Oral Presentations
The name appeared in the following program is the first author of each paper.
Notifications on presenter is recorded by the organization committee and they are used to check the registration status of each paper.

September 29 (Mon.) 13:00-15:30
1-1. Reactor Analysis Method (Mizuho_A)
Session Chair : Tarmer Bahadir(Studsvik Scandpower, Inc.), Masayuki Tojo(GNF-J)
13:00
13:20
13:40
14:00
14:20
14:40
15:00

Hideo Soneda
Hitachi's Advanced Technologies
Development of Enhanced SPH Method for Pin-by-Pin Core Calculations Satoshi Takeda
STUDY ON CROSS SECTION CORRECTION USING SPH METHOD
Akinori Giho
FOR A WHOLE CORE MOC CALCULATIONASYMPTOTIC, MULTIGROUP FLUX RECONSTRUCTION AND
Travis John Trahan
CONSISTENT DISCONTINUITY FACTORSA POSTERIORI RECONSTRUCTION OF THE FLUX PROFILE IN THE
CASE OF LOCALIZED AXIAL HETEROGENEITIES: AN APPLICATION Enrico Girardi
TO THE MODELING OF PWR MIXING GRIDS
Effects of Advanced Radial Submeshing Methods on Pin Power
Reconstruction for an EPR Core Design
STUDY ON ROBUST ENERGY GROUP STRUCTURE TO SPECTRALINTERFERENCE FOR PWR PIN-BY-PIN CORE ANALYSIS

Hitachi-GE Nuclear Energy, Ltd.
NUCLEAR FUEL INDUSTRIES, LTD

20000021103613

Shikoku Electric Power Co., Inc.

1084323

Los Alamos National Laboratory

1106814

Electricite de France - R&D

1101880

Petra Mala

Paul Scherrer Institute1105669

Satoshi Wada

Osaka university

1105989

SS2-1. Reactor Physics and Criticality Safety Activities in OECD/NEA Working Party (Mizuho_B)
Session Chair : Mark DeHart(INL), Teruhiko Kugo(JAEA)
13:00
13:20
13:40
14:00
14:20
14:40
15:00

Activities of OECD/NEA on Scientific Issues of Reactor Systems andCritical Safety - Current Status and Future Plan
UNCERTAINTY ANALYSIS OF THE OECD/NRC OSKARSHAMN-2
BWR STABILITY BENCHMARKANALYSIS OF THE OECD/NEA OSKARSHAMN-2 FEEDWATER
TRANSIENT AND STABILITY BENCHMARK WITH SIMULATE-3K
Data Assimilation for Kinetic Parameters Uncertainty Analysis
Criticality and Reactor Physics Benchmark Experiments: Influence of
Nuclear Data Uncertainties
THE EVALUATION OF THE SUBCRITICAL EXPERIMENTSPERFORMED IN THE IPEN/MB-01 RESEARCH REACTOR FACILITY
FOR THE IRPhE PROJECTPOLARIS: A NEW TWO-DIMENSIONAL LATTICE PHYSICSANALYSIS CAPABILITY FOR THE SCALE CODE SYSTEM

TBDOECD/NEA

2000000

Ivan Angelov Gajev

Royal Institute of Technology

1093239

Abdelhamid Dokhane

Paul Scherrer Institute (PSI)

1101708

Evgeny Ivanov
Winfried Zwermann
Adimir dos Santos
Matthew A Jessee

Institut de Radioprotection et Surete
Nucleaire
Gesellschaft fuer Anlagen- undReaktorsicherheit (GRS) mbH
Instituto de Pesquisas Energeticas e
Nucleares IPEN-CNEN/SPOak Ridge National Laboratory

1105973110385210956911127872

SS1-1. Molten Salt Reactors (Hiei)
Session Chair :Imre Pazsit(Chalmers Univ. of Tech.), Ritsuo Yoshioka(International Thorium Molten-Salt Forum)
13:00
13:20
13:40
14:00
14:20
14:40
15:00

EXPERIMENTAL MODELLING AND NUMERICAL ANALYSIS OF A

Bogdán Keisuke Yamaji
MOLTEN SALT FAST REACTOR
Victor Dykin
Remark on the propagating neutron noise in a MSR
Victor Dykin
The two-group point-kinetic component of neutron noise in an MSR
NEUTRONICS OF FLUID FUEL SYSTEM WITH PERFECT REMIXING Sandra Dulla
An innovative approach to dynamics modeling and simulation of the
Matteo Zanetti
Molten Salt Reactor Experiment
Ritsuo Yoshioka
SAFETY CRITERIA AND GUIDELINES FOR MSR ACCIDENT
Reactivity-insertion-transient Analysis of a Fluoride Salt Cooled HighYang Yang
Temperature Reactor

Institute of Nuclear Techniques, Budapest
University of Technology and Economics
Chalmers University of Technology
Chalmers University of Technology
Politecnico di Torino
Politecnico di Milano, Department of Energy,
Nuclear Engineering DivisionInternational Thorium Molten-Salt Forum
Shanghai Institute of Applied Physics

11069111127633
11059941105851
110409010814431087405

SS5. Multiscale, Multiphysics Approaches in Nuclear Science and Engineering Applications (Atago)
Session Chair : Richard Martineau(INL), Takeshi Mitsuyasu(Hitachi)
13:00
13:20
13:40
14:00
14:20
14:40
15:00

Influence of an SN solver in a fine-mesh neutronics/thermal-hydraulics
framework
High-Fidelity Multi-Physics Calculations for Light Water Reactors Using
Coupled CTF/TORT-TD/FRAPTRANTHE COUPLING OF the NEUTRONIC TRANSPORT APPLICATIONRATTLESNAKE TO THE NUCLEAR FUELS PERFORMANCEAPPLICATION BISON UNDER THE MOOSE FRAMEWORKA model of two-stage core calculation method coupled with subchannel
analysis for boiling water reactors
SUBSPACE METHODS FOR MULTI-PHYSICS REDUCED ORDER
MODELING IN NUCLEAR ENGINEERING APPLICATIONSExtension of the Entropy Viscosity Method to Flows with Friction Forces
and Source Terms
EFFICIENT FINITE ELEMENT FIELD INTERPOLATION FOR
MULTIPHYSICS APPLICATIONS

Klas Jareteg

Division of Nuclear Engineering, Department
of Applied Physics, Chalmers University of

1104064

Technology
Jeffrey William Magedanz

The Pennsylvania State University

1106221

Frederick N Gleicher

Idaho National Laboratory

1106401

Takeshi Mitsuyasu

Hitachi, Ltd.

1100830

Bassam A. Khuwaileh

Department of Nuclear Engineering North
Carolina State University, Raleigh, NC, USA

1126241

Jean C Ragusa

Texas A&M University

1106855

Jean C Ragusa

Texas A&M University

1106423

SCK-CEN
Jozef Stefan Institute
Alternative Energies and Atomic Energy

11047881102618

11-1. Research Reactors and Spallation Sources (Cosmos)
Session Chair : Ronald J. Ellis(ORNL), Gaillot Stephan(CEA)
13:00
13:20
13:40
14:00
14:20

An updated core design for the multi-purpose irradiation facility MYRRHA Gert Van den Eynde
Luka Snoj
FEASIBILITY STUDY OF INSTALLING A THERMAL TO 14 MeV
Development and Validation of a New APOLLO2-Based Calculation
Florent Chevallier
Scheme Dedicated to Ex-Core Rod Irradiations in the OSIRIS MTRNEUTRONIC DESIGNS AND ANALYSES OF A NEW COREMODERATOR ASSEMBLY AND NEUTRON BEAM PORTS FOR THE Dundar Ucar
PENN STATE BREAZEALE REACTOR
Design Studies for a Multiple Application Thermal Reactor for Irradiation
experiments (MATRIX)

Michael A Pope

Commission (CEA)

1106200

The Pennsylvania State University

1084741

Idaho National Laboratory

1085708

14:40
15:00

SIMULATED IRRADIATION OF SAMPLES IN HFIR FOR USE AS
POSSIBLE TEST MATERIALS IN THE MPEX (MATERIAL PLASMAEXPOSURE EXPERIMENT) FACILITYMCNPX ANALYSIS OF DELAYED NEUTRON FRACTION IN
BERYLLIUM REFLECTED CORES

Ronald J. Ellis

Oak Ridge National Laboratory

1178823

SILVA KALCHEVASCK-CEN

1077740#REF!

Georgia Institute of Technology

1106236

Gabriel Kooreman

Georgia Institute of Technology

1106275

Coline Brosselard

EDF R&D/SINETICS

1094770

Jaakko Leppanen

VTT Technical Research Centre of Finland

1103983

Maurice Grimod

CEA de Saclay

1105216

Dmitriy Anistratov

North Carolina State University

1126229

Aldo Dall'Osso

AREVA NP1086844

September 29 (Mon.) 15:45-18:15
1-2. Reactor Analysis Method (Mizuho_A)
Session Chair : Richard Sanchez(CEA), Hideki Matsumoto(MHI)
15:45

16:05
16:25
16:45
17:05
17:25
17:45

APPLICATION OF THE EFFICIENT CONSISTENT SPATIAL HOMOGENIZATION METHOD IN NEUTRON TRANSPORT THEORY TO AGAS COOLED THERMAL REACTOR PROBLEM
APPLICATION OF THE HYBRID DIFFUSION-TRANSPORT SPATIALHOMOGENIZATION METHOD TO A HIGH TEMPERATURE TESTREACTOR BENCHMARK PROBLEMNORMALIZATION METHODS FOR DIFFUSION CALCULATIONSWITH VARIOUS ASSEMBLY HOMOGENIZATIONSON THE PRACTICAL FEASIBILITY OF CONTINUOUS-ENERGYMONTE CARLO IN SPATIAL HOMOGENIZATIONA Dynamic Homogenization Model for Pebble Bed Reactors
HOMOGENIZATION OF THE STEP CHARACTERISTIC SCHEME IN
PHASE SPACE
Spatial Rehomogenization of Cross Sections and Discontinuity Factors
for Nodal Calculations

SS2-2. Reactor Physics and Criticality Safety Activities in OECD/NEA Working Party (Mizuho_B)
Session Chair : Mark DeHart(INL), Teruhiko Kugo(JAEA)
15:45
16:05
16:25
16:45
17:05
17:25
17:45

EVALUATION OF LARGE 3600MWth SODIUM-COOLED FAST
REACTOR OECD NEUTRONIC BENCHMARKS
Evaluation of Medium 1000 MWth Sodium-cooled Fast Reactor OECDneutronic Benchmarks
SFR WHOLE CORE BURNUP CALCULATIONS WITH TRIPOLI-4
MONTE CARLO CODE
SUMMARY AND STATUS OF OECD/NEA UAM-LWR BENCHMARKUNCERTAINTY AND SENSITIVITY ANALYSIS OF OECD/NEA UAM
FUEL THERMAL BEHAVIOUR BENCHMARK USING A
FALCON/URANIE METHODOLOGYNew PSI methodology for manufacturing and technological uncertainty
quantification
RE-EVALUATION AND CONTINUED DEVELOPMENT OF SHIELDINGBENCHMARK DATABASE SINBAD

#REF!

CEA

1103961

Nicolas Emile Stauff

Argonne National Laboratory

1104111

Yi-Kang Lee

CEA-Saclay

1106242

Maria Nikolova Avramova

The Pennsylvana State University

1104813

Younsuk Yun

Paul Scherrer Institut

1105847

Marco Pecchia

Paul Scherrer Institut (PSI)

1105694

Ivan Alexander Kodeli

Jozef Stefan Institute1106058

PSI SwitzerlandUJV Rez / Czech Technical University in

1102672

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

HYBRID SPECTRUM MOLTEN SALT REACTOR

Jiri Krepel

16:05

Thorium Conversion Optimization in Two-Fluid Molten-Salt Reactor

Jan Frybort

16:25
16:45
17:05
17:25

Development of Computer Code Systems for Molten Salt Reactor Core
AnalysisUse of MCDancoff Factor Correction for Multi-group Fuel Depletion
Analyses of Liquid Salt Cooled Reacotors
COMPARATIVE STUDIES ON PLUTONIUM AND 233U UTILIZATIONIN MINIFUJI MSR
ON AN OPTIMIZED NEUTRON SHIELDING FOR AN ADVANCEDMOLTEN SALT FAST REACTOR DESIGN

Yongjin Jeong

PragueUlsan National Institute of Science andTechnology

1105569
1106010

Michael Huang

Georgia Institute of Technology

1106115

Abdul Waris

Bandung Institute of Technology

1108342

Bruno Merk

Helmholtz-Zentrum Dresden-Rossendorf e.V. 1105793

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

16:05
16:25
16:45
17:05
17:25
17:45

CEA CONTRIBUTION TO THE ANALYSIS OF THE CONTROL ROD

Vincent PASCAL

CEA, DEN, DER, SPRC/LEDC

1096440

Performed During PHENIX End-of-Life Experiments - JAEA’s Calculation Kazuya Takano

Japan Atomic Energy Agency

1090681Indira Gandhi Centre for Atomic Research1102267

WITHDRAWAL TEST PERFORMED DURING PHENIX END-OF-LIFE
EXPERIMENTS (IAEA Common Research Program)
IAEA Benchmark Calculations on Control Rod Withdrawal Test

Results
BENCHMARK ANALYSIS OF PHENIX CONTROL ROD WITHDRAWAL
END-OF-LIFE EXPERIMENTS
CALCULATION OF THE PHENIX END-OF-LIFE TEST “CONTROLROD WITHDRAWAL” WITH THE ERANOS CODE
Simulation of PHENIX Control Rod Withdrawal Experiments withSIMMER-IV
STUDY OF THE EFFECT OF HETEROGENEITY OF THE CONTROLRODS IN THE PHENIX REACTOR
IAEA BENCHMARK CALCULATIONS ON CONTROL ROD
WITHDRAWAL TEST PERFORMED DURING PHENIX END-OF-LIFE

Devan Kunhiraman
Evgeny Ivanov
Vladimir Kriventsev
FABRIZIO GABRIELLIInstitut de Radioprotection et Surete
Nucleaire
Karlsruhe Institute of Technology (KIT)
Karlsruhe Institute of Technology (KIT),
Institute for Nuclear and Energy

11040991104108
1106079

Vincent PASCAL

CEA, DEN, DER, SPRC/LEDC

1096461

Sandra Dulla

POLITECNICO DI TORINO

1076430

Sandra Dulla

Politecnico di Torino

1107269

Cheolho Pyeon

Kyoto University

1111120

Tadafumi Sano

Kyoto University Research Reactor Institute1126220

EXPERIMENTS - BENCHMARK RESULTS AND COMPARISONS

11-2. Research Reactors and Spallation Sources (Cosmos)
Session Chair : Gert Van den Eynde(SCK/CEN), Cheolho Pyeon(KURRI)
15:45
16:05
16:25
16:45

A method for reactivity monitoring in subcritical source-driven systems
INTEPRETATION OF EXPERIMENTAL MEASUREMENTS ON THE
SC-1 CONFIGURATION OF THE VENUS-F CORE
Neutronic Characteristics of Solid Targets in Accelerator-Driven System
at Kyoto University Critical Assembly
EVALUATION OF NEUTRON SPECTRUM AT IN-CORE IRRADIATION EQUIPMENTS IN KUR WITH LOW ENRICHED URANIUM FUEL

17:05

17:25
17:45

ACCULACY OF THORIUM-LOADED ACCELERATOR-DRIVENSYSTEM EXPRIMENTS AT KYOTO UNIVERSITY CRITICAL

Masao Yamanaka

ASSEMBLY
SHUTDOWN TRANSIENTS ANALYSIS FOR REFLECTOR DEVICES

POWER CALCULATIONS IN JULES HOROWITZ MATERIAL TESTING Patrizio Console Camprini
REACTOR (JHR)
Neutronic Analysis of the PULSTAR Reactor Using Monte Carlo
Simulations

Ayman I. Hawari

Graduate school of Energy Science, Kyoto
University

1090583

ENEA

1104755

North Carolina State University

1127882

September 30 (Tue.) 8:00-10:05
1-3. Reactor Analysis Method (Mizuho_A)
Session Chair : Han Gyu Joo(SNU), Wei Shen(CCSN)
8:00
8:20
8:40
9:00
9:20
9:40

Research Reactor In-Core Fuel Management Optimisation Using The
Multiobjective Cross-Entropy Method
Transient Cycle Fuel Management Optimization of a Pressurized Water
ReactorA MULTI-LEVEL PARALLEL COMPUTATION OF REACTOR CORES
USING GPU FOR LOADING PATTERN OPTIMIZATION
Exact-to-Precision Generalized Perturbation Theory for Reactor Design
Caculations
Depletion GPT-Free Sensitivity Analysis of the TMI Reactor Eigenvalue
Model
The "Virtual Density" Theory of Neutronics: A Generic Method forGeometry Distortion Reactivity Coefficients

Evert B. Schlunz

South African Nuclear Energy Corporation
SOC Ltd (Necsa)

1101850

tongkyu park

FNC Technology Co., Ltd.

1105495

Takuya Okubo

Nagoya University

1101184

Congjian Wang

North Carolina State University

1127680

Christopher Brandon Kennedy

North Carolina State University

1106622

Mark Reed

TerraPower

1106665

Matthew S Hodges

University of Nevada, Las Vegas (UNLV)

1105240

Franco Michel-Sendis

OECD Nuclear Energy Agency

1123747

Pierre LECONTECEA Cadarache

1104030

Akiyuki Tsuchiya

Hitachi-GE Nuclear Energy, Ltd.

1107488

Anssu Ranta-aho

Teollisuuden Voima Oyj

1106047

Marcel TARDY

AREVA

1106235

Maxime Dion

Ecole Polytechnique de Montreal

1080046

Nuclear Engineering, Ltd.

1105664

University of Michigan

1105191

Massachusetts Institute of Technology

11053195-1. Nuclear Criticality Safety (Mizuho_B)
Session Chair : John Bess(INL), Toshihiro Yamamoto(KURRI)
8:00
8:20
8:40
9:00
9:20
9:40

NUCLEAR CRITICALITY SAFETY IN THE UNITED STATES: RECENT
EVENTS, TRENDS AND A REVIEW OF THE SAFETY CULTURE
A NEW OECD/NEA DATABASE OF NUCLIDE COMPOSITIONS OF
SPENT NUCLEAR FUEL
OECD EGBUC Benchmark VIII ? Comparison of calculation codes andmethods for the analysis of small-sample reactivity experiments
Criticality Calculation of Fuel Debris in Fukushima Daiichi Nuclear Power
StationDesign of an efficient calculation model of BWR cold critical experimentsfor validation
FIRST BURNUP CREDIT APPLICATION FOR TRANSPORT AND
STORAGE CASK USING FRENCH EXPERIMENTS

2-1. Deterministic Transport Theory (Hiei)
Session Chair : Nam Zin Cho(KAIST), Masato Tabuchi(NEL)
8:00
8:20
8:40
9:00
9:20
9:40

A Collision Probability Based Method to Compute Cross Sections
Sensitivities for the Subgroup Self-Shielding Technique
Improvement of a convergence technique for MOC calculation with large

Masato Tabuchi
negative self-scattering cross section
Shane Gray Stimpson
Boundary Acceleration Techniques for CMFD-Accelerated 2D-MOC
A Low Order Nonlinear Transport Acceleration Scheme for the Method of
Lulu Li
Characteristics
p-CMFD ACCELERATION AND NONOVERLAPPING LOCAL/GLOBAL
Seungsu YukITERATIVE TRANSPORT METHODS WITH 2-D/1-D FUSION KERNELAPPLICATION OF THE SDD-CMFD ACCELERATION TECHNIQUE TO
Brendan M. Kochunas
PARALLEL 3-D METHOD OF CHARACTERISTICS TRANSPORT

Korea Advanced Institute of Science andTechnology

1126428

University of Michigan

1104795

Jaakko Leppanen

VTT Technical Research Centre of Finland

1103981

Toni Kaltiaisenaho

VTT Technical Research Centre of Finland

1106126

Tomohiro Endo

Nagoya University

1126624

Bryan R. Herman

Massachusetts Institute of Technology

1094801

Sean E Carney
Dennis Mennerdahl

University of Michigan
E Mennerdahl Systems

11053121106471

Fausto Franceschini

westinghouse

1126928

Nicholas R Brown

Brookhaven National Laboratory

1104783

Argonne National Laboratory

1128293

University of California, Berkeley

1105391

Oak Ridge National Laboratory

1127750MIT1126323

3-1. Monte Carlo Methods (Atago)
Session Chair : Jaakko Leppänen(VTT), Takanori Kitada(Osaka Univ.)
8:00
8:20
8:40
9:00
9:20
9:40

UNSTRUCTURED MESH BASED MULTI-PHYSICS INTERFACE FOR
CFD CODE COUPLING IN THE SERPENT 2 MONTE CARLO CODEANALYZING THE STATISTICS OF GROUP CONSTANTS
GENERATED BY SERPENT 2 MONTE CARLO CODE
Theoretical Prediction on Underestimation of Statistical Uncertainty for
Fission Rate Tally in Monte Carlo Calculation
ANALYSIS OF TALLY CORRELATION IN LARGE LIGHT WATER
REACTORS
Higher-Mode Applications of Fission Matrix Capability for MCNP
A Symmetric View Hiding the Ugly Truth

SS8. Reactor Physics of Non-Traditional LWR Fuel Design (Cosmos)
Session Chair : Bojan Petrovic(Georgia Tech.), Yoichiro Shimazu(Fukui Univ.)
8:00
8:20
8:40
9:00
9:20
9:40

I2S-LWR EQUILIBRIUM CYCLE CORE ANALYSISUranium nitride composite fuels in a pressurized water reactor:
exploration of multi-batch cycle length and UB4 admixture for reactivity

control
Florent Heidet
Impact of Coating on Nitride Fuel Performance in PWRs
OPTIMIZATION OF FULLY CERAMIC MICRO-ENCAPSULATED FUEL
Massimiliano Fratoni
ASSEMBLY FOR PWR
Fully Ceramic Microencapsulated Fuels: Characteristics and Potential
Jeffrey J Powers
LWR Applications
NEUTRONIC CHALLENGES OF ADVANCED BOILING WATER
Koroush Shirvan
REACTOR DESIGNS

September 30 (Tue.) 10:20-12:00
1-4. Reactor Analysis Method (Mizuho_A)
Session Chair : Jess Gehin(ORNL), Scott Palmtag(Core Physics Inc.)

10:20
10:40
11:00
11:20
11:40

Watts Bar Unit 1 Cycle 1 Zero Power Physics Tests Analysis with VERACS
AP1000R PWR REACTOR PHYSICS ANALYSIS WITH VERA-CS AND
KENO - PART I: ZERO POWER PHYSICS TESTS
AP1000 PWR REACTOR PHYSICS ANALYSIS WITH VERA-CS AND
KENO - PART II: POWER DISTRIBUTION
SOLUTION OF THE BEAVRS BENCHMARK USING THE nTRACERDIRECT WHOLE CORE TRANSPORT CODECALCULATION OF THE PRESSURE VESSEL FLUENCE IN THE
HUNGARIAN VVER-440 PLANTS FOR THE LIFETIME EXTENSION

Jess C. Gehin

Oak Ridge National Laboratory

1127867

Fausto Franceschini

westinghouse

1126938

fausto franceschini

westinghouse

1126207

Min Ryu

Seoul National University

1104549

Gyorgy Hegyi

Magyar Tudomanyos Akademia
Energiatudomanyi Kutatokozpont

10840935-2. Nuclear Criticality Safety (Mizuho_B)
Session Chair : Bo Feng(ANL), Tomohiro Endo(Nagoya Univ.)
10:20
10:40
11:00
11:20
11:40

UNCERTAINTY EVALUATION OF REACTIVITY IN SINGLE AND
MULTI-REGION TSUNAMI MODELING ANALYSIS FOR DRY CASKSTORAGETransient Analysis in Super Critical Condition for Several Fuel-solution
Tanks System with Different Layout
Comparison of Gamma Dose Rate Calculations for PWR Spent Fuel
Assemblies
SPENT FUEL CANISTER CRITICALITY CALCULATION IN
GROUNDWATER PENETRATION ACCIDENTFAVORABLE FEATURES IN KINETICS OF FAST REACTORS WITH
PHYSICALLY THICK 208Pb-REFLECTOR

Quentin T Newell
Haruka Kikuchi
Bo Fengsi yuan wu
Gennady Genrikhovich Kulikov

University of Nevada, Las Vegas (UNLV)
Department of Nuclear Engineering, Tokyo
Institute of Technology
Argonne National Laboratory
East China Institute of Technology, Ministryof Education
National Research Nuclear University
Moscow Engineering Physics Institute1087608
1119923110628510922741092833

2-2. Deterministic Transport Theory (Hiei)
Session Chair : Farzad Rahanema(Georgia Tech. Univ.), Wu Hongchun(Xi'an Jiaotong univ.)
10:20
10:40
11:00
11:20

Iterative Properties of the Integral Transport Matrix Method for the DD
Scheme in 2D Cartesian Geometry
Neutron Leakage Treatment in Reactor Physics: Consequences on SFR
Characteristics Prediction
REVISIT BOUNDARY CONDITIONS FOR THE SELF-ADJOINTANGULAR FLUX FORMULATION
Accuracy Preserving Surrogate for Neutron Transport Caculations

Dmitriy Anistratov

North Carolina State University

1106264

Gerald Rimpault

CEA

1104627

Yaqi Wang

Idaho National Laboratory1102705

Congjian Wang

North Carolina State University

1127699

13. Radiation Applications and Nuclear Safeguards (Atago)
Session Chair : Alexis C. Kaplan(Michigan Univ.), Yasunori Kitamura(JAEA)
10:20
10:40
11:00
11:20
11:40

DESIGN OF LONG NEUTRON COUNTER FOR INTENSIFIED D-T
NEUTRON SOURCE
Utilizing Simulated Rossi-Alpha Distributions to Develop New Methods ofCharacterizing Spent Nuclear Fuel
ANTINEUTRINO EMISSION FROM FUELS WITH HIGHPROLIFERATION RESISTANCE
Validation of the Implicit Correlation Method in MCPNX-PoliMi using
Plutonium Cross-Correlation Measurements
A UNIQUE TUNGSTEN-BASED TAGGING APPROACH FOR
MAINTAINING OF CONTINUITY OF KNOWLEDGE OF NUCLEAR

Yanan Li

Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team1126839

Alexis C Kaplan

LANL / University of Michigan

1104803

Tomooki SHIBASUBATECH Laboratory

1123449

Matthew James Marcath

University of Michigan, Ann Arbor

1105483

Dina Chernikova

Chalmers University of Technology

1127570

youqi zheng

Xi'an Jiaotong University

1084871

Julio C.L. Fernandes

Politecnico di Torino

1102531

Samuel Shaner

Massachusetts Institute of Technology

1105035

Ludovic Maas

IRSN

1094627

Brian C Kiedrowski

Los Alamos National Laboratory

1107334

Takanori Kajihara

Hokkaido University

1105724

Ryota Katano

Nagoya University

1101303

Hokkaido university

1068216

Wu Mingyu

1065760

Nuclear Fuel Industries, Ltd.

1105480

EDF R&D

1127599

WASTE COPPER CANISTERS

9-1. Transient and Safety Analysis (Cosmos)
Session Chair : Kostadin Ivanov(Pennsylvania Univ.), Yuichiro Ban(Toshiba)
10:20
10:40
11:00
11:20
11:40

DEVELOPMENT OF NEUTRON KINETIC CODE FOR MOLTEN SALTREACTORSTUDY OF NEUTRON PROPAGATION IN MULTIGROUP
TRANSPORT BY SPACE ASYMPTOTIC METHODS
Sensitivity Analysis and Performance of the Adiabatic, Theta, andMultigrid Amplitude Function Kinetics Methods in 2D MOC Neutron
Transport
Computations of heterogeneous dilution transients using CFX andHEMERA V1
Prompt Behavior of Generalized-Eigenvalue Point Kinetics Models

September 30 (Tue.) 13:30-15:40
1-5. Reactor Analysis Method (Mizuho_A)
Session Chair : Myung Hyun Kim(Kyung Hee University), Masahiro Tatsumi(NEL)
13:30
13:50
14:10
14:30
14:50
15:10

AUTOMATIC CONSTRUCTION OF A SIMPLIFIED BURN-UP CHAIN
MODEL BY THE SINGULAR VALUE DECOMPOSITION
Generation of simplified burnup chain using contribution matrix of nuclide
production
IMPORTANT FISSION PRODUCT NUCLIDES IDENTIFICATION

Go CHIBAMETHOD FOR SIMPLIFIED BURNUP CHAIN CONSTRUCTION
Mingyu Wu
Application of Backtracking Algorithm to Depletion Calculations
Modeling the Cross Section of Gadolinia Pins in the Depletion for Pin-byYasuhiro Kodama
Pin Core Calculations
THE OPTIMIZED ALGORITHM FOR THE MICROSCOPIC DEPLETION
Matthieu Raju Guillo
MODEL IN THE COCAGNE CORE CODE A 2-LEVEL CORE
PARTITIONING APPROACH

4-1. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Hany Abdel-Khalik(NSCU), Kenji Yokoyama(JAEA)
13:30
13:50
14:10

Validation and benchmarking of calculation methods for photon andneutron transport at cask configurations
RECENT ADVANCES IN THE V&V OF THE NEW FRENCH CEA
APOLLO3R NEUTRON TRANSPORT CODE : BENCHMARKSANALYSIS OF THE FLUX SOLVERSDECART CODE VERIFICATIONS BY NUMERICAL BENCHMARKCALCULATIONS OF HTTR

Luc Schlomer

WTI Wissenschaftlich-Technische
Ingenieurberatung GmbH

1086809

Jean-Marc Palau

CEA

1086829

Chag Joon Jeong

Korea Atomic Energy Research Institute1101312

14:3014:50
15:10

DEVELOPMENT AND VERIFICATION OF THREE-DIMENSIONAL
HEX-Z BURNUP SENSITIVITY SOLVER BASED ON GENERALIZED

Kenji Yokoyama

Japan Atomic Energy Agency

1106037PERTURBATION THEORY
Validation of HELIOS for ATR Core Follow Analyses
PERSENT: NEED OF A DETERMINISTIC CODE FOR SENSITIVITY

Samuel E Bays

Idaho National Laboratory

1081829GERARDO ALIBERTI

Argonne National Laboratory

1100850

Shane Gray Stimpson

University of Michigan

1105193

Yuki KATO

Nagoya University

1106157

Victor F. Boyarinov

NRC Kurchatov Institute1102516

Roland Lenain

CEA France

1127828

William BoydMIT1104798

Nikolay V. Sultanov

NRC "Kurchatov institute"

1103956

Andrea Zoia

CEA/Saclay

1099127

ANALYSIS IN 3D GEOMETRY AND TRANSPORT THEORY

2-3. Deterministic Transport Theory (Hiei)
Session Chair : Yousry Azmy(North Carolina State Univ.), Kazuya Yamaji(MHI)
13:30
13:50
14:10
14:30
14:50
15:10

Axial Transport Solvers for the 2D/1D Scheme in MPACT
Development of Legendre Expansion of Angular Flux Method for 3D
MOC Calculation
BENCHMARK ON DETERMINISTIC TIME-DEPENDENT TRANSPORTCALCULATIONS WITHOUT SPATIAL HOMOGENISATION
Coarse-grained parallelism for full-core transport calculations
PARALLEL PERFORMANCE RESULTS FOR THE OPENMOCMETHOD OF CHARACTERISTICS CODE ON MULTI-CORE
PLATFORMS
MAKING MORE PRECISE THE SURFACE PSEUDOSOURCESMETHOD FOR RBMK CLUSTER CELLS

3-2. Monte Carlo Methods (Atago)
Session Chair : Kan Wang(Tsinghua Univ.), Yasushi Nauchi(CRIEPI)
13:30
13:50
14:10
14:30
14:50
15:10

A MONTE CARLO METHOD FOR PROMPT AND DELAYED ALPHAEIGENVALUE CALCULATIONSGEOMETRY NAVIGATION ACCELERATION BASED ON AUTOMATIC
NEIGHBOR SEARCH AND ORIENTED BOUNDING BOX IN MONTECARLO SIMULATION
Continuous-Energy Monte Carlo Methods for Calculating Generalized
Response Sensitivities using TSUNAMI-3D
Enhancements in Continuous-Energy Monte Carlo Capabilities in SCALE
6.2
Leakage-corrected fast reactor assembly calculation with Monte-Carlocode TRIPOLI4 and its validation methodology
Impact of Nearest Neighbor Distribution of Fuel Particle on Neutronics
Characteristics in Statistical Geometry Model

Zhenping Chen

Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team1104233

Christopher Michael Perfetti

Oak Ridge National Laboratory

1127880

Bradley Thomas Rearden

Oak Ridge National Laboratory

1142583

li cai

CEA,DEN, DER/SPRC

1102576

Takahiro Koide

Nagoya University

1101295

12-1. Fuel Cycle and Actinide Management (Cosmos)
Session Chair : Nicolas Brown (BNL), Naoyuki Takaki(Tokyo City Univ.)
13:30
13:50
14:10
14:30

SIMULATION OF FUEL CYCLES WITH MINOR ACTINIDEMANAGEMENT USING A FAST BURNUP CALCULATION TOOL
Variations in activity, toxicity and decay heat of nuclear waste of various
fuel cycles
EFFECT OF HETEROGENEITY IN PLUTONIUM RECYCLING IN
STEADY STATE PWREVALUATION METHOD OF EQUIVALENCE FACTORS FOR MOX
FUEL AND NON-LINEAR “Equivalent Pu-239” FORMULA

Mate Szieberth

Budapest University of Technology andEconomics, Institute of Nuclear Techniques

1127550

Nicolas Emile Stauff

Argonne National Laboratory

1104106

Marc Ernoult

Institut de Physique Nucleaire

1105805

Mikio TOKASHIKI

Nuclear Fuel Industries, Ltd.

1105974

14:50

Development of a Fuel Performance Code for Thorium-Plutonium Fuel

Klara Linnea Insulander Bjork

15:10

Two-Stage Fuel Cycles with Accelerator-Driven Systems

Florent Heidet

Thor Energy and Chalmers University ofTechnology
Argonne National Laboratory

11275731106238

September 30 (Tue.) 15:55-18:05
1-6. Reactor Analysis Method (Mizuho_A)
Session Chair : Mohamed Elsawi (Khalifa University) , Yunzhao Li(Xi'an Jiaotong University)
15:55
16:15
16:35
16:55
17:15
17:35

MODELLING OF SHUTDOWN COOLING REACTIVITY EFFECTS
WITH SIMULATE
IMPLEMENTATION AND VERIFICATION OF THE SDM IN THE TITAN
3-D Sn TRANSPORT CODE
TRANSPORT CORE SOLVER VALIDATION FOR THE ASTRIDCONCEPTUAL DESIGN STUDY WITH APOLLO3R
METHODOLOGY ASSESSMENT FOR THE EVALUATION OF THE
COOLANT VOID WORTH IN SODIUM FAST REACTORS WITH A
LOW VOID EFFECT CORE DESIGN
Results of verification of computer codes used for analysis of BN-1200
reactor core neutronics
APOLLO3R based method for 3D warped cores calculations ;Application to flowering tests of Phenix

Tamer BahadirStudsvik Scandpower, Inc.

1106193

Nathan John Roskoff

Virginia Tech

1106648

Jean-Francois Vidal

CEA Cadarache

1101764

SARA BORTOTPaul Scherrer Institut (PSI)

1099616

Elena Marova

OKBM

1104321

Cyril Patricot

CEA1106232

Kensuke Kojima

Japan Atomic Energy Agency

1126354

Fabrice Hoareau

EDF R&D

1094607

Geoffrey A Gunow

Massachusetts Institute of Technology

1106258

Peter Grimm

Paul Scherrer Institute (PSI)

1104066

Wim Haeck

IRSN

1102701

4-2. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Bassam Khuwaileh(NCSU), Kensuke Kojima(JAEA)
15:55
16:15
16:35
16:55
17:15
17:35

BENCHMARK CALCULATION WITH MOSRA-SRAC FOR BURNUP OF
A BWR FUEL ASSEMBLY
VERIFICATION OF THE COCAGNE CORE CODE USING CLUSTER
DEPLETION CALCULATIONSLWR Fuel Reactivity Depletion Verification Using 2D Full Core MOC andFlux Map Data
CASMO-4E and CASMO-5 Analysis of the Isotopic Compositions of the
LWR-PROTEUS Phase II Burnt PWR UO2 Fuel Samples
EXPERIMENTAL VALIDATION OF DECAY HEAT CALCULATIONSWITH VESTA 2.1
DEVELOPMENT AND VALIDATION OF AD HOC ORIGEN-ARP
LIBRARIES FOR VERY HIGH BURNUP UO2 PWR FUEL WITH

Stefano Caruso

SCALE/TRITON

National Cooperative for the Disposal ofRadioactive Waste (NAGRA)

1095732

2-4. Deterministic Transport Theory (Hiei)
Session Chair : Ricardo Barros(Univ. do Estado do Rio de Janeiro ), Tesuo Matsumura(CRIEPI)
15:55

Phase Space Bases for Response Matrix Methods

Jeremy A Roberts

Kansas State University

1107355

16:15
16:35
16:55
17:15

17:35

The Drift Diffusion Limit of Thermal Neutrons: Theoretical and Numerical
Results
Adequacies of Different Convergence Accuracy Measures in Full-Core
Nodal Flux Computations
FLEXIBLE SEMI-ANALYTICAL CALCULATION METHOD OF ESCAPEPROBABILITY
CORRECTED DIAMOND DIFFERENCE METHOD FOR COUPLINGFROM THE METHOD OF CHARACTERISTICS TO DISCRETE
ORDINATESENERGY MULTIGROUP SPECTRAL GREEN´S FUNCTIONCONSTANT NODAL METHOD FOR FIXED-SOURCE Sn PROBLEMS

Pablo A Vaquer

Texas A&M University

1126267

Rene van Geemert

AREVA GmbH1137347

Tetsuo Matsumura

CRIEPI1094291

Mitchell T.H. Young

University of Michigan

1106304

Ricardo C. Barros

Universidade do Estado do Rio de Janeiro

1121396

Aleksandar Stoyanov Ivanov

Karlsruher Institut fur Technologie

1127876

Guillaume Truchet

CEA, DEN, DER/SPRC/LEPh, Cadarache

1104427

Byoung Kyu Jeon

Seoul National University

1105946

Bechtel Marine Propulsion Corporation

1142507

Tsinghua University

1104699

Korea Atomic Energy Research Institute1105031

Florent Heidet

Argonne National Laboratory

1106234

Nicholas R Brown

Brookhaven National Laboratory

1104784

Baptiste Leniau

Subatech1105816

IN X,Y-GEOMETRY

3-3. Monte Carlo Methods (Atago)
Session Chair : Daniel J. Kelly(Knolls Atomic Power Laboratory ), Zeguang LI(Tsinghua University)
15:55
16:15

LARGE-SCALE MONTE CARLO CALCULATIONS WITH THERMALHYDRAULIC FEEDBACK
SODIUM VOID REACTIVITY EFFECT ANALYSIS USING THE NEWLY
DEVELOPED EXACT PERTURBATION THEORY IN MONTE-CARLOCODE TRIPOLI-4

16:35
16:55
17:15
17:35

Monte Carlo Perturbation Analysis on Isothermal Temperature Reactivity
Coefficient of Light-Water Moderated and Reflected Critical Assembly

Daniel F. Gill
Monte Carlo and Thermal-Hydraulic Coupling via PVMEXEC
Perturbation Based Monte Carlo Criticality Search in Density, Enrichment
Zeguang LI
and Concentration
Monte Carlo Perturbation Method for Geometrical Uncertainty Analysis
Ho Jin Park
using McCARD

12-2. Fuel Cycle and Actinide Management (Cosmos)
Session Chair : Alberto Talamo(ANL), Kazufumi Tsujimoto(JAEA)
15:55
16:15
16:35
16:55
17:15
17:35

Thorium-Fueled Breed-and-Burn Fuel Cycle
Fuel Cycle Analysis of a Self-Sustaining Light Water Cooled Reactor with
232Th/233U Fuel and Impact of 233U (n,γ) Cross Section Evaluations
PRODUCTION OF 232U FROM IRRADIATION OF STANDARD AND
THORIUM-BASED FUELS IN PWR REACTORSFUEL CYCLE SCHEME DESIGN AND EVALUATION FOR THORIUMURANIUM BREEDING RECYCLE IN CANDU REACTORSAn Inventory Analysis of Thermal-Spectrum Thorium-Fueled Molten Salt
Reactor Concepts
Agent-Based Dynamic Resource Exchange in Cyclus

Bo YANG

Shanghai Nuclear Engineering Research and
Design Institute1087546

Jeffrey J. Powers

Oak Ridge National Laboratory

1126252

Matthew J. Gidden

UW - Madison1104469

October 1 (Wed.) 8:00-10:05
1-7. Reactor Analysis Method (Mizuho_A)
Session Chair : Eleodor Nichita(University of Ontario Institute of Technology), Chang Joon Jeong(KAERI)
8:00
8:20
8:40
9:00
9:20
9:40

High order source approximaion for the EFEN method
EXTENSION OF LINEAR SOURCE MOC METHODOLOGY TO
ANISOTROPIC SCATTERING IN CASMO5FINITE DIFFERENCE EQUATIONS FOR NEUTRON FLUX AND
IMPORTANCE DISTRIBUTION IN A HETEROGENEOUS REACTORWITHOUT HOMOGENIZATION AND DIFFUSION APPROXIMATIONEFFICIENT SUBSPACE CONSTRUCTION FOR REDUCED ORDER
MODELING IN REACTOR ANALYSIS
Variational Acceleration of Fission Source Iteration for Subcritical
Source-Driven Systems
An Incident Flux Coupling Calculation Study for Nodal Method and Monte
Carlo Method

Yunzhao Li
Rodolfo M. Ferrer
Aleksandr V Elshin
Bassam A. Khuwaileh

School of Nuclear Science and Technology,
Xi'an Jiaotong University
Studsvik Scandpower, Inc.
Alexandrov Research Institute of Techology,
Rosatom
Department of Nuclear Engineering North
Carolina State University, Raleigh, NC, USA

1106187
1099700
11046101126237

Bilge Ozgener

Istanbul Technical University

1124804

Xinzhe Wang

China Institute of Atomic Energy

1104763

Tomohiro Endo

Nagoya University

1084668

Kuniharu Kinoshita

Nagoya University

1101192

Akio Yamamoto

Nagoya University

1090063

Tomoaki Watanabe

Nagoya University

1098930

Ting Zhu

Paul Scherrer Institut

1102712Axel Hoefer

AREVA GmbH Offenbach, Germany

1104105

Kevin L Manalo

Georgia Institute of Technology

1106684

Ce Yi

Georgia Institute of Technology

1106722

Jason HouThe Pennsylvania State University

1104276

4-3. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Nuria Garcia Herranz(Univ. Politecnica de Madrid), Go Chiba(Hokkaido Univ.)
8:00
8:20
8:40
9:00
9:20
9:40

Confidence interval estimation by bootstrap method for uncertainty
quantification using random sampling method
Uncertainty quantification of neutronics characteristics using Latin
Hypercube Samplig method
Uncertainty Quantification of BWR Core Characteristics using Latin
Hypercube Sampling Method
Applicability of the cross section adjustment method based on random
sampling technique for burnup calculation
NUSS-RF: Stochastic Sampling-Based Tool for Nuclear Data Sensitivity
and Uncertainty Quantification
MOCABA: A General Monte Carlo-Bayes Procedure for Improved
Predictions of Integral Functions of Nuclear Data

SS3. Hybrid Particle Transport Methods for Solving Complex Problems in Real-Time (Hiei)
Session Chair : Alireza Haghighat(Virginia Tech Univ.), Kazuya Yamaji(MHI)
8:00
8:20
8:40

DISCRETIZED MESH TOOLS AND RELATED TREATMENT FOR
HYBRID TRANSPORT APPLICATION WITH 3D DISCRETE
ORDINATES AND MONTE CARLOA NOVEL HYBRID WEIGHTING SCHEME FOR MULTI-GROUPCROSS SECTION COLLAPSING
Development of An Iterative Lattice-Core Coupling Method Based on
MICROX-2 Cross Section Libraries

9:00

9:20
9:40

SOLUTION OF A STYLIZED EUROPEAN PRESSURIZED REACTOR(EPR) BENCHMARK PROBLEM USING THE COARSE MESH RADIA- Daniel Lago
TION TRANSPORT METHOD (COMET)COMPUTATIONAL EFFICIENCY AND ACCURACY OF THE FISSIONCOLLISION SEPARATION METHOD IN 3D HTTR BENCHMARKPROBLEMS
Use of the Fission Matrix Method for Solution of the Eigenvalue Problem
in a Spent Fuel Pool

Georgia Institute of Technology

1106263

Dingkang Zhang

Georgia Institute of Technology

1106314

William J Walters

Virginia Tech

1107387

John Robert Tramm

Argonne National Laboratory

1106774

Naoyuki Nakadozono

Hitachi Research Laboratory, Hitachi, Ltd.

1126312Daniel J. Kelly

Bechtel Marine Propulsion Corporation

1138691

Nicole Simone Guilliard

University Stuttgart - IKE

1127863

Dan Kotlyar

University of Cambridge

1126263

Nicholas E Horelik

Massachusetts Institute of Technology

1106210

3-4. Monte Carlo Methods (Atago)
Session Chair : Ho Jin Park(KAERI), Toshihiro Yamamoto(KURRI)
8:00
8:20
8:40
9:00
9:20
9:40

XSBENCH - THE DEVELOPMENT AND VERIFICATION OF APERFORMANCE ABSTRACTION FOR MONTE CARLO REACTORANALYSIS
Development of Neutron Current Connection Method for Whole Core
Analysis Based on Monte Carlo Method
Analysis of Select BEAVRS PWR Benchmark Cycle 1 Results Using
MC21 and OpenMC
Monte Carlo Neutronics Analysis of Sodium-cooled fast Reactor
Benchmark with OTF Temperature and Burnup Treatment
VERIFICATION OF COUPLED 3D FUEL CYCLE ANALYSIS WITH
STABLE MONTE CARLO BASED CODE, BGCORE, AGAINST THE
NODAL DIFFUSION DYN3D CODE
Domain Decomposition and Terabyte Tallies with the OpenMC Monte
Carlo Neutron Transport Code9-2. Transient and Safety Analysis (Cosmos)
Session Chair : Miriam Daeubler(KIT), Shigeaki Aoki(MNF)
8:00
8:20
8:40
9:00
9:20
9:40

RECRITICALITY RISK IN PWR SPENT FUEl POOLSDEMONSTRATION OF FULLY COUPLED SIMPLIFIED EXTENDEDSTATION BLACK-OUT ACCIDENT SIMULATION WITH RELAP-7
CODE SCALING APPLICABILITY TO A COLD LEG SBLOCA SCENARIO IN A NUCLEAR POWER PLANT
TRANSIENT SIMULATION OF GAS BUBBLE IN A MEDIUM SIZEDLEAD COOLED FAST REACTOR
Power Ramp Transient in a Sodium-Cooled Fast Reactor used for Minor
Actinides Transmutation
STEP TOWARDS INTEGRAL VALIDATION OF ENERGETIC RECRITICALITY PREDICTION FOR SODIUM COOLED FAST REACTOR

Guillaume Grandjean

Institut National des Sciences et TechniquesNuclÃľaires

1081521

Haihua Zhao

Idaho National Laboratory

1100842

Andrea Querol

Universitat Politecnica de Valencia

1106251

Carl Fredrik Hellesen

Uppsala University

1106269

Sara Perez-Martin

Karlsruhe Institute of Technology

1127822

Tatiana Ivanova

Institut de Radioprotection et de Surete
Nucleaire (IRSN)

1126197

October 1 (Wed.) 10:20-12:00
1-8. Reactor Analysis Method (Mizuho_A)
Session Chair : Youqi Zheng(Xi'an Jiaotong University), Tatsuya Iwamoto(GNF-J)
10:20
10:40
11:00
11:20
11:40

Bradley T. Rearden
Modernization Enhancements in SCALE 6.2
Hadrien Leroyer
SALOME-CORE platform: uses for EDF R&D neutronic studies
A STEADY-STATE CORE ANALYSIS CODE FOR THE MODELING OF
Shengcheng Zhou
ACCELERATOR-DRIVEN SUBCRITICAL REACTORSGood Practice in Development of Advanced Assembly/Core Calculation
Masahiro Tatsumi
Methods and Implementations of AEGIS/SCOPE2
VALIDATION OF LANCR01/AETNA01 BWR CODE PACKAGE
Tatsuya Iwamoto
AGAINST FUBILA MOX EXPERIMENTS AND FUKUSHIMA DAIICHINUCLEAR POWER PLANT UNIT 3 MOX CORE

Oak Ridge National Laboratory
EDF R&D

11537411086739

Xi'an Jiaotong University

1085680

Nuclear Engineering Ltd.

1105705

Core Design Group, Global Nuclear FuelJapan

1100982

4-4. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Oscar Cabellos(Univ. Politecnica de Madrid), Hiroki Iwamoto(JAEA)
10:20
10:40
11:00
11:20
11:40

Pinpower Uncertainty Quantification of LWR-PROTEUS Phase III
Experiments
QUANTIFICATION OF CODE, LIBRARY AND CROSS-SECTION

Mathieu Hursin

Paul Scherrer Institut

1099077

Paul Scherrer Institut

1100815

Hiroki Iwamoto

Japan Atomic Energy Agency

1104912

Nuria Garcia-Herranz

Universidad Poltecnica de Madrid

1106021

Oscar Cabellos

Universidad Politecnica de Madrid

1105867

Paolo Balestra

ENEA

1126097

Yusuke Kuroda

TEPCO SYSTEMS CORPORATON

1086934

Antonio Sargeni

IRSN

1099602

Kai FAN

Tsinghua University

1093593

Wim Uyttenhove

SCK-CEN, Belgian Nuclear Research Centre 1104770

UNCERTAINTY EFFECTS ON THE VOID REACTIVITY COEFFICIENT Olivier Leray
OF A BWR UO2 ASSEMBLYSENSITIVITY AND UNCERTAINTY ANALYSIS OF BURNUP
REACTIVITY FOR AN ACCELERATOR-DRIVEN SYSTEMSENSITIVITY/UNCERTAINTY ANALYSIS FOR BWRCONFIGURATIONS OF EXERCISE I-2 OF UAM BENCHMARKIMPACT OF THE FISSION YIELD NUCLEAR DATA UNCERTAINTIESIN THE PIN-CELL BURN-UP OECD/NEA UAM BENCHMARK

8-1. Reactor Operation and Safety (Hiei)
Session Chair : Gerald Rimpault(CEA), Hiroshi Akie(JAEA)
10:20
10:40
11:00
11:20

Effects of cross sections libraries parameters on the OECD/NEA
Oskarshamn-2 Benchmark solution
EVALUATION OF OPERATIONAL EXPERIENCES AND REACTORPHYSICS TESTS OF MOX LOADED BWR CORES
Coupling effects in large reactor cores: the impact of heavy andconventional reflectors on power distribution perturbations
IMPROVE THE ACCURACY OF THE POWER DISTRIBUTION
RECONSTRUCTION USING POWER DISTRIBUTIONS OF
DIFFERENT STATUS AS THE FUNDAMENTAL HARMONIC

6-1. Reactor Physics Experiments (Atago)
Session Chair : Patrick Blaise(CEA), Takuya Umano(Toshiba)
10:20

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

10:40

11:00
11:20
11:40

12 YEARS OF FRANCO-JAPANESE INTERNATIONAL PROGRAMSIN EOLE FOR THE VALIDATION OF 100%MOX RECYCLING IN LWRS
Application of the best representativity method to a future PWR fuelassembly calculation using four critical experiments of different facilities
CHARACTERIZATION OF IRRADIATION FIELDS IN THE

DEN,CAD, Department of Reactor Studies,
Patrick BLAISE

Takuya UMANOEXPERIMENTAL FAST REACTOR JOYO FOR FUEL AND MATERIAL Shigetaka Maeda
IRRADIATION
Reaction Rate, Fission Product Yield, and Rossi-Alpha Measurements
Using a HEU Metal, Copper Reflected Critical Assembly

Rene G. Sanchez

Experimental Physics Section - SPEX /DEN,CAD, Department of Nuclear
Technology, - STRI
Toshiba Corporation Power Systems
Company

11040681104323

Japan Atomic Energy Agency

1098227

Los Alamos National Laboratory

1105192

9-3. Transient and Safety Analysis (Cosmos)
Session Chair : Silva Kalcheva(SCK/CEN), Mikio Tokashiki(NFI)
10:20
10:40

11:00

11:20
11:40

SIMMER-III MODELING OF GAS COOLED FAST REACTORVALIDATION OF THE SUBCHANNEL CODE CTF AGAINST THE
BENCHMARK DATA OF THE OECD/NEA PSBTVALIDATION OF CASMO5 / SIMULATE-3K USING THE SPECIAL
POWER EXCURSION TEST REACTOR III E-CORE: COLD STARTUP, HOT START-UP, HOT STANDBY AND FULL POWER
CONDITIONSVALIDATION OF THE NODAL KINETICS CODE SYSTEMGALAXY/COSMO-K USING THE SPERT-III E-CORE EXPERIMENTS
Propagation of Nuclear Data Uncertainty for a Control Rod Ejection
Accident using the Total Monte-Carlo Method

Xue-Nong Chen

Karlsruhe Institute of Technology (KIT),
Institute for Nuclear and Energy

1106163

Patricio Hidalga

Universitat Politecnica de Valencia

1106259

Gerardo M. GrandiStudsvik Scandpower, Inc.

1099703

Kazuya Yamaji

Mitsubishi Heavy Industries, Ltd.

1119885

Dirceu F. da Cruz

Nuclear Research and Consultancy Group
NRG

1102512October 1 (Wed.) 13:30-15:40
1-9. Reactor Analysis Method (Mizuho_A)
Session Chair : Richard Sanchez(CEA), Baocheng Zhang(WH)
13:30
13:50
14:10
14:30
14:50
15:10

Resonant Upscattering Effects on U238 Absorption Rates
The up-scattering treatment in the fine-structure self-shielding method in
APOLLO3
Target Motion Sampling Temperature Treatment Technique with Tracklength Estimators in OpenMC - Preliminary Results
Problem-Dependent Doppler Broadening of Continuous-Energy Cross
Sections in the KENO Monte Carlo Computer CodeVERIFICATION OF DOUBLY-HETEROGENEOUS SELF-SHIELDINGMETHOD BASED ON EQUIVALENCE THEORY
Quantification of Resonance Interference Effect for Multi-Group Effective
Cross-Section in Lattice Physics Calculation

CLAUDE MOUNIER

Commissariat a l'energie atomique et aux
energies alternatives

1106006

Li LEI-MAO

CEA

1107332

Tuomas Viitanen

VTT Technical Research Centre of Finland

1106078

Shane William Daniel Hart

The University of Tennessee

1101879

Sooyoung Choi
Hiroki Koike

Ulsan National Institute of Science andTechnology

1106138

Mitsubishi Heavy Industries, Ltd.

1105375

Vanessa Vallet

CEA/DEN

1104036

Ryan Nathaniel Bratton

The Pennsylvania State University

1127866

jean-christophe cyrille Sublet

United Kingdom Atomic Energy Authority

1103900

Joshua Hykes

Studsvik Scandpower, Inc.

1104131

Luca Fiorito

SCKCEN

1106036Paul Scherrer Institut

1104088

Donny Hartanto

KAIST

1104420

Yunlong Xiao

Xi an Jiaotong University

1092268

Bruno Merk

Helmholtz-Zentrum Dresden-Rossendorf e.V. 1096317

Wu Seung YOU

Kyung Hee Univsersity

1093377

Francesco Lodi

University of Bologna

11044224-5. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Maria Avramova(Penn State University), Willen F. G. Rooijen(Fukui Univ.)
13:30
13:50
14:10
14:30
14:50

15:10

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
Bias and Uncertainty Assessment of Pressurized Water Reactor Fuel
Isotopics
INVENTORY UNCERTAINTY QUANTIFICATION AND PROPAGATIONUSING TENDL COVARIANCE DATA IN FISPACT-II
Validation of CASMO5 spent fuel isotopics with decay and fission yield
uncertainties
UNCERTAINTY PROPAGATION AND SENSITIVITY ANALYSIS IN
THE ALEPH MONTE CARLO BURNUP CODE: APPLICATIONS TOFISSION PULSE DECAY HEAT CALCULATIONSUNCERTAINTY QUANTIFICATION OF SPENT FUEL NUCLIDECOMPOSITIONS DUE TO CROSS SECTIONS, DECAY CONSTANTS Olivier Leray
AND FISSION YIELDS

7-1. Reactor Concepts and Designs (Hiei)
Session Chair : Bojan Petrovic(Georgia Tech. Univ.), Akiyuki Tsuchiya (Hitachi GE)
13:30
13:50
14:10

14:30

14:50
15:10

CONCEPTUAL STUDY OF A LONG-LIFE PGSFRCORE DESIGN STUDIES ON THE FAST REACTOR WITH FLEXIBLEBREEDING RATIOENHANCED FEEDBACK EFFECTS IN SODIUM COOLED FAST
REACTORS USING MODERATING MATERIAL - THE EFFECT OF
THE PLUTONIUM CONTENT IN THE FUELADVANCED SODIUM COOLED REACTOR CORES HAVING
THORIUM BLANKETS FOR EFFECTIVE BURNING OF
TRANSURANIC NUCLIDESPHISICS IMPROVEMENTS AND COMPARATIVE STUDY WITHERANOS 2.2 ON THE GEN-IV LEAD-COOLED FAST REACTORCONCEPT, ALFRED
Velocity characteristic and stability of wave solutions for a candle reactor
with thermal feedback

Volodymyr Khotyayintsev

Department of Physics, T. Shevchenko
National University of Kyiv

1106054

3-5. Monte Carlo Methods (Atago)
Session Chair : Bradley Rearden(ORNL), Macro Pecchia(PSI)
13:30
13:50
14:10

Effects of Stochastic Noise on a Three-Dimensional Monte Carlo
Depletion Analysis of the H.B. Robinson Reactor
EFFICIENT ESTIMATION OF ADJOINT-WEIGHTED KINETICSPARAMETERS IN THE MONTE CARLO WIELANDT CALCULATIONSHIGHER-ORDER CHEBYSHEV RATIONAL APPROXIMATIONMETHOD (CRAM)

Scott J SpychalaSung Hoon Choi
Maria Pusa

Bettis Atomic Power Laboratory
Korea Electrical Engineering & Science
Research Institute
VTT Technical Research Centre of Finland

114246611059651119422

14:30
14:50
15:10

Efficiency and Accuracy Evaulation of the Windowed Multipole Direct
Doppler Broadening Method
DIFFUSION THEORY MONTE CARLO METHOD WITH TRANSPORTCORRECTIONSGENERATION OF ONE-GROUP SELF SHIELDED CROSS SECTIONSWITH MULTI-GROUP APPROACH FOR MONTE CARLO BURNUP

Colin Josey
Argala Srivastava
Dan Kotlyar

Massachusetts Institute of Technology
BHABHA ATOMIC RESEARCH CENTRE,
MUMBAI, INDIA

11266511105538

Ben Gurion University

1084485

Konrad Schauer

AREVA

2000001

Alan H. Wells

Electric Power Research Institute (EPRI)

1106299

IRSN

1085679

Japan Atomic Energy Agency

1127170

CODES

15. Research Related to Fukushima Accident (Cosmos)
Session Chair : Alireza Haghighat(Virginia Tech Univ.), Akio Yamamoto(Nagoya Univ.)
13:30
13:50
14:10
14:30
14:50
15:10

AREVA Dismantling and Decommisioning experience and Fuel Debris
removal approach for Fukushima Dai-ichi
RE-CRITICALITY POTENTIAL AT FUKUSHIMI DAI ICHI UNIT 4
FOLLOWING THE MARCH 2011 EARTHQUAKE AND TSUNAMI

Why a criticality excursion was possible in the Fukushima spent fuel pool Antonio SARGENICRITICAL EXPERIMENTS FOR FUEL DEBRIS USING MODIFIEDSTACYCONCEPT OF CAPTURE CREDIT BASED ON NEUTRON INDUCEDGAMMA RAY SPECTROSCOPY
A methodology for fast and accurate decay heat calculations for in-pool

Kazuhiko IZAWA
Yasushi Nauchi

Industry1105423

AREVA NC Recycling

1125870

Thomas G. Saller

University of Michigan

1106112

Thomas G. Saller

University of Michigan

1106177

DIFFUSION EQUATION EIGENVALUE PROBLEM WITH THE FINITE

Alvaro Bernal

Universitat Politecnica de Valencia

1106303

VOLUME METHODA Generalized Multigroup Method Based on Finite Elements
AN ASYMPTOTIC, HOMOGENIZED, ANISOTROPIC, MULTIGROUP

Andrew T Till

Texas A&M University

1127687

Travis John Trahan

Los Alamos National Laboratory

1105474

Carolyn N McGraw

Texas A&M University

1127761

used fuel assemblies developed at AREVA La Hague reprocessing

Agnes Launay

Central Research Institute of Electric Powerfacility

October 2 (Thu.) 8:00-10:05
1-10. Reactor Analysis Method (Mizuho_A)
Session Chair : Aldo Dall'Osso(AREVA NP), Christophe Demaziere(Chalmers Univ. of Tech.)
8:00
8:20
8:40
9:00
9:20
9:40

An Asymptotic Homogenized SP2 Approximation to the Boltzmann
Equation. I. Derivation
An Asymptotic Homogenized SP2 Approximation to the Boltzmann
Equation. II. Discontinuity Factors and Numerical Testing
GENERALIZED AND STANDARD MULTIGROUP NEUTRONDIFFUSION APPROXIMATION TO THE NEUTRON TRANSPORTEQUATION
ACCURACY OF THE LINEAR DISCONTINUOUS GALERKIN METHODFOR REACTOR ANALYSIS WITH RESOLVED FUEL PINS

4-6. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Laurent Chabert(AREVA TA), Kenji Nishihara(JAEA)
8:00

SENSITIVITY AND UNCERTAINTY CALCULATIONS METHODS OF

NEUTRONICS PARAMETERS IN PWR CORES PART I: THEORY AND Toshikazu TAKEDASENSITIVITY CALCULATIONS

8:20
8:40
9:00
9:20
9:40

Constrained Quantities in Uncertainty Quantification: Ambiguity and Tips
to Follow
ADJOINT-BASED SENSITIVITY AND UNCERTAINTY ANALYSIS OF
LATTICE PHYSICS CALCULATIONS WITH CASMO-4

Zoltan Perko

Assimilation Analysis?
Sensitivity Analysis via Reduced Order Adjoint Method

University of Fukui
TU Delft, Department of Radiation Scienceand Technology, Nuclear Energy and11003981084579

Radiation Applications
Maria Pusa

VTT Technical Research Centre of Finland

1106024

CEA

1106197

Paridhi AtheNorth Carolina State University

1126688

Youngsuk BangFNC Technology, Co., Ltd

1106183

Dan Ilas

Oak Ridge National Laboratory

1154037

Odmaa Sambuu

Tokyo Institute of Technology

1105504

Marat Margulis

Ben Gurion University

1091293

HwanYeal Yu

KAIST1105145

Benjamin A. Lindley

University of Cambridge

1068763

Francesco Venneri

Ultra Safe Nuclear Corporation

1127098

Nuclear data uncertainty propagation on power maps in large LWR cores Alain Santamarina
Are Modeling Uncertainties Properly Considered in Neutronics Data

Research Institute of Nuclear Engineering,

7-2. Reactor Concepts and Designs (Hiei)
Session Chair : Jan L. Kloosterman(Delft University of Technology), Deokjung Lee(UNIST)
8:00
8:20
8:40
9:00
9:20
9:40

SmAHTR-CTC Neutronic Design
SUPPRESSION OF EXCESS REACTIVITY OF SMALL LONG-LIFE
PRISMATIC HTGR WITH PASSIVE DECAY-HEAT REMOVALCONCEPTUAL DESIGN OF A SELF-SUSTANABLE PRESSURIZEDWATER REACTOR WITH BOILING CHANNELS
Feasibility Study of Application of BigT
Axially Homogeneous Thorium Fuel Designs for Transuranic Burning in
Reduced-moderation BWRs
Neutronic Analysis of A Micro Modular Reactor

6-2. Reactor Physics Experiments (Atago)
Session Chair : Jacques Di Salvo(CEA), Kenichi Yoshioka(Toshiba)
8:00
8:20

AN IMPROVED FEYNMAN-α CORRELATION ANALYSIS WITH A
MOVING-BUNCHING TECHNIQUE
AN ALTERNATIVE SOURCE JERK METHOD IMPLEMENTATION FOR
THE SUBCRITICALITY ESTIMATION OF THE VENUS-F

Ryohei Okuda
Anatoly Kochetkov

SUBCRITICAL CORE IN THE FREYA PROJECTPULSED NEUTRON AND SOURCE JERK EXPERIMENTS FOR
8:40

9:00
9:20
9:40

REACTIVITY ASSESSMENT IN DEEP SUBCRITICALCONFIGURATION: A CASE STUDY WITHIN THE FRAMEWORK OF
THE FREYA PROJECT
Reactivity Measurement of the Lead Fast Subcritical VENUS-F Reactor
using Beam Interruption Experiments
REACTIVITY MEASUREMENTS AT GUINEVERE FACILITY USING
THE INTEGRAL kp METHODESTIMATION OF THE DELAYED NEUTRON FRACTION β_eff OF THE
MAESTRO CORE IN MINERVE ZERO POWER REACTOR

Giorgia Mila

Thibault Matthieu Chevret
Sebastien Chabod
Erez Giladlnterdisciplinary Graduate School of Scienceand Engineering, Kinki University
SCK-CEN, Belgium Nuclear ResearchCentreINFN - Section of Turin
LPC Caen, ENSICAEN/Universite deCaen/CNRS-IN2P3
CNRS
Department of Nuclear Engineering, BenGurion University of Negev

110224711058641128133

10874371104440
1090478

10-1. Nuclear Data (Cosmos)
Session Chair : Pierre Leconte(CEA), Junichi Hori(KURRI)
8:00
8:20
8:40
9:00
9:20
9:40

COMAC: Nuclear Data Covariance Matrices Library for Reactor
Applications
Model-based Generation of Neutron Induced Fission Yields up to 20
MeV by the GEF Code
EXPERIMENTS TO IMPROVE UNCERTAINTY OF THE 1st DELAYEDNEUTRON GROUP ABUNDANCE IN FAST FISSIONS OF U238 ANDSENSITIVITY STUDIES OF THE RELATIVE PARAMETERSOBSERVATION OF NEUTRON THERMALIZATION IN GRAPHITE
USING THE SLOWING-DOWN-TIME TECHNQIUEMEASUREMENT OF NEUTRON CAPTURE CROSS SECTION OF
232TH IN THE LOW ENERGY REGION
New Revisions of Reactor Physics Standards

Pascal Archier

CEA, DEN, DER, SPRC, LEPh

1077966

Kilian P. Kern

Karlsruhe Institute of Technology

1106067

Heejun Chung

University of Florida

1104110

Ayman I. Hawari

North Carolina State University

1105487

Jun-ichi Hori

Research Reactor Institute, Kyoto University

1127786

Dimitrios Cokinos

Brookhaven National Laboratory

1113347

Alain Hebert

Ecole Polytechnique de Montreal

1104443

Justin Michael Pounders
Jaeha Kim

Bettis Atomic Power Laboratory
KAIST

11276461105707

Yurii Bilodid

Helmholtz-Zentrum Dresden-Rossendorf

1106131

Ansar CALLOOEDF R&D/SINETICS

1092925

October 2 (Thu.) 10:20-12:00
1-11. Reactor Analysis Method (Mizuho_A)
Session Chair : Alain Hebert(Ecole Polytechnique de Montreal), Ren-Tai Chiang(Energy Engineering Service)
10:20
10:40
11:00
11:20
11:40

Revisiting the Clio perturbative approach for analyzing systems infundamental mode conditions
A Generalization of Lambda-Mode Xenon Stability Analysis
Preliminary Study of the Impact of Xe-135m on the PCR of CANDU
Verification of the spectral history correction method with fully coupled
Monte-Carlo code BGCore
CORE NEUTRONICS METHODOLOGIES APPLIED TO THE MOXLOADED KAIST 1A BENCHMARK: REFERENCE TO INDUSTRIALCALCULATIONS

4-7. Verification, Validation and Uncertainty Analysis (Mizuho_B)
Session Chair : Maria Pusa(VTT), Tadafumi Sano(KURRI)
10:20
10:40
11:00
11:20
11:40

Sensitivity and uncertainty analysis on reactor kinetic parameters using
perturbation theory
Selecting benchmarks for reactor calculations
DICE 2013: NEW CAPABILITES AND DATAOECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic
Compositions for World Reactor Designs
Effective Physics-Based Uncertainty Quantification for ZrHx Thermal
Neutron Scattering in TRIGA Reactors

Cyrille Bouret
Erwin Alhassan

Commissariat Energie Atomique et aux
Energies Alternatives
Division of Applied Nuclear Physics,

1126308

Department of Physics and Astronomy

1106307

Ian P Hill

Uppsala University
OECD/NEA

1105736

Ian C Gauld

Oak Ridge National Laboratory

1106253

Weixiong Zheng

Texas A&M University

1104568

Cristina Montalvo

Technical University of Madrid (UPM)

1099698

Imre Pazsit

Chalmers Univ. of Technology

1104708

Glenn E. Sjoden

Georgia Institute of Technology

1106266

Hae-Yong Jeong

Sejong University

1122850

Klaes-Hakan Bejmer

Vattenfall AB Sweden

1203930

Nicholas R Brown

Brookhaven National Laboratory

1104785

John Darrell BessIdaho National Laboratory

1084758

Jesson Hutchinson

LANL

1105154

8-2. Reactor Operation and Safety (Hiei)
Session Chair : Fausto Franceschini(WH), Naoyuki Nakadozono(Hitachi)
10:20
10:40
11:00
11:20
11:40

ADVANCED SURVEILLANCE OF RESISTANCE TEMPERATURE
DETECTORS IN NUCLEAR POWER PLANTS.
REFINED METHOD FOR SURVEILLANCE AND DIAGNOSTICS OF
THE CORE BARREL VIBRATIONS OF THE RINGHALS PWRs
ASSESSMENT OF FLOW INDUCED VIBRATION LIMITS IN
PRELIMINARY I2S-LWR FUEL DESIGNSBLOCKAGE INDEX FOR THE DETECTION OF FLOW BLOCKAGE INA SUBASSEMBLY OF SODIUM-COOLED FAST REACTORSECOND GENERATION SHIELDING ASSEMBLIES - NEUTRON FLUXIMPACT ON REACTOR PRESSURE VESSEL AND CORE DESIGN

6-3. Reactor Physics Experiments (Atago)
Session Chair : Nicholas Brown(BNL), Hironobu Unesaki(KURRI)
10:20
10:40
11:00
11:20
11:40

BNL Metal Fuel Lattice Experiments: Candidates for Reactor Physics
Benchmark Evaluation
Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor
Upgraded LEU-Fuel CoreCALIBAN AND GODIVA-IV MEASUREMENTS USING HELIUM-3
DETECTOR SYSTEMSPOWER SPECTRAL ANALYSIS FOR A SUBCRITICAL REACTORSYSTEM DRIVEN BY A PULSED SPALLATION NEUTRON SOURCE
Equivalencies of Open & Closed Loop Reactivity Measurements

Atsushi Sakon
Benjamin A Baker

lnterdisciplinary Graduate School of Scienceand Engineering, Kinki University
Idaho State University

11025821130969

SS4. Advanced Geometry Processing in Deterministic and Monte Carlo Methods (Cosmos)
Session Chair : Hyung Jin Shim(Seoul National Univ.), Shinya Kosaka(MHI)
10:20
10:40
11:00
11:20
11:40

DEVELOPMENT OF A MULTI-GROUP SN TRANSPORTCALCULATION CODE WITH UNSTRUCTURED TETRAHEDRALMESHESDEVELOPMENT OF AN UNSTRUCTURED MESH BASEDGEOMETRY MODEL IN THE SERPENT 2 MONTE CARLO CODEA CAD BASED AUTOMATIC MODELING METHOD FOR PRIMITIVESOLID BASED MONTE CARLO CALCULATION GEOMETRY
Developments and Applications of the Geometry Conversion Tool
McCad for Monte Carlo Particle Transport Simulation
A MEMORY EFFICIENT ALGORITHM FOR CLASSIFYING UNIQUEREGIONS IN CONSTRUCTIVE SOLID GEOMETRIESSER GI HONG

Kyung Hee Univsersity

1093366

Jaakko Leppanen

VTT Technical Research Centre of Finland

1101586

Dong Wang

Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team1105335

Lei LuKarlsruhe Institute for Technology (KIT)1106105

Derek M Lax

Massachusetts Institute of Technology

1106231

Chalmers University of Technology

1106213

October 2 (Thu.) 13:30-15:40
1-12. Reactor Analysis Method (Mizuho_A)
Session Chair : Dimitrios Cokinos(BNL), Rong-Jiun Sheu(Institute of Nuclear Engineering and Science)
13:30

THE ROLE OF THE EIGENVALUE SEPARATION IN REACTOR
DYNAMICS AND NEUTRON NOISE THEORY

Imre Pazsit

13:50

14:10
14:30
14:50
15:10

INVESTIGATION OF CONDITIONAL TRANSPORT UPDATE IN
METHOD OF CHARACTERISTICS BASED COARSE MESH FINITE
DIFFERENCE TRANSIENT CALCULATION
IMPLEMENTATION OF AN A PRIORI TIME STEP ESTIMATOR FOR
THE MULTIGROUP NEUTRON DIFFUSION EQUATION IN
ASYNCHRONOUSLY COUPLED RELAP5-3D
DEVELOPMENT OF A THREE-DIMENSIONAL KINETICS CODE FORCOMMERCIAL-SCALE FBR FULL CORE ANALYSISUSE OF ADJOINT FUNCTIONS FOR COMPARING MEASURED AND
CALCULATED PARAMETERS IN THE SUBCRITICAL SYSTEMSDEVELOPMENT OF THE NEUTRON SOURCE EVALUATIONMETHOD AND PREDICTOR OF SRM/SRNM COUNT RATE IN BWR

Yeon Sang Jung

Seoul National University

1106109

Michael Wayne Hackemack

Texas A&M University

1127681

Yoichiro Shimazu

University of Fukui Research Institute of
Nuclear Engineering

1081523

Sergey Shevchenko

SEC NRS1091810Masayuki Tojo

Global Nuclear Fuel-Japan

1097414

Kouji Hiraiwa

TOSHIBA Corporation

2000003

Benoit Forget

Massachusetts Institute of Technology

1105147

Luka Snoj

Jozef Stefan Institute1106257

Christophe R. Demaziere

Chalmers University of Technology

1099675

ganglin yu

Tsinghua University

1106245

Han Gyu Joo

Seoul National University

1127896

SIMULATOR

14. Education in Reactor Physics (Mizuho_B)
Session Chair : Ben Forget(MIT), Takanori Kameyama(Tokai Univ.)
13:30
13:50
14:10
14:30
14:50
15:10

Education programs for students and graduate students withexperimental facilities for nuclear energy in Toshiba
Past, Present and Future of MIT Reactor Physics
NEW PRACTICAL EXERCISES AT THE JSI TRIGA MARK II
REACTOR
Developing a course in nuclear reactor modelling and going from
campus-based to web-based teaching
Introduction to the status of Reactor physics Education in Tsinghua
University
Reactor Physics Education at Seoul National University

7-3. Reactor Concepts and Designs (Hiei)
Session Chair : Liangzhi Cao(Xi'an Jiaotong U.), Naoto Aizawa(Tohoku Univ.)
13:30

13:50
14:10
14:30
14:50
15:10

DEVELOPMENT OF THE 900 SECOND SPECIFIC IMPULSECARBIDE LOW ENRICHED URANIUM NUCLEAR THERMAL ROCKET
Molybdenum-99 Production in the Oregon State TRIGA Reactor:

Paolo Francesco Venneri

Korea Advanced Institute of Science andTechnology

1126817

Analysis of Multiple Smaller Core Designs Using a New LEU Target as

Andrew J Hummel

Oregon State University

1106761

Fuel
Preliminary Design of the Delft Isotope Production Reactor (DIPR)A STUDY OF SAFETY CORE DESIGN ON BEAM TRANSIENT FOR

Jan Leen Kloosterman

Delft University of Technology

1104016

Naoto Aizawa

Tohoku University

1105750

Liangzhi Cao

Xi'an Jiaotong University

1080898

Frank Jozef Wols

Delft University of Technology

1069253

John Darrell BessIdaho National Laboratory1102703

Mohamed Ouisloumen

Westinghouse Electric Company LLC

1101855

Vicente Becares

CIEMAT

1127823

Toru Yamamoto

Nuclear Regulation Authority

1103841

Jacques DI SALVO

CEA, DEN, DER/SPEx, Cadarache

1099684

ACCELERATOR DRIVEN SYSTEMNEW INVERTED HYDRIDE FUEL DESIGN CONCEPT FOR
PRESSURE TUBE TYPE SUPER CRITICAL WATER REACTORS
Preliminary safety analysis of a Thorium Breeder Pebble Bed Reactor

6-4. Reactor Physics Experiments (Atago)
Session Chair : Mohamed Ouisloumen(WH), Toru Yamamoto(NRA)
13:30
13:50
14:10
14:30

14:50
15:10

Overview of the 2014 Edition of the International Handbook of Evaluated
Reactor Physics Benchmark Experiments (IRPhEP Handbook)Analysis of Tungsten Gray Rods Critical Experiments Using PARAGON
with Ultra-Fine Energy Mesh Methodology
Monte Carlo assessment of spatial and energy effects in the VENUS-F
subcritical configurations and application for reactivity determination.
VALIDATION OF ORIGEN2 COUPLED WITH JENDL-4.0 BASE LIBRARIES FOR ISOTOPIC COMPOSITIONS OF IRRADIATED LIGHT
WATER REACTOR FUELSTHE AMMON EXPERIMENT IN EOLE FACILITY: A CHALLENGINGPROGRAM DEDICATED TO THE EXPERIMENTAL VALIDATION OF
JHR NEUTRONIC AND PHOTONIC CALCULATION TOOLS
Monte Carlo analysis of reactivity effect measurements in the AMMON
experimental program dedicated to JHR neutron studies

Claire Vaglio-Gaudard

CEA - Commissariat a l'Energie Atomique etaux Energies Alternatives

1101619

10-2. Nuclear Data (Cosmos)
Session Chair : Nicolas E. Stauff (ANL), Pascal Archier (CEA)
13:30
13:50
14:10
14:30

14:50

15:10

Impact of the interpolation mode on the secondary particles for shielding
and criticality benchmarks with TRIPOLI-4 Monte Carlo CodeImpact of the differences in nuclear data on estimated k-effective of SFR
cores
NUCLEAR DATA SENSITIVITY ANALYSIS FOR ISOTOPIC
GENERATION USING JENDL-4.0, ENDF/B-VII.1 AND JEFF-3.1.1
VALIDATION OF A POINTWISE ENERGY NEUTRON CROSS

cedric Jouanne

CEA Saclay

1104046

Nicolas Emile Stauff

Argonne National Laboratory

1104107

Yosuke Kawamoto

Nuclear Reactor Engineering Laboratory

1105258

SECTION LIBRARY GENERATED BY RXSP-BETA2.0 USING ENDFB- Jiankai YU
VII.0
CRITICALITY ANALYSES OF URANIUM ZIRCONIUM CARBONNITRIDE LEU FUEL WITH ENDF/B-VII.1, JEFF-3.1.2, AND JENDL-4
NUCLEAR DATA FILESFEEDBACK ON 239PU AND 240PU NUCLEAR DATA ANDASSOCIATED COVARIANCES THROUGH THE CERES INTEGRALEXPERIMENTS

Department of Engineering Physics,
Tsinghua University

1102714

Alberto Talamo

Argonne National Laboratory

1068258

Pierre LECONTECEA Cadarache

1104041