Poster 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.

October 1 (Wed.) 16:15-18:15 (Mizuho_C, Mizuho_D)
Note on the poster session:-Size of poster board: 90cm(width) by 200cm(height)-Preparation of poster display : Oct.1 (Wed.), 14:00 - 16:00-Removal of poster : Oct.1 (Wed.), 18:15 - 19:00-Best poster award will be selected through the evaluation of technical paper and poster presentation

Track1 Reactor Analysis Method
108947610920071093007110261111026221103763110447711045191105008
1105869
11059241106049
110612711061561106178

A Parametric Study and Comparison of BWR Fuel Depletion Calculations
using CASMO-4, MCNPX, and SCALE/TRITON
NEUTRON NOISE INDUCED BY FLUCTUAUTIONS OF THE BORIC
ACID CONTENT IN PRESSURIZED WATER REACTORSAPPLICATION OF WESTINGHOUSE NEXUS/ANC9 CROSS-SECTION
MODEL FOR PWR ACCIDENT ANALYSESNUMERICAL DISPERSION AND DISSIPATION ANALYSIS OF NODALEXPANSION METHOD

Ching-Sheng LinHessam Malmir
Baocheng Zhang

Westinghouse Electric Company

Xiafeng Zhou

Institute of Nuclear and New Energy
Technology, Tsinghua University
Institute of Nuclear and New Energy
Technology, Tsinghua University
South African Nuclear Energy
Corporation (Necsa)

The Integration of Control Rod Calculation and VSOPJiong Guo

REFLECTOR MODELLING WITH MULTI-GROUP NODAL

Suzanne Anel

EQUIVALENCE THEORY FOR THE SAFARI-1 RESEARCH REACTOR
Groenewald
RELATIONSHIP BETWEEN COMPUTED ANSI/ANS-5.1 AND ORIGEN-S
Ren-Tai Chiang
DECAY HEAT POWERS FOR BWR LOCA SAFETY ANALYSISA New Method To Measuring The α Eigenvalue of A subcritical reactor
yun baisystem
PROGRESS TOWARDS AN ACCURATE LATTICE-HOMOGENIZATIONEleodor Nichita
TECHNIQUE FOR PRESSURE-TUBE SUPERCRITICAL WATER
COOLED REACTOR NEUTRONIC CALCULATIONSPROPOSAL OF SUBCRITICAL PWR CORE BENCHMARK PROBLEMSA NEW MONTE CARLO-DETERMINISTIC TWO-STEP METHOD FOR
FAST REACTOR DIFFUSION ANALYSISExplicit Transverse Leakage Treatment Using an Analytic Basis Function
Expansion
Method for calculation capturing reactions contribution to total energy
release in nuclear reactors.
A PRELIMINARY ANALYSIS OF THE ACCURACY OF HOMOGENIZED2D CROSS SECTION IN 3D NODAL CALCULATIONS FOR BWRS
Deterministic Lattice Code Development at UNISTKAISTSteven A Thompson

The Pennsylvania State University

Rynat BekirovichBahdanovich

The Belarusian State University

Andrew M. Ward

University of Michigan

Chidong Kong

Ulsan National Institute of Science andTechnology

MONTE CARLO MODELLING OF VR-1 REACTOR CORE

1126966

ASSESSMENT OF THE 2D/1D IMPLEMENTATION IN MPACT
Benjamin S Collins
Coupled Neutronics and Thermal-Hydraulic Solution of a Full-Core PWR
Scott Palmtag
using VERA-CS
ASSESSMENT OF THE WIMS9A/PARCS CODE SYSTEM FOR POWER
Mohamed A. Elsawi
DENSITY CALCULATIONS OF THE WESTINGHOUSE AP1000
REACTOR
Feasibility of Nodal Equivalence Theory Using Functionalized Discontinuity
Woosong KimFactors
THE MULTIGROUP NEUTRONIC MODEL OF NUSTAR’S 3D CORE
Shaohong Zhang
CODE EGRETSIMULATION OF WATTS BAR INITIAL STARTUP TESTS WITH
Andrew T. Godfrey
CONTINUOUS ENERGY MONTE CARLO METHODS

1106738
111983711275241127710112777311278191128004

University of Ontario Institute ofTechnology

Woong Heo

11258171106273

IAPCM

Osaka university

11252541106272

Energy Engineering Service

Takanori Kitada

Implications of mesh refinement in lattice physics on BWR core analysis
Petri Forslundand nuclear design
Guimaraes
GENERATING MULTIGROUP DATA STOCHASTICALLY FOR A HIGHLYKevin John Connolly
HETEROGENEOUS VHTR PROBLEMREFERENCE SOLUTION FOR CORE PHYSICS
Radim Vocka
Development of A Cross Section Library Applicable to Various Reactor
Changho LeeTypes
Monte Carlo Analysis of Doppler Reactivity Coefficient for UO2 Pin Cell
Yasunobu Nagaya
Geometry
Assessment of the depletion capability in MPACT
Ang Zhu

1106220

Institute of Nuclear Engineering andScienceDepartment of Energy Engineering,
Sharif University of Technology

Tomas Bily

Westinghouse Electric Sweden ABGeorgia Institute of Technology
UJV Rez, a.s.
Argonne National Laboratory
Japan Atomic Energy Agency
University of Michigan
Czech Technical University in Prague,FNSPE, Department of Nuclear Reactors
University of Michigan
Core Physics Inc.
Khalifa University of Science, Technologyand Research
KAIST
Shanghai NuStar Nuclear PowerTechnology Co., Ltd.
Oak Ridge National Laboratory

Track2 Deterministic Transport Theory
103908210837971104084

The application and performance of ACMFD acceleration in 2D/3D full core
Zhiyong LI
MOC transport fuse method
A COUPLING METHOD OF SUBGROUP AND WAVELET EXPANSION
hongchun Wu
FOR THE RESONANCE PARAMETER CALCULATIONVerification of Ray Effect Elimination Module in the Transport Code ARES MengTeng CHEN

Track3 Monte Carlo Methods

China Nuclear Power Technology
Research Institute Shanghai Branch
Xi'an Jiaotong university
North China Electric Power University

1084078
1099694110186411024071104378

3-D MONTE CARLO NEUTRON-PHOTON TRANSPORT CODE JMCTAND ITS ALGORITHMS
NEUTRON CHANNELS SHIELD DESIGN ANALYSES OF KIPTNEUTRON SOURCE FACILITY
Development of a new convergence criterion for Monte Carlo coupledsimulation with thermal-hydraulics feedback
JCOGIN: A PARALLEL PROGRAMMING INFRASTRUCTURE FOR
MONTE CARLO PARTICLE TRANSPORT
Void transit time calculations by neutron noise of propagating perturbation

1105995

using complex-valued weight Monte Carlo
Simulating Fast Transients with Fuel Behavior Feedback using theSerpent~2 Monte Carlo code
STATUS OF MONTE CARLO CODE DEVELOPMENT AT UNIST1106262

Criticality Benchmarking of ANET Monte Carlo Code1105897

1123217
1228305

Effective diffusion Homogenization of Cross Sections with the Monte Carlomethod
ESTIMATING LOCAL IN- AND EX-CORE RESPONSES WITHIN MONTECARLO SOURCE ITERATION EIGENVALUE CALCULATIONSLi Deng

Institute of Applied Physics andComputational Mathematics

Zhaopeng Zhong

Argonne National Laboratory

XU WU

University of Illinois at UrbanaChampaign
Institute of Applied Physics andComputational Mathematics

Baoyin Zhang
Toshihiro Yamamoto

Kyoto University

Ville Valtavirta

VTT Technical Research Centre ofFinland
UNIST
NCSR 'Demokritos', Institute of Nuclear
and Radiological Sciences & Technology,
Energy & Safety

Hyunsuk Lee
Thalia Xenofontos
Dusan Calic

ZEL-EN d.o.o. / Institute Jozef Stefan

Kenneth William Burn

ENEA

Track4 Verification, Validation and Uncertainty Analysis
1067489

Review of Neutronic Assessments applied to small reactor core physics

Laurent Chabert

1084893

A new neutronics analysis code system for fast reactors

Toshikazu Takeda

AREVA TA
Research Institute of Nuclear
Engineering, University of Fukui

Ryan P. Kelley

University of Florida

Basma FOAD

University of Fukui

Go Chiba

Hokkaido university

Noriyuki Takemoto

Japan Atomic Energy Agency

Alexander Aures

Gesellschaft fur Anlagen- undReaktorsicherheit (GRS) mbH

Sebastien Lahaye

CEA/DEN/DANS/DM2S/SERMA/LLPR

Anton Peregudov

Institute for Physics and PowerEngineering

Jing Song
Gerald Rimpault

Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team
CEAMarco Pecchia

Paul Scherrer Institut (PSI)

Damar Wicaksono
Kyung-Hoon Lee
Jonatan Hejzlar

Ecole polytechnique federale deLausanne
Korea Atomic Energy Research InstituteUJV Rez, a.s.

Zhiwen Xu

TerraPower LLC

Xiaotong SHANG

Department of Engineering Physics
Tsinghua University,REAL team

1100864
110280911028101103703
11039241104067110431611046701104787110570911058251106101
1106250
11094321126078

Uncertainty Analysis of Delayed Neutron Fissile Material Assay Using a
Genetic Algorithm
IR APPROXIMATION FOR CALCULATING SENSITIVITY AND
UNCERTAINTY OF PWR CELLS BY TAKING ACCOUNT OF SELFSHIELDING EFFECTUNCERTAINTY QUANTIFICATION OF REACTOR KINETICSPARAMETERS USING JENDL-4.0 COVARIANCE DATASurvey on Effect of Crystal Texture of Beryllium on Total Cross-section toImprove Neutronic Evaluation in JMTR
UNCERTAINTY AND SENSITIVITY ANALYSIS FOR AN OECD/NEA
HTGR BENCHMARK WITH XSUSAFirst Verification and Validation Steps of MENDEL Release 1.0 CodeSystemApplication of the GRS method for estimation of uncertainties of LMFBR
type reactor physics parameters with taking into account macroscopicexperiments
BENCHMARKING OF PHOTON AND COUPLED NEUTRON ANDPHOTON PROCESS OF SUPERMC 2.0
In Depth Uncertainty Estimation of the Neutron Computational Tools
UPDATE OF THE PSI CRITICALITY SAFETY EVALUATIONMETHODOLOGY USING MCNPX2.7 AND ENDF/B-VII.1
VALIDATION OF TWO MONTE CARLO CODES FOR LWR BURNUPCALCULATIONSBENCHMARKING OF DeCART2D AGAINST CRITICAL EXPERIMENTS
AUTOMATED REACTOR RECORDS EVALUATION FRAMEWORKMonte Carlo Based Equilibrium Cycle Analysis of One-Dimensional Breedand Burn Benchmark Problem
Validation of the Monte Carlo Code RMC with a PWR MOXUOX CoreBenchmark1126219

INVERSE UNCERTAINTY QUANTIFICATION FOR NUCLEAR DATAASSESSMENTBassam A. Khuwaileh

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

1127444

ON HOW SENSITIVE THE CROSS-SECTION SENSITIVITYCALCULATIONS ARE TO PN ORDER APPROXIMATIONS

Ivan Alexander Kodeli

Jozef Stefan Institute1137713

VALIDATION OF NUSTAR’S PWR CORE ANALYSIS SYSTEM

Shaohong Zhang

Shanghai NuStar Nuclear PowerTechnology Co., Ltd.

Track6 Reactor Physics Experiments
1085573
1086633
1094641110432911048171105052

DETERMINATION OF THE 58Ni (n,p) 58Co REACTION CROSS
Zdena Lahodova
SECTION FOR BOTH GROUND AND ISOMERIC STATESInvestigation on Subcriticality Mwasurement Using Inherent NeutronTakeshi Shiozawa
Source in Nuclear Fuel
BENCHMARK CALCULATIONS OF SODIUM FAST CRITICALElena Mitenkova
EXPERIMENTS
CRITICAL EXPERIMENTS FOR BWR FUEL ASSEMBLIES WITH THEKenichi YOSHIOKA
CLUSTER OF GADOLINIA RODSTHE CALCULATION AND MEASUREMENT OF FAST NEUTRON REFLECTION IN THE VVER-1000 MOCK-UP MODEL PLACED IN THE LR-0 Michal Kostal
REACTORDEVELOPMENT OF REACTIVITY METER WITH NOVELTY NEUTRONShoichi TashiroSOURCE INTENSITY EVALUATION MODEL FOR BWR APPLICATION

1106265

Boron Carbide Neutron Screen for GRR-1 Neutron Spectrum Tailoring

Nefeli Chrysanthopoulou

1106850RESEARCH AND DEVELOPMENT ACTIVITIES FOR TRANSMUTATIONTakanori Sugawara
PHYSICS EXPERIMENTAL FACILITY IN J-PARC

Research Centre Rez Ltd.
Nagoya University
Nuclear Safety Institute of Russian
Academy of Sciences
TOSHIBA Corporation
Research Center Rez
Global Nuclear Fuel-Japan Co., Ltd.
NCSR "Democritos", INRaSTES,
Research Reactor Lab
JAEA

1108237
1126731112735511276881127907

Analysis of TCA criticality,βeff and βeff/l using CASMO-4 and CASMO-5
Measurement of Subcriticality using Delayed Neutron Source Combinedwith Pulsed Neutron Accelerator
ROSSI-α PARAMETER MEASUREMENT OF DALAT NUCLEARREACTOR BY ANALYSIS OF CROSS POWER SPECTRAL DENSITYOBTAINED FROM 2 ION CHAMBERSANALYSIS OF INTEGRAL EXPERIMENT FOR THORIUM FUEL CYCLEAT KYOTO UNIVERSITY CRITICAL ASSEMBLYSTUDIES OF POTENTIAL FOR CONVERSION OF THE IDAHONATIONAL LABORATORY TREAT TRANSIENT TEST REACTOR TOLOW-ENRICHMENT FUEL

shigeaki aoki
Tsuyoshi Misawa

MNFKyoto Univerisity Research ReactorInstituteTuan Minh Nguyen

Nuclear Research Institute, Dalat

Yoshiyuki Takahashi

Research Reactor Institute, Kyoto
UniversityDimitrios C.
Kontogeorgakos

Idaho National Laboratory

Benjamin A. Lindley

University of Cambridge

Toshikazu Takeda

Research Institute of Nuclear
Engineering, University of Fukui
Xi'an Jiaotong University
KEPCO International Nuclear Graduate
School (KINGS)
KAISTNational Autonomous University ofMexico
Ulsan National Institute of Science andTechnology
Center Nuclear Technology NRC“Kurchatov Institute”

Track7 Reactor Concepts and Designs

1090085

Axially Heterogeneous Thorium Fuel Designs for Transuranic Burning in
Reduced-moderation BWRs
Method Development and Reactor Physics Data Evaluation for Improving
Prediction Accuracy of Fast Reactors' Minor Actinides Transmutation

1094719

Performance
A long life sodium cooled fast reactor concept with radial shuffling

Zhipeng Li

1099566

Critical Boron Concentration Reduction Method in a Core DesignChang Joo Hah

1104423

BigT - A New Burnable Absorber Concept for PWR

1104448PRELIMINARY DESIGN OF A SPHERICAL BREED/BURN REACTOR

Mohd-Syukri Yahya
Elias Yammir Garcia-

1104495

OPTIMIZATION OF ULTRA-LONG CYCLE FAST REACTOR CORE

1068764

110618911270321127409

THE MAIN CHARACTERISTICS OF THE EVOLUTION PROJECT VVERS WITH SPECTRUM SHIFT REGULATIONALTERNATIVE CORES FOR A MULTIPURPOSE EXPERIMENTALSODIUM-COOLED FAST REACTOR WITH U-ZR FUELPRELIMINARY EVALUATION OF COOLANT VOID REACTIVITY OF A
RE-ENTRANT CHANNEL PRESSURE-TUBE SUPERCRITICAL WATER
COOLED REACTOR

Cervantes
Taewoo Tak
Pavel S Teplov
Tae-Yang Noh

Kyung Hee University

Peter Schwanke

UOIT

Wenhuai Li

China Nuclear Power Technology
Research Institute
Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team

Track8 Reactor Operation and Safety
1092230
11059901127006

The Development of SOPHORA -PHYSOR2014
DEVELOPMENT OF RISK MONITOR RISKANGEL FOR RISKINFORMED APPLICATIONS IN NUCLEAR POWER PLANTS
STEADY-STATE SUBCHANNEL ANALYSIS OF PARTIALLY BLOCKEDCOOLANT CHANNELS IN A POOL-TYPE TRIGA REACTORFang Wang
Jean C Ragusa

Texas A&M University

Track9 Transient and Safety Analysis
10847171094483
1101709110376011046191105969

IMPROVEMENT OF SPACE-TIME KINETICS CAPABILITY IN THE
Maxime Guyot
SNATCH SOLVER AND COMPARISON TO KIN3D/PARTISN RESULTS
Different methods to model the MSLB without primary cooling pumps using
Benoit NORMANDHEMERA V1 system codes
SOME RESULTS OF STUDYING OF SPATIAL KINETICS IN FASTIrina Panova
REACTORSATUCHA 2 OBLIQUELY INSERTED CONTROL RODS RELAP5Raul Gonzalez Gonzalez
3D/NESTLE MODELThe SIMMER/PARTISN Capability for Severe Accident Analyses
Marco Marchetti
Development of a high-fidelity Monte Carlo thermal-hydraulics coupled
Miriam Daeubler
code system Serpent/SUBCHANFLOW - First results

CEA Cadarache
IRSN
Nuclear Safety Institute of Russian
Academy of Sciences
San Piero a Grado Nuclear ResearchGroup (GRNSPG)
Karlsruhe Institute of Technology
Karlsruhe Institute of Technology

Track10 Nuclear Data
10686711081806
1105105
1106244ANALYSIS OF RADIOACTIVITY RATIOS OF FISSION PRODUCTNUCLIDES DEPOSITED TO SOIL IN FUKUSHIMA DAI-ICHI NUCLEARPOWER PLANT ACCIDENTEXPERIMENTAL UNCERTAINTY ESTIMATION IN PROFIL ANDPROFIL-2 SAMPLE IRRADIATION EXPERIMENTS IN PHENIXUPDATED MULTI-GROUP CROSS SECTIONS OF MINOR ACTINIDESWITH IMPROVED RESONANCE TREATMENTPARAMETERIZED REPRESENTATION OF MACROSCOPIC CROSSSECTION FOR PWR REACTOR CONSIDERING WITH 12 BURNABLEABSORBER FUEL RODS IN THE FUEL ELEMENTGo ChibaHokkaido universityEdwin Privas

CEA CadaracheMuhammad SohailKyung Hee UniversityJoao Claudio Batista Fiel

Military Institute of Engineering

Track11 Research Reactors and Spallation Sources
1087763
11061211126142112773811280781128332JULES HOROWITZ REACTOR.FRANCE EXPERIMENTAL LOOP
Stephane GAILLOT
DEVELOPMENT ACCORDING OPTIMIZED IRRADIATION PROCESS.
Preliminary Neutronic Design for the Conceptual Fluid Granular Spallation
Jinyang Li
TargetAccumulation of tritium in beryllium slab under neutron irradiation

Sandybek Kunakov

A PRELIMINARY STUDY OF AN IMPROVED AREA METHOD, ADAPTED
Paolo Saracco
TO SHORT TIME TRANSIENTS IN SUB-CRITICAL SYSTEMSPRELIMINARY NEUTRONICS ANALYSIS OF A SPALLATION TARGETBin Wu
FOR TRANSMUTATIONPRELIMINARY OPTIMIZATION ANALYSIS OF THE RADIATION
Qi Yang
SHIELDING OF THE CHINA LEAD-BASED RESEARCH REACTORCEA CadaracheInstitute of Modern Physics, Chinese
Academy of Sciences
Physico-Technical faculty, Al-Farabi
Kazakh National University
INFN
Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team
Institute of Nuclear Energy Safety
Technology, CAS ・ FDS Team1233696

Preliminary Analysis of Radioactive Source Term for Normal Operation ofChina Lead-Alloy Cooled Research Reactor

Tongqiang Dang

Institute of Nuclear Energy Safety
Technology, CAS ・ FDS TeamGennady Genrikhovich
Kulikov

National Research Nuclear University
Moscow Engineering Physics Institutealberto talamo

Argonne National Laboratory

Evgeny Seleznev

Nuclear Safety Institute of Russian
Academy of Sciences

Kazuki Kora

Kyushu University

Baptiste MOUGINOT

CNRSRobert Ryan Flanagan

University of Texas at Austin

Hiroshi Akie

Japan Atomic Energy Agency

Track12 Fuel Cycle and Actinide Management
1092834
1100837
11017271105292110580211063241106382

FUSION HYBRIDS FOR GENERATION OF ADVANCED(231Pa+232U+233U+234U)-FUEL IN CLOSED (U-Pu-Th)-FUEL CYCLE
233U FUEL PRODUCTION AND 30-YEAR UTILIZATION WITHOUTREPROCESSING AND REFUELLING USING HEAVY WATER COOLANT
UNCERTAINTY ANALYSIS FOR FUEL FLUX CALCULATIONS OF FAST
REACTORS WITH EXTERNAL FUEL CYCLESTUDY ON TRANSMUTATION AND STORAGE OF LLFP USING A
HIGH-TEMPERATURE GAS-COOLED REACTOR
CORE LIBRARY FOR ADVANCED SCENARIO SIMULATION, C.L.A.S.S.
: PRINCIPLE & APPLICATION.
Fuel Composition Generation Techniques of Nuclear Fuel Cycle
Simulators
Core Burnup Calculation of Uranium Rock-like Oxide Fuel PWR for SpentFuel Composition Estimation

Track14 Education in Reactor Physics
1085651

MULTI-COLLISION THEORY FOR EDUCATED PEDESTRIANSPaolo Picca

1073641

Virtual Labs on unique experimental equipment
PINSPEC: A MONTE CARLO CODE FOR PIN CELL SPECTRALIvan S Saldikov

Department of System and Industrial
Engineering, the University of Arizona
NRNU MEPHI

William BoydMITTatjana Jevremovic

The University of Utah

Taro Ueki

Japan Atomic Energy Agency

11048021127873

CALCULATIONS FOR EDUCATIONAL APPLICATIONSUNIQUE APPROACHES IN EMPHASIZING THE ROLE OF REACTORLABORATORIES AND FACILITIES FOR TRAINING AND EDUCATIONOF FUTURE NUCLEAR ENGINEERS ‘WITHOUT THE BORDERS’

Track15 Research Related to Fukushima Accident
1102869

Spatial Correlation Modeling of Macroscopic Cross Section withWeierstrass Function

SpecialSession2 Reactor Physics and Criticality Safety Activities in OECD/NEA Working Party
110984811211531195260

EFFECTS OF NUCLEAR DATA LIBRARY AND ULTRA-FINE GROUPCALCULATION FOR LARGE SIZE SODIUM-COOLED FAST REACTOR

Teruhiko KUGO

Japan Atomic Energy Agency

Mari Marianne Uematsu

JAEA

Chris Perfetti

Oak Ridge National Laboratory

OECD BENCHMARKSEVALUATION OF OECD/NEA/WPRS BENCHMARK ON MEDIUM SIZEMETALLIC CORE SFR BY DETERMINISTIC CODE SYSTEM: MARBLE
AND MONTE CARLO CODE: MVP
Quantifying the Effect of Undersampling in Monte Carlo Simulations using
SCALE

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

Research on SCWR Core Characteristics Utilizing Pin-Wise Neutronics
Thermal-Hydraulic Coupling Method

Qian HONG

Shanghai Nuclear
Engineering Design & Research InstituteSpecialSession6 Nuclear Criticality Safety of Fuel Debris
1087714

POST-ACCIDENT DEFUELING PROCEDURE AND ITS CRITICALITY
SAFETY EVALUATION OF THE FUKUSHIMA-DAIICHI NUCLEAR
POWER PLANTS

Naoyuki TakakiTOKYO CITY UNIVERSITY