutronic simulations in this study were performed
using Monte Carlo SERPENT code with ENDF/B-VII.0 library.

PAPER 1104448

Preliminary Design of a Spherical Breed/Burn Reactor
E.Y.García-Cervantes, J.L.François, R.C.López-Solís
National Autonomous University of Mexico, Morelos, México
A breed/burn reactor (BBR) is a reactor that through transmutation
process can burn depleted uranium (or other fertile material as
thorium) with only a small amount of fissile material. The advantagesof BBR over other type of nuclear reactors are wide, taking into
account its long life operation without refueling. In this paper we
present the simulation results done with MCNPX code and JEFF3.1 cross sections library at 1200 K. The purpose of this study is tocompare two different types of BBR: a spherical core and a cylindricalcore, with the same nuclear fuel in both reactors. We found that bothreactors maintained a supercritical state during 4500 days (12.3
years) without refueling. The plutonium production was achieved
along the reactor in both cases. We also find that the spherical
reactor operated as a stationary wave reactor, where the neutron flux
uniformly increases in the sphere along the breeding zone with time,
meanwhile the cylindrical reactor operated as a traveling wave reactor
where neutron flux propagates as a wave in the cylinder along burnup.
Further analysis will include Th-U fuel cycles and a more detailed
model.

PAPER 1104495

Optimization of Ultra-Long Cycle Fast Reactor Core
T.Tak, D.Lee(1), T.K.Kim(2)1)Ulsan National Institute of Science and Technology, Ulsan, Korea, 2)
Argonne National Laboratory, Argonne, USA
An optimization of an ultra-long cycle fast reactor (UCFR) design witha power rate of 1000 MW (electric), UCFR-1000, has been performed
to increase the safety of UCFR. Firstly, geometric optimization has
been performed to decrease its peaking factors so that the peaktemperatures measured by thermal hydraulic feedback are within
the limit of design basis event (DBE). Secondly, fuel composition
optimization has been performed by adopting Pressurized Water
Reactor (PWR) spent fuel as a blanket material instead of naturaluranium. Lastly, a small-size UCFR with a power rate of 100 MWe,
UCFR-100, has been proposed for developing a short term deployable
nuclear reactor. The major optimization process for UCFR-100 is
decreasing maximum neutron flux and fast neutron fluence.
The optimized UCFR-1000 has been enlarged radially and shortened
axially from the initial UCFR design and this modification makes theburning speed of active core movement slower. It has been confirmedthat a full-power operation of 60 years without refueling is feasible forboth UCFR-1000 and UCFR-100 core designs by a breed-and-burnstrategy. By the design optimization study, the reductions of maximum
neutron flux, fast neutron fluence, and axial power peaking have beenachieved, which are favorable for the safety of the UCFR.

PAPER 1106189

The Main Characteristics of the Evolution Project
VVER-S with Spectrum Shift Regulation
P.Teplov, A.Chibiniaev, E.Bobrov, P.Alekseev
NRC “Kurchatov Institute”, Moscow, Russia
The researches on the increase of achievable fuel burn-up in lightwater reactors (LWR) with so-called «spectrum shift» have been
carried out in the world for a long time. The main idea of the «spectrum
shift» is based on neutron spectrum shifting from the resonanceenergy region at the beginning of the cycle to the thermal region at theend of the cycle. There are many different ways of such regulation in
the core – starting from coolant density variation during reactor cycle,
to changing water-uranium ratio with some mechanical equipment.
Spectrum shift reactivity regulation with zirconium displacers ispresented in this paper for uranium and mixed uranium and plutoniumoxide (MOX) fuel loadings for modernized reactor core VVER-S withnew fuel assemblies with special guide tubes for displacers, developed
for these purposes.

89

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
PAPER 1127032

Alternative Cores for a Multipurpose Experimental
Sodium-Cooled Fast Reactor with U-Zr Fuel
T.Noh, M.H.Kim
Kyung Hee University, Gyeonggi-do, Korea
The conceptual design of a Multipurpose Experimental Sodium cooled
Fast Reactor (MESOF) based on U-Zr driver fuel was proposed at theprevious study. Core performance characteristics and safety features
are evaluated for alternative cores with U-TRU-Zr fuel and UO2 fuel.
All nuclear design calculations were done with TRANSX / DANTSYS/ REBUS3(with DIF3D module) system, except effective delayed
neutron fraction (βeff) which was evaluated with MCNPX.
The inherent safety feature is proven by negative sodium void worth
for all cores. The shutdown margin of U-TRU-Zr core is enough tocontrol large excess reactivity at BOC, but the safety feature for UO2
core will be evaluated in later study.

PAPER 1127409

Preliminary Evaluation of Coolant Void Reactivity ofa Re-Entrant Channel Pressure-Tube Supercritical
Water Cooled Reactor
P.Schwanke, E.Nichita
UOIT, Ontario, Canada
The Pressure-Tube Supercritical-Water Cooled Reactor (PT-SCWR)
is one of the reactor concepts investigated by the Generation-IV
International Forum as a reactor concept capable of achieving an
overall nuclear power plant efficiency in the vicinity of 50%, higher
than the approximately 35% that characterizes current nuclear power
plants. The development process includes studying the neutronic
methods that can provide an adequately-accurate modelling of thecore. The PT-SCWR evolved from the conventional pressurized
heavy-water reactor but presents a more heterogeneous core
because of the much higher discharge burnup (~40 MWd/kg vs.
~7.5 MWd/kg) and greater variation of coolant density. The latest PTSCWR concept features re-entrant fuel channels, each consisting ofa pressure tube lined by a ceramic thermal insulator, and a central
inner flow-tube. The fuel assembly is located in the space between
the ceramic insulator and the central flow-tube. The re-entrant channel
configuration presents unique coolant voiding characteristics becauseof the separate central and outer coolant regions. There may be
the possibility, under certain postulated accident conditions, that thecoolant in the region between the flow-tube and insulator can void
before the coolant in the flow-tube. This work presents preliminary
calculations of the resulting coolant-void reactivity for the case when
only the outer coolant voids as well as for the case when the entire
coolant in the channel voids. Lattice as well as full-core calculations
are performed. Results show that voiding of only the outer coolant
region yields a small, ~2 mk, positive reactivity, while the subsequent
complete coolant voiding induces a large (>20 mk) negative reactivity.

Track8: Reactor Operation and Safety
PAPER 1092230

Development and Preliminary Verification of thePWR On-Line Core Monitoring Software System:
SOPHORAW.Li, X.Zhang, Z.Dang, Y.Guan, H.Lu, N.Bai, K.Huang, J.Li(1),Y.Wu, X.Fan, S.Cai(2)1)China Nuclear Power Technology Research Institute, Shenzhen,
China, 2)Information Technology Center, Shenzhen, China
This paper presents an introduction to the development andpreliminary verification of a new on-line core monitoring softwaresystem (CMSS), named SOPHORA, for fixed in-core detector(FID) system of PWR. Developed at China General Nuclear Power
Corporation (CGN), SOPHORA integrates CGN’s advanced PWR
core simulator COCO and thermal-hydraulic sub-channel code
LINDEN to manage the real-time core calculation and analysis.
Currents measured by the FID are re-evaluated and used as bases
to reconstruct the 3-D core power distribution. The key parameterssuch as peak local power margin and minimum DNBR margin are
obtained by comparing with operation limits. Pseudo FID signals
generated by data from movable in-core detector (MID) are used to90

verify the SOPHORA system. Comparison between predicted power
peak and the responding MID in-core flux map results shows that the
SOPHORA results are reasonable and satisfying. Further verificationand validation of SOPHORA is undergoing and will be reported later.

PAPER 1105990

Development of Risk Monitor RiskAngel for RiskInformed Applications in Nuclear Power Plants
F.Wang, J.Wu, Y.Wu, L.Hu, Y.Li, J.Wang, J.Wang, R.Hu(1), Y.Yin(2),Z.Yang(3)1)Institute of Nuclear Energy Safety Technology, Anhui, China, 2)
Shenzhen University, Guangdong, China, 3)Nuclear and Radiation
Safety Center, Beijing, China
This paper presents the development of risk monitor software
RiskAngel at FDS Team and its applications as a plant specific risk
monitor, which supports risk-informed configuration risk management
for the two CANDU 6 units at the Third Qinshan Nuclear Power Plant
in China. It also describes the regulatory prospective on risk-informed
PSA applications and the use of risk monitor at operating nuclear
power plants, high level technical and functional requirements for thedevelopment of CANDU specific risk monitor software, and future
development trends.

PAPER 1127006

Steady-State Subchannel Analysis of Partially
Blocked Coolant Channels in a Pool-Type TRIGA
Reactor
C.M.Chance, J.I.C.Vermaak, J.C.Ragusa
Texas A&M University, Texas, USA
Additional instrumentation tubes (neutron detectors, thermocouples)
are scheduled for deployment in Texas A&M University’s TRIGA corein order to collect measurements for multiphysics code validations.
This instrumentation will displace water and reduce coolant flow insome areas between fuel bundles. Thermal-hydraulic analyses arecarried out using COBRA-TF and STAR-CCM+ to analyze the change
in cooling capabilities. Results show that the coolant exit temperature
increases by ~9°C and that DNBR margins are not affected.
Additionally, since the instrumentation tubes act as flow straighteners,
an overall reduction in hot spots is observed. The study also provided
insight into the comparison of COBRA-TF with the higher fidelity code;
STAR-CCM+.

Track9: Transient and Safety Analysis
PAPER 1084717

Improvement of Space-Time Kinetics Capability in the
SNATCH Solver and Comparison to KIN3D/PARTISN
Results
M.Guyot, R.L.Tellier, P.Archier(1), F.Gabrielli(2)1)CEA Cadarache, Saint-Paul-lez-Durance, France, 2)Karlsruhe Instituteof Technology, Eggenstein-Leopoldshafen, Germany
Efforts are going on both at the CEA Cadarache and at the Karlsruhe
Institute of Technology (KIT) to develop an efficient and robust threedimensional neutron kinetics solver to be employed as a new neutronic
tool for simulating severe accident transients with the SIMMER code.
To this aim, the SNATCH space-time kinetics has been recently
extended and coupled with the SIMMER code in a multiple-channel
approach. On the other hand, the PARTISN code was coupled to
KIN3D, a time-dependent model for neutron kinetics of the ERANOS
code system and applied to transient analyses. In this paper, therecent SNATCH extensions are introduced and a comparison to the
KIN3D/PARTISN code is provided. SNATCH implementation of the
Improved Quasi-Static Method (IQM) relies on a special treatment inwhich Point Kinetics (PK) parameters are computed for each micro
time step by assuming a linear variation of the flux shape over a
shape time step. This procedure provides a better coupling between
the PK parameters and material perturbations. A new formulation for
the reactivity computation is implemented in SNATCH, as the original
formulation does not account for reactivity variation for source-induced
transients. Comparative results show that, overall, SNATCH neutron
kinetics results are coherent with KIN3D/PARTISN for a fast transientinduced by a source-jerk in a three-dimensional subcritical system

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
driven by an external source. It is also found that reactivity variations
are more correctly assessed by the new formulation on reactivity.

PAPER 1094483

Different Methods to Model the MSLB without Primary
Cooling Pumps Using HEMERA V1 System Codes
B.Normand

faulty withdrawal.

PAPER 1104619

The SIMMER/PARTISN Capability for Transient
Analysis
M.Marchetti, F.Gabrielli, A.Rineiski, W.Maschek

Some Results of Studying of Spatial Kinetics in Fast
Reactors

Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany
SIMMER-III is extensively used in many institutions for reactor
transient and safety analyses. It is a two dimensional fluid-dynamics
code coupled with a spatial neutron kinetics model. It features a
modular structure. A three dimensional version, SIMMER-IV, is alsoavailable. The recent trend to parallel computations has prompted
the parallelization of the fluid-dynamics module, yet both versionsof SIMMER still rely on non-parallel neutron transport solvers, the
TWODANT and THREEDANT codes, respectively.
In this paper, we consider substituting these two solvers with theirparallel evolution: PARTISN. PARTISN stands for PArallel TImedependent SN and it is a 3D neutron transport code. Our aim is toestablish a SIMMER version with a parallelized neutronic module. We
describe here the coupling procedure and the modifications needed
in both SIMMER and PARTISN to make the coupling possible. We
then test our SIMMER-III/PARTISN and SIMMER-IV/PARTISN
coupled codes for several cases. These first tests provide an initial
validation of the coupled codes. Comparison with the standard
SIMMER-III and SIMMER-IV codes results proves that the couplingis correct. The advantage, in term of computational time of using aparallel neutron transport solver is found to be quite substantial. In
one case the speedup of the total computational time is ≈ 40% with
only 8 processors, with a reduction in the cumulative time spent into
the neutron transport solver of a factor of 3 or more. This may be
useful especially for three-dimensional cases that feature a very large
number of meshes and many energy groups.

I.S.Panova, E.F.Seleznev, A.A.Belov(1), A.M.Zhukov,
I.P.Matveenko(2)PAPER 1105969

IRSN, Fontenay-aux-Roses, France
HEMERA.V1 (Highly Evolutionary Methods for Extensive Reactor
Analyses), is a fully coupled 3D computational chain developed jointly
by IRSN and CEA. It is composed of CRONOS2 (core neutronics),
FLICA4 (core thermal-hydraulics) and the system code CATHARE2.
Multi-level and multi-dimensional models are developed to account forneutronics, core thermal-hydraulics, fuel thermal analysis and system
thermal-hydraulics, dedicated to best-estimate and conservative
simulations and sensitivity analysis. In the IRSN, the HEMERA.V1
chain is adopted to investigate several types of reactivity accidents.
This paper presents an example of HEMERA.V1 utilization for thesimulation of the Main Steam Line Break (MSLB) without primarycooling pumps. For this transient, two different methods to model
the water flow at the core entrance are compared. The model using
a matrix flow which takes into account the flow in the loops leads tothe loss of natural circulation which may occur several DNBR (andthe minimum one may not be the first one) and the model using an
uniform distribution of flow rates leads to an unique DNBR and theintact loops have an asymptotic flow rate (there is no loss of natural
circulation).

PAPER 1101709

1)Nuclear Safety Institute of Russian Academy of Sciences, Moscow,
Russia, 2)Institute for Physics and Power Engineering, Obninsk, Russia
The paper presents an analysis of the solution to the spatial nonstationary equation of neutron transport the example of a fast reactor.
To solve the spatial kinetics equation in three-dimensional geometry
in diffusion approximation a calculated code TIME was created. In a
direct problem the neutron flux density and its derivatives (for example,
reactor power) are determinate at each time step. Solution of a direct
problem allows to investigate a deformation of a steady neutron fieldform which is one of the main reasons of spatial effects occurrence
in a reactor. Such deformations affect spatial distribution of delayedneutrons that is reflected in behavior of reactivity and in related with it
results of experiments. The paper also considers some questions of
reducing of spatial effects through a choice of a point of reactivity input
or detector location. In the inverse problem the reactor reactivity is
calculated using the known dependence of reactor power on time. The
paper describes some problems of solving the inverse problem taking
into account spatial effects.

PAPER 1103760

Atucha-2 Obliquely Inserted Control Rods RELAP53D/NESTLE Model
R.G.Gonzalez, F.D'Auria(1), C.Parisi(2),M.Pecchia(3),
O.Mazzantini(4)1)GRNSPG, Pisa, Italy, 2)ENEA, Rome, Italy, 3)PSI, Villingen,
Switzerland, 4)NA-SA, Buenos Aires, Argentina
Atucha-2 is a Siemens-designed PHWR reactor in phase ofcommissioning in the Republic of Argentina. Its geometrical complexityand peculiarity (e.g., oblique control rods, positive void coefficient)required a developed and validated complex three dimensional (3D)
neutron kinetics (NK) model. In the framework of the agreement
between NA-SA and University of Pisa a detailed NES-TLE (threedimensional neutron kinetics code) model of the Atucha-2 NPP
was developed. This document summarizes the procedures forthe implementation of the oblique control rods into the RELAP5/NESTLE model: a particular arrangement of RELAP5/NESTLE control
rods insertion mode for such kind of oblique control rods and an
implementation into the homogenized two group cross sections of adhoc calculated correction factors (these parameters were obtained bypreviously executed Monte Carlo calculations) was developed. Someapplications, among the scenarios selected to perform safety analysisof the Atucha-2 NPP (CNA-II), are also reported: preliminary Scram
Rod Worth, analysis of a Control Rod Ejection Accidents and a CR

Development of a High-Fidelity Monte Carlo
Thermal-Hydraulics Coupled Code System Serpent/
SUBCHANFLOW - First Results
M.Daeubler, J.Jimenez, V.Sanchez
Karlsruhe Institute of Technology, Karlsruhe, Germany
Worldwide efforts to develop high-fidelity or ab-initio, high performance
multi-physics tools are on-going due to the availability of relatively
cheap, large-scale parallel computers. In order to arrive at a coupledneutronics and thermal-hydraulics tool with a higher spatial resolution
and accuracy than currently used Best-Estimate tools, an external
coupling between the Monte Carlo Reactor Physics code Serpent andthe sub-channel code SUBCHANFLOW has been developed. The
coupled code system is intended to serve as reference for deterministic
reactor dynamics code developments in future exploiting the fact that
Serpent was conceived as a lattice code for such deterministic tools.
The achieved coupling is based on Serpent’s recently introduced
universal multi-physics interface. Enabling the interface, Serpent treats
temperature dependence of nuclear data using target motion sampling.
Furthermore, variance reduction techniques and under-relaxation
schemes were utilized. The development is verified by code-to-code
comparison with the deterministic tool DYNSUB. Simulation resultsof both code systems for a pressurized water reactor fuel assembly
under hot full power conditions were found in good agreement. Finally,
the impact of the thermal-hydraulic feedback on Serpent’s numericalperformance was assessed. For a hot full power simulation, Serpent
ran roughly ten times slower than with pre-broadened continuous
energy cross sections for homogeneous hot zero power thermalhydraulic conditions. However, the developments presented here
simply represent a first step towards Serpent/SUBCHANFLOW as
a comprehensive reference for deterministic safety analysis tool
development as many issues such as combining target motion
sampling with bound atom scattering have yet to be resolved.

91

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Track10: Nuclear Data

improvement from systematic resonance treatment especially for 237Np
and 243Am.

PAPER 1068671

PAPER 1106244

G.Chiba, Y.Kawamoto, M.Tsuji, T.Narabayashi

J.B.Fiel(1), L.C.Leal(2)Analysis of Radioactivity Ratios of Fission ProductNuclides Deposited to Soil in Fukushima Dai-Ichi
Nuclear Power Plant Accident
Hokkaido university, Sapporo, Japan
The present study focuses on radioactivity ratios between different
fission product isotopes being the same element in nuclear fuelloaded in the Fukushima Dai-Ichi nuclear power plant. Using themeasured radioactivity data for the soil of the power plant site, weestimate the radioactivity ratios at the reactor shutdown and comparethem with numerically calculated values. Detail irradiation histories foreach unit and axial profiles of void fraction and power distribution areapproximately considered.
Calculation values of the following radioactivity ratios over a whole
core, Cs-134/Cs-137, Cs-136/Cs-137, Sr-89/Sr-90 and Te-132/
Te-129m, agree with the measurement-based values within 50%
difference. Especially the measurement-based radioactivity ratio of
Cs-134 to -137 and that of Cs-136 to -137 are well reproduced bynumerical calculations within 25% difference. On the radioactivity
ratios of Te-132 to -129m, good agreement between calculation andmeasurement-based values is obtained when the JEFF-based nuclear
data library is used.

PAPER 1081806

Experimental Uncertainty Estimation on the Effective
Capture Cross Sections Measured in the PROFIL
Experiments in Phenix
E.Privas, G.Noguère, C.De Saint Jean, J.Tommasi, P.Archier
CEA Cadarache, Saint-Paul-lez-Durance, France
A desire of increasing nuclear system safety and fuel depletion is
directly translated by a better knowledge on nuclear data. PROFIL and
PROFIL-2 experiments give integral information on capture and (n,2n)cross sections and cumulative fission yields for several isotopes (95Mo,
97
Mo, 101Pd, 105Pd, 133Cs, 143Nd, 144Nd, 145Nd, 147Sm, 149Sm, 151Eu, 233U,
234
U, 235U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 244Cm ...). Interpretation havebeen done many times in the past but without experimental uncertainty
estimation. The cross section library JEFF-3.1.1, the covariancedata base COMAC and the code system ERANOS-2.2 are used forthis updated interpretation. This study is focusing on the uncertainty
estimation on experimental values sensitive to capture cross sections.
Three steps are required: the fluence scaling, the uncertainty
propagation on the fluence and finally the uncertainty estimation on
ratio variation of interest. This work is done with CONRAD using
Bayesian adjustment and marginalization method. Mean C/E results
and conclusions are identical to the previous interpretation. A fluence
uncertainty of 1.4% is found for the two experimental pins of PROFIL-2
and 1.9% for PROFIL. Propagating this new information on the
fluence to ratio variation of interest gives experimental uncertainties
between 1% to 2.5% for the isotopes present in the experimental pins.
One of the main results are for 238Pu, 239Pu, 240Pu, 241Pu and 242Pu
capture cross sections: C/E are respectively equal to 1.03, 0.98, 0.97,
1.08 and 1.14 with an uncertainty lower than 2.5%. All the results will
provide feedback on variance-covariance matrices for further works.

PAPER 1105105

Updated Multi-Group Cross Sections of Minor
Actinides with Improved Resonance Treatment
M.Sohail, O.Shim, M.Kim
Kyung Hee University, Korea
The study of minor actinide in transmutation reactors and other futureapplications makes resonance self-shielding treatment a significant
issue for criticality and isotope depletion. Resonance treatment
for minor actinides has been carried out by subgroup method withimproved interference effect through interference correction. Subgroupdata was generated using RMET21 and GENP codes along withmulti-group cross section data by NJOY nuclear data processing
system. Updated multi-group cross section data library for a neutrontransport code nTRACER was compared with solutions from MCNPX.
The resonance interaction of uranium with minor actinides has been
included by modified interference treatment of interference correction
in subgroup methodology. The comparison of cross sections andmultiplication factor in pin and assembly problems showed significant

92

Parameterized Representation of Macroscopic
Cross Section for PWR Reactor Considering with 12
Burnable Absorber Fuel Rods in the Fuel Element
1)Military Institute of Engineering, Rio de Janeiro, Brazil, 2)Oak Ridge
National Laboratory, Oak Ridge, TN, USA
The purpose of this work is to describe, by means of Tchebychev
polynomial, a parameterized representation of the homogenizedmacroscopic cross section for PWR fuel element as a function ofsoluble boron concentration, moderator temperature, fuel temperature,
moderator density and 23592U enrichment, considering burnable
absorber rods in the fuel element. The cross-section data nalyzed
are fission, scattering, total, transport, absorption and capture. The
parameterization enables a quick and easy determination of problemdependent cross-sections to be used in few groups calculations. The
methodology presented in this paper will allow generating group crosssection data to perform PWR core calculations without the need togenerate them based on computer code calculations using standard
steps. The results obtained by the proposed methodology whencompared with results from either the SCALE code calculations or with
the MCNPX code calculations show very good agreement.

Track11: Research Reactors and Spallation Sources
PAPER 1087763

Jules Horowitz Reactor.France Experimental Loop
Development According Optimized Irradiation
Process.
S.Gaillot, S.Vitry, T.Dousson
CEA Cad- FRANCE, Saint-Paul-lez-Durance, France
Experimental reactors in the world enable researchers to meet theneeds of industry and institutions not only by providing support to theexisting nuclear infrastructure (Gen.2) but also by preparing the future
generation (Gen.3, Gen.4) or even by responding to other needs aswell (supports with fusion, medical applications). It is for this specificpurpose that the Jules Horowitz Polyvalent Irradiation Reactor is now
being built at the CEA Cadarache Research Center (located in thesouth of France). This Material Testing Reactor (MTR type) is designed
to irradiate materials or fuel samples for various experimental tests.
The reactor will also produce Mo99 radioelements that will supply 25%
to 50% of current European needs.
The goal of this paper is to describe a fuel irradiation loop, now understudy, that will be designed to carry out power ramps tests to provide
technical support to the Generation 2 and 3 nuclear power plants. In
order to increase its irradiation capacity (2 to 3 per cycle), this loop
takes into account the requirements that will lead to the optimizationof all experimental processes in the facility (such as non-destructive
examinations before and after the test, specific loading tools). All
these considerations are being taken into account in order to offerto the customer’s complete and optimized conditions in terms ofexperimental irradiations processes.

PAPER 1106121

Preliminary Neutronic Design for the Conceptual
Fluid Granular Spallation Target
J.Y.Li, L.Gu, D.W.Wang(1), W.Jiang(2)1)Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou,
China, 2)University of Science and Technology of China, Hefei, China
A new concept of spallation target design for Accelerator-Driven
Subcritical system (ADS) was performed. The main material of thetarget was tungsten, Helium was used as coolant, proton beam
current is 4mA, and energy is 250MeV. Some physical parameterssuch as the filling mode, neutron leakage spectrum, the averageleakage neutron yield, leakage neutron distribution along the axial of
spallation target, the stability of spallation target, and differences inspallation target energy deposition distribution have been analyzed,
based on the differences between granular particle size and the formof arrangement. From the analyzed result, when the filling mode

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
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Poster Session
is face-center cubic arrangement, and the affection by diameter of
granular is minimal, the perturbation is no more than 16%. And thisperturbation will be lower with the smaller size of granular. The radiusof spallation target should be a little larger than the beam radius, it isappropriate to choose 10cm for spallation target radius and 20cm forits length, respectively.

PAPER 1128332

Preliminary Optimization Analysis of the Radiation
Shielding of the China Lead-Based Research Reactor
Q.Yang, B.Li, T.Li(1), C.Liu, Y.Bai, J.Song(2)Physico-Technical faculty, Al-Farabi Kazakh National University, Almaty,
Kazakhstan
In the present work the programming code is created on the basisof which the accumulation kinetics of tritium and isotope of He4 in
the Be9 sample is analyzed depending on the time. The program iswritten in C++ programming language and for the calculations Monte
Carlo method was applied. This program scoped on the calculation ofconcentration of helium and tritium in beryllium samples depending on
the spectrum of the neutron flux in different experimental reactors such
as JMTR, JOYO and IPEN/MB. The processes of accumulation ofhelium and tritium for each neutron energy spectrum of these reactors
were analyzed.

1)University of Science and Technology of China / Institute of Nuclear
Energy Safety Technology, Chinese Academy of Sciences, Anhui, China,2)Institute of Nuclear Energy Safety Technology, Chinese Academy of
Sciences, Anhui, China
Accelerator Driven subcritical System (ADS) is recognized as anefficient nuclear waste transmutation device. Supported by theStrategic Priority Research Program of “the Future Advanced Nuclear
Fission Energy-ADS transmutation system”, the China LEAd-based
Research Reactor (CLEAR-I) is proposed. Along with the approachingof the CLEAR-I design, the radiation shielding for CLEAR-I is updated
and optimized step by step to meet with new shielding requirements.
Employing the modeling program MCAM and calculation system
VisualBUS developed by FDS Team, the shielding capability wasverified using Monte Carlo method. As shown from the results, the fastneutron flux for components in reactor vessel is under the limitationand the neutron radiation for mechanism in containing room has been
as low as possible. After shutdown for 7 days, the dose rate in mostarea of containing room is lower than 100 μSv/hr, allowing hands on
operation. Replacement of components such as the spallation target
in containing room is possible.

PAPER 1127738

PAPER 1233696

P.Saracco(1), R.Marotta, G.Lomonaco, D.Chersola(2), L.Mansani(3)T.Dang(1), L.Mao(1,2), Q.Zeng(1), Y.Wu(1,2)

PAPER 1126142

Accumulation of Tritium in Beryllium Material under
Neutron Irradiation
S.Kunakov, N.Takibayev, N.Kenzhebayev

A Preliminary Study of an Improved Area Method,
Adapted to Short Time Transients in Sub-Critical
Systems
1)INFN Genova, Genova, Italy, 2)INFN Genova / Universita di Genova,
Genova, Italy, 3)Ansaldo Nucleare SpA, Genova, Italy
Since the ’90s interest has grown in the characterization of subcritical systems. Because they should not have control/shutdown
devices, it is necessary to prevent by design neutron flux divergencein all conditions. Most of the available theoretical models on neutron
kinetic behavior in multiplying systems are tailored on the descriptionof critical systems; then they have decreasing validity as well as thesystem is far and far from criticality; as a consequence many of theavailable monitoring methods are of questionable use for sub-criticalsystems and give contradictory informations: this is true, among theothers, for the Area Method. One of the difficulties in understanding
the kinetic behavior of a sub-critical system is the scarcity ofexperimental measurements available. Then we decided to develop a
full time-dependent MCNP6 simulation of a sub-critical core to obtainsome ”phenomenological-like” data whose analysis leads to conclude
that geometrical inhomogeneity of the system plays a key role indetermining its time behavior over short time scales. Starting from an
original observation by Weinberg and Wigner, we conclude that, forthe short time scale considered (~ μs), we cannot use the ”classical”
diffusion approximation but a little bit more complicated formulation,
similar to what is known as telegrapher’s equation: along this line
we find a reasonable justification for the idea that, before using any
analysis method grounded on the assumption that the system hasrelaxed into its fundamental mode, we could have to wait for a longer
time than the prompt neutrons lifetime. We conclude that from anexperimentalist’s point of view it should be at lest necessary to cut
from the analysis the lowest part of the time interval before using the
Area Method, but more refined analyses are needed.

PAPER 1128078

Preliminary Neutronics Analysis of a Spallation Target
for Transmutation
B.Wu, T.Dang, C.Liu, J.Zou, M.Wang, J.Jiang
Institute of Nuclear Energy Safety Technology, Anhui, China
Accelerator Driven subcritical System (ADS) was recognized as aneffective nuclear waste transmutation device. Target in liquid or solidin an independent loop bombarded by the charged particle beam wasconsidered as the neutron source. Heavy metal was chosen as target
material or coolant. The present work was to discuss the possibility of
taking Minor Actinides as part of spallation target material, for a bettertransmutation performance of entire ADS. According to the thermalcooling and irradiation time limitation, a conceptual design of target fortransmutation was proposed. And preliminary neutronics analysis fortarget performance assessment including neutron flux, neutron yieldas well neutron spectrum is shown in this work.

Preliminary Analysis of Radioactive Source Term
for Normal Operation of China Lead-Based Research
Reactor (CLEAR-I)
1)Institute of Nuclear Energy Safety Technology, Chinese Academy of
Sciences, Hefei, Anhui, China, 2)University of Science and Technology
of China, Hefei, Anhui, China
China LEAd-based Research Reactor (CLEAR-I) is a 10 MW leadbismuth cooled research reactor, which serve as ADS and Leadcooled Fast Reactor technology verification platform. The evaluationof radioactive source term is needed for reactor decommissioning,
worker and public dose rate. In this contribution, radioactive source
term of CLEAR-I was calculated for normal operation in reactor
core, primary coolant, cover gas and secondary coolant by CADbased Multi-Functional 4D Neutronics & Radiation Simulation System(VisualBUS) developed by FDS Team. This calculation is a guidance
for reactor decommissioning, worker and public dose and radioactive
waste disposal and transportation.

Track12: Fuel Cycle and Actinide Management
PAPER 1092834

Fusion Hybrids for Generation of Advanced(231Pa+232U+233U+234U)-Fuel in Closed (U-Pu-Th)-Fuel
Cycle
G.G.Kulikov, A.N.Shmelev, E.G.Kulikov, V.A.Apse, E.F.Kryuchkov
National Research Nuclear University MEPhI (Moscow Engineering
Physics Institute), Russia
Technology of controlled thermonuclear fusion (CTF) is traditionallyregarded as a practically inexhaustible energy source. However,development, mastering, broad deployment of fast breeder reactors
and closure of nuclear fuel cycle (NFC) can also extend fuel base of
nuclear power industry (NPI) up to practically unlimited scales. Under
these conditions, it seems reasonable to introduce into a circle of the
CTF-related studies the works directed towards solving some principal
problems which can appear in a large-scale NPI in closed NFC.
The first challenge is a large scale of operations in NFC back-end thatshould be reduced by achieving substantially higher fuel burn-up in
power nuclear reactors. The use of 231Pa-232Th-232U-233U fuel in light-water
reactor (LWR) opens a possibility of principle to reach very high (about
30% HM) or even ultra-high fuel burn-up.
The second challenge is a potential unauthorized proliferation offissionable materials. As is known, a certain remarkable quantity of232
U being introduced into uranium fraction of nuclear fuel can produce
a serious barrier against switching the fuel over to non-energy
purposes.

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Involvement of hybrid thermonuclear reactors (HTR) into NPI structure
can substantially facilitate resolving these problems. If HTR will be
involved into NPI structure, then main HTR mission consists not inenergy generation but in production of nuclear fuel with a certain
isotope composition.
The present paper analyzes some neutron-physical features in
production of advanced nuclear fuels in thorium HTR blankets.
The obtained results demonstrated that such a nuclear fuel may becharacterized by very stable neutron-multiplying properties during full
LWR operation cycle and by enhanced proliferation resistance too.
The paper evaluates potential benefits from involvement of HTR withthorium blanket into the international closed NFC.

PAPER 1100837
233

U Fuel Production and 30-Year Utilization without
Reprocessing and Refuelling Using Heavy Water
Coolant

A.Talamo, Y.Gohar
Argonne National Laboratory, Chicago, USA
This study examines the physics of a thorium fuel cycle based on
generating the initial fissile (233U) fuel inventory in a Deuterium-Tritium
fusion device and on operating a 600 MWth fission reactor. For bothphases of the fuel cycle, the fuel form is an aqueous slurry consistingof thorium oxide micro-particles dispersed into heavy water. The slurryis the fuel carrier and the coolant. After 180 full power days in thefusion driven device, the fuel enrichment is 1.4%. The enrichments
is defined as the ratio between the fissile actinides mass and thetotal actinides mass. After the removal of fission products, the 1.4%
enriched slurry thorium-uranium fuel can be used for longer than 30
full power years in a 600 MWth critical reactor core, without adding
any fissile material. The critical reactor has three zones: inner fissile,central fertile, and outer reflector.

PAPER 1101727

Uncertainty Analysis for Fuel Flux Calculations ofFast Reactors with External Fuel Cycle
E.F.Seleznev, A.A.Belov(1), G.N.Manturov, A.M.Tsybulya(2)1)Nuclear Safety Institute of Russian Academy of Sciences, Moscow,
Russia, 2)Institute for Physics and Power Engineering, Obninsk, Russia
The paper focuses on the results of uncertainty analysis whencalculating nuclide composition in fuel of fast reactors and on
uncertainties of determining nuclide composition in the external fuelcycle. As demonstrated, the main contributions to the uncertainty of
nuclide composition are due to:● uncertainties in operation of the reactor and in the fuel-cycle time;● uncertainties in nuclide clean-up factors at the Closed Nuclear Fuel
Cycle (CNFC) stages when reprocessing spent nuclear fuel;● uncertainties in isotopic-kinetics cross-sections;● uncertainties in nuclide decay data.

PAPER 1105292

Study on Transmutation and Storage of LLFP Using a
High-Temperature Gas-Cooled Reactor
K.Kora, H.Nakaya, K.Kubo, H.Matsuura(1), S.Shimakawa, M.Goto,
S.Nakagawa(2)1)Kyushu University, Fukuoka, Japan, 2)Japan Atomic Energy Agency,
Ibaraki, Japan
There is a need to temporally store high-level radioactive waste (HLW)
until the location of final disposal is decided. HLW contains several
types of long-lived fission product (LLFP) which stay radioactive for
hundreds of thousands of years. In addition, they tend to be chemicallymobile when dissolved into ground water thus may not be suited forgeological disposal. A facility that is able to store and incinerate LLFP
simultaneously is desirable. The high-temperature gas-cooled reactor(HTGR) is one of the fourth generation nuclear reactors currently
under research and it has some favorable characteristics that allow
the reactor to destroy LLFP through nuclear transmutation. In thisstudy, the capability of HTGR as LLFP transmuter was evaluated interms of neutron economy. Considering gas turbine high-temperature
reactor with 300 MWe nominal capacity (GTHTR300) as HTGR,
transmutations of four types of LLFP nuclide were estimated using
Monte Carlo transport code MVP and ORIGEN. In addition, burnup simulations for whole-core region were carried out using MVPBURN. It was numerically shown that the neutron fluxes change
significantly depending on the arrangement of LLFP in the core. When
15 t of LLFP is placed in an ideal manner, the GTHTR300 can sustain

94

sufficient reactivity for one year while transmuting up to 30 kg per year.
Additionally, there are more space available for storing larger amountof LLFP without affecting the reactivity. These results suggest that
there is a possibility of using GTHTR300 as both LLFP storage andtransmuter.

PAPER 1105802

Core Library for Advanced Scenario Simulation,
C.L.A.S.S. : Principle & Application.
B.Mouginot, B.Leniau, N.Thiollière (1) , M.Ernoult, S.David,
X.Doligez(2), A.Bidaud, O.Méplan, R.Montesanto(3), G.Bellot, J.B.Clavel,
I.Duhamel, E.Letang, J.Miss(4)1)SUBATECH, Nantes, France, 2)Institut de Physique Nucléaire
d'Orsay, Orsay, France, 3)Laboratoire de Physique Subatomique etCosmologique, Grenoble, France, 4)Institut de Radioprotection et deSûreté Nucléaire, Fontenay-aux-Roses, France
The global warming, the increase of world population and thedepletion of fossil resources have lead us in a major energy crisis.
Using electronuclear energy could be one of the means to solve apart of these issues. The way out of this crisis may be enlightenedby the study of transitional scenarios, guiding the political decisions.
The reliability of those studies passes through the wide variety of thesimulation tools and the comparison between them.
From this perspective and in order to perform complex electronuclear
scenario simulation, the open source Core Library for Advance
Scenario Simulation (CLASS) is being developed. CLASS main
asset is its ability to include any kind of reactor, whether the system
is innovative or standard. A reactor is fully described by its evolution
database that must contain a set of different fuel compositions inorder to simulate transitional scenarios. CLASS aims at being a useful
tool to study scenarios involving Generation IV reactors as well as
innovative fuel cycles, like the Thorium cycle.
The following contribution will present in detail the CLASS software.
Starting with the working principle of this tool, one will explain theworking process of the different modules such as the evolution
module. It will be followed by an exhaustive presentation of the UOXMOX bases generation procedure. Finally a brief analysis of the error
made by the CLASS evolution module will be presented.

PAPER 1106324

Fuel Composition Generation Techniques of Nuclear
Fuel Cycle Simulators
R.R.Flanagan, E.A.Schneider, C.Bagdatlioglu
University of Texas at Austin, Texas, USA
Nuclear fuel cycle simulators track the flow of materials through thefacilities that comprise a nuclear energy system. The composition ofthese materials, which simulators specify at the elemental or isotopic
level, is driven by the neutronic characteristics of the reactors in thesystem. Therefore, all simulators include a method for generating input
and output compositions for the reactor fuel they track, widely known
as recipes. This paper surveys the recipe generation approaches
taken by five simulators, which range from pre-computed reactor
physics modeling to on-the-fly calculations. It concludes with anillustrative example of the canonical parametric recipe generation
problem simulators are called upon to solve.

PAPER 1106382

Core Burnup Calculation of Uranium Rock-Like Oxide
Fuel PWR for Spent Fuel Composition Estimation
H.Akie, K.Nishihara, Y.Nakano, N.Shirasu(1), T.Iwamura(2)1)Japan Atomic Energy Agency, Ibaraki, Japan, 2)University of Fukui,
Fukui, Japan
Burnup calculations of uranium rock-like oxide (U-ROX) fuel werecarried out using a PWR core model, in order to obtain the spent
U-ROX fuel composition, which is to be used for the analysis of the
LWRs ”replace once” scenario in Japan. Fuel assemblies loading andshuffling were simulated for the power peaking reduction, and U-ROX
fuels were burnt for 3 times of 400 EFPD cycle. As a result, plutonium
production amount in U-ROX is much less than that in UO2, and thefraction of 238Pu in total plutonium becomes large. This is due to less238
U conversion to 239Pu in U-ROX than UO2, while the burnup of 235U
and the conversion to 238Pu is as much as that in UO2. The power
peaking factor of the U-ROX PWR core tends to be large, and the
reduction of the peaking factor seems important.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
Track14: Education in Reactor Physics
PAPER 1085651

Multi-Collision Theory for Educated Pedestrians
P.Picca, R.Furfaro(1), P.Ravetto(2)1)University of Arizona, AZ, USA, 2)Politecnico di Torino, Torino, Italy
The aim of the paper is to present the linear transport theory in a form
which can be easily accessible also to person with a basic graduate
knowledge in mathematics and physics. The time dependent linear
Boltzmann equation is approached using a multi-collision methodology
and the solution for each propagation direction is sought via classicalmethods used for linear ODE, such as Laplace transform and Green’s
function method. Some results that can be fully managed analyticallyare presented and a general strategy for the numerical implementation
proposed.

PAPER 1073641

Virtual Labs on Unique Experimental Equipment
I.S.Saldikov, V.V.Afanasyev, E.F.Kryuchkov, V.I.Petrov,
M.Y.Ternovykh, G.V.Tikhomirov
NRNU MEPhI, Moscow, Russian Federation
A description of a database on the support of laboratory work on aunique experimental equipment, concept development and use ofvirtual labs in the learning process are present.

PAPER 1104802

PINSPEC: A Monte Carlo Code for Pin Cell Spectral
Calculations for Educational Applications
W.Boyd, N.Gibson, S.Shaner, L.Li, J.Hunter, K.Smith, B.Forget
MIT, Massachusetts, USA
Students in many reactor physics courses are exposed to canonicalreactor physics concepts through theoretical problems simplified
to allow for tractable analytical solutions. Such problems typicallyrequire tedious mathematical derivation which is often not the mosteffective approach to teaching basic reactor physics concepts. A newcomplementary methodology to introduce these concepts is madepossible with PINSPEC, a pin cell Monte Carlo code for educationaluse. PINSPEC enables students to simulate pin cell models forvarious reactor types with a simple-to-use Python interface. PINSPEC
uses point-wise cross section data and includes a module for SingleLevel Breit-Wigner cross-section generation and Doppler broadening.
The PINSPEC code supports a variety of tallies which students may
use to compute resonance integrals, multi-group cross sections,and more for various materials and pin configurations. PINSPEC isundergoing review for open source release in the near future such that
it will be a free and accessible tool for instructors developing reactor
physics curricula with an applied and interactive approach to learning.

PAPER 1127873

Unique Approaches in Emphasizing the Role ofReactor Laboratories and Facilities for Training and
Education of Future Nuclear Engineers ‘without theBorders’
T.Jevremovic, R.Schow, L.McDonald IV(1), A.Rey(2)1)University of Utah, Utah, USA, 2)UdelaR, Montevideo, Uruguay
The 21st century brings numerous challenges into daily lives ofnuclear engineers world-wide, such as at nuclear power plants (facing
new safety paradigm influenced by 2011-unfortunate Japanese
Fukushima incident), at the universities in educating and training new
generations of nuclear engineers in capturing new expectations ofsafety standards, and in reaching out to countries that are interested
to develop new or revitalize their decades-old education and training
programs as pertaining to nuclear engineering and science. In thispaper we present two novel aspects we have developed: (a) theestablishment of novel educational approaches pertaining to training
and practices of nuclear safety culture in university curricula, and (b)development of novel digital-type class focused on the basic aspectsof nuclear science and engineering shared between a class in theState of Utah (USA) and Uruguay.

Track15: Research Related to Fukushima Accident
PAPER 1102869

Spatial Correlation Modeling of Macroscopic Cross
Section with Weierstrass Function
T.UekiJapan Atomic Energy Agency, Tokai, Ibaraki, Japan
Material media formed via sedimentation are under the random
fluctuation of constituent atoms/molecules with spatial correlation.
Extreme physical disorder is known to lead to the 1/f fluctuation. In
order to model these characteristics inherent in the formation of fueldebris, a randomized Weierstrass function (RWF) in fractal geometryis examined in this work. Here, RWF is a version of Weierstrass
function, is able to confine the influence of correlation within a certain
range of coordinate increments, and is globally under a fixed variance.
Therefore, RWF is capable of modeling the macroscopic cross section
with some unknown spatial distribution of constituent isotopes. The
potential impact of RWF modeling on neutron transport simulation isassessed using Woodcock tracking. Results obtained indicate thatthe distance-to-collision sampling over realizations of a continuously
varying random medium will be necessary for the Monte Carlocalculation if the influence range of spatial correlation is larger thanthe mean free path of the corresponding non-random homogeneous
medium.

SpecialSession2: Reactor Physics and Criticality Safety
Activities in OECD/NEA Working Party
PAPER 1109848

Effects of Nuclear Data Library and Ultra-Fine GroupCalculation for Large Size Sodium-Cooled Fast
Reactor OECD Benchmarks
T.Kugo, K.Sugino, M.M.Uematsu(1), K.Numata(2)1)Japan Atomic Energy Agency, Ibaraki, Japan, 2)NESI incorporation,
Ibaraki, Japan
The present paper summarizes calculation results for an international
benchmark proposed by the Sodium-cooled Fast Reactor core Feedback and Transient response (SFR-FT) under the framework of theWorking Party on scientific issues of Reactor Systems (WPRS) of
the Nuclear Energy Agency of the OECD. It focuses on the large size
oxide-fueled SFR. Library effect for core performance characteristics
and reactivity feedback coefficients is analyzed using sensitivity
analysis. The effect of ultra-fine energy group calculation in effective
cross section generation is also analyzed. The discrepancy is about
0.4% for a neutron multiplication factor by changing JENDL-4.0 with
JEFF-3.1. That is about -0.1% by changing JENDL-4.0 with ENDF/
B-VII.1. The main contributions to the discrepancy between JENDL-4.0
and ENDF/B-VII.1 are 240Pu capture, 238U inelastic scattering and 239Pu
fission. Those to the discrepancy between JENDL-4.0 and JEFF-3.1
are 23Na inelastic scattering, 56Fe inelastic scattering, 238Pu fission,
240
Pu capture, 240Pu fission, 238U inelastic scattering, 239Pu fission and239
Pu nu-value. As for the sodium void reactivity, JEFF-3.1 and ENDF/
B-VII.1 underestimate by about 8% compared with JENDL-4.0. The
main contributions to the discrepancy between JENDL-4.0 and ENDF/
B-VII.1 are 23Na elastic scattering, 23Na inelastic scattering and 239Pu
fission. That to the discrepancy between JENDL-4.0 and JEFF-3.1
is 23Na inelastic scattering. The ultra-fine energy group calculation
increases the sodium void reactivity by 2%.

PAPER 1121153

Evaluation of OECD/NEA/WPRS Benchmark on
Medium Size Metallic Core SFR by Deterministic Code
System: MARBLE and Monte Carlo Core: MVP
M.Uematsu, T.Kugo(1), K.Numata(2)1)Japan Atomic Energy Agency, Ibaraki, Japan, 2)NESI incorporation,
Ibaraki, Japan
In the frame work of the working party on reactor and system (WPRS)of the OECD nuclear Energy Agency (NEA), the benchmark on
sodium cooled fast reactor (SFR) was conducted. Within the OECD/
NEA/WPRS benchmark, study on medium size metallic fuel core was95

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Wednesday October 1, 2014 - 16:15 - 18:15 PM - Mizuho_C,D

Poster Session
performed using code system for fast reactor core calculation withdeterministic method MARBLE and with Monte Carlo method MVP.
The latest nuclear library JENDL-4.0 is used for calculation of eigenvalues(keff) and reactivity (sodium void, Doppler and control rod worth). The
differences of calculation results between the analysis methods are
summarized in this paper. Sensitivity studies of eigenvalue and sodium
void reactivity for the medium size metallic fuel benchmark core areconducted to determine the main reactions contributing to the difference
between JENDL-4.0 and other libraries JEFF-3.1 and ENDF/B-VII.1.

PAPER 1195260

Quantifying the Effect of Undersampling in Monte
Carlo Simulations Using SCALE
C.M.Perfetti, B.T.Rearden
Oak Ridge National Laboratory, Oak Ridge, USA
This study explores the effect of undersampling in Monte Carlocalculations on tally estimates and tally variance estimates for burnup
credit applications. Steady-state Monte Carlo simulations were
performed for models of several critical systems with varying degreesof spatial and isotopic complexity, and the impact of undersampling on
eigenvalue and flux estimates was examined. Using an inadequate
number of particle histories in each generation was found to produce
a bias of about 100 pcm in eigenvalue estimates and biases thatexceeded 10% in fuel pin flux estimates.

SpecialSession5: Multiscale, Multiphysics Approaches in
Nuclear Science and Engineering Applications
PAPER 1102894

Research on SCWR Core Characteristics Utilizing PinWise Neutronics Thermal-Hydraulic Coupling Method
Q.Hong, G.Niu, C.Tang, L.Lu, G.Shi
Shanghai Nuclear Engineering Research & Design Institute, Shanghai,
China
According to the intense flow heterogeneity and strong Neutronics /Thermal-Hydraulic feedback of SCWR, in present work, a pin-wiseneutronics and thermal-hydraulic coupling method is developed and acore analysis code package SCAP is established. The package utilizes
sub-channel program NCED-SCWR, lattice program PARAGON,
nodal diffusion program MRAPS, and multi-functional code COUPLE.
Furthermore, 3D core power distribution & flow distribution, reactivity
parameters and isotope density are studied for a typical core with 121
fuel assemblies. The results show that both the coupling method andcode package are applicable to the SCWR core analysis. Based on
the analysis, it is pointed out that the fuel rod with the highest cladding
temperature may be not the one with peaking power in SCWR, which
is different from PWRs. The obtained conclusion will be benefit for
SCWR core design.

SpecialSession6: Nuclear Criticality Safety of Fuel Debris
PAPER 1087714

Post-Accident Defueling Procedure and Its Criticality
Safety Evaluation of the Fukushima-Daiichi Nuclear
Power Plants
N.Takaki
Tokyo City University, Tokyo, Japan
Post-accident defueling procedure for the Fukushima-Daiichi power
plants is discussed. The authors propose a rotating plug defuelingsystem that enables access to deep vessel bottom and a watercleanup system to avoid loss of visibility in the water during defuelingworks. Criticality of fuel debris accumulated at the bottom of reactor
vessel was evaluated based on conservative assumptions for wide
range of water volume ratio. Required amount of neutron absorbing
materials to prevent re-criticality accident throughout defuelingprocesses was also evaluated. Boron pellet and boron gel aredeveloped and some fundamental experiments were performed toobtain the properties.

96

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Mizuho_A

Track1-10 Reactor Analysis Methods
Session Chair: Aldo Dall'Osso(AREVA NP), Christophe Demaziere(Chalmers Univ. of Tech.)
8:00 AM

9:00 AM

T.G.Saller, E.W.Larsen, T.Downar

A.T.Till, J.E.Morel, M.L.Adams

University of Michigan, Ann Arbor, USA
Because many of the current-generation reactor analysis codes
provide neutron fluxes based on diffusion theory, there is considerable
interest in methods that provide a more accurate diffusion solution
without significantly increasing computational costs. In this work,
an asymptotic analysis, previously used to derive a homogenizeddiffusion equation for lattice systems, is generalized to derive a onedimensional, one-group homogenized SP 2 equation as a moreaccurate alternative to the standard homogenized diffusion equation.
This analysis results in new diffusion coefficients and an improved
formula for flux reconstruction. Five fixed-source lattice functions
are required to calculate the diffusion coefficients, three of which
are also used in the flux reconstruction. In a homogeneous medium,
these lattice functions reduce to simple polynomials in angle, and thehomogenized SP2 equation reduces to its standard form. The new flux
reconstruction formula includes three “correction” terms that accountfor spatial derivatives of the scalar flux and differences between theinfinite-lattice eigenvalues and the core eigenvalue.

Texas A&M University, TX, USA
The standard multigroup (MG) method for energy discretization of thetransport equation can be sensitive to approximations in the weighting
spectrum chosen for cross-section averaging. As a result, MG often
inaccurately treats important phenomena such as self-shielding
variations across a fuel pin. From a finite-element viewpoint, MG
uses a single fixed basis function (the pre-selected spectrum) withineach group, with no mechanism to adapt to local solution behavior.
To address these issues, we introduce a Petrov-Galerkin finiteelement multi-group (PG-FEMG) method, a generalization of the MG
method that is related to the family of multiband (MB) methods. The
only approximation in PG-FEMG is that the angular flux is a linear
combination of basis functions. The coefficients in this combination are
the unknowns. A basis function is non-zero only in the discontiguous
set of energy intervals associated with its band. We implement the
PG-FEMG method for several realistic pin-cell problems and find it
to be significantly more accurate per degree of freedom than MG for
the quantities of interest that we tested, including criticality eigenvalue
and power profile shape. We find PG-FEMG is much less sensitive
to errors in weighting spectra compared to standard MG. We discuss
straightforward generalizations to multi-dimensional problems ofpractical interest.

An Asymptotic Homogenized SP2 Approximation to
the Boltzmann Equation. I. Derivation

8:20 AMAn Asymptotic Homogenized SP2 Approximation to
the Boltzmann Equation. II. Discontinuity Factors andNumerical Testing
T.G.Saller, E.W.Larsen, T.Downar
University of Michigan, Ann Arbor, USA
The current generation of LWR reactor analysis codes based on
homogenized diffusion theory has provided sufficient accuracy for awide range of reactor application. However, there is interest in methods
that can improve the accuracy of these codes by using a more
rigorous diffusion coefficient and a more robust flux reconstruction
method. In a companion paper, a one dimensional, one-grouphomogenized SP2 equation is asymptotically derived, and an improved
formula for flux reconstruction is obtained. In the present paper, theSP2 equation is tested, and the flux reconstruction formula is used toobtain flux and current discontinuity factors for the SP2 equation. For atest problem, the asymptotic SP2 formulation is compared to standard
SP2, asymptotic diffusion, and standard diffusion. The eigenvalue andreconstructed fluxes are both examined. In general, the asymptotic
equations are more accurate than the standard equations, and SP2
is more accurate than diffusion theory, especially for optically small
systems.

8:40 PM

Generalized and Standard Multigroup Neutron
Diffusion Equation Eigenvalue Problem with the Finite
Volume Method
A.Bernal, R.Miro, G.Verdu, D.Ginestar
Universitat Politecnica de Valencia, Valencia, Spain
Most of the neutron diffusion codes use numerical methods giving
accurate results in structured meshes. However, the applicationof these methods in unstructured meshes to deal with complex
geometries is not straightforward and it may cause problems ofstability and convergence of the solution. By contrast, the Finite
Volume Method (FVM) is easily applied to unstructured meshes andis typically used in the transport equations due to the conservationof the transported quantity within the volume. In this paper, the FVM
algorithm implemented in the ARB Partial Differential Equations
Solver has been used to discretize the multigroup neutron diffusion
equation to obtain the matrices of the generalized eigenvalue problem,
which has been solved by means of the SLEPc library. Nevertheless,these matrices could be large for fine meshes and the eigenvalue
problem resolution could require a high calculation time. Therefore, atransformation of the generalized eigenvalue problem into a standard
one is performed in order to reduce the calculation time.

A Generalized Multigroup Method Based on Finite
Elements

9:20 AMAn Asymptotic, Homogenized, Anisotropic,
Multigroup Diffusion Approximation to the Neutron
Transport Equation
T.J.Trahan(1, 2), E.W.Larsen(2)1)Los Alamos National Laboratory, New Mexico, USA, 2)University ofMichigan, Ann Arbor, USA
Due to its relatively low computational cost, the neutron diffusion
approximation remains one of the most commonly-used computational
tools for reactor analysis. Although diffusion requires more
approximations than higher-order “transport” methods such as SN andMonte Carlo, the computational cost of these methods prohibits theiruse for routine reactor analysis. It has long been known that diffusion
is anisotropic in heterogeneous reactors, i.e., neutrons diffuse more
rapidly in some directions than others. While in many reactors theanisotropic diffusion effects are negligible, in lattices containing voided
or optically-thin channels these effects are significant. It is therefore
desirable that the homogenized diffusion coefficient be a tensor. In
the past, monoenergetic diffusion tensors have been asymptoticallyderived from the exact, continuous energy Boltzmann transport
equation for large systems with a periodic lattice structure. We have
now derived multigroup diffusion tensors using asymptotic analysisof the multigroup, homogenized diffusion equations. The multigroupdiffusion tensors are defined such that the asymptotic limits of thecontinuous energy transport equation and the multigroup diffusion
equation are the same. Our numerical results indicate that theasymptotic, multigroup diffusion tensors are more accurate than thestandard scalar multigroup diffusion coefficients for certain problems.
In particular, the asymptotic diffusion tensor is more accurate for large
lattice systems, and systems containing optically-thin channels.

9:40 AM

Accuracy of the Linear Discontinuous Galerkin
Method for Reactor Analysis with Resolved Fuel Pins
C.N.Mcgraw, M.L.Adams, W.D.Hawkins, M.P.Adams, T.Smith
Texas A&M University, Texas, USA
Significant literature exists on the accuracy of the Method of
Characteristics (MOC) for solving the transport equation for reactors
with realistic representations of geometries. The same is not true
for Discontinuous Finite Element Methods. We present a resolution
study and error analysis detailing how the Linear DFEM (LD) spatial
discretization method performs on the well-known two-dimensional
C5G7 benchmark problem as a function of spatial and angularresolution, for spatial meshes that conform to the pin geometries.
We compare pin powers and k-eigenvalues against reference MCNP
results, as a function of spatial and angular resolution. We use “product”
Gauss-Chebyshev quadrature sets that range from 2 to 24 polar levels

97

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Mizuho_A

Track1-10 Reactor Analysis Methods
Session Chair: Aldo Dall'Osso(AREVA NP), Christophe Demaziere(Chalmers Univ. of Tech.)and 32 to 512 azimuthal angles. Our spatial resolution ranges from64 to 900 quadrilateral cells per pincell. We find that the LD method
performs well, with k and pin-power results consistent with third-ordertruncation error. The LD error on our quadrilateral cells appears to be
between the errors of “lumped” and standard (un-lumped) versions ofPieceWise Linear DFEM (PWLD). (LD has 3 unknowns per polygonal
cell; the PWL methods have four per quadrilateral cell, five perpentagonal cell, etc.) On a coarse spatial mesh with three rings in fueland one in water per pin cell, the LD spatial truncation error is <10
pcm in k and well below 0.5% in each of the 1056 pin powers. Ourerror analysis indicates that our high-resolution solution (900 cells per
pincell, 24 polar angles, and 256 azimuthal angles) is accurate to <1 pcm in k and to <0.004% in each pin power. We provide these pin
powers. These high-resolution solution errors are significantly smaller
than the statistical precision reported for the reference Monte Carlo
results.

98

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Mizuho_B

Track4-6 Verification, Validation and Uncertainty Analysis
Session Chair: Laurent Chabert(AREVA TA), Kenji Nishihara(JAEA)
8:00 AM

which require some modifications to the standard equations ofgeneralized perturbation theory. Numerical results are presented andanalyzed for a PWR fuel assembly with control rods out and inserted.
The computational efficiency of the calculations is discussed.

T.Takeda, B.Foad, H.Katagiri(1), H.Matsumoto, K.Kirimura(2)9:00 AM

Sensitivity and Uncertainty Calculations Methods of
Neutronics Parameters in PWR Cores Part I: Theory
and Sensitivity Calculations
1)University of Fukui, Fukui, Japan, 2)Mitsubishi Heavy Industry, Kobe,
Japan
Sensitivity and uncertainty calculations methods of neutronics
parameters in pressurized light water reactors have been developed.
The sensitivity is composed of three terms; the first is the sensitivityof cell-averaged multi-group cross-sections relative to multi-groupinfinite dilution cross-sections, the second is the sensitivity ofassembly averaged few-group macroscopic cross-sections relative
to cell-averaged multi-group cross-sections, and the third is thesensitivity of neutronics parameters in PWR cores relative to fewgroup macroscopic cross-sections. Combining the three sensitivities,
the sensitivity of neutronics parameters in PWR cores relative to multigroup infinite dilution cross-sections is obtained. The discussion ofthis method will be presented in two papers; the present paper is partI, where the theory and some numerical results for typical pin cells,
fuel assemblies and a simple PWR core are shown. The present
method gives us multi-group sensitivities for individual nuclides ineach reaction type, and wide ranges of applications are possible to thefields such as cross-section adjustment and uncertainty reduction.

8:20 AM

Constrained Quantities in Uncertainty Quantification:
Ambiguity and Tips to Follow
Z.Perko, D.Lathouwers, J.L.Kloosterman, T.H.J.J.van der Hagen
Delft University of Technology, Faculty of Applied Sciences, Departmentof Radiation Science and Technology, Section of Nuclear Energy andRadiation Applications, Delft, The Netherlands
The nuclear community relies heavily on computer codes andnumerical tools. The results of such computations can only be trusted
if they are augmented by proper sensitivity and uncertainty (S&U)
studies. This paper presents some aspects of S&U analysis whenconstrained quantities are involved, such as the fission spectrum or
the isotopic distribution of elements.
A consistent theory is given for the derivation and interpretation of
constrained sensitivities as well as the corresponding covariancematrix normalization procedures. It is shown that if the covariancematrix violates the “generic zero column and row sum” condition,
normalizing it is equivalent to constraining the sensitivities, but sinceboth can be done in many ways different sensitivity coefficientsand uncertainties can be derived. This makes results ambiguous,
underlining the need for proper covariance data. It is also highlighted
that the use of constrained sensitivity coefficients derived witha constraining procedure that is not idempotent can lead tobiased results in uncertainty propagation. The presented theory
is demonstrated on an analytical case and a numerical example
involving the fission spectrum, both confirming the main conclusions of
this research.

8:40 PM

Adjoint-Based Sensitivity and Uncertainty Analysis of
Lattice Physics Calculations with CASMO-4
M.PusaVTT Technical Research Centre of Finland, Espoo, Finland
The topic of this paper is the development of sensitivity and uncertainty
analysis capability to the reactor physics code CASMO-4 in the UAM(Uncertainty Analysis in Best-Estimate Modelling for Design, Operationand Safety Analysis of LWRs) benchmark. The developed calculation
system enables the uncertainty analysis of homogenized multi-groupcross-sections, diffusion coefficients and pin powers with respect tonuclear data. The uncertainty analysis methodology is deterministic,
meaning that the sensitivity profiles of the responses are computed
first, after which uncertainty is propagated by combining the sensitivity
profiles with the covariance matrices of the uncertain nuclear data.
The sensitivity analysis is based on perturbation theory which enables
computing the sensitivity profiles efficiently by solving one generalizedadjoint system for each response. The mathematical backgroundof this work is reviewed and the main conclusions related to theimplementation are summarized. Special emphasis is placed on thesensitivity analysis of two-group homogenized diffusion coefficients

Nuclear Data Uncertainty Propagation on Power Maps
in Large LWR Cores
A.Santamarina
CEA, Cadarache, France
Nuclear data uncertainty propagation is more and more required insafety calculations of large NPP cores. In this study, nuclear data
uncertainties have been propagated to LWR cores using accuratetransport calculations. PWR benchmarks representative of the French
NPP fleet were selected: one 900MWe (Fessenheim), one 1500MWe(CHOOZ), as two complementary GEN-III benchmarks belonging to
the OECD/UAM benchmark, representative of large PWR UOX and–partially loaded - MOX cores surrounded by thick stainless steel
reflector. Reactivity and assembly power map sensitivity coefficientsto multigroup cross sections have been calculated using perturbation
theory tools implemented in ERANOS2/SNATCH transport code.
Different sources of nuclear data uncertainty were compared:
covariance matrices from BOLNA, and French COMAC associatedto JEFF3.1.1 evaluations, were used for uncertainty propagation.
The total uncertainty on the multiplication factor keff is not sensitive tothe size and the loading of the PWR cores; this uncertainty amounts
to 560pcm (1σ) with COMAC, compared to 780 pcm with BOLNA
covariances. The main component of keff uncertainty is ν U235, followed
by 238U capture, σc U235 and σf U235.
The uncertainty on the radial power map increases with the core size,due to higher Eigen Value Separation factor (EVS>20). For example,
using COMAC covariances, the uncertainty in central assemblies
increases from 1.5% for PWR-900MWe to 3.4% for large N4 andGEN-III PWRs. In the peripheral assemblies which can host the power
peak, the uncertainty increases from 1.1% for 900MWe to 2% for large
PWRs. Therefore, the potential power swing bias in JEFF3.1.1 corecalculations varies from 2.6% for PWR-900MWe to 5-6% for N4 andlarge GEN-III PWRs. Using BOLNA covariance files, the uncertainty
on the Pcentre/Pperiph power ratio amounts to 10% (5% with COMAC) for
GEN-III PWR core, and up to 24% in the Gen-III MOX benchmark (12%with COMAC). Our powerful deterministic calculation scheme allowed
the decomposition of the LWR parameter uncertainty in isotopic
partial cross-section components. The main contributor to the power
uncertainty is the 238U inelastic scattering: 3% and 5%, respectively
for COMAC and BOLNA, in GEN-III central assemblies. However, 16O
and 1H elastic scattering cross-sections should be also validated andimproved in order to reduce their individual 1%-2% current contribution
to the central power uncertainty.

9:20 AMAre Modeling Uncertainties Properly Considered in
Neutronics Data Assimilation Analysis?
P.Athe, H.Abdel-Khalik
North Carolina State University, Raleigh, USA
Data assimilation employs the body of available experimental data
to reduce nuclear data uncertainties. The assimilation algorithm
is based on the assumption that nuclear data constitute the major
source of simulation uncertainty, implying that modeling uncertainties
are negligible. This manuscript questions this basic assumption andproposes a further extension of the assimilation algorithm to allow
one to differentiate between modeling and nuclear data-introduced
uncertainties. This is particularly important when the objective of theanalysis is to improve predictions at reactor conditions that are notcovered by the available body of experimental data, e.g., predictionof hot reactor conditions based on the assimilation of cold benchmark
experiments. We employ representative BWR pin cell models todemonstrate the challenges of the current assimilation algorithm, andcompare its performance to the proposed algorithm.

99

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Mizuho_B

Track4-6 Verification, Validation and Uncertainty Analysis
Session Chair: Laurent Chabert(AREVA TA), Kenji Nishihara(JAEA)
9:40 AM

Sensitivity Analysis via Reduced Order Adjoint
MethodY.Bang(1), H.S.Abdel-Khalik(2)1)FNC Technology Ltd., Yongin, South Korea, 2)North Carolina State
University, NC, USA
Notwithstanding the voluminous literature on adjoint sensitivity
analysis, it has been generally dismissed by practitioners as
cumbersome with limited value in realistic engineering models. This
perception reflects two limitations about adjoint sensitivity analysis:a) its most effective application is limited to calculation of first-ordervariations; when higher order derivatives are required, it quickly
becomes computationally inefficient; and b) the number of adjoint
model evaluations depends on the number of responses, which
renders it ineffective for multi-physics model where entire distributions,such as flux and power distribution, are often transferred between
the various physics models. To overcome these challenges, thismanuscript employs recent advances in reduced order modelingto re-cast the adjoint model equations into a form that renders its
application to real reactor models practical. Past work applied reducedorder modeling techniques to render reduction for general nonlinear
high dimensional models by identifying mathematical subspaces,
called active subspaces, that capture all dominant features of themodel, including both linear and nonlinear variations. We demonstrate
the application of these techniques to the calculation of first-orderderivatives, or as commonly known sensitivity coefficients, fora fuel assembly model with many responses. We show that thecomputational cost becomes dependent on the physics model itself,
via the so-called rank of the active subspace, rather than the numberof responses or parameters.

100

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Hiei

Track7-2 Reactor Concepts and Designs
Session Chair: Jan L. Kloosterman(Delft University of Technology), Deokjung Lee(UNIST)
8:00 AMSmAHTR-CTC Neutronic Design
D.Ilas, D.E.Holcomb, J.C.Gehin
Oak Ridge National Laboratory, Oak Ridge, USA
Building on prior experience for the 2010 initial Small Advanced HighTemperature Reactor (SmAHTR) neutronic design and the 2012
neutronic design for the Advanced High-Temperature Reactor (AHTR),
this paper presents the main results of the neutronic design effort for
the newly repurposed SmAHTR-CTC reactor concept. The results are
obtained based on full-core simulations performed with SCALE 6.1.
The dimensionality of the SmAHTR design space is reduced by using
constraints originating in material fabricability, fuel licensing, molten
salt chemistry, and thermal-hydraulic and mechanical considerations.
The new design represents in many regards a substantial
improvement from the neutronic performance standpoint over the2010 SmAHTR concept. Among other results, it is shown that a fuelcycle length of over two years or discharged fuel burnup of 40 gigawatt-day per metric ton of initial heavy metal are possible with an 8%
fuel enrichment in a once-through fuel cycle, while eight-year oncethrough fuel cycle lengths are possible at higher fuel enrichments.

8:20 AM

Suppression of Excess Reactivity of Small Long-Life
Prismatic HTGR with Passive Decay-Heat Removal
O.Sambuu, T.Obara
Tokyo Institute of Technology, Tokyo, Japan
Since the accident at Fukushima Daiichi Nuclear Power Plant in 2011,
design concepts for nuclear reactors have been reconsidered withmuch greater emphasis placed upon passive systems for decay-heatremoval. The present study focuses on the passive safety featuresof decay-heat removal in HTGRs. Generally, the feasibility of suchfeatures of HTGRs depends on reactor design parameters such asreactor power, maximum power density, initial temperature of coreand core size. In a previous work, conditions for design parameters
for small, prismatic HTGRs with passive decay-heat removal to be
operated continuously for only one-year without temporarily shutting
down for regular safety inspection were obtained by residual-heattransfer calculation using equations for fundamental heat transfer
mechanisms. In the present study, analogue conditions for long-termoperation reactors were obtained, and comparisons of the resultswith those previous results were discussed. The appropriate size ofreactor core for a 100 MWt reactor having a maximum power densityof 3 W/cm3 and initial core temperature of 1123 K were obtained.
Consequently, criticality and burnup analyses for the proposed reactor
with 20 wt% uranium enrichment were performed to confirm thepossibility of designing a long-life core with a core size and reactor
power which meets the condition of removing decay heat successfully.
Optimizations for core burn-up with uniformly distributed burnablepoison (BP) particles with various diameters and concentrations werecarried out; the objective in these optimizations was to suppress theexcess reactivity at the beginning of the cycle and thus flatten theburnup reactivity.

8:40 PM

Conceptual Design of a Self-Sustanable Pressurized
Water Reactor with Boiling Channels
M.Margulis(1), E.Shwageraus(2)1)Ben Gurion University, Beer Sheva, Israel, 2)Cambridge University,
Cambridge, UK
Parametric studies have been performed on a seed-blanket Th-U233 fuelconfiguration in a pressurized water reactor (PWR) with boiling channels
to achieve high conversion ratio. Previous studies on seed-blanket
concepts required substantial reduction of the core power density in orderto operate under nominal PWR system conditions. Boiling flow regime inthe seed area allows better heat removal, which in turn, may potentiallyallow increasing the power density of the core. In addition, the reduced
moderation improves the breeding performance. A 2-dimensional design
optimization study was carried out with BOXER and SERPENT codes
in order to determine the most attractive fuel assembly configurationthat would ensure breeding. Effects of various parameters, such as void
fraction, blanket fuel form, number of seed pins and their dimensions, on
the conversion ratio were examined. The obtained results, for which thepower density was set to 104 W/cc, created a map of designs with theircorresponding fissile inventory ratio (FIR) values. It was found that several

options have the potential to achieve the main objective a self-sustainable
Thorium fuel cycle in PWRs without significant reduction in the core
power density.

9:00 AM

Application of the BigT Burnable Absorber to a
Soluble Boron-Free PWR Core
H.Yu, M.Yahya, Y.Kim
KAIST, Daejeon, Korea
A new burnable absorber concept named “Burnable absorberIntegrated control rod Guide Thimble” (BigT) was recently proposed
for the Pressurized Water Reactor (PWR). This paper presents theapplication of the BigT burnable absorber to a soluble boron-free (SBF)
Korean PWR core design. Preliminary lattice calculations based on
the PLUS7 fuel assembly installed with three types of BigT burnable
absorbers were performed to characterize the BigT with gadolinium
and boron carbide burnable absorbers. A 3D SBF OPR1000 coreloaded with the BigT absorbers was subsequently modeled anddepleted to find its equilibrium cycle corresponding to a 3-batch
fuel management. In the current configurations, both the BigT andconventional erbia-urania burnable absorbers were simultaneously
used to control the core excess reactivity. Preliminary results indicate
the promising potentials of the BigT absorbers to realize a SBF PWR
as the core reactivity was shown to be properly managed. All neutronic
calculations were completed by using the continuous energy Monte
Carlo SERPENT code with ENDF/B-VII.0 library.

9:20 AM

Axially Homogeneous Thorium Fuel Designs forTransuranic Burning in Reduced-Moderation BWRs
B.A.Lindley, G.T.Parks
University of Cambridge, Cambridge, UK
Reduced-moderation Boiling Water Reactors (RBWRs) can allow
sustained burning of transuranics (TRU), such that full actinide recycle canbe achieved. However, the void coefficient (VC) tends to become positive
with reduced moderation and high TRU loading, which can severely limit
the design feasibility and performance. This motivates use of thorium (Th)as an alternative feed to uranium (U), as this tends to result in a morenegative VC. Using coupled neutronic-thermal-hydraulic core calculations,Th-fuelled RBWRs are shown to be capable of achieving high discharge
burn-ups, while incinerating an external supply of TRU. This gives themsome flexibility to incinerate TRU with different isotope vectors. Spatial
separation of Th-TRU and Th-U3 into regions of the order of a few
thermal neutron diffusion lengths greatly improves neutronic performance.
This can be accomplished radially or axially, but radial separation resultsin significantly easier fuel fabrication. A companion paper considers axiallyheterogeneous fuel designs, which are not found to deliver improved
performance. A radially heterogeneous RBWR core design, can achieve
135 kg/GWthyr waste incineration rate, and at least 70–80 GWd/t
discharge burn-up, depending on the TRU cooling time, and whether asingle pass through a conventional PWR is used as an initial step before
full recycle in an RBWR. Alternatively, a micro-heterogeneous assembly
design with Th-Pu and Th-U3-MA pins can reduce fuel fabrication costs,
although this limits the TRU incineration rate to ~100 kg/GWthyr. ~200 cm
and ~120 cm core heights are considered, with the former being generally
preferred due to the possibility to match both the rating and pressure
vessel size of an ABWR.

9:40 AM

Neutronic Analysis of a Micro Modular Reactor
F.Venneri(1), C.Jo, J.Chang(2), A.Hawari(3)1)USNC, Los Alamos, USA, 2)KAERI, Daejeon, Korea, 3)NC State
University, Raleigh, USA
Small Modular Reactor of few mega watts range meets power needat remote area when it can be “factory manufactured” and extremely
safe. We have developed a concept of Micro Modular Reactor basedon the FCM compact fuel which has superior safety characteristics. We
have made a preliminary neutronic analysis to confirm the long life time
which is essential for remote application, and to check the controllability
based on the criticality analysis. A core of 200 cm diameter and 200 cm
or 300 cm long cylinder shape can achieve the core life of 5 ~ 10 years
with 20 ~ 10% LEU. When the rated power is reduced by design, theservice period can be extended to 20 ~ 40 years.

101PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Atago

Track6-2 Reactor Physics Experiments
Session Chair: Jacques Di Salvo(CEA), Kenichi Yoshioka(Toshiba)
8:00 AMAn Improved Feynman-α Correlation Analysis with aMoving-Bunching Technique
R.Okuda, A.Sakon, S.Hohara, W.Sugiyama, K.Hashimoto
lnterdisciplinary Graduate School of Science and Engineering, Kinki
University, Osaka, JapanThe bunching technique has been widely utilized in Feynman-α
neutron correlation analysis to synthesize neutron counts of longer
gate widths by bunching neutron counts stored in adjacent MCS
channels. An alternative technique referred to as “moving-bunching
technique” was proposed to reduce a statistical scatter of variance-tomean ratio of neutron counts. The conventional bunching technique
has no overlap of adjacent bunches, while the present technique has
a much longer overlap to filter white noise similarly to the movingaverage technique. A Feynman-α experiment was performed in the
UTR-KINKI, to confirm the applicability and advantage of the proposed
bunching technique. When a neutron detector was placed far from thecore, a Feynman-α analysis with the conventional bunching technique
led to a scattered variance-to-mean ratio from which prompt-neutron
decay constant was never determinable. However, another analysis
with the proposed moving-bunching technique gave a successful
result even for such a remote detector. For a neutron detector close
to the core, the proposed technique resulted in a slight reduction of
statistical error of the decay constant.

8:40 PM

An Alternative Source Jerk Method Implementation
for the Subcriticality Estimation of the VENUS-F
Subcritical Core in the FREYA Project
A.Kochetkov, P.Baeten, W.Uyttenhove, G.Vittiglio, J.Wagemans(1),A.Billebaud, S.Chabod(2), J.-L.Lecouey, G.Ban, F.-R.Lecolley,
N.Marie, G.Lehaut(3), X.Doligez(4), F.Mellier(5)1)SCK-CEN, Mol, Belgium, 2)CNRS-IN2P3, Grenoble Cedex, France,
3)LPC Caen, ENSICAEN/Université de Caen/CNRS-IN2P3, Caen,
France, 4)IPNO, CNRS-IN2P3/Université Paris sud, Orsay, France, 5)
CEA/DEN/DER/SPEX/LPE Cadarache, Saint-Paul-lez-Durance, France
As a prolongation of the FP-6 GUINEVERE (Generation of
Uninterrupted Intense NEutron pulses at the lead VEnus REactor)
project at the VENUS-F facility at SCK• CEN, the FP-7 FREYA (Fast
Reactor Experiments for hYbrid Applications) project was launched in
2011. The FREYA project aims at validating the methodology for subcriticality monitoring of a subcritical core as proposed in the MUSE
experiments conclusions. This is needed for the reactivity monitoringof a powerful ADS (Accelerator Driven System) like MYRRHA.
Experiments are also needed for sub-critical and critical ADS andLFR (Lead Fast Reactor) core characterisation for benchmarking andlicensing purposes. Several levels of sub-criticality of the first subcritical core were measured applying the integral Source Jerk method
and compared to the reference results obtained with the MSM (Modified
Source Multiplication) method.

8:20 AM

Pulsed Neutron and Source Jerk Experiments
for Reactivity Assessment in Deep Subcritical
Configuration: A Case Study within the Framework of
the FREYA Project
G.Mila(1), S.Argirò, S.Beolè, M.Masera(2), S.Di Maria, Y.Romanets,
P.Teles, P.Vaz, (3), M.Osipenko, M.Ripani, P.Saracco(4)1)INFN - Section of Turin, Torino, Italy, 2)INFN - Section of Turin/ University of Turin, Torino, Italy, 3)Universitade de Protecção e
Segurança Radiológica / Universidade Técnica de Lisbona (IST/CTN),
Sacavém, Portugal, 4)INFN - Section of Genova, Genova, Italy
The FREYA project (Fast Reactor Experiment for hYbrid Applications)investigates various aspects connected to the physics and operationof Accelerator Driven Systems (ADS). The experimental facility
consists of the GENEPI-3C deuteron accelerator coupled to the
VENUS-F fast reactor. The study of robust and convenient methods
for online reactivity monitoring in different operating conditions is oneof the primary goals of the FREYA project. A study is presented of two
such methods applied to a deep-subcritical configuration (keff ≈ 0.90)
of the VENUS-F reactor: the area method, which can be performedin Pulsed Neutron Source (PNS) experiments, and the source jerk
technique, suitable in case of a continuous source. Both strategies

102

require, to some extent, input from models or simulation. Theoreticaland experimental uncertainties are discussed.

9:00 AM

Reactivity Measurement of the Lead Fast Subcritical
VENUS-F Reactor Using Beam Interruption
Experiments
T.M.Chevret, J.-L.Lecouey, N.Marie, F.-R.Lecolley, G.Lehaut,
G.Ban(1), A.Billebaud, S.Chabod(2), X.Doligez(3), A.Kochetkov,
P.Baeten, W.Uyttenhove, G.Vittiglio, J.Wagemans(4), F.Mellier(5),V.Bécares, D.Villamarin(6)1)LPC Caen, ENSICAEN/Université de Caen/CNRS-IN2P3, Caen,
France 2)LPSC CNRS-IN2P3/UJF/INPG, Grenoble, France, 3)IPNO,
CNRS-IN2P3/Université Paris Sud, Orsay, France, 4)SCK-CEN, Mol,
Belgium, 5)CEA/DEN/DER/SPEX/LPE Cadarache, Saint-Paul-lezDurance, France, 6)CIEMAT, Madrid, Spain
In order to incinerate minor actinides and thus to reduce the issues
linked to nuclear waste management, Accelerator-Driven Systems(ADS) are being under study. An ADS consists in the coupling of aparticle accelerator with a sub-critical fast reactor. The on-line reactivity
monitoring is a serious issue regarding safety, therefore several
methods to estimate the reactivity of such sub-critical systems have to
be investigated. Here, we present one method based on the study ofthe neutron population evolution during beam interruption experiments
carried out in the framework of the FREYA FP7 program [1,2] at the
GUINEVERE facility, which couples the fast lead sub-critical reactor
VENUS-F with the deuteron accelerator GENEPI-3C at SCK-CEN in
Mol, Belgium. After describing the facility, the analysis based on point
kinetics theory and preliminary results of the reactivity measurements
will be presented. Then, spatial effects that are not taken into accountby point kinetics theory will be highlighted using MCNP simulations,and correction factors to raw results will be calculated. In the end, final
results will be compared to reference reactivity values obtained with
the Modified Source Multiplication (MSM) method.

9:20 AM

Reactivity Measurements at GUINEVERE Facility
Using the Integral kp Method
S.Chabod, A.Billebaud (1) , F.-R.Lecolley, J.-L.Lecouey,
G.Lehaut, N.Marie, G.Ban(2), X.Doligez(3), A.Kochetkov, P.Baeten,
A.Krása, W.Uyttenhove, G.Vittiglio, J.Wageman(4), F.Mellier(5),D.Villamarin(6)1)LPSC CNRS-IN2P3/UJF/INPG, Grenoble, France, 2)LPC Caen,
ENSICAEN/Université de Caen/CNRS-IN2P3, Caen, France, 3)IPNO,
CNRS-IN2P3/Université Paris Sud, Orsay, France, 4)SCK-CEN, Mol,
Belgium, 5)CEA/DEN/DER/SPEX/LPE Cadarache, Saint-Paul-lezDurance, France, 6)CIEMAT, Madrid, Spain
In the framework of the GUINEVERE (EUROTRANS-IP FP6) and
FREYA (FP7) projects, a two-level procedure is developed to measure
the time variations of an ADS reactivity ρ(t). First, the time variations of
the reactor power and neutron source intensity are monitored online,
allowing to extract the fast fluctuations of ρ(t) around a reference
value. Then, absolute measurements of ρ are periodically performed toreadjust the relative measurements. These calibration measurements
require the analysis of the decay of the neutron population triggeredby programmed beam interruptions. In this paper, we describe an
innovative method developed to measure the prompt multiplication
factor, kp, of a subcritical reactor (and therefore its absolute ρ).
Unlike conventional methods used for reactivity measurements, ourapproach only relies on the prompt decay of the neutron population,
therefore enabling the use of shorter beam interruptions (< 50 μs) thanusual (~ 1 ms). After a review of the method’s mechanics, we apply
it to measurements performed at the GUINEVERE facility, with the
VENUS-F reactor, a zero power lead subcritical fast reactor. During
these measurements, the neutron source was operated in pulse
mode. Our results, promising, are compared with reference reactivity
values provided by a MSM analysis.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Atago

Track6-2 Reactor Physics Experiments
Session Chair: Jacques Di Salvo(CEA), Kenichi Yoshioka(Toshiba)
9:40 AM

Estimation of the Delayed Neutron Fraction βeff of the
MAESTRO Core in MINERVE Zero Power Reactor
E.Gilad, O.Rivin, H.Ettegui, I.Yaar(1), B.Geslot, A.Pepino, J.Di
Salvo, A.Gruel, P.Blaise(2)1)Nuclear Research Center NEGEV, Beer-Sheva, Israel, 2)CEA DEN/
CAD/DER/SPEx/LPE, Saint Paul-les-Durance, France
A method for determining the effective delayed neutron fraction βeff
using in-pile reactivity oscillations and Fourier analysis is presented.
This method is based on measurements of the reactor's power
response to small periodic in-pile reactivity perturbations and utilizes
Fourier analysis for reconstruction of the reactor zero power transfer
function.This approach enables the estimation of βeff using multiparameter nonlinear weighted least-squares fit. The method extends
previous work (Yedvab et al. 2006) by accounting for higher harmonics
excitation in the frequency domain by the trapezoidal reactivity signal,
both in the reactivity perturbation and in the reactor power response.
We show that by using this new approach it is possible to obtain
the reactor transfer function in a wide range of frequencies, using
only a single oscillation frequency. This method is applied to a set ofmeasurements of the MAESTRO core configuration in the MINERVE
Zero Power Reactor located at the Cadarache center (CEA). The
derived value of βeff, using this method, is 711±17 (1σ) pcm, which is inagreement with the calculated value of 716 pcm.

103

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Cosmos

Track10-1 Nuclear Data
Session Chair: Pierre Leconte(CEA), Junichi Hori(KURRI)
8:00 AMCOMAC: Nuclear Data Covariance Matrices Libraryfor Reactor Applications
P.Archier, C.De Saint Jean, G.Noguère, O.Litaize, P.Leconte, C.
Bouret
CEA Cadarache, Saint Paul-lez-Durance, France
Uncertainty quantification for new reactor concepts such as the
ASTRID project (sodium-cooled FBR studies at CEA) requires a
precise evaluation of nuclear data covariances. Since 2005, the SPRC
at CEA-Cadarache has developed a covariance matrices library, called
COMAC (COvariance MAtrices from Cadarache, version 0.1), for themost important nuclear data: cross sections, multiplicities (prompt anddelayed neutrons), prompt fission neutron spectrum (PFNS), fission
yields. This paper presents the methodology used to produce thosecovariance matrices, illustrated with representative cases (23Na, 238U,
239
Pu, 240Pu). We have used the CONRAD evaluation code to analyse
simultaneously differential and specific integral experiments and to
generate a coherent and realistic set of covariances. The qualityof those covariances are illustrated with uncertainty propagation
calculations on key parameters for ASTRID.

8:20 AM

Model-Based Generation of Neutron Induced Fission
Yields Up to 20 MeV by the GEF CodeK.Kern, M.Becker, C.Broeders, R.Stieglitz
Karlsruhe Institute of Technology, Leopoldshafen, Germany
Model-based fission product yields from the fission of various
important target nuclides have been calculated for incident neutronenergies up to 20 MeV, divided into a 77 energy group structure. The
calculations have been performed with two versions of the GEF code,
which have been externally coupled to TALYS-1.4. In this application,
the TALYS-1.4 code calculates any pre-fission nucleon or gamma
emission from the compound nucleus as well as the probabilities
of excitation states at the time it undergoes fission. The obtained
quantities, fully characterizing the fissioning nucleus, are then passed
to GEF, which generates the corresponding primary fission productyields in a Monte Carlo calculation. Cumulative fission product yields
have been calculated using these primary fission product yields
together with evaluated radioactive decay data as input. The interim
and final results from the modelling, i. e. cross-sections, independentand cumulative fission yields, have been compared to experimental
data. Important results from this, as well as sensitivities and reliabilitiesof the models, are discussed in this paper. The objective of this work
was to generate energy dependent fission product yields data to serve
as a basis for further investigations on potential improvements ofevaluated data for nuclear reactor applications, which are beyond thescope of this publication.

8:40 PM

Experiments to Improve Uncertainty of the 1st
Delayed Neutron Group Abundance in Fast Fissionsof 238U and Sensitivity Studies of the Relative
Parameters
H.Chung, K.A.Jordan
University of Florida, Florida, USA
The delayed neutron nuclear data are limited by its short lifetime,
limited yield fraction, difficulty to obtain a well-defined sample, aninefficient (uncharacterized) experimental system, and numericalinstabilities of fitting methods. The uncertainties on the relative
abundances of the longer-lived delayed neutron groups are up to
13% for thermal fissions of 235U and 16% for fast fissions of 238U.
To evaluate more accurate delayed neutron nuclear data, a novel
approach to experimentally measure the relative uncertainties on the
delayed neutron group abundances has been proposed by Jordan
and Perret. This approach combines delayed gamma and delayed
neutron fission rate measurement techniques; with two independent
estimates of the same fission rate, higher uncertainty delayed neutron
parameters can be linked to lower uncertainty delayed gamma
parameters. The present authors have designed, optimized, andbuilt an experimental apparatus capable of producing high accuracy
measurements for delayed gamma-rays and neutrons, with the goalof implementing the above approach. Finally, sensitivity studies ofthe parameters that have non-negligible contributions to the overall
uncertainty were carried out. With regards to the designed setup and104

the combined technique, the relative uncertainty of the longest-lived
delayed neutron group is calculated. The calculation shows that theuncertainty of the longest-lived delayed neutron group of 238U can be
reduced by a factor of two to 8.81%.

9:00 AM

Observation of Neutron Thermalization in Graphite
Using the Slowing-Down-Time Technique
A.I.Hawari, B.W.Wehring
North Carolina State University, NC, USA
Experimental measurements and computational simulations were
performed to study the slowing down and thermalization of neutrons
in graphite. The experiment was conducted at the ORELA facilityof Oak Ridge National Laboratory and utilized the slowing-downtime technique to observe the behavior of ORELA neutron pulsesthat were injected into a rectangular reactor-grade graphite pile withdimensions of 70×70×70 cm3. The time dependent reaction rateof neutrons leaking from the pile was recorded in two Li-6 glass
scintillation detectors that were placed on top and in the back of thepile. Simulations of the experiment were performed using the MCNP5Monte Carlo code to perform basic design and to reproduce theexperimental results.To examine neutron thermalization effects, thesimulation invoked thermal neutron scattering, i.e., S(α,β), treatment
for graphite by using the thermal scattering data libraries as givenin the ENDF/B-VII database. The comparison of experimental andcomputational data showed good agreement at early times that
correspond to leaking neutron energies in the epithermal and fast
range. However, noticeable differences were found between themeasured and calculated time spectra as the neutrons reached thethermal energy range. These findings motivate examining the validityof using the ENDF/B-VII graphite S(α,β) libraries to capture accurately
the phenomenon of neutron thermalization in reactor-grade graphite,which is used in various nuclear systems including advanced andGeneration IV nuclear reactors.

9:20 AM

Measurement of Neutron Capture Cross Section of
232
Th in a Low Energy Region
J.Hori, T.Sano, Y.Takahashi, H.Unesaki, K.Nakajima
Research Reactor Institute, Kyoto University, Kumatori, Japan
We have measured the neutron capture cross section of 232Th by
the neutron time-of-flight (TOF) method using a 12.7 m flight pathand a 46-MeV electron linear accelerator (linac) at the Research
Reactor Institute, Kyoto University (KURRI). For the capture crosssection measurement, an assembly of 12 pieces of Bi4Ge3O12 (BGO)
scintillators was used as a total energy absorption detector. The
incident neutron spectrum on the sample was determined with the10
B(n,αγ) standard reaction and the relative capture cross section of232
Th was normalized to the evaluated values of JENDL-4.0 at the
21.8-eV resonance. The present results were compared with theexisting experimental values and the evaluated data in JENDL-4.0.

9:40 AM

New Revisions of Reactor Physics Standards
D.Cokinos
Brookhaven National Laboratory, New York, USA
This paper presents an overview of two of the most basic reactor
physics standards, the newly revised “Steady State Neutronics
Methods for Power Reactor Analysis” and its companion standard,“Nuclear Data for Reactor Design”. These two popular state-ofthe-art standards provide important guidance for developing thenecessary input data needed to calculate reactor lattice and coreparameters such as reaction rates spatial distributions, reactivity
and flux distributions in power reactors for all currently used reactor
types, from fast to thermal reactors. The standards provide guidance
for the selection of cross section data and libraries, the developmentof nuclear data sets suitable for specific applications, energy groupstructures and group collapsing. Key elements in the complex
sequence of power reactor calculations are outlined. The effectsof simplifications and approximations made in the treatment of theneutronic and geometric models and the biases and uncertainties
resulting from such simplifications and assumptions are discussed. In
the newly-revised standard on neutronics methods, clear distinction ismade between the two important tools for assessing the reliability of

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 8:00 AM - Cosmos

Track10-1 Nuclear Data
Session Chair: Pierre Leconte(CEA), Junichi Hori(KURRI)
the results of the calculations, verification and validation. To provide
an auditable path in the verification and validation processes, thestandard requires detailed documentation including methods used,
selection of calculation models and experimental data and results ofhigher order calculations.

105

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 10:20 AM - Mizuho_A

Track1-11 Reactor Analysis Methods
Session Chair: Alain Hebert(Ecole Polytechnique de Montreal), Ren-Tai Chiang(Energy Engineering Service)
10:20 AM

Revisiting the Clio Perturbative Approach for
Analyzing Systems in Fundamental Mode Conditions
A.Hébert
École Polytechnique de Montréal, Montréal, Canada
The Clio perturbative approach is a powerful tool available in the
APOLLO2 lattice code for performing a detailed reactivity analysis
between two heterogeneous systems in fundamental mode conditions.
It is possible to compute group wise and isotope wise reactivity effects
due to a modification in one of the two systems. This modification
can be related to isotopic composition, temperature, cross-section
uncertainties, or even a swap of solution algorithm. The Clio
perturbative approach is an exact method based on the knowledge ofthe adjoint flux of the unperturbed system. Our contribution consists in
a new normalization scheme for distributing the group wise reactivity
effects due to the leakage term of the transport equation in a way thatis consistent to both the direct and adjoint equations.

10:40 AMA Generalization of λ-Mode Xenon Stability Analysis
J.M.Pounders, J.D.Densmore
Bettis Atomic Power Laboratory, PA, USA
A new method for analyzing xenon transients in nuclear reactors basedon λ-mode stability analysis is developed. Like previous work, the new
analysis method is based on a linearization of the coupled diffusion,
iodine, and xenon equations, but the approach is generalized here topermit flux perturbations that include a component in the fundamental
spatial mode and thus a complete expansion of any perturbation. Thisextension also allows power to be explicitly constrained. The typicalconstant-reactivity approach to λ-mode analysis appears as a special
case of the new generalized theory.

11:00 AM

Preliminary Study of the Impact of Xe-135m on the
PCR of CANDU
J.Kim, Y.Kim
KAIST, Daejeon, Korea
In the standard analysis model of the Xenon worth in thermal reactors,
Xe-135m has been neglected due to its short half-life (15.29 min).
The impact of Xe-135m during a transient was briefly analyzed using
point reactor model with the TENDL-2011 and TENDL-2013 data that
it will lead the Xenon worth to be increased. These results are thenapplied to the PCR measurement based on the measured data from
Wolseong2 reactor in order to analyze the potential impacts of Xe135m on the PCR of CANDU reactors. When the Xenon worth isincreased, the measured PCR is generally decreased by 0.54pcm/%Pwith TENDL-2011, and by 0.20pcm/%P with TENDL-2013 on average.

11:20 AM

Verification of the Spectral History Correction Methodwith Fully Coupled Monte-Carlo Code BGCore
Y.Bilodid, E.Fridman(1), M.Margulis(2), D.Kotlyar, E.Shwageraus(3)1)Helmholtz-Zentrum Dresden-Rossendorf, Germany, 2)University ofCambridge, Cambridge, UK, 3)Ben Gurion University of the Negev, Be’er
Sheva, Israel
Recently, a new method for accounting for burnup history effects on
few-group cross sections was developed and implemented in thereactor dynamic code DYN3D. The method relies on the trackingof the local Pu-239 density which serves as an indicator of burnup
spectral history. The validity of the method was demonstrated in PWR
and VVER applications. However, the spectrum variation in BWR
core is more pronounced due to the stronger coolant density change.
Therefore, the purpose of the current work is to further investigate
the applicability of the method to BWR analysis. The proposed
methodology was verified against recently developed BGCore system,
which couples Monte Carlo neutron transport with depletion andthermal hydraulic solvers and thus capable of providing a reference
solution for 3D simulations. The results clearly show that neglecting
the spectral history effects leads to a very large deviation (e.g. 2000
pcm in reactivity) from the reference solution. However, a very good
agreement between DYN3D and BGCore is observed (on the order of
200 pcm in reactivity), when the Pu-correction method is applied.

106

11:40 AM

Core Neutronics Methodologies Applied to the MOXLoaded KAIST 1A Benchmark: Reference to Industrial
Calculations
A.Calloo, H.Leroyer, M.Fliscounakis, D.Couyras
EDF R&D/SINETICS, Clamart, France
EDF R&D is presently developing a new, state-of-the-art calculation
chain called ANDROMEDE including the APOLLO2/JEFF3-based
CEA multigroup library/REL2005 scheme package for assemblycomputations and COCAGNE 3D code for core computations.
The goal of this paper is to validate the calculation chain and its
methodologies on a numerical benchmark of a small PWR which has
been loaded with mixed fuel, KAIST 1A. The latter is challenging,
being highly heterogeneous as it has assemblies with burnablepoison, offers a rodded configuration and includes both UOX-MOX
and core-reflector interfaces. Thus, we will test the capabilities of themodels used in ANDROMEDE to compute such cores. The validation
methodology employed is as follows: stochastic calculations are used
to validate the ability of assembly schemes SHEM-MOC and REL2005
for the computation of 2D full cores. Afterwards, industrial two-groupdiffusion calculations were set up. Reactivity coefficients and pinby-pin power distributions were compared with those obtained fromREL2005. Finally, the last section gives the prospects of the use ofmultigroup SPn for industrial calculations. They raise several questions
such as the energy meshes to be used as well the 2D reflector model
to be applied. A reflector model is set up to test the SPn solver onfull-core calculations with results compared to those of the REL2005
scheme.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 10:20 AM - Mizuho_B

Track4-7 Verification, Validation and Uncertainty Analysis
Session Chair: Maria Pusa(VTT), Tadafumi Sano(KURRI)
10:20 AM

Sensitivity and Uncertainty Analysis on Reactor
Kinetic Parameters Using Perturbation Theory
C.Bouret, L.Buiron, G.Rimpault
CEA Cadarache, France
The analysis of nuclear reactors behavior in transient depends on
the neutronic parameters including the effective delayed neutron
fraction (βeff) and its family dependent components (βeff), theneutron generation lifetime (lp), and the decay constants (λi) for eachdelayed neutron family. With the aim of an accurate assessmentof neutron parameters used to simulate transient, uncertainties onthese parameters must be estimated by considering the contributionsof different nuclear data, including those from the delayed data
(fission yield νd and fission spectrum χd). This paper presents amethodology based on the generalized perturbation theory whichallows the assessment of kinetic parameters sensitivities to nuclear
data. The methodology is applied to the heterogeneous core
design "CFV" considered for sodium-cooled reactor ASTRID. The
contributions of indirect term and direct terms (σfission ν and νd, χ andχd) are decomposed by isotopes and reactions and uncertainties
on the kinetic parameters are calculated using the most recent
data covariance suggested in COMAC (COvariance Matrices from
Cadarache, version 0.1).

10:40 AM

Selecting Benchmarks for Reactor Calculations
E.Alhassan, H.Sjöstrand.J.Duan, P.Helgesson, S.Pomp,
M.Österlund(1), D.Rochman, A.Koning(2)1)Division of Applied Nuclear Physics, Department of Physics andAstronomy, Uppsala University, Uppsala, Sweden, 2)NRG, Petten, The
Netherlands
Criticality, reactor physics, fusion and shielding benchmarks are
expected to play important roles in GENIV design, safety analysis andin the validation of analytical tools used to design these reactors. For
existing reactor technology, benchmarks are used to validate computercodes and test nuclear data libraries. However the selection of thesebenchmarks are usually done by visual inspection which is dependent
on the expertise and the experience of the user and thereby resulting
in a user bias in the process. In this paper we present a method for theselection of these benchmarks for reactor applications and uncertainty
reduction based on Total Monte Carlo (TMC) method. Similarities
between an application case and one or several benchmarks arequantified using the correlation coefficient. Based on the method, we
also propose two approaches for reducing nuclear data uncertainty
using integral benchmark experiments as an additional constraint
in the TMC method: a binary accept/reject method and a method of
uncertainty reduction using weights. Finally, the methods were applied
to a full Lead Fast Reactor core and a set of criticality benchmarks.

11:00 AMDICE 2013: New Capabilites and Data
I.Hill, J.Gulliford, N.Soppera, M.BossantOECD Nuclear Energy Agency, Issy-les-Moulineaux, France
The validation of the codes and nuclear data libraries for reactor
applications relies widely on the International Handbook of EvaluatedCriticality Safety Benchmark Experiments. DICE, the Database for
ICSBEP, is distributed with the Handbook to assist users to identify
relevant validation benchmarks. This paper describes and provides
examples of updates made to the DICE software in 2013 thatsignificantly improve the tool. Data available to handbook users vastly
increased between the 2012 and the 2013 edition, as the numberof sensitivity data files distributed grew from approximately 700 to
3500, or from 15% to 75% coverage of cases. In conjunction, a new
searching feature for three group sensitivity data was implemented
allowing users to identify the set of benchmarks with the mostsensitivity to a particular nuclear reaction. The new sensitivity data has
been compared to existing DICE data, illustrating the agreement ofsensitivity data and neutron balance data within DICE. This is shown tobe an effective tool for both searching evaluations and detecting data
entry errors. Additionally, DICE has implemented a trend plot feature tocreate plots similar to those appearing in validation publications using
ICSBEP data. Users can incorporate their own data into these plots, tocompare different codes or nuclear data libraries. Finally, DICE is now
publically available as a web-start via the NEA website.

11:20 AMOECD NEA Benchmark Database of Spent Nuclear
Fuel Isotopic Compositions for World Reactor
Designs
I.C.Gauld(1), N.Sly(2), F.Michel-Sendis(3)1)Oak Ridge National Laboratory, Tennessee, USA, 2)University ofTennessee, Tennessee, USA, 3)OECD NEA Data Bank, Issy-LesMoulineaux, France
Experimental data on the isotopic concentrations in irradiatednuclear fuel represent one of the primary methods for validating
computational methods and nuclear data used for reactor and spent
fuel depletion simulations that support nuclear fuel cycle safety
and safeguards programs. Measurement data have previously not
been available to users in a centralized or searchable format, andthe majority of accessible information has been, for the most part,
limited to light-water-reactor designs. This paper describes a recent
initiative to compile spent fuel benchmark data for additional reactor
designs used throughout the world that can be used to validate
computer model simulations that support nuclear energy and nuclear
safeguards missions. Experimental benchmark data have been
expanded to include VVER-440, VVER-1000, RBMK, graphitemoderated MAGNOX, gas-cooled AGR, and several heavy-watermoderated CANDU reactor designs. Additional experimental data forpressurized light water and boiling water reactor fuels have also been
compiled for modern assembly designs and more extensive isotopic
measurements. These data are being compiled and uploaded to a
recently revised structured and searchable database, SFCOMPO, toprovide the nuclear analysis community with a centrally accessible
resource of spent fuel compositions that can be used to benchmark
computer codes, models, and nuclear data. The current version of
SFCOMPO contains data for eight reactor designs, 20 fuel assembly
designs, more than 550 spent fuel samples, and measured isotopic
data for about 80 nuclides.

11:40 AM

Effective Physics-Based Uncertainty Quantification
for ZrHx Thermal Neutron Scattering in TRIGA
Reactors
W.Zheng, R.G.McClarren
Texas A&M University, Texas, USA
The thermal neutron scattering cross sections of ZrHx are heavily
affected by the solid frequency distributions, also called “phonon
spectra”, of Zr and H in ZrHx. Although the phonon spectra vary for
different x in ZrHx, current reference data, e.g. ENDF, are based on
ZrH2. This may introduce non-negligible errors in the simulations
for TRIGA reactors. In the previous work, we have proposedparameterized phonon spectrum (PPS) models to explore the effects
of changing the spectra by varying the parameters and investigatedthe effects on reactivity and fuel fission rate density on TRIGA latticemodel.
In this work, we extend the analyses to quantities of interest (QOIs)
on the realistic full-core geometry of the TRIGA reactor at Texas
A&M University. In this work, we sampled the parameters with Latin
Hypercube sampling designs (LHS) in a novel way and generated
corresponding phonon spectra. NJOY and MCNP were used to carryout the calculations. We investigated reactivity (ρ), neutron mean
generation time (Λ), fuel temperature feedback coefficient αTFuel(293.6 600 K) and ex-core/in-core detector reaction rates. Statistical analyses
indicate ρ, Λ and αTFuel are sensitive to the parameters while other QOIs
are not.
We calibrated the parameters for ENDF-VII as a surrogate forexperimental data. Results show the feasibility of the parameter
calibrations. Future work will perform experiments to archive QOIs
and to calibrate the parameters in the PPS model to generate thermal
scattering cross sections used in future simulations.

107

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 10:20 AM - Hiei

Track8-2 Reactor Operation and Safety
Session Chair: Fausto Franceschini(WH), Naoyuki Nakadozono(Hitachi)
10:20 AM

Advanced Surveillance of Resistance Temperature
Detectors in Nuclear Power Plants
C.Montalvo, A.G.Berrocal(1), J.A.Bermejo(2), C.Queral(3)1)Applied Physics to Natural Resources Department. Technical
University of Madrid (UPM), Madrid, Spain, 2)Nuclear Generation
Division. Technical University of Madrid (UPM), Madrid, Spain, 3)
Department of Energy Systems. Technical University of Madrid (UPM),
Madrid, Spain
The dynamic response of several RTDs located at the cold leg ofa PWR has been studied. A theoretical model for the heat transfer
between the RTDs and the surrounding fluid is derived. It proposesa two real poles transfer function. By means of noise analysis
techniques in the time domain (autoregressive models) and the
Dynamic Data System methodology, the two time constants of thesystem can be found. A Monte Carlo simulation is performed in orderto choose the proper sampling time to obtain both constants. The twopoles are found and they permit an advance in situ surveillance of thesensor response time and the sensor dynamics performance. One ofthe poles is related to the inner dynamics whereas the other one islinked to the process and the inner dynamics. So surveillance on theprocess and on the inner dynamics can be distinguished.

10:40 AM

Refined Method for Surveillance and Diagnostics of
the Core Barrel Vibrations of the Ringhals PWRs
I.Pazsit(1), H.Nylen(1,2), C.Montalvo Martin(3)1)Chalmers University of Technology, Gothenburg, Sweden, 2)Ringhals
AB, Väröbacka, Sweden, 3)Technical University of Madrid (UPM),
Madrid, Spain
Surveillance and diagnostics of core barrel vibrations has been
performed in the Swedish Ringhals PWRs for several years, with main
focus on the pendular motion (beam mode). The monitoring of thebeam mode showed that the amplitude of the corresponding peak inthe ex-core neutron spectra increases along the cycle, and decreases
after refueling. Previous investigations on the reason of this behaviour,
i.e. whether it is due to the increase of the core barrel vibration
amplitude or to the increase of the neutron physics coupling between
vibrations and neutron noise, were not decisive. The objective of thework reported here is to clarify this question. From frequency analysis,
two modes of vibration have been identified in the frequency rangeof the beam mode. Several results coming from the trend analysis
performed during recent years indicate that one of the modes is dueto the core barrel motion itself and remains constant during cycle,
and the other is due to the individual flow induced vibrations of thefuel elements, showing an increasing trend during the cycle. In thiswork, the method to separate the contributions from the two modes
has been refined, and the results of this approach to the latest
measurements are presented. The results confirm the origin of thetwo vibration modes and show constant amplitude of the core barrel
motion throughout the cycle.

11:00 AM

Assessment of Flow Induced Vibration Limits in
Preliminary I2S-LWR Fuel Designs
G.E.Sjoden, M.Chin, C.Yi, B.Petrovic
Georgia Institute of Technology, Atlanta, USA
The Integral Inherently Safe Light Water Reactor (I2S-LWR) is a novel
PWR concept being developed by a multi-institutional team, led by
Georgia Tech, under the auspices of the Department of Energy’s
Nuclear Energy University Programs Integrated Research Projects(DOE NEUP IRP). The I2S-LWR aims at delivering an electric power of~1 GW while, simultaneously, achieving an overall level of safety thatis enhanced with respect to GW-class Generation III+ LWRs, including
considerations for “accident tolerant fuels”. The adoption of inherent
safety features and unconventional materials for the main corecomponents are key design features intended to permit the I2S-LWR
to achieve the design objectives. This work summarizes a preliminary
approach to identify Flow Induced Vibration (FIV) limits for new
fuel designs proposed for the I2S-LWR. While not a substitute for a
detailed, system-level Computational Fluid Dynamics (CFD) analysis,
the approach presented here provides a methodology assembled
from “best practices” documented in the literature to establish design
thresholds and enable designs that resist FIV damage. This is an108

essential task to evaluate early in the design process, simply because
the best designs considering other perspectives may inherently faildue to FIV related causes. Because different fuels behave differently,
we identify the essential design considerations and focus on the main
fuel candidates for the I2S-LWR, i.e. UO2 and U3Si2, augmenting thefuel material selection process and the fuel rod design in general.

11:20 AM

Blockage Index for the Detection of Flow Blockage ina Subassembly of Sodium-Cooled Fast Reactor
H.Y.Jeong, M.G.Park(1), S.H.Seong, J.H.Jeong(2)1)Sejong University, Seoul, Korea, 2)Korea Atomic Energy ResearchInstitute, Daejeon, Korea
An early detection of flow blockage in a subassembly is one of theimportant concerns for the safety of a Sodium-cooled Fast Reactor.
We propose a concept of blockage index based on the temperaturechange at measurement point for each subassembly and combining
it with a kernel function for the relative position of subassemblies.
The effectiveness of the proposed method as a reactor protection
parameter is evaluated and compared with the existing methods. We
see the possibility of increasing the reliability of blockage detection
and also of detecting the occurrence of blockage at an early stage if
we use the blockage index.

11:40 AM

Second Generation Shielding Assemblies - Neutron
Flux Impact on Reactor Pressure Vessel and Core
DesignK.H.Bejmer, J.Loberg(1), U.Sandberg(2)1)Vattenfall AB, Stockholm, Sweden, 2)Ringhals AB, Väröbacka,
SwedenIn order to increase the protection of the reactor pressure vessel
beltline weld against fast neutron flux a second generation shielding
assemblies are planed to be designed and installed in the nearest
future at Ringhals 3 and 4. These assemblies might be designed withthree axial zones. The top and bottom zones contain low enriched
235
U while the central shielding part contains only steel rods. Different
designs with varying 235U enrichment and length of the central zone isstudied in order to find a balance between the desired shielding factor
at the beltline weld and core design feasibility. The most favorable
design was found to have 7 steel nodes (106.68 cm), 5 bottom nodes(76.20 cm) and 12 top nodes with standard fuel rods containing 1.5 w/
o 235U. This design results in an estimated shielding factor of 7.8 ± 0.7
and has small impact on cycle length and core design. The reduced
number of fuel pins due to the shielding steel zone will result in a F∆H
penalty of 2.2 %.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 10:20 AM - Atago

Track6-3 Reactor Physics Experiments
Session Chair: Nicholas Brown(BNL), Hironobu Unesaki(KURRI)
10:20 AMBNL Metal Fuel Lattice Experiments: Candidates for
Reactor Physics Benchmark Evaluation
N.R.Brown, A.Aronson, M.Todosow
Brookhaven National Laboratory, NY, USA
The paper presents an argument that some legacy experiments
conducted by Brookhaven National Laboratory and Westinghouse
Atomic Power Division may be interesting candidates for evaluationas part of the OECD/NEA International Reactor Physics Experiment
Evaluation Project. Material buckling measurements from several
experiments with water-moderated lattices of uranium metal rods (cladin Al) are compared with scoping calculations from two well-known
lattice tools: the TRITON/NEWT deterministic code from the SCALE
package and the Serpent Monte Carlo code. In the case of Serpentthe measured material buckling agrees well with the calculations, withthe exception of cases where both the moderator-to-fuel ratio is high(4.0) and the fuel rod diameter is large (0.600”). The SCALE results
show a small systematic bias when compared with the data from theexperiments. Measured quantities that could be considered as partof a future benchmark evaluation include: criticality, material buckling,
migration area, four factor formula parameters, intercell and intracell
flux traverses, and kinetics parameters.

10:40 AM

Benchmark Evaluation of the Neutron Radiography(NRAD) Reactor Upgraded LEU-Fuel Core
J.BessIdaho National Laboratory, Idaho, USA
Benchmark models were developed to evaluate the cold-critical startup measurements performed during the fresh core reload of the
Neutron Radiography (NRAD) reactor with Low Enriched Uranium(LEU) fuel. The final upgraded core configuration with 64 fuel elements
has been completed. Evaluated benchmark measurement data
include criticality, control-rod worth measurements, shutdown margin,and excess reactivity. Dominant uncertainties in keff include themanganese content and impurities contained within the stainless steel
cladding of the fuel and the 236U and erbium poison content in the fuelmatrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data areapproximately 1.4% greater than the benchmark model eigenvalue,
supporting contemporary research regarding errors in the crosssection data necessary to simulate TRIGA-type reactors. Uncertainties
in reactivity effects measurements are estimated to be ~10% withcalculations in agreement with benchmark experiment values within
2σ. The completed benchmark evaluation details are available inthe 2014 edition of the International Handbook of Evaluated Reactor
Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD
LEU cores containing 56, 60, and 62 fuel elements have also been
completed, including analysis of their respective reactivity effects
measurements; they are also available in the IRPhEP Handbook but
will not be included in this summary paper.

11:00 AMCALIBAN and Godiva-IV Measurements Using
Helium-3 Detector Systems
J.Hutchinson, A.Sood, M.Smith-Nelson, J.Goda, J.Bounds,
W.Myers, T.Cutler, B.Richard, B.Rooney(1), A.Chapelle, P.Casoli,
N.Authier(2)1)LANL, NM, USA, 2)CEA, France
CALIBAN and Godiva-IV are both fast burst reactors with HEU cores.
A description and comparison of the two assemblies is presented.
Joint LANL and CEA measurements were conducted to measureconfigurations on the CALIBAN and Godiva assemblies using He-3
detector systems. Measured configurations included many subcriticalconfigurations at various reactivity states as well as configurationsat and above delayed critical. The subcritical and delayed criticalmeasurements were analyzed using the Hage-Cifarelli formalism of
the Feynman Variance-to-Mean method. This analysis can be used
to calculate the multiplication of each configuration, from which themultiplication factor and reactivity are determined. These results
are compared to the reactivity states based upon control rod worth
curves and MCNP simulations. For each delayed critical configuration
an estimate of the delayed neutron fraction is also calculated. Traditionally subcritical neutron noise methods have not been used on
systems at, near, or above delayed critical. This work discusses theuse and problems encountered of such methods in this regime. Countrate measurements from the positive period configurations were also
very accurate but correlated neutron analysis for these configurations
was not successful. Very few data sets which can be analyzed using
neutron multiplicity analysis exist over such a large range of reactivity
values (-$20 to +$0.14).

11:20 AMPower Spectral Analysis for a Subcritical Reactor
System Driven by a Pulsed Spallation Neutron Source
A.Sakon, S.Hohara, W.Sugiyama, K.Hashimoto(1), C.Pyeon,
T.Sano, T.Yagi(2)1)lnterdisciplinary Graduate School of Science and Engineering, Kinki
University, Osaka, Japan, 2)Nuclear Engineering Science Division,
Research Reactor Institute, Kyoto University, Osaka, Japan
A series of power spectral analyses for a thermal subcritical reactor
system driven by a pulsed spallation neutron source was carried out at
Kyoto University Critical Assembly (KUCA), to determine the promptneutron decay constant of the Accelerator-Driven System (ADS). Highenergy protons (100MeV) obtained from the fixed field alternating
gradient accelerator were injected onto a lead-bismuth target,
whereby the spallation neutrons were generated. In the cross-power
spectral density between time-sequence signal data of two neutron
detectors, many delta-function-like peaks at the integral multiple ofpulse repetition frequency could be observed. However, no continuous
reactor-noise component could be measured. This is because thesedetectors have too high count-rate to be placed closely to the core.
From the point data of these delta-function-like peaks, the promptneutron decay constant could be determined. At a slightly subcriticalstate, the decay constant was consistent with that obtained by aprevious power spectral analysis for a pulsed 14MeV neutron sourceand by a pulsed neutron experiment. At another deeply subcriticalstate, however, the present analysis leads to an underestimate of the
decay constant.

11:40 AM

Equivalency of Open Loop and Closed Loop
Reactivity Measurement Techniques
B.A.Baker, G.R.Imel
Idaho State University, Idaho, USA
There is a need for integral physics data on reactivity worth of minor
actinides (transmutation studies) and fission products (burn-up credit)
for validation of differential data. In France, the MINERVE facility has
been used to determine low-worth reactivity measurements using
a closed-loop oscillator technique. However, it has been deemed
unfeasible to perform such measurements in a simulated fast reactor
spectrum in MINERVE as was done in the past. We propose thatreactivity worth measurements could be directly performed in thefast reactor MASURCA using an open loop oscillator technique.
Theoretically, these two methods should be equal in how well ameasurement can be known. This paper compares open loop
techniques (pile oscillator method analyzed with harmonic analysis
and inverse kinetics) with the closed loop technique (reactivity
oscillator method). Open and closed loop techniques are defined in
the classic control sense – closed loop utilizes feedback while open
loop has no imposed feedback. Experiments were performed to show
variations based on frequency and power as well as the determinationof a small worth sample on the order of 0.04 cents with standard
deviations on the order of 0.003 cents. All techniques were shown to
produce results that were limited only by reactor noise.

109

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 10:20 AM - Cosmos

SS4 Advanced Geometry Processing in Deterministic and Monte Carlo
Methods
Session Chair: Hyung Jin Shim(Seoul National Univ.), Shinya Kosaka(MHI)
10:20 AM

11:20 AM

S.G.Hong(1), J.W.Kim, D.H.Kim, Y.O.Lee(2)L.Lu, U.Fischer, Y.Qiu

1)Kyung Hee University, Gyeonggi-do, Korea, 2)Korea Atomic Energy
Research Institute, Daejon, Korea
This paper reviews the computational methods used in the MUST(Multi-group Unstructured geometry SN Transport) code for solving
the multi-group SN transport equation in general geometries anddescribes the status of development of MUST. MUST solves the multigroup transport equation with unstructured tetrahedral meshes formodeling complicated geometrical problems. For tetrahedral meshgeneration, input generation, and output visualization, we developed amanagement program where the mesh generation is based on Gmsh
and TetGen that are open softwares. The geometrical modeling isdone with the commercial CAD softwares such as CATIA. MUST uses
the discontinuous finite element method (DFEM) and two-sub cellbalance methods with linear discontinuous expansion (LDEM-SCB) to
spatially discretize the transport equation. We applied MUST to threeneutron and gamma coupled test problems for testing MUST.

Karlsruhe Institute for Technology (KIT)., Karlsruhe, Germany
McCad is a three-dimension (3D) geometry conversion tool thathas been developed at KIT to enable the bi-directional conversions
between CAD models and the semialgebraic geometry representation
utilized by most Monte Carlo particle transport codes. This paper
introduces the recent improvements of core conversion algorithms,
as well as the developments of new interfaces. The use of McCad forneutronics applications is illustrated on the examples of some fusion
facilities with complicated geometries. The current status of McCad,
including its capabilities and limitations, and the future development
plans are discussed as well.

Development of a Multi-Group SN Transport
Calculation Code with Unstructured Tetrahedral
Meshes

10:40 AM

Development of an Unstructured Mesh Based
Geometry Model in the Serpent 2 Monte Carlo Code
J.Leppanen(1), M.Aufiero(2)1)VTT Technicl Research Centre of Finland, VTT, Finland, 2)Politecnico
di Milano, Milano, Italy
This paper presents a new unstructured mesh based geometry type,developed in the Serpent 2 Monte Carlo code as a by-product ofanother study related to multi-physics applications and coupling to
CFD codes. The new geometry type is intended for the modeling ofcomplicated and irregular objects, which are not easily constructedusing the conventional CSG based approach. The capability is put
to test by modeling the “Stanford Critical Bunny” – a variation of awell-known 3D test case for methods used in the world of computer
graphics. The results show that the geometry routine in Serpent 2
can handle the unstructured mesh, and that the use of delta-tracking
results in a considerable reduction in the overall calculation time as
the geometry is refined. The methodology is still very much underdevelopment, with the final goal of implementing a geometry routine
capable of reading standardized geometry formats used by 3D design
and imaging tools in industry and medical physics.

11:00 AMA CAD Based Automatic Modeling Method for
Primitive Solid Based Monte Carlo Calculation
Geometry
D.Wang(1), S.Yu(2), Q.Gan(1), B.Wu(2), J.Song(3), M.Cheng(2),Y.Yuan(4), P.Long, L.Hu, Y.Wu(1)1)Institute of Nuclear Energy Safety Technology, CAS FDS Team, China,2)Institute of Nuclear Energy Safety Technology, Chinese Academy of
Sciences, Hefei, Anhui, China, 3)University of Science and Technology
of China, Hefei, Anhui, China, 4)Shenzhen University, Shenzhen,
Guangdong, China
The Multi-Physics Coupling Analysis Modeling Program (MCAM),
developed by FDS Team, China, is an advanced modeling tool aiming
to solve the modeling challenges for multi-physics coupling simulation.
The automatic modeling method for SuperMC, the Super Monte Carlo
Calculation Program for Nuclear and Radiation Process, was recently
developed and integrated in MCAM5.2. This method could bi-convert
between CAD model and SuperMC input file. While converting from
CAD model to SuperMC model, the CAD model was decomposed
into several convex solids set, and then corresponding SuperMC
convex basic solids were generated and output. While inverting fromSuperMC model to CAD model, the basic primitive solids was created
and related operation was done to according the SuperMC model.
This method was benchmarked with ITER Benchmark model. The
results showed that the method was correct and effective.

110

Developments and Applications of the Geometry
Conversion Tool McCad for Monte Carlo Particle
Transport Simulations

11:40 AMA Memory Efficient Algorithm for Classifying Unique
Regions in Constructive Solid Geometries
D.Lax, W.Boyd, N.Horelik, B.Forget, K.Smith
Massachusetts Institute of Technology, Massachusetts, USA
Monte Carlo simulations continue to grow in size and fidelity.
Accordingly, with millions of core regions and materials and billions oftallies the need arises for efficient and scalable indexing. This paper
presents a novel method for geometry and material management
based on hierarchical geometries within the Constructive Solid
Geometry (CSG) paradigm. The required algorithms are presented
in detail with a series of performance evaluations. Finally, thememory demands are compared between theoretical and empiricalbenchmarks. The presented indexing scheme will be at the core of
future efforts to overcome the obstacles preventing the routine use of
fully detailed Monte Carlo calculations for nuclear reactor analysis.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Mizuho_A

Track1-12 Reactor Analysis Methods
Session Chair: Dimitrios Cokinos(BNL), Rong-Jiun Sheu(Institute of Nuclear Engineering and Science)
13:30 PMThe Role of the Eigenvalue Separation in Reactor
Dynamics and Neutron Noise Theory
I.Pàzsit, V.Dykin

multiplication factor and reactivity. The paper describes the direct
and adjoint stationary problems and the problems with coefficientsdepending on time. As examples in first part of the paper weconsidered the comparing of measured and calculated values in thesubcritical system consisting of plutonium, and in the second part ofthe paper – the calculated and measured reactivity of VVER-1000
reactor.

Chalmers University of Technology, Göteborg, Sweden
The eigenvalue separation has been used in the past forcharacterisation of the space-time kinetics of reactor transients, andthe stability properties of large loosely coupled cores. In this paper
we explore the role of the eigenvalue separation on the neutronic
response of a critical core to small stochastic perturbations, inparticular the spatial characteristics of the arising neutron noise.

15:10 PM

13:50 PM

M.Tojo, H.Suzuki, H.Sato, T.Iwamoto

Investigation of Conditional Transport Update in
Method of Characteristics Based Coarse Mesh Finite
Difference Transient Calculation
Y.S.Jung, H.G.Joo
Seoul National University, Seoul, Korea
As an effort to achieve efficient yet accurate transport transient
calculations for power reactors, the conditional transport update
scheme in method of characteristics (MOC) based coarse meshfinite difference (CMFD) transient calculation is developed. In thisscheme, the transport calculations serves as an online group constantgenerator for the 3-D CMFD transient calculation and the solution of
3-D transient problem is mainly obtained from the 3-D CMFD transient
calculation. In order to reduce the computational burden of theintensive transport calculation, the transport updates is conditionallyperformed by monitoring change of composition and core condition.
This efficient transient transport method is applied to 3x3 assemblyrod ejection problem to examine the effectiveness and accuracy of theconditional transport calculation scheme.

Development of the Neutron Source Evaluation
Method and Predictor of SRM/SRNM Count Rate in
BWR Simulator
Global Nuclear Fuel-Japan, Yokosuka, JapanThe source range monitors (SRMs) and the start-up range neutron
monitors (SRNMs) are important instruments from the BWR criticality
safety viewpoints. There is a limitation of the minimum count rate (3cps)
to guarantee the normality of the SRMs/SRNMs. After the long outage,
this limitation is critical for the fuel shuffling due to the decay of theneutron sources in the fuel. The neutron source intensity evaluation
method based on a micro burn-up model and the predictor functionof the SRM/SRNM count rate are developed in AETNA01, GNF’s
three dimensional neutronic-thermal hydraulic BWR core simulator.
These new functions are validated through the comparisons between
operating BWR’s measured data after shut-down and during shuffling.
Through these comparisons, high accuracy of the SRM/SRNM countrate predictor of AETNA01 was presented.

14:10 PM

Implementation of an a Priori Time Step Estimator
for the Multigroup Neutron Diffusion Equation in
Asynchronously Coupled RELAP5-3D
M.W.Hackemack(1), J.M.Pounders(2)1)Texas A&M University, TX, USA, 2)Bettis Atomic Laboratory, PA, USA
In this work, an a priori time step estimation scheme based on local
truncation error analysis is developed for the multigroup neutron
diffusion equations. The estimator is used to optimize the time step
size of the neutron diffusion solution, which is asynchronously coupled
to the thermal-hydraulic solution. Numerical results using RELAP5-3D
are shown.

14:30 PM

Development of a Three-Dimensional Kinetics Codefor Commercial-Scale FBR Full Core Analysis
Y.Shimazu, T.Takeda, W.F.G.van Rooijen
University of Fukui, Tsuruga, Japan
A transient analysis code has been developed for the analysis
of future, commercial-size Fast Breeder Reactors. The codeuses diffusion theory, a nodal expansion method for the spatial
discretization, while the temporal discretization uses either a direct
numerical solution, the Improved Quasi-Static method, or the adiabatic
method. Numerical acceleration is used to obtain practical calculation
times on large problems. Trial calculations have shown satisfactorycode performance.

14:50 PMUse of Adjoint Functions for Comparing Measured
and Calculated Parameters in the Subcritical Systems
A.Popykin, S.Shevchenko, R.Shevchenko, N.Zhylmaganbetov(1),V.Kulikov(2)1)SEC NRS, Moscow, Russia, 2)VNIPIET, Saint Petersburg, Russia
The report examines the use of adjoint functions for comparing
the measured and calculated values in subcritical systems. We
will consider the quantities expressed in terms of fractional-linear
functionals of solutions of the corresponding tasks: the effective

111

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Mizuho_B

Track14 Education in Reactor Physics
Session Chair: Ben Forget(MIT), Takanori Kameyama(Tokai Univ.)
13:30 PM

Education Programs for Students and Graduate
Students with Experimental Facilities for Nuclear
Energy in Toshiba
Guest Speaker: K.Hiraiwa
TOSHIBA Corporation, Japan

students also had the possibility to attend the tutorials in a classicalcampus-based teaching set-up. An evalua-tion of the web-basedteaching methods was made possible by the fact that the course was
earlier run in a campus-based format. Despite a limited number ofstudents making difficult to draw any systematic conclusions, a deeper
learning to the course concepts was perceived.

14:50 PM

13:50 PM

Introduction to the Status of Reactor Physics
Education in Tsinghua University

B.Forget, K.Smith

K.Wang, G.Yu, Z.Li, G.Shi

Past, Present and Future of MIT Reactor Physics
Massachusetts Institute of Technology, MA, USA
The reactor physics curriculum at MIT has been undergoing lots ofchanges in recent years. This paper describes the course sequenceand philosophy that we have implemented and continue to develop.
Additionally, thoughts on future direction with respect to modularity andweb-based learning are presented.

14:10 PM

New Practical Exercises at the JSI TRIGA Mark II
Reactor
L.Snoj, S.Rupnik, A.Jazbec
Jožef Stefan Institute, Ljubljana, Slovenia
Since the 1990s the Jožef Stefan Institute (JSI) TRIGA reactor has
been extensively used for performing training in experimental reactor
physics. In 2012 we upgraded some of the existing and introduced
some new exercises. The pulse mode operation exercise was
upgraded by installation of new data acquisition system. The criticalexperiment exercise was improved by adding a new detector inside
the reactor core and changing the data acquisition system. Now wemonitor neutron population with two independent fission chambers on
different locations. In the past the void reactivity coefficient exercise
was performed by inserting Al tube into various positions in the reactor
core and measuring the corresponding reactivity changes. In order tomake the exercise more realistic, we installed a pneumatic system forgenerating air bubbles just below the core. The aim of the exercise is
to measure reactivity changes versus flow rate and air bubble position.
The second new exercise was measurement of water activation.
In this exercise we installed special system which pumps the water
through the core at a constant flow rate to the reactor platform, where
the water activity is measured. The purpose of the exercise is tomeasure the 16N and 19O gamma line intensity and dose rate versus
reactor power. The third new exercise, named in core flux mapping,
was performed by measuring the axial fission rate distribution at
various radial positions in the core. We used CEA – developed mini
fission chambers and a special home developed system for moving
the fission chamber in axial direction and measuring the count rateversus fission chamber position. In the paper the experiments arepresented together with results.

14:30 PM

Developing a Course in Nuclear Reactor Modellingand Going from Campus-Based to Web-Based
Teaching
C.Demazière, K.Jareteg
Chalmers University of Technology, Gothenburg, Sweden
This paper presents the development of a course in nuclear reactor
modelling at Chalmers University of Technology in Gothenburg,
Sweden. This course, part of an international master programin nuclear engineering, deals with the modelling of deterministic
neutron transport, fluid dynamics, and heat transfer in nuclear cores.
The objective of the course is to present the methods applied in
the simulation tools used by the industry for both steady-state andtransient calculations. The originality of the course lies with the factthat all the above fields are tackled in a single course and that all
the fundamental algorithms, together with their approximations andlimitations, are detailed. The paper also reports on a web-basedteaching framework lately adopted for the course. In this configuration,
the course was entirely based on using the web for all lectures,
tutorials, examination, and communication with the teaching staff.
All course materials were accessible on the web to all students at
any time, with the lectures recorded in advance and the tutorials live
broadcasted and also available for on-demand viewing. The on-site

112

Tsinghua University, Beijing, China
This paper introduces the general situation of reactor physics
education in Tsinghua university, including its history, the course
series of the undergraduates and graduates and the reactor physics
experiment education. This paper also simply introduces the students
studying in nuclear engineering field and other nuclear engineering
related courses.

15:10 PM

Reactor Physics Education at Seoul National
University
H.G.Joo, H.J.Shim
Seoul National University, Seoul, Korea
The reactor physics education and research programs of Seoul
National University (SNU), which focus on high fidelity and efficient
reactor simulation and uncertainty evaluation, are presented. In orderto foster the students to have proper knowledge and experience inboth deterministic and probabilistic reactor analysis methods withclear understanding of the physical behaviors of nuclear reactors, theundergraduate and graduate courses cover various mathematicaland numerical methods as well as the principles of nuclear
characteristics and physical behaviors. The research areas spanfrom the development of the methods and computer programs for
direct whole core calculation involving the method of characteristics
transport calculation to the Monte Carlo uncertainty analysis. Those
covers cross section generation, resonance treatment, depletion
method, advanced nodal methods, space-time kinetics method, Monte
Carlo whole core calculation with thermal feedback and et cetera. The
reactor physics curriculum, the contents of the relevant courses, andthe cutting edge research topics and the achievements of SNU reactor
physics education are detailed.

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Hiei

Track7-3 Reactor Concepts and Designs
Session Chair: Liangzhi Cao(Xi'an Jiaotong Univ.), Naoto Aizawa(Tohoku Univ.)
13:30 PM

Development of the 900 Second Specific Impulse
Carbide Low Enriched Uranium Nuclear Thermal
RocketP.F.Venneri, Y.Kim(1), P.Husemeyerr, S.Howe(2)1)KAIST, Daejeon, Korea, 2)CSNR, Idaho, USA
This study presents the Carbide Low Enriched Uranium Nuclear
Thermal Rocket (LEU-NTR) in its latest form. First, a background on
the design methodology is given along with a summary of previous
work on the mass optimization of the core. The issue of the rather
large power peaking is then raised and addressed through theimplementation of radial enrichment zoning and increasing theradial reflector thickness. The advantages and disadvantages ofeach variation are then addressed before the issue of how to raise
the coolant exit temperature is raised. Two methods to do so arethen discussed and combined to produce three configurations of
the Carbide LEU-NTR that operate with a specific impulse of 900
sec. Finally, the new performance characteristics are presented andcompared with previous iterations of the design.

13:50 PM

Molybdenum-99 Production in the Oregon StateTRIGA Reactor: Analysis of Multiple Smaller Core
Designs Using a New LEU Target as Fuel
A.J.Hummel, T.S.Palmer
Oregon State University, Oregon, USA
The most widely used and versatile medical radioisotope today
is technetium-99m. Roughly 30 million people depend on thisradioisotope for diagnostic imaging procedures each year, and thisdemand is expected to grow. Although there are numerous waysof procuring this isotope, the most common is from fission productmolybdenum-99. Mo-99 is produced in all nuclear reactors fueled with
U-235 as a fission fragment with a yield of around 6.1%. Mo-99 has
a half-life of just over 2.5 days, and it will decay to Tc-99m 87% of thetime. The Reduced Enrichment for Research Test Reactors (RERTR)
program was established at Argonne National Laboratory in 1978 toinvestigate technology that would aid in converting High Enriched
Uranium (HEU) facilities to Low Enriched Uranium (LEU) fuel. Since
the majority of all Mo-99 produced currently comes from the irradiation of HEU fuel targets, there has been a growing effort to design LEU
targets that can yield comparable quantities of high Specific Activity(SA) Mo-99. Recently, a novel LEU target design has been developedfor use in TRIGA reactors for production of Mo-99, and this work
examines the reactor behavior of three different core configurations
fueled solely with this new target. MCNP5 was the simulation tool
used to perform this analysis.

14:10 PM

Preliminary Design of the Delft Isotope Production
Reactor (DIPR)
J.L.Kloosterman, M.V.Huisman, M.Rohde
Delft University of Technology, Delft, The Netherlands
The abundant and strongly growing use of Technetium-99m in medical
diagnostics depends on just a few producers and processors of
Molybdenum-99 world-wide, making the molybdenum supply chain
very sensitive to interruptions. New production routes are needed,
especially because some of the reactors used for the production ofthis isotope will soon reach the end of their economic lifetime. In thispaper a preliminary design of a special purpose isotope production
reactor, named DIPR, is presented, which could produce about 8% of
the world-wide demand.
The reactor design is based on aqueous homogeneous reactors
studied in the past and has been evaluated using coupled neutronics
and CFD calculations. Both steady state and transient analyses havebeen carried out, showing the mild behavior of the reactor in various
situations. The consequences of an operation error leading to anincrease of the uranium concentration in the fuel solution has to beinvestigated in greater detail, taking into account a more complete
physics model. Until now, no show stopper has been identified andthe DIPR seems a promising reactor for securing the isotope supply
chain.

14:30 PMA Study of Safety Core Design on Beam Transient for
Accelerator Driven System
N.Aizawa, T.Iwasaki
Tohoku University, Sendai, Japan
Beam transient is the inherent transient on accelerator driven system,and safety analyses on beam transient are performed in the present
study. The safety of the core is evaluated by the cladding temperatureand the stress state on a cladding from the aspects of short term andlong term. Beam shape change and beam incident position change
are analyzed. As the result of safety analyses, the rapid increase of
the load for a cladding is seen especially in beam incident positionchange, and the original core design does not fulfill the criteria
specified in the present study. The safety core design is investigatedin response to the results of safety analyses. The modification of thefuel composition is adopted as the core design approach, and safety
analyses are performed for the designed cores. By the introductionof various fuel compositions, the cladding temperature is reduced byabout 110 degC at most on the normal beam state, and the load fora cladding is relieved to a great degree on the transient conditions.
The core design with the modification of the fuel composition is foundto be able to satisfy all the criteria. However, the requirements for an
accelerator specification are increased, and further development isneeded in the accelerator to assure the safety of the core only by themodification of core design.

14:50 PMNew Inverted Hydride Fuel Design Concept for
Pressure Tube Type Super Critical Water Reactors
A.Ahmad, L.Cao, H.Wu
Xi'an Jiaotong University, Xi'an, China
In this study, an innovative core design having inverted configuration
has been proposed for pressure tube type supercritical water
reactors. In this design the relative positions of fuel and coolant havebeen inverted and U-Th-Zr-hydride fuel has been used. A coupledneutronics and thermal hydraulics analysis was done for the proposed
Inverted Pressure Tube Type (IPTT) SCWR. The neutronics analysis
was carried out by using a 3D fine mesh diffusion theory code andthermal hydraulics calculations were done by using single channel
model. These two codes were coupled with each other by a link code.
The average outlet temperature for the proposed IPTT-SCWR was
found to be 625 degC with maximum clad surface temperature (MCST)
under the design limits i.e. below 850 degC. Moreover a core loading
pattern has also been proposed to achieve uniform radial power
distribution and lower cladding surface temperature.

15:10 PM

Preliminary Safety Analysis of a Thorium HighConversion Pebble Bed Reactor
F.J.Wols, J.L.Kloosterman, D.Lathouwers, T.H.J.J.van der Hagen
Delft University of Technology, Delft, The Netherlands
An inherently safe thorium High-Conversion Pebble Bed Reactor
would combine the inherent safety characteristics of the Pebble
Bed Reactor with the favourable waste characteristics and resource
availability of the thorium fuel cycle. Previous work by the authors
showed that high conversion ratio’s can be achieved within a thorium
Pebble Bed Reactor (PBR) at a practical operating regime. The
thorium PBR core design consists of a cylindrical core with a central
driver zone surrounded by a breeder zone. The breeder pebbles havea 30 g heavy metal (HM) loading to enhance conversion of Th-232 into
U-233, while the driver pebbles (10 w% U-233) contain a lower metalloading to enhance fission.
In previous studies, thorium PBR designs were presented for threecore diameters, using a 7.5 g heavy metal (HM) loading for the driver
pebbles. The current paper investigates the safety of these thorium
PBR designs in terms of reactivity coefficients and possible reactivity
insertion due to water ingress. Early results indicated that the values of
the reactivity coefficients for the three designs with 7.5 g HM loading
per driver pebble were rather small and the possible reactivity insertiondue to water ingress was very large. Therefore, also a lower HM
loading per driver pebble (4 g) was investigated to reduce the impact
of water ingress, since the core becomes less under-moderated.
For the three core diameters investigated, it is shown that reducing themetal loading in the driver pebbles to 4 g is indeed advantageous in

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Hiei

Track7-3 Reactor Concepts and Designs
Session Chair: Liangzhi Cao(Xi'an Jiaotong Univ.), Naoto Aizawa(Tohoku Univ.)
terms of safety, water ingress leads to a smaller reactivity increase but
also the reactivity coefficients become stronger negative. Secondly,
the breeding performance of the cores with a 4 g driver pebble HM
loading improves. On the downside, the driver pebble residence times
become shorter, which could increase fuel reprocessing costs. Fuelpebbles would have to be recycled at an increased rate, which might
be more challenging from a practical perspective.

114

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Atago

Track6-4 Reactor Physics Experiments
Session Chair: Mohamed Ouisloumen(WH), Toru Yamamoto(NRA)
13:30 PM

Overview of the 2014 Edition of the International
Handbook of Evaluated Reactor Physics Benchmark
Experiments (IRPhEP Handbook)
J.D.Bess, J.B.Briggs(1), J.Gulliford, I.Hill(2)1)Idaho National Laboratory, Idaho, USA, 2)OECD NEA, Issy-lesMoulineaux, France
The International Reactor Physics Experiment Evaluation Project(IRPhEP) is a widely recognized world class program. The work of the
IRPhEP is documented in the Interna-tional Handbook of Evaluated
Reactor Physics Benchmark Experiments (IRPhEP Hand-book).
Integral data from the IRPhEP Handbook is used by reactor safetyand design, nuclear data, criticality safety, and analytical methods
development specialists, worldwide, to perform necessary validationsof their calculational techniques. The IRPhEP Hand-book is among
the most frequently quoted reference in the nuclear industry and is
expected to be a valuable resource for future decades.

13:50 PM

Analysis of Tungsten Gray Rods Critical Experiments
Using PARAGON with Ultra-Fine Energy Mesh
Methodology
M.Ouisloumen, H.Huria(1), K.Yoshioka, T.Umano, T.Kikuchi,
S.Gunji, H.Matsumiya, S.Sugahara(2)1)Westinghouse Electric Company LLC, PA, USA, 2)Toshiba Corporation
Power & Industrial Systems R&D Center, Kawasaki, Japan
New critical experiments using gray control rods with tungsten wererecently performed at the Toshiba NCA critical facility. This paper
presents analyses of these experiments using both stochastic anddeterministic codes. We used the continuous energy Monte Carlocode MCNP and the Westinghouse lattice physics code PARAGON.
The basic nuclear data source for the cross-sections is ENDF/B-VII.1.
First, the tungsten data is validated against Monte Carlo calculations.
The ultra-fine energy mesh with 6064 group cross-sections library
was used in PARAGON to extend the validation of the methodology
to the cold temperature conditions. Comparisons focused on reactivity
and the measured fission rate distributions. The results show that the
ENDF/B-VII:1 data adequately reproduces the measured tungsten
gray rod reactivity worth. Also, the energy mesh and the methodology
used in PARAGON are seen to be adequate in predicting thereactivity and fission rate distributions for these challenging and highly
heterogeneous experiments.

14:10 PMMonte Carlo Assessment of Spatial and Energy
Effects in the VENUS-F Subcritical Configurations
and Application for Reactivity Determination
V.Bécares, D.Villamarín, E.M.González-Romero(1), A.Kochetkov,
P.Baeten, W.Uyttenhove, G.Vittiglio, J.Wagemans(2), A.Billebaud,
S.Chabod(3), J.-L.Lecouey, F.-R.Lecolley, G.Lehaut, N.Marie(4),X.Doliguez(5), F.Mellier(6)1)Nuclear Fission Division, CIEMAT, Madrid, Spain, 2)Belgian Nuclear
Research Centre (SCK-CEN), Mol, Belgium, 3)Laboratoire de Physique
Subatomique et de Cosmologie, Grenoble, France, 4)Laboratoire dePhysique Corpusculaire de Caen, ENSICAEN/Université de Caen,
Caen, France, 5)Institut de Physique Nucléaire d’Orsay, Orsay, France, 6)
CEA/DEN/DER/SPEX/LPE Cadarache, Saint-Paul-lez-Durance, France
Experiments with a subcritical core and an external pulsed neutron
source (PNS) have been performed at the VENUS-F reactor of the
SCK-CEN during the year 2012 within the frame of the FREYA projectof the 7th European Framework Programme. Different reactivity levels,
obtained by varying the position of the control rods and the safetyrods, have been investigated, with values of kef f ranging between 0.9
and 0.96. The area-ratio and the prompt decay constant techniques
are well-known techniques to determine the reactivity of the systemfrom the results of PNS experiments. However, both techniques are
based on the point kinetics model and hence they are affected by thespatial and energy effects present in real systems, which will cause
them to produce biased values of the reactivity. In this work, we apply
a methodology based on Monte Carlo simulations to assess the spatialand energy effects present in the VENUS-F core and to determine
relationships between the reactivity and the measured parameters(area-ratio and prompt decay constant) that take into account theseeffects and allow the application of the reactivity determination
techniques to obtain accurate reactivity values beyond the limits of thepoint kinetics model. The results of the validation of this methodology
against the experimental results obtained in the VENUS-F reactor will
also be discussed.

14:30 PM

Validation of ORIGEN2 Coupled with JENDL-4.0 Base
Libraries for Isotopic Compositions of Irradiated Light
Water Reactor Fuels
T.Yamamoto
Japan Nuclear Nuclear Safety Organization, Minato-ku, Tokyo, JapanA set of cross section libraries for ORIGEN2, ORLIBJ40, have beendeveloped based on JENDL-4.0 by the Japan Atomic Energy Agency.
It has been applied to the analysis of measured isotopic composition
data of PWR and BWR irradiated fuels. The number of fuel samples
referred in the present study was 33 both for PWR and BWR UO2
fuels. The fuel samples of the PWR fuels were taken from 15x15,
17x17 and 18x18 fuel assemblies, and those of the BWR fuels from
8x8 and 9x9 fuels. The enrichments of the PWR sample were from 1.7
to 4.5 wt%, those of the BWR fuels from 2.1 to 4.9 wt%. The ranges
of fuel burnups were 14 to 78 GWd/t for the PWR fuels and 17 to 69
GWd/t for the BWR fuels. Nuclides studied were 19 actinides and 42
fission product nuclides. The calculations were performed taking into
account initial fuel compositions and irradiation histories which wereconsistent with the sample burnups determined based on measured
isotopic compositions. The calculation results were compared withthe measured results and the biases (C/E - 1's) were obtained. The
C/E-1's were plotted against sample burrnups and checked about
anomaly and trend with burnups. It was noted that the compositions of
235
U of samples taken from corner rods of a BWR 9x9 fuel assemblies
and those of 237Np of part of samples from a PWR 17x17 assembly
were considerably overestimated. After screening of the data, thebiases were statistically analyzed to obtain the average and standard
deviation for each nuclide.

14:50 PMThe AMMON Experiment in EOLE Facility: A
Challenging Program Dedicated to the Experimental
Validation of JHR Neutronic and Photonic Calculation
Tools
J.Di Salvo, A.Gruel, B.Geslot, P.Blaise, F.Mellier, A.Pepino,
A.Roche, S.Arnaud, K.Blandin, A.Foucras, D.Garnier, J.-F.Ledoux,
C.Morel, J.Bonora, J.-M.Girard, C.Jammes, H.Philibert, N.Thiollay,
J.-F.Villard(1), C.Vaglio-Gaudard, A.-C.Colombier(2)1)CEA, DEN, DER/SPEx, Cadarache, Saint Paul les Durance, France, 2)
CEA, DEN, DER/SPRC, Cadarache, Saint Paul les Durance, France
The new AMMON integral experiment has been dedicated to theanalysis of the Jules Horowitz Reactor (JHR) neutron and photon
physics. The last shutdown of this 3-years program, in the EOLE zero
power experimental reactor at CEA Cadarache, occurred at the endof March 2013. The core configurations consisted of an experimentalzone of 6 or 7 JHR assemblies with U3Si2-Al 27% 235U enrichedcurved fuel plates, surrounded by a driver zone with enough standardPWR UOx fuel pins to reach criticality. The purpose of this paper is tosynthesize the main features of this program. Experimental methods
and their associated uncertainties will be presented, with a focus onsome new results obtained for nuclear heating measurements. Itemphasizes the importance of the measurements performed in criticalmock-ups for the calibration of biases and associated uncertainties forneutron and photon calculation tools.

15:10 PMMonte Carlo Analysis of Reactivity EffectMeasurements in the AMMON Experimental Program
Dedicated to JHR Neutron Studies
C.Vaglio-Gaudard, A.C.Colombier, M.Lemaire(1), J.Di Salvo,
A.Gruel(2)1)CEA, DEN, DER/SPRC, Cadarache, Saint Paul les Durance, France,
2)CEA, DEN, DER/SPEx, Cadarache, Saint Paul les Durance, France
The AMMON integral experiment was loaded in the EOLE criticalmock-up at CEA Cadarache: it aims at providing experimental

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PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Atago

Track6-4 Reactor Physics Experiments
Session Chair: Mohamed Ouisloumen(WH), Toru Yamamoto(NRA)
data for assessing the calculation biases due to nuclear data on
the Jules Horowitz Reactor (JHR) neutron and photon parameters.
The experiment has a very specific design in comparison withthe previous cores (MISTRAL, EPICURE, FUBILA, PERLE, etc.)
consisting in the loading of UOX or MOX fuel pins in EOLE. The
AMMON core contains an experimental zone composed of JHR
U3Si2-Al circular assemblies and a driver zone of several hundredsof UOX 3.7% 235U enriched fuel pins. This experimental program
was very ambitious, since five main configurations – a reference
configuration and 4 perturbed configurations - had to be studied for
2 years, with a lot of different measurement types. The experimental
program was finished last year, in early 2013; all the main foreseenmeasurements were achieved, with a very good control of all theexperimental (measurement+technological) uncertainties. Almost all
the measurements have been today analyzed with TRIPOLI-4® Monte
Carlo reference calculations, using the JEFF3.1.1 nuclear data library.
This paper gives a general overview of the analysis of the criticalstates studied in the AMMON experiment: the comparison between
Calculation and Experiment concerns the core excess reactivity andthe reactivity effect induced by the introduction of perturbations in thereference configuration. The analysis shows a good prediction of thereactivity excess +120 pcm ±300 pcm (at one standard deviation).
The reactivity effects, resulting from perturbations introduced in the
AMMON core – hafnium, beryllium, water hole, isothermal temperature- are also well calculated; in particular, the bias on hafnium reactivity
worth, supposed to be only due to nuclear data, amounts to +1.0%± 1.7% (at one standard deviation). The calculation bias on theisothermal temperature coefficient in the AMMON experimentalzone reaches -0.4 ± 0.8 pcm/°C on the 20°C-35°C range, and -1.5
± 1.0 pcm/°C on the 35°C-50°C range. These results will be usedin the Verification & globale Validation – Uncertainty Quantification(V&V&UQ) process implemented for the JHR neutron and photon
calculation tools.

116

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Cosmos

Track10-2 Nuclear Data
Session Chair: Nicholas Stauff(ANL), Pascal Archier(CEA)
13:30 PM

Impact of the Interpolation Mode on the Secondary
Particle Energies for Shielding and Criticality
Benchmarks with TRIPOLI-4® Monte Carlo Code
C.Jouanne, O.Petit
CEA Saclay, Gif-sur-Yvette, France
The ENDF-6 format used for evaluation files allows different
interpolation modes between distributions related to discrete incident
energies. It is the case for angular distributions, energy distributions,energy-angle distributions and continuous photon energy spectra.
The linear-linear interpolation is very often used particularly in the
JEFF and ENDF/B libraries. This has no major impact when the
distribution covers all the range (e.g. for angular distributions andfor prompt fission neutron spectra), but, in all the other cases it
could induce a violation of kinematic energy conservation. Certain
codes for nuclear data processing (e.g. NJOY) and particle transport(e.g. MCNP) reinterpret data of evaluation files by changing theinterpolation mode, in order to improve compliance with the kinematicsof the physical reactions, even if this approach does not allow theevaluators to subsequently test the quality of their work. A new option
for the TRIPOLI-4 ® code offers the users the opportunity to choose
between the raw evaluation and a more physical treatment of thesedata for shielding and criticality benchmarks using neutron or coupledneutron-photon calculations. Thus, it is now possible to quantify the
impact of such an interpretation, moreover this allows the evaluators
to really test the relevance of their evaluation. The most sensitive
configurations appear to be high energy neutron and coupled neutronphoton calculations.

13:50 PM

Impact of the Differences in Nuclear Data on
Estimated K-Effective of SFR Cores
N.E.Stauff, G.Aliberti, T.K.Kim, T.A.Taiwo
Argonne National Laboratory, Argonne, USA
Neutronic analyses rely on accurate cross-sections from nuclear datalibraries. The differences in the cross-sections from the JEFF-3.1
and ENDF/B-VII.0 nuclear data libraries are evaluated in this paperby estimating their impact on the eigenvalue of Sodium-cooled Fast
Reactors (SFR). This analysis is performed with three SFR corescharacterized by a variety of power ranges, fuel types (uranium andthorium) and fuel forms (oxide and metal). It was observed that thereactivity value is consistently larger when JEFF-3.1 is used insteadof ENDF/B-VII.0 and the discrepancy can be as high as 1,900 pcm.
A perturbation analysis was performed with the ERANOS2.2 code
system to evaluate the contributions by isotope, reaction and energy
range to the observed reactivity changes. It was found that the capture
cross sections of Th-232 and U-233 are both responsible for as highas 700 pcm of the discrepancy in the case of the thorium-fueled core.
Plutonium and sodium are the two main contributors to the observed
reactivity change in the case of uranium-fueled cores. The reactivity
impact due to the differences in the cross-sections from ENDF/
B-VII.0 and JEFF3.1 is also compared to the uncertainties calculated
respectively with the COMMARA-2.0 and the COMAC.V0 covariancematrixes.

14:10 PM

Nuclear Data Sensitivity Analysis for Isotopic
Generation Using JENDL-4.0, ENDF/B-VII.1 andJEFF-3.1.1
Y.Kawamoto, G.Chiba, M.Tsuji, T.Narabayashi
Hokkaido University, Sapporo, Japan
In burn-up calculations, accuracy of isotopic generation prediction
strongly depends on the nuclear data, such as neutron cross sections,fission yields and decay constants. In this study, we perform burnup calculations using JENDL-4.0, ENDF/B-VII.1 and JEFF-3.1.1,
and compare the calculation results with PIE (Post Irradiation
Examination) data. We also perform burn-up sensitivity analysis
based on the generalized perturbation theory to clarify the cause ofdifference on isotopic generation between the libraries. As a result,
there are large discrepancies between JEFF-3.1.1 and the others
generally. Furthermore, we clarify their causes for each nuclide andenergy group. For neutron cross sections, some nuclides have large
discrepancies between JEFF-3.1.1 and the others, and they give large
impacts on specific isotopic generation predictions. On fission yields,

ones from Pu-239 and Pu-241 have large discrepancies between
JEFF-3.1.1 and the others, and they give large impacts on specificisotopic generation predictions, especially Gd-160. Decay constantdiscrepancies do not give any large impacts on isotopic generation
predictions.

14:30 PM

Validation of a Pointwise Energy Neutron Cross
Section Library Generated by RXSP-BETA2.0 Using
ENDF/B-VII.0
J.Yu, W.Li, K.Wang(1), J.Vujic(2)1)Tsinghua University, Beijing, China, 2)University of California at
Berkeley, California, USA
A new pointwise energy neutron cross section library named
ENDFb7_r in ACE format for Reactor Monte Carlo code RMC has
been generated by Reactor Cross Section Processing code RXSP
using ENDF/B-VII.0. The pointwise energy cross section library called
ENDFb7_n generated by NJOY has also been constructed for intercomparison of results. Benchmark tests for series of criticality reactor
cores and assemblies including both uranium and plutonium fuels withthermal, intermediate and fast neutron spectrum have been performed
with the code RMC using these two libraries ENDFb7_r and ENDFb7_
n. The k-effective and neutron flux calculated with two libraries show
very good agreement with each other. Moreover, another practical
PWR fuel assembly depletion model is further constructed andsimulated by RMC. The calculated results of k-effective and isotopic
concentration swings with burnup agree very well with each other.
It has been proved that the self-processed library named ENDFb7_
r is accurate enough to be used for both criticality and depletioncalculations.

14:50 PM

Criticality Experiments and Analyses of Uranium
Zirconium Carbon Nitride LEU Fuel
A.Talamo, Y.Gohar, Z.Zhong (1) , S.N.Sikorin, S.G.Mandzik,
S.A.Polazau, T.K.Hryharovich, I.S.Holubeva(2)1)Argonne National Laboratory, Chicago, IL, USA, 2)Joint Institute forPower and Nuclear Research – Sosny, Minsk, Belarus
MCNP version 6.1 was used to model the Giacint facility of Belarus
with uranium zirconium carbon nitride low enriched uranium (LEU)
fuel. The fuel rods have either stainless steel or niobium clad. Niobium
absorbs less neutrons than stainless steel. The fuel rods are immerged
into a water tank. In the experiments, the water level is changed
according to the fuel loading pattern to obtain critical condition. The
MCNP numerical simulations used nuclear data libraries based on
ENDF/B-VII.1, JEFF-3.1.2, or JENDL-4 and the obtained results werecompared. The analyses included: the effective multiplication factor,
the delayed neutron fraction, the neutron generation time, and thesensitivity coefficients. The ACE format of the JENDL-4 nuclear datalibrary has been generated at Argonne National Laboratory using
NJOY version 2012.

15:10 PM

Feedback on 239Pu and 240Pu Nuclear Data andAssociated Covariances through the CERES Integral
Experiments
P.Leconte, G.Truchet, G.Noguere, E.Privas, P.Archier, C.De SaintJean(1), J.Gulliford(2), D.Hanlon(3)1)CEA Cadarache, Saint Paul-Lez-Durance, France, 2)OECD/NEA,
Issy-les-Moulineaux, France, 3)AMEC, Dorset, UK
Benchmark measurements of irradiated and un-irradiated fuelsamples were performed in the framework of the CERES collaborative
programme between AEA and CEA. These experiments provide
relevant data for the validation of fuel burn-up and criticality-safetycalculations for the whole fuel cycle. As part of this programme, pileoscillation measurements were carried out on a range of mixed oxide
samples with Plutonium of various mass and isotopic contents, bothin the MINERVE and DIMPLE reactors. Four core configurations,
two over-thermalized situations and two PWR-type situations, wereconstituted with different forward and adjoint flux spectra, emphasizing
fission and/or capture contributions. The experiments were analyzed
using reference TRIPOLI4 calculations with the JEFF-3.1.1 library,using exact 3-dimensional models of the core configurations. In a

117

PHYSOR2014 - The Role of Reactor Physics towards a Sustainable Future
Thursday October 2, 2014 - 13:30 PM - Cosmos

Track10-2 Nuclear Data
Session Chair: Nicholas Stauff(ANL), Pascal Archier(CEA)
first step, calculations of each DIMPLE configuration were performed
and compared with the experiment, showing very good agreementswith a maximum C-E of -230 pcm. In the second step, reactivity
worth experiments were analyzed, using recently developed exactperturbation capabilities in TRIPOLI4. A consistent assimilation of
Calculation over Experiment discrepancies was performed with
the CONRAD code, using the Integral Data Assimilation method.
Covariance matrix on multigroup neutron cross sections andmultiplicities were generated and significant trends were identified,
especially on the 239Pu and 240Pu capture cross sections in the thermalenergy range (E < 0.1 eV). Further investigations should be required
to confirm these conclusions, due to the strong dependence of thesetrends and of posterior covariances to prior covariances.

118

PHYSOR 2014 Program/Proceedings CD-ROM
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PHYSOR 2014 Program
7:30

8:00-12:00

September
12:00-13:30
28(Sun.)
13:30-17:35

September
29(Mon.)

Registration OpenWorkshop 5 (Mizuho_A)

Workshop 3 (Mizuho_B)

Workshop 1 (Mizuho_A)

Workshop 4 (Mizuho_B)

Workshop 7 (Mizuho_C)

Workshop 8 (Hiei)

16:00-20:00

Welcome Cocktail (Mizuho_D)

7:30

Registration Open

8:30-8:50

Opening Session (Mizuho_A, B)
Reiko Fujita, President, Atomic Energy Society of Japan
Michaele C. Brady Raap, President, American Nuclear Society

8:50-10:00

Plenary Session 1 (Mizuho_A, B)

10:20-11:30

Plenary Session 2 (Mizuho_A, B)

11:30-13:00

Hosted Lunch (Mizuho_C, D)
Etsuro Saji, Mitsubishi Heavy Industries

13:00-15:30

Track1-1
Reactor Analysis Methods(Mizuho_A)

SS2-1
Reactor Physics and Criticality
Safety Activities in OECD/NEA
Working Party(Mizuho_B)

SS1-1
Molten Salt Reactors(Hiei)

SS5
Track11-1
Multiscale, Multiphysics
Research Reactors and Spallation
Approaches in Nuclear Science
Sourcesand Engineering Applications(Cosmos)(Atago)15:45-18:15

Track1-2
Reactor Analysis Methods(Mizuho_A)

SS2-2
Reactor Physics and Criticality
Safety Activities in OECD/NEA
Working Party(Mizuho_B)

SS1-2
Molten Salt Reactors(Hiei)

SS7
Track11-2
Control Rod Withdrawal Tests
Research Reactors and Spallation
Performed During the PHENIX
SourcesEnd-of-Life Experiments(Cosmos)(Atago)Registration Open

8:00-10:05

Track1-3
Reactor Analysis Methods(Mizuho_A)

Track5-1
Nuclear Criticality Safety(Mizuho_B)

Track2-1
Deterministic Transport Theory(Hiei)

Track3-1
Monte Carlo Methods(Atago)SS8Reactor Physics of Non-Traditional
LWR Fuel Design(Cosmos)

10:20-12:00

Track1-4
Reactor Analysis Methods(Mizuho_A)

Track5-2
Nuclear Criticality Safety(Mizuho_B)

Track2-2
Deterministic Transport Theory(Hiei)

Track13
Radiation Applications andNuclear Safeguards(Atago)Track9-1
Transient and Safety Analysis(Cosmos)

September 12:00-13:30
30(Tue.)

Hosted Lunch (Mizuho_C, D)
Sun-Doo Kim, KEPCO Nuclear Fuel

13:30-15:40

Track1-5
Reactor Analysis Methods(Mizuho_A)

Track4-1
Verification, Validation andUncertainty Analysis(Mizuho_B)

Track2-3
Deterministic Transport Theory(Hiei)

Track3-2Monte Carlo Methods(Atago)Track12-1
Fuel Cycle and Actinide
Management(Cosmos)

15:55-18:05

Track1-6
Reactor Analysis Methods(Mizuho_A)

Track4-2
Verification, Validation andUncertainty Analysis(Mizuho_B)

Track2-4
Deterministic Transport Theory(Hiei)

Track3-3Monte Carlo Methods(Atago)Track12-2
Fuel Cycle and Actinide
Management(Cosmos)

18:30-21:00

Conference Banquet (Mizuho_C, D)
Ronald J. Ellis, Chair, ANS Reactor Physics Division
Toshikazu Takeda, Honorary Chair, Univ. of Fukui

7:30

Registration Open

8:00-10:05

Track1-7
Reactor Analysis Methods(Mizuho_A)

Track4-3
Verification, Validation andUncertainty Analysis(Mizuho_B)

SS3
Hybrid Particle Transport Methods
for Solving Complex Problems in
Real-Time(Hiei)

10:20-12:00

Track1-8
Reactor Analysis Methods(Mizuho_A)

Track4-4
Verification, Validation andUncertainty Analysis(Mizuho_B)

Track8-1
Reactor Operation and Safety(Hiei)

12:00-13:30
13:30-15:40

Track3-4Monte Carlo Methods(Atago)Track1-9
Reactor Analysis Methods(Mizuho_A)

Track4-5
Verification, Validation andUncertainty Analysis(Mizuho_B)

Track6-1
Track9-3
Reactor Physics Experiments Transient and Safety Analysis(Atago)(Cosmos)

Track7-1
Reactor Concepts and Designs(Hiei)

Track3-5Monte Carlo Methods(Atago)Registration Open

8:00-10:05

Track1-10
Reactor Analysis Methods(Mizuho_A)

Track4-6
Verification, Validation andUncertainty Analysis(Mizuho_B)

Track6-2
Track7-2
Reactor Concepts and Designs Reactor Physics Experiments(Hiei)(Atago)10:20-12:00

Track1-11
Reactor Analysis Methods(Mizuho_A)

Track4-7
Verification, Validation andUncertainty Analysis(Mizuho_B)

Track8-2
Reactor Operation and Safety(Hiei)

Track1-12
Reactor Analysis Methods(Mizuho_A)

Track14
Track7-3
Track6-4
Education in Reactor Physics Reactor Concepts and Designs Reactor Physics Experiments(Mizuho_B)(Hiei)(Atago)

12:00-13:30

16:00-17:00

Track15
Research Related to Fukushima
Accident(Cosmos)

Poster Session (Mizuho_C, D)

7:30

13:30-15:40

Track9-2
Transient and Safety Analysis(Cosmos)

Lunch (on own)

16:15-18:15

October2(Thu.)

Workshop 2 (Hiei)

Lunch (on own)

7:30

October1(Wed.)

Workshop 6 (Mizuho_C)

Track10-1
Nuclear Data
(Cosmos)

SS4
Advanced Geometry Processing inTrack6-3
Deterministic and Monte Carlo
Reactor Physics Experiments
Methods(Atago)(Cosmos)

Lunch (on own)

Closing (Mizuho_C, D)

Track10-2
Nuclear Data
(Cosmos)