list of abstracts
 
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log #
Main Author (Org)
Title and authors
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01
Deterministic Transport Theory Methods
001
Alain Mazzolo
CEA
Properties of neutron trajectories in bounded domains
A. Mazzolo

002
Makoto Tsuiki
IFE, OECD Halden Reactor
Project
A variational nodal transport method for pressurized water reactor core calculations
M. Tsuiki

003
Jose Ignacio Duo
Pennsylvania State University
Global Error Analysis of the Spatial Approximation in Discrete Ordinates Methods
J.I. Duo

 
004
Romain Le Tellier
Ecole Polytechnique de Montreal
Spectral Analysis of an Algebraic Collapsing Acceleration for the Characteristics Method
R. Le Tellier

 
005
Toshikazu Takeda
Osaka University
Development of 3-D Detailed FBR Core Calculation Method Based on Method of Characteristics
T. Takeda

 
006
Yousry Y. Azmy
Penn State University
Properties of the Sn-Equivalent Integral Transport Operator in Heterogeneous Slabs
Y.Y. Azmy

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007

Barry Ganapol
University of Arizona

Mining the Multugroup Discrete-Ordinates for High Quality Solutions
B. Ganapol

075
Gregory Davidson
University of Michigan
Finite Element Transport Using Wachspress Rational Basis Functions on Quadrilaterals in Diffusive Regions
G. Davidson

085
Guy Marleau
École Polytechnique de Montréal
Multigroup Adjoint Transport Solution Using the Method of Cyclic Characteristics
G. Marleau

 
086
Patrick S. Brantley
Lawrence Livermore National Laboratory
Spatially Continuous Mixed Simplified P_2-P_1 Solutions for Multidimensional Geometries
P.S. Brantley

 
090
Emiliano Masiello
CEA
A discontinuous finite element method for transport calculations in two-dimensional geometries with heterogeneous cells
E. Masiello, R. Sanchez

 
105
Nam Zin Cho
Korea Advanced Institute of Science and Technology
CMADR Acceleration and Its Convergence Analysis of the Method of Characteristics for Neutron Transport Problems
N.Z. Cho

 
115
Patrick S. Brantley
Lawrence Livermore National Laboratory
Angularly Adaptive P1-Double P0 Diffusion Solutions of Non-Equilibrium Grey Radiative Transfer Problems in Planar Geometry
P.S. Brantley

 
125
Viktor F. Tsibulskiy
RRC Kurchatov Institute
The correction on the limited size of fixed zones to equations of a characteristics method
V.F. Tsibulskiy

 
141
Felix C. Difilippo
Oak Ridge National Laboratory
Applications of the Two-Dimensional Burnup Analysis Capabilities of SCALE-5
F.C. Difilippo

 
179
Laurent Plagne
EDF R&D
Generic Programming for Deterministic Neutron Transport Codes
L. Plagne

 
180
Gert Van den Eynde
SCK-CEN
The Boundary Sources Method with arbitrary order anisotropic scattering
G. Van den Eynde

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186
M. Erradi
Mohammed V University
Analysis of a MOX-UO2 Interface by the Methods of Characteristics
M. Erradi

193
Serge Van Criekingen
CEA
Generalizing Raviart-Thomas Elements to PN Transport
S. Van Criekingen

201
Shunichi Wakana
Research laboratory for nuclear reactors, Tokyo Institute of technology

Development of Lattice Program to Analyze HTR
S. Wakana

 
203
Ricardo Barros
Universidade do Estado do Rio de Janeiro
Discrete ordinates Albedo Boundary Conditions for Indented Nuclear Reactor Global Calculations
R. Barros

 
204
Boris Dmitrievich Abramov
IPPE
On positive solvability of generalized askew's coarse-mesh method
B.D. Abramov

 
215
Julien Cartier
CEA/DAM
Mixed and Hybrid Finite Element Method for the Transport Equation
Julien Cartier

 
236
Ruben Panta Pazos
Universidade de Santa Cruz do Sul
Variational Formulation for Nodal Methods in Transport Theory
R. Panta Pazos

 
253
Marvin L. Adams
Texas A&M University
Adaptive Discrete-Ordinates Algorithms and Strategies
M. L. Adams

 
269
Marvin L. Adams
Texas A&M University
Behavior of Continuous Finite Element Discretizations of the Slab-Geometry Transport Equation
M. L. Adams

 
283
James Holloway
Department of Nuclear Engineering and Radiological Sciences, College of Engineering, University of Michigan

An Implicit Riemann Solver for the Time-Dependent Pn Equations
R. McClarren, J.P. Holloway

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284
James Holloway
Department of Nuclear Engineering and Radiological Sciences, College of Engineering, University of Michigan

Establishing an Asymptotic Diffusion Limit for Riemann Solvers on the Time-Dependent Pn Equations
R. McClarren, J.P. Holloway

290
Edward Larsen
University of Michigan
A Monte Carlo-Deterministic Method for Global Transport Calculations
A.B. Wollaber, E.W. Larsen

293
D. Scott Lucas
Idaho National Laboratory
Core Modeling of the Advanced Test Reactor with the Attila Code
D. S. Lucas

 
298
D. Lathouwers
Delft University of Technology
Calculation of the dominant time-eigenvalues in neutron transport
D. Lathouwers

 
307
Brian David Lansrud
Los Alamos National Laboratory
A Spatial Multigrid Iterative Method for Two-Dimensional Discrete-Ordinates Transport Problems
B. D. Lansrud

 
316
A-M Baudron
CEA

MINOS : A SPN Solver for Core Calculation in the DESCARTES System
A-M Baudron, J-J Lautard

 
327
Taewan Noh
Hongik University
Dual Diffusion Synthetic Acceleration for Self-Adjoint Angular Flux Transport Equation
T. Noh

 
340
Marvin L. Adams
Texas A&M University
New spatial discretization methods for transport on unstructured grids
H.G. Stone, M.L. Adams

 
02
Radiative Transfer and Biomedical Applications
 
008
Jean Gouriou
CEA/DRT/LMD
Monte Carlo Simulations used to Establish a French National HDR Brachyterapy Reference
J. Gouriou, G. Douysset

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009
Gennady N. Malyshkin
Russian Federal Nuclear Center - VNIITF
Development of a Computer Model of the NG-12I Neutron Generator for a Fast Neutron Therapy Planning System
G. N. Malyshkin

010
Nikolay Borisov
Institute of Biophysics
An Experimental Test on Large Animals of MCNP Application for Whole Body Counting
N. Borisov

196
Edwige Buffard
FEMTO-St/Dpt CREST
Theoretical determination of the dose perturbations caused by a hip prosthesis during a pelvic irradiation
E. Buffard

 
270
Joel Herault
Centre A.Lacassagne
MCNPX proton transport simulations for a therapy set-up
J.Hérault, N.Iborra, B.Serrano, P.Chauvel

 
304
Benjamin Serrano
LPES/CRESA
Monte Carlo simulation of a medical linear accelerator for Radiotherapy use
B. Serrano, E. Franchisseur, J. Hérault, A. Hachem, R.J. Bensadoun

 
03
Reactor Core Methods
 
011
Roger de Wouters
Tractebel Engineering
Calculation of Ex-core Detector Responses
R. de Wouters

 
012
Marcos Pimenta de Abreu
IPRJ/UERJ
Leakage conditions for multigroup anisotropic scattering models of thermal nuclear reactors
M.P. de Abreu

 
013
Augusto Gandini
University of Rome
The Generalized Mode Method (GMM) for Reactor Calculations
A. Gandini

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014
Wei Shen
AECL
Recent Progress in the Development of Cross-Section Models in the Reactor Fuelling Simulation Program (RFSP)
W. Shen

142
Jasmina Vujic
University of California, Department of Nuclear
New Monte Carlo Procedures and Cross Section Libraries for Fuel Burnup in Innovative Reactor Designs
J. Vujic

153
Jean Ragusa
Texas A&M University
A technique for reflector homogenization
J. Ragusa

 
175
Mike Ferrer
Penn State University
Error Analysis of the Multidimensional Nodal Integral Method for Solving the Neutron Diffusion Equation
M. Ferrer

 
194
José M. Aragonés
Universidad Politecnica de Madrid
The Analytic Coarse-Mesh Finite-Difference Method for Multi-Group and Multi-Dimensional Diffusion Calculations
José M. Aragonés

 
232
Tamer Bahadir
Studsvik Scandpower, Inc.
SIMULATE-4 Multi-group Nodal Code with Microscopic Depletion Model
T. Bahadir

 
273
Jean-Jacques Lautard
CEA
A component Mode Synthesis Method for D Cell by Cell SPN Core Calcualtion Using the Mixed Dual Finite Element Solver MINOS

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294
Djordje Tomasevic
NECSA
A Multigroup Analytic Nodal Method in Hexagonal-z Geometry
D.Tomasevic

310
Dmitriy Anistratov
North Carolina State University

Strongly Consistent Coarse-Mesh Discretization Method for the Multidimensional Low-Order Quasidiffusion Equations Based on the Equations for Subregion Averaged Scalar Fluxes
D. Anistratov

04
Monte Carlo Methods
 
015
Thomas A. Brunner
Sandia National Labboratories
Comparison of Four Parallel Algorithms for Domain Decomposed Implicit Monte Carlo
T. A. Brunner

 
016
G. Anil Kumar
Indian Institute of Technology
Monte Carlo simulation of the response of some ionization chambers
G. Anil Kumar

 
017
Wang Ruihong
Institute of Applied Physics and Computational
Monte Carlo Statistical Estimation Method for Time-Absorption(Alpha ) Eigenvalue Calculation
Wang Ruihong, Pei Lucheng, Ying yangjun

 
018
M. Nima
Damghan University of Basic Sciences
Monte Carlo Modelling of Light Collection in Scintillators with Different Surface Coverings
N GHAL-EH and R KOOHI-FAYEGH

 
019

Jeffery D. Densmore
Los Alamos National Laboratory

Discrete Diffusion Monte Carlo for Grey Implicit Monte Carlo Simulations
J.D. Densmore, T.J. Urbatsch, T.M. Evans, M.W. Buksas

 
020

Gérald Samba
CEA

Asymptotic Diffusion Limit of the Symbolic Monte-Carlo Method for the Transport Equation
J-F Clouet, G. Samba

 
081
Toshihisa Yamamoto
Osaka University
Acceleration of Monte Carlo Solution by Conjugate Gradient Method
T. Yamamoto

 
089
John S. Hendricks
LANL
The MCNPX Radiation Transport Code
J.S. Hendricks

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100
Evgeny Ivanov
RRC Kurchatov Institute
Error propagation in Monte-Carlo burn-up calculations
E. Ivanov

107
Yen-Wan Hsueh Liu
National Tsing Hua University
Dose Calculation for the BNCT Treatment Room at THOR by the "TORT-coupled MCNP" Technique
U. Y. Chen, Y-W H. Liu

124
Viktor F. Tsibulskiy
RRC Kurchatov Institute
Monte-Carlo method in UNK complex
V.D. Davidenko ,V.F. Tsibulskiy

 
134
Winfried Zwermann
Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH
Monte Carlo and Deterministic Transport Methods Applied for Reactor Calculations
W. Zwermann

 
160
Anatoli Tsiboulia
Institute of Physics and Power Engineering
Algorithm of Sensitivity keff to Group Cross Sections Calculation Using Monte Carlo Method and Features of Its Implementation in the MMKKENO Code
A. Tsiboulia

 
174
Chanatip Tippayakul
Pennsylvania State University
Multi-Group Diffusion Cross Section Generation using the Monte Carlo Method
C. Tippayakul

 
192
Jaakko LeppŠnen
VTT Technical Research Centre of Finland
A New Assembly-level Monte Carlo Neutron Transport Code for Reactor Physics Calculations
J. LeppŠnen

 
198
Beomseok Han
Seoul National University
Monte Carlo Analysis of Be-Free He-cooled Solid Breeder Fusion Blanket
B. Han

 
207
Kohei Iwanaga
Tokyo institute of technology

Treatment of kinetics of subcritical system
Kohei Iwanaga

 
223
Stelios Zimeras
University of the Aegean
Markov random fields based models and MCMC methods: theory and applications in medical image biology
S. Zimeras

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229
S.C. van der Marck
NRG
Temperature interpolation in MCNP
S.C. van der Marck

244
Anna G.Tsarina
Obninsk State Technical University
Monte Carlo Approach to Radiation Induced Defects Growth Problem
A. G.Tsarina

247
Nuria Garcia-Herranz
Polytechnic University of Madrid
Applicability of the MCNP-ACAB System to Inventory Prediction in High-Burnup Fuels: Sensitivity/Uncertainty Estimates
N. Garcia-Herranz

 
255
Victor I. Belousov
Obninsk State Technical University
The Electron Transport Problem Sampling by Monte Carlo Individual Collision Technique
V. I. Belousov

 
278
J. Eduard Hoogenboom
Delft University of Technology
Optimization of path length stretching in Monte Carlo calculations for non-leakage problems
J. E. Hoogenboom

 
285
Gumersindo Verdś
Universidad Politécnica de Valencia
Monte Carlo Simulation for dosimetry Studies of a Heterogeneous Water Phantom
B. Juste, R. Miro, S. Gallardo, A. Santos, G. Verdś, S. Calzada

 
05
Computational Methods on Advanced Computers

 

 

021
Roberto Orsi
ENEA
BOT3P: A Mesh Generation Software Package for Transport Analysis with Deterministic and Monte Carlo Codes
R. Orsi

022
Richard J. Procassini
Lawrence Livermore National Laboratory
Load Balancing of Parallel Monte Carlo Transport Calsculations
R.J. Procassini, M. O'Brien, J. Taylor

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111
Jose E. Roman
Universidad Politecnica de Valencia
Accurately Bi-orthogonal Direct and Adjoint Lambda Modes via Two-sided Eigensolvers
J.E. Roman, V. Vidal, G. Verdu

158
Kenji Yokoyama
Japan Nuclear Cycle Development Institute
Restructuring of Burnup Sensitivity Analysis Code System By Using an Object-Oriented Design Approach
K. Yokoyama

172
Victor M. Garc’a
Polytechnic University of Valencia
Sequential and Parallel resolution of the 3D Transient Neutron Diffusion Equation
V. M. Garc’a

 
181
Mohamed Dahmani
Ecole Polytechnique de Montreal
Hybrid shared/distributed parallelism for 3D Characteristics Transport Solvers
M. Dahmani

 
250
Najib Guessous
Ecole Normale Supérieure de Fès
Three efficients higher order analytical nodal methods for the numerical solution of the multigroup neutron diffusion equations
N. Guessous

 
292
David Weber
Argonne National Laboratory
The Numerical Nuclear Reactor -- A High Fidelity, Integrated Neutronic, Thermal-Hydraulic and Thermo-Mechanical Code
D.P. Weber, T. Sofu, P. Pfieffer, W.S. Yang, K.S. Kim, T.H. Chun, T. Downar, J. Thomas, Z. Zhong, H.G. Joo, C.H. Kim

 
06
Physics and Computational Methods
for Advanced Reactor Concepts
 
023
Chang Keun Jo
Korea Atomic Energy Research
A Feasibility Study on the Thorium Oxide and Nitride Fuels in a CANDU Reactor
C.K. Jo, C.J. Park, C.J. Jeong

 
024
Chang Joon Jeong
Korea Atomic Energy Research
A Dry Process Fuel Cycle Analysis for a Sodium-Cooled Fast Reactor
Chang Joon Jeong, Gyu Hong Roh and Hangbok Choi

 
025
Yuxiang Gu
University of Illinois at Urbana-Champaign
A Hybrid Method for Multi-group Neutron Diffusion Equations in Arbitrary geometry
Y. Gu, Rizwan-uddin

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067
Gérald Rimpault
CEA
Gas Cooled Fast Reactor Benchmarks for JNC and CEA Neutronic Tools Assessment
G. Rimpault,Kazuteru SUGINO and Hideyuki HAYASHI

074
Bertrand Stepnik
Framatome-ANP
Qualification of the HTR neutronics tool NEPTHIS
B. Stepnik, C. Cavalier, C. Trakas, S. Thareau

106
Sandra Dulla
Politecnico di Torino
Analysis of the Role of Fluid-Dynamic Phenomena in the Neutronics of Molten-Salt Systems
S. Dulla, P. Ravetto

 
118
Nicolas Huot
CEA
The JHR Neutronic Calculation Scheme Based on the Method of Characteristics
N. Huot, A. Aggery, D. Blancher, C.D'Aletto, J. Di Salvo, C. Doderlein, P. Sireta, G. Willermoz

 
136
Frédéric Damian
CEA
VHTR neutronic calculation scheme: validation elements using MCNP and TRIPOLI4 Monte-Carlo codes
F. Damian

 
137
Xavier Raepsaet
CEA
TRIPOLI-4.3 Modeling of Pebble Bed Configuration of ASTRA Critical Facility of the Kurchatov Institute
X. Raepsaet

 
138
Imed Limaiem
CEA
VHTR Core Modelling: Coupling between Neutronic and Thermal-hydraulics
I. Limaiem

 
139
Boris P. Kochurov
Inst. of Theor.&Experim. Physics
The Influence of Inter Cell Currents on Burn up Characteristics of VVER Type Light Water Reactor
B.P. Kochurov

 
159
Anatoly D. Klimov
Research and Development Institute of Power Engineering

System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels
A.D. Klimov

 
166
Gray S. Chang
Idaho National Laboratory
HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis
G. S. Chang

 
188
Yunlin Xu
Purdue University
Optimal Perturbation Size for Matrix Free Newton /Krylov Methods
Y. Xu; T. J. Downar

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189
Yunlin Xu
Purdue University
The Implementation of Matrix Free Newton /Krylov Methods Based on a Fixed Point Iteration
Y. Xu; T. J. Downar

231
Tetsuo Sawada
Tokyo Inst. of Technology/RLNR
Percolation-Fission Model Calculations of Spallation Residues in the 1-GeV Proton-Induced Reaction on 208Pb Targets
T. Sawada

282
Hans D. Gougar
Idaho National Laboratory
Validation of the Neutronic Solver within the PEBBED Code
H. Cougar, A. Ougouag, W. Terry

 
301
Christos Trakas
FRAMATOME-ANP
Optimization of the Deep Burner-Modular Helium Reactor (DB-MHR) concept for actinide incineration
C. Trakas, G.B. Bruna

 
311
Andrei Rineiski
Forschungszentrum Karlsruhe
Kinetics and cross-section developments for analyses of innovative reactor transmutation concepts with SIMMER
A. Rineiski

 
322
Alexandre Mourogov
EDF R&D
The Lead-cooled Fast Reactor BREST-300: Analysis with Sensitivity Method
V. Smirnov, V. Orlov, A. Mougorov, D. Lecarpentier, T. Ivanova

 
326
Francesco Cadini
Polytechnic of Milan

An adaptive model-free fuzzy controller
M. Marseguerra, E. Zio, F. Cadini

 
07
Perturbation Theory and Variational Methods
 
026
Jeffrey A. Favorite
Los Alamos National Laboratory
Variational estimates of ratios of neutron-induced gamma lines
J.A. Favorite

 
088
Patrick Erhard
EDF R&D and CERFACS
Use of Data Assimilation techniques in neutronics: benefits and first results
P. Erhard, G. Gacon, S. Buis, S. Massart

 
129
Tatiana Ivanova
Institut de Physique Nucléaire Orsay/CNRS
Comparison of M/C and SN Eigenvalue Sensitivity Methods by the Analysis of VENUS-2 Benchmark and Thorium Molten Salt Reactor
T. Ivanova

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306
A. Bidaud
IPN Orsay
Reaction rate Tatio Sensitivity and Uncertainty Analysis Using Generalised Perturbation Theory : Application to VENUS2 Benchmark
A. Bidaud

08
Reactor Analysis
027
Ren-Tai Chiang
GE Energy
Impacts of Exposure, Xenon and Feedwater Temperature on BWR Stability Behavior
R.T. Chiang, A.K. Chung, R. D. McCord

028
Aldo Dall'Osso
Framatome ANP
Core monitoring and prediction with a 3D on line model
A. Dall'Osso

 
072
Francesco Ganda
University of California, Berkeley
A Simplified Method for Multi-Batch PWR Core Analysis Based on SAS2H Unit Cell Calculations
F. Ganda, E. Greenspan

 
082
Elisabeth Varin
Ecole Polytechnique de Montréal
Parallel DRAGON/DONJON calculations for consistent history-based method
E. Varin, R. Chambon, G. Marleau

 
097
Tadashi Ushio
Nuclear Engineering, Ltd.
Development of Advanced Neutronics Design System of Next Generation, AEGIS
Tadashi Ushio, Naoki Sugimura and Masaaki Mori

 
109
Hyun Chul Lee
KAERI - Korea Atomic Energy Research Institute
Fourier Convergence Analysis Applied to Neutron Diffusion Eigenvalue Problem
Hyun Chul Lee, Jae Man Noh, Hyung Kook Joo, Deokjung Lee, Thomas J. Downar

 
127
Hakim Ferroukhi
Laboratory for Reactor Physics and Systems Behaviour
Evaluation of the PWR REA Pulse Width for Realistic UO2/MOX Cores
H. Ferroukhi, M.A. Zimmermann

 
187
M. Erradi
Mohammed V University
Monte Carlo Analysis of Experiments on the Reactivity Temperature Coefficient for UO2 and MOX Light Water Moderated Lattices
M. Erradi

 
219
Benoit Lance
Belgonucleaire
WIMS-MCNP methodology for gamma heating calculations
Benoit Lance, Paul Van den Hende, Adolfo La Fuente

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230
Tomasz Kozlowski
Purdue University
Full-Core Three-Dimensional Pin-by-pin Rod Ejection Simulation Based on Multi-group SP3 Approximation in PARCS
T. Kozlowski

264
J.C. Kuijper
NRG
Propagation of Cross Section Uncertainties in Combined Monte Carlo Neutronics and Burnup Calculations
J.C. Kuijper

275
Jean Koclas
IGN Ecole Polytechnique de Montreal
Analysis of Power Variations Due to a Loss of Moderator in CANDU-6
Y. Fan, J. Koclas

 
300
Manfred Clemente
Gesellschaft fźr Anlagen- und Reaktorsicherheit GRS - mbH
Analysis of the "Loss of Feedwater" Scenario for RBMK-1500 by the Coupled Code ATHLET-QUABOX/CUBBOX Taking into Acount Late Reactor Trip Signals and ATWS Conditions
M. Clemente, L. Kuriene, S. Langenbuch, J.P. Weber

 
320
L. Erradi
Mohammed V University

Preliminary Analysis of the 2 MW Moroccan Triga MARK II Research Reactor
L. Erradi, E. Chakir, T. Elbardouni, M. Tabet, A. Htet, H. Boukhal, A. Chetaine

 
09
Lattice Physics Methods
 
029
Dave Knott
GE Nuclear Energy / Global Nuclear Fuel - Americas
Description and Validation of the GE Lattice Physics Code LANCER02
D. Knott, V.W. Mills, E. Wehlage

 
030
Albert Gu
AREVA Framatome-ANP
BWR Lattice Perturbation Theory in BALO
Albert G. Gu, Ralph G. Grummer

 
066
Akio Yamamoto
Nagoya University
Non-equidistant Ray Tracing for the Method of Characteristics
A.Yamamoto, N. Sugimura, T. Ushio, M. Tabuchi

 
068
Gérald Rimpault
CEA
Enhanced Streaming Algorithms for GCFR Core Depressurization Effect
Gérald Rimpault, Danièle Plisson-Rieunier, Cyrille De Saint Jean, Jean TommasiI

 
112
Kang-Seog Kim
Korea Atomic Energy Research Institute
Forced Structured Coarse Mesh Finite Difference Method for the Characteristics Nethod Applied to the Complex Geometry
K.S. Kim

 
162
Andreas Pautz
Framatome ANP GmbH
Application of the Discrete Ordinates Code DORT to Fuel Lattices: Benchmark Results and Improved Cross Section Generation Capabilities
A. Pautz

 
268
Karthikeyan Ramamoorthy
Ecole Polytechnique de Montreal
Effects of Advanced self shielding and burnup models on fuel temperature coefficient estimation for CANDU-6 assembly
K. Ramamoorthy

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312
Alain Santamarina
CEA

Determination of the optimized SHEM mesh for neutron transport calculations
A. Santamarina

10
Neutron Kinetics and Dynamics
031
Carl Sunde
Chalmers University of Technology
Investigation of the neutron noise induced by shell-mode core-barrel vibrations in a reflected reactor
C. Sunde, D. Demazière, I. Pazsit

032
Damian Ginestar
Universidad Politecnica de Valencia
A nodal collocation method for the PL approximation of the transport equation
M. Capilla, C.F. Talavera, D. Ginestar, G. Verdu

 
033

Christophe Demazière
Chalmers University of Technology, Department of Reactor Physics

Investigation of the frequency-dependence of the MTC noise estimator
C. Demazière

 
065
E.A. Ivanov
RRC Kurchatov Institute
Soft model of accidental self-sustaining chain reaction in nuclear facility
E.A. Ivanov, S.V. Tchernov

 
079
Bruno Merk
Forschungszentrum Karlsruhe
Multi Scale Approximation Functions for the P1 and P3 Approximation of the Time Dependent Boltzmann Equation
B. Merk, D. G. Cacuci

 
087
J. B. Doshi
Indian Institute of Technology, Bombay
Analysis of transients in pressurised heavy water reactors using coupled neutronics-thermal-hydraulics model
J.B. Doshi, Tirthankar Roy, and P.K. Baburajan

 
130
B.D. Abramov
IPPE
On Reactur Kinetics Simulation with Different Sets of Delayed neutron Data
B.D. Abramov

 
155
Thomas Downar
nuclear engineering
Analysis of the OECD/NEA Ringhalls Instability Benchmark with TRACE/PARCS
T. Downar

 
216
Cristian Rabiti
IKET
Time step control for solving the transient even-parity neutron transport equation
Cristian Rabiti, G. Lohnert, W. Maschek, A. Rineiski

 
267
Kengo Hashimoto
Kinki University
Derivation of an Improved Correction Formula for Dynamic Rod Worth Measurement
K. Hashimoto

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308
Akio Yamamoto
Nagoya University

Efficient Calculation of Time-Dependent Neutron Transport Equation Using the Constrained Interpolated Profile (CIP) Method
S. Khano, T. Endo, A. Yamamoto, Y. Yamane, Y. Kitamura

11
Criticality and Safety Analysis
034
Mihaly Makai
KFKI Atomic Energy Res. Inst.
Uncertainty Estimation of Neutronics Codes
L. P‡l and M. Makai

165
Alexander Vasiliev
Paul Scherrer Institut - PSI
Analysis of MCNPX/ JEF-2.2 and JENDL-3.3 results for a suite of low-enriched thermal compound uranium benchmarks
A. Vasiliev

 
176
Ulrich Hesse
Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH

Status of the GRS Reactivity and Burn-up Code System KENOREST
U. Hesse

 
185
Geoff Dobson
Serco Assurance
Modelling of dropped rods for criticality analysis
G. Dobson

 
217
Benoit Lance
Belgonucleaire
Preliminary analysis of the REBUS-PWR results
Benoit Lance, Daniel Marloye, Alfred Renard, Jacques Basselier, Peter Baeten, Leo Sannen, Yves Parthoens, Mireille Gysemans

 
218
Benoit Lance
Belgonucleaire
DANTSYS and MCNP as versatile tools for the safety aspects of the Belgonucleaire MOX plant
Marc Labilloy, Paul Van den Hende, Benoit Lance Jérémie Moerenhout, Bert Lievens, Henri Libon

 
12
Accelerators and Subcritical Systems
 
035
Andrei M. Voloschenko
Keldysh Institute of Applied Mathematics
An Experience in the Use the Sn Method for 1D/2D/3D Spallation Target Neutronics and Schielding Calculations
Vyacheslav P. Kryuchkov, Jonghwa Chang, Young-Sik Cho, Andrei M. Voloschenko, Oleg V. Sumaneev

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069
Sandra Dulla
Politecnico di Torino
Neutron importance in source-driven systems
S. Dulla, F. Cadinu, P. Ravetto

102
Daniel Ballester
Polytechnic University of Valencia
Stochastic transport theory of the deterministica and stochastic pulsing CPSD methods
D. Ballester, J-L Munoz-Cobos

202
Cheol Ho Pyeon
Kyoto University Research Reactor Institute
Experimental Analyses for Accelerator Driven Subcritical Reactor in Kyoto University Critical Assembly by using Foil Activation Method
C.H. Pyeon

 
209
Hervault Morgan
Kyoto University
Monte Carlo analysis of subcriticality for Accelerator Driven Subcritical Reactor mock up in Kyoto University Critical Assembly
M. Hervault, C.H. Pyeon, T. Misawa, H. Unesaki and S. Shiroya

 
259
Jacques Maillard
CNRS/IN2P3/MAPS/IDRIS
A constant shape solution of evolution equation in hybrid reactor: the soliton hybrid reactor
J. Maillard

 
279
Imre P‡zsit
Chalmers University of Technology
Neutron kinetics in subcritical source driven cores with applications to the source modulation method
J. Wright, I. P‡zsit

 
328
Mitja Majerle
Nuclear Physics Institute of CAS
MCNPX Benchmark Tests for "Energy Plus Transmutation" Setups
M. Majerle

 
13
Fuel loading optimization and fuel design
 
078
Frank Popa
Westinghouse
The Pearls(TM) Loading Pattern Search Tool
S. Si, F. D. Popa

 
151
Juan-Luis François
Facultad de Ingenieria - UNAM
Development of a Scatter Search Optimization Algorithm for BWR Fuel Lattice Design
J.L. François

 
152
M. Cecilia
Engineering School, UNAM
Development of a Fuzzy Logic Method to Build Objective Functions in Optimization Problems: Application to BWR Fuel Lattice Design
M. Cecilia

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210
Elisabeth Varin
Ecole Polytechnique de Montreal
Optimization of CANDU fuel management with gradient methods using generalized perturbation theory
R. Chambon, E. Varin and D. Rozon

258
Matthew Jessee
North Carolina State University
Coupled Bundle-Core Design using Rod-by-Rod Optimization
M. Jessee

277
Alejandro Castillo
Instituto Nacional de Investigaciones Nucleares
BWR Control Rod Patterns and Fuel Loading Optimization using Heuristic Methods
J.J. Ortiz, A. Castillo, J.L. Montes, R. Perusquia

 
14
Radiation Protection and Shielding
 
036
Yixue Chen
Institute for Plasma Physics of Chinese Academy of Sciences
Three-dimensional coupled Monte Carlo-Discrete Ordinates computational scheme for shielding calculations of large and complex nuclear facilities
Y. Chen, U. Fischer

 
191
Christophe Furstoss
The Institute of Radiological Protection and Nuclear
Numerical absorbed dose distributions inside a mathematical anthropomorphic phantom irradiated by monoenergetic photon fields
C. Furstoss, S. Ménard

 
227
Xiaosong Li
research reactor FRM-II
Radioactivity and dose rate evaluation for neutron irradiation at the new research reactor FRM-II
X. Li

 
233
Stéphanie Ménard
IRSN
Interest of the mesh dose distributions for the development of an instrumented phantom measuring the effective dose E
S. Ménard, C. Furstoss

 
257
Marie-Laure Hervé
IRSN/DRPH/SDE/LDRI
Relation between whole body and organ doses and local doses measured by ESR for different external exposures
M.L. Hervé

 
260
Loic de Carlan
IRSN
OEDIPE: a voxel based tool for the assesment of internal dose: Applications and new developments
L. de Carlan, I. Aubineau-lanièce, N. Borissov, N. Pierrat, S. Chiavassa, S. Lamart, D. Franck

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323
G‡bor Hord—sy
KFKI

Coupled neutron-gamma calculations for the VVER-1000 benchmark on the LR-0 experimental facility
G. Hord—sy

15
Nuclear Data Analysis and Methods
037
YongDeok Lee
Korea Atomic Energy Research Institut
Neutron Data Files for Pd-107, I-129 and Cs-135
Y.D.Lee, J.h.Chang

 
167
Arnaud Courcelle
ORNL
Statistical tests of U238 resonance parameters and structure of cross section in the unresolved range
A. Courcelle

 
177
Abdelhamid Dokhane
Ecole Polytechnique Federale de Lausanne
Application and Assessment of the PSI Reactor Signal Analysis Methodology to Stability Tests performed at KKL
A. Dokhane

 
178
Abdelhamid Dokhane
Ecole Polytechnique Federale de Lausanne
A Systematic and Consistent Reactor Noise Analysis Methodology for BWR Core Stability Evaluation
A. Dokhane

 
211
Vicent Garcia i Llorens
Univ. Politech. València
Neutronic Signal Conditioning Using Shannon Functions
V.G. i Llorens and G. Verdu

 
241
Natalia Borisova Janeva
Institute for Nuclear Research and Nuclear Energy

Neutron Cross Sections in the Unresolved Resonance Region
N. Borisova Janeva

 
251
Russell D. Mosteller
Los Alamos National Laboratory
Reactivity Impact of Deuterium Cross Sections for Heavy-Water Benchmarks
R. D. Mosteller

 
262
Luiz Leal
Oak Ridge National Laboratory
Covariance Data for 233U in the Resolved Resonance Region for Criticality Safety Applications
L. Leal

 
265
Khaled Ibrahim
Laboratoire Théma
A computer program for macroscopic cross section parameterisation for reactor core diffusion calculation
K. Ibrahim

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296
David Gerts
Air Force Institute of Technology
Piecewise Average, Anisotropic, Multi-Group Cross Sections
D.Gerts, K. Mathews, Wright-Patterson

297
R. Dagan
Forschungzentrumkarlsruhe
Imroved S(a,b) tables for TRIGA REactivity Feedback Calculations
R. Dagan

16
Computational Fluid Dynamics
 
038
Ivan Otic
Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies
Turbulent time scales and the temperature variance dissipation rate in natural convection in lead-bismuth


 
039
Adrian Tentner
Argonne National Laboratory
Computational Fluid Dynamics Modeling of Two-Phase Flow in a BWR Fuel Assembly (Log number 324)
Andrey Ioilev, Simon Lo, Maskhud Samigulin, Adrian Tentner, Vasily Ustinenko

 
092
Nikolay Kolev
Framatome-ANP
Do we have appropriate constitutive sets for sub-channel and fine-resolution 3D-analyses of two-phase flows in rod bundles?
N. I. Kolev

 
145
Sunchai Nilsuwankosit
Chulalongkorn University
Pressure Estimation by Least Squared Fit
S. Nilsuwankosit

 
149
Suneet Singh
Department of Nuclear, Plasma and Radiological Engineering

Parallel Modified Nodal Integral Method for Three-Dimensional Incompressible Navier-Stokes & Energy Equations
S. Singh

 
199
Hakan Mattsson
Chalmes University of Technology
CFD Simulation of the Pulsed Neutron Activation Technique for Water Flow Measurements
H. Mattsson

 
213
Brian E. Mays
Framatome ANP, Inc.
An Improved Nodal Integral Method for the Shallow Water Equations in Cartesian Geometries
Brian E. Mays

 
214
Brian E. Mays
Framatome ANP, Inc.
A Nodal Integral Method for the Shallow Water Equations on the Surface of a Sphere
Brian E. Mays, J. Dorning

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224
Yizhou Yan
University of Illinois at Urbana-Champaign

CFD Simulation of a Research Reactor
Y. Yan

238
Rafael Mir—
Universidad Politecnica de Valencia
Detailed Simulation of a PWR Fuel Rod and Fuel Element
R. Mir—

303
Queral Cesar
Polytechnic Unversity of Madrid
Conversion of the Steam Generator Model of Almaraz NPP from RELAP5 into TRAC-M, TRAC-P and TRACE
C. Queral, I. Gonzalez, A. Exposito, I. Gallego, A. Concejal

 
17
Models and Methods for Material Sciences
 
17




 
18
Environmental and Bioscience
 
154
Peane P. Maleka
Kernfysisch Versneller Instituut
Tests and calibration of a pulsed-neutron monitoring tool using Monte Carlo simulations
P.P. Maleka

 
256
Mary F. Wheeler
The University of Texas at Austin
Dynamically Adaptive Discontinuous Galerkin Methods for Contaminant Transport in Porous Media
M. F. Wheeler

 
19
Computer Codes and Benchmarks (poster session)
 
040
Frédéric Laugier
EDF
Very fast isotopic inventory and decay heat calculations in PWR spent nuclear fuel for industrial application
F. Laugier, E. Cabrol, C. Garzenne, S. Marguet and M. Dussartre

 
041
Kenneth A. Van Riper
White Rock Science
Parameterized Combinatorial Geometry Modeling in Moritz
K.A. Van Riper

 
042
Stanislav Stepanek
Skoda JS, a.s., Pilsen
Application of the code for hexagonal geometry based on composite nodal finite elements method and its comparison with a classical finite difference program
S. Stanislav, J.S. Skoda

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043
Petri Kotiluoto
VTT Processes
VENUS-2 MOX-fuelled reactor dosimetry benchmark calculations at VTT
P. Kotiluoto, F. Wasastjerna

044
Richard Procassini
Lawrence Livermore National Laboratory
Update on the Development and Validation of MERCURY: A Modern, Monte Carlo Particle Transport Code
R. Procassini, J. Taylor, S. McKinley, G. Greenman, D. Cullen, M. O'brien, B. Beck, C. Hagmann

045
Victor F. Boyarinov
RRC Kurchatov Institute
New Code System SUHAM-U, Variant SUHAM-U-VVER-01
V.F. Boyarinov, V.D. Davidenko, A.A. Polismakov, V.F. Tsybulsky

 
046
Victor F. Boyarinov
RRC Kurchatov Institute
Verification of Code System SUHAM-U on Benchmark Calculations of VVER-1000 Fuel Assemblies with Uranium or MOX Fuel
V.F. Boyarinov, V.D. Davidenko, V.F. Tsybulsky

 
073
Francesco Ganda
University of California, Berkeley
OECD Benchmark A of MOX Fueled PWR Unit Cells Using SAS2H
F. Ganda, D. Barnes, E. Greespan

 
083
Yousry Azmy
Penn State University
NuclearGUI: A Graphical User Interface for 3D Discrete Ordinates Neutral Particle Transport Codes in the DOORS and BOT3P Packages
Pierre-Wilfried Saintagne, Yousry Y. Azmy

 
098
Massimo Pescarini
ENEA
ENEA-Bologna Production and Testing of JEF-2.2 Multigroup Cross Section Libraries for Nuclear Fission Applications
M. Pescarini, R. Orsi, T. Martinelli, A.I. Blokhin, V. Sinitsa

 
113
Andrei M. Voloschenko
Keldysh Institute of Applied Mathematics
The CNCSN: One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Discrete Ordinates Code Package
A. M. Voloschenko, S. V. Gukov, V. P. Kryuchkov, A. A. Dubinin, O. V. Sumaneev

 
121
Guy Marleau
École Polytechnique de Montréal
DRAGON Analysis of the VENUS-2 MOX Benchmark
G. Marleau, C. Bibowski

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123
Akiko Kanda
Japan Nuclear Energy Safety Organization - JNES
Verification of the three-dimensional neutron kinetics code SKETCH-INS
A. Kanda, T. Nakajima

126
Thomas W. Laub
Sandia National Laboratories
The Integrated TIGER Series Vestion 5.0
T.W. Laub

146
Naoki Yamano
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology

Development of Integrated Nuclear Data Utilization System for Innovative Reactors
N. Yamano

 
170
Yaroslav Kozmenkov
Forschungszentrum Rossendorf e. V. - FZR
Comparative Assessment of Coupled RELAP5/PARCS and DYN3D/RELAP5 Codes Against the Kozloduy-6 Pump Trip Test
Y. Kozmenkov

 
173
Oliver Kšberl
CEA
OECD/NEA International Uranium and Plutonium Benchmarks of a Modular HTR
O. Kšberl

 
183
Luc Van Den Durpel
Argonne National Laboratory
Towards an Integrated Nuclear Energy Systems Model
L. Van Den Durpel

 
206
Hong-Chul KIM
Hanyang University
Benchmark Calculations for VENUS-2 MOX-Fueled Reactor Dosimetry
Hong-Chul KIM, Chang-ho SHIN, Chi Young HAN, Jong Kyung KIM, and Byung-Chan NA

 
245
Evgeny Romanov
State Scientific Centre of Russia Research Institute of Atomic Reactors

ORIP_XXI Computer Programs for Isotope Transmutation Simulations
E. Romanov

 
246
Oliver Kšberl
CEA
GT-MHR Benchmark Calculations using MCNP(X) and TRIPOLI4 Monte Carlo Codes
O. Kšberl

 
261
Yi-Kang Lee
CEA
TRIPOLI-PEPIN Depletion Code and Its First Numerical Benchmarks for PWR High-Burnup UO2 and MOX Spent Fuel
Y.K. Lee

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263
Nikolay Laletin
RRC Kurchatov Institute
Complex SVS for nutron-physical calculations in uranium water reactors
N. Laletin

291
Imre P‡zsit
Chalmers University of Technology
Benchmarking the invariant embedding method against analytical solutions in model transport problems
M. Wahlstrom, I. P‡zsit

299
Clif Drumm
Sandia National Laboratories
Coupled Electron-Photon Transport with the CEPTRE Code
C. Drumm

 
302
Christophe Calvin
CEA
DESCARTES: A new generation system for neutronic calculations
C. Calvin

 
315
Gennadi Jerdev
Institute of Physics and Power Engineering
SKALA- Computational System for Nuclear Criticality and Radiation Safety Analysis
G. Jerdev

 
318
Nadia Messaoudi
SCK/CEN
Three-dimensional MCNP Simulations for the OECD/NEA VENUS-2 MOX-fuelled Reactor Dosimetry Benchmark Calculations
N. Messaoudi, W. Haeck

 
321
L. Erradi
Mohammed V University
Constitution and validation of new neutron cross section libraries for MCNP code using NJOY system and some critical benchmarks
T. EL Bardouini, L. Erradi, E. Chakir, O. Meroun, B. EL Bakkari, M. Azahra, T. EL Khoukhi

 
339
Michael L. Hall
Continuum Dynamics
Higher-Fidelity Yet Efficient Modeling of Radiation Energy Transport Through Three-Dimensional Clouds
M.L. Hall, A.B. Davis

 
342
Matthieu Longeot
GaliNéo
On an optimal exponential transform parameter calculation for biasing neutral particle deep-penetration Monte Carlo transport
L. Bindel, M.M. Idrissi, B. Gonzalez

 
344
Charles Boyle
Particle Technologies UK
GERALD - A General Environment for radiation Analysis and Design
C. Boyle, P. de Oliveira, C. de Oliveira, J.M. Galan, M. Adams

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20
Advanced Computational Methods for Research Reactors
047
Federico E. Teruel
University of Illinois at Urbana-Champaign
An Alternative model for neutron flux maximization in Research Reactors
F.E. Teruel

084
S. Todd Keller
Oregon State University
Modeling a TRIGA Mark II Reactor Using the Attila(TM) Three-Dimensional Deterministic Transport Code
S.T. Keller, T. S. Palmer

 
140
David Blanchet
CEA - Technicatome
Analysis of gamma-ray dosimetry experiments in the EOLE mock-up for the validation of the JHR nuclear heating calculation scheme
D. Blanchet

 
148
Jianwei Hu
Department of nuclear engr. UIUC

A multipurpose research reactor design
J. Hu

 
168
Kursat Bekar
Penn State University
TORT Modelling of the Beam Port Facility of PSBR and Comparison with MCNP
K. Bekar

 
182
Jean-Michel Ruggieri
CEA
Unstructured characteristic method embedded with variational nodal method using domain decomposition techniques
J.M. Ruggieri

 
266
Andr‡s Keresztśri
KFKI Atomic Energy Research Institute
3D kinetic response matrix calculations of the Budapest Research Reactor
A. Keresztśri

 
21
Advanced Methods in 3D Deterministic Transport
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048
Robert C. Ward
CCS-4, Los Alamos National Laboratory
A Diffusion Synthetic Acceleration Method for Block Adaptive Mesh Refinement
R.C. Ward, R.S. Baker, J.E. Morel

049
Masahiro Tatsumi
Nuclear Fuel Industries, Ltd.
Development and Verification of SCOPE2: Advanced Core Calculation Code for PWRs
M. Tatsumi, H. Hyudou, H. Nagano

050
Gianluca Longoni
University of Florida
Evaluation of a New Preconditioning Algorithm based on the 3-D Even-Parity Simplified Sn Equations for the Discrete Ordinates Method in Parallel Computing Environments
G. Longoni, A. Haghighat, G. E. Sjoden

 
094
Jim E. Morel
Los Alamos National Laboratory
Sn Finite-Element Lumping on Quadrilateral Meshes in X-Y Geometry
J.E. Morel, J.S. Warsa

 
114
Glenn E. Sjoden
University of Florida
The New Exponential Directional Iterative 3-D Sn Scheme for Parallel Adaptive Differencing
G.E. Sjoden, D. Shedlock, A. Haghighat, C. Yi

 
128
Elmer E. Lewis
Northwestern University
The Variational Nodal Method : Some History and Recent Activity
E.E. Lewis, M. A. Smith, G. Palmiotti

 
143
Kang-Seog Kim
Korea Atomic Energy Research Institute
Depletion Capability of 3-dimensional Whole Core Transport Code DeCART
K.S. Kim

 
150
Farzad Rahnema
Georgia Institute of Technology
3-D Monte Carlo Based Coarse- Mesh Transport Method
F. Rahnema

 
252
Marvin L. Adams
Texas A&M University
A PieceWise Linear Finite Element Discretization of the Diffusion Equation for Arbitrary Polyhedral Grids
M. L. Adams

 
280
Heath L. Hanshaw
University of Michigan
Linear-Solution-Preservation and Diffusive Solutions in Multidimensional S_N Calculations
H. L. Hanshaw, E.W. Larsen

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332
Glenn E. Sjoden
University of Florida

A New Two-Level Dynamic Parallel Data Scheme for Large 3-D Sn Calculations
Glenn E. Sjoden, D. ShedLock, A. Haghighat, C. Yi

22
Advances in System Modeling
for Probabilistic Risk Assessment (PRA)
116
Tunc Aldemir
The Ohio State University
An On-Line Probabilistic Diagnostic and Prognostic Approach to Accident Management
Tunc Aldemir, Aram Hakobyan, Kyle Metzroth

 
117
Tunc Aldemir
The Ohio State University
A Finer Mesh is Not Always a Better Mesh - The Case of Non-Compact Support in Probabilistic Dynamics
T. Aldemir

 
119
Enrico Zio
Politecnico di Milano
Multi-objective Genetic Algorithm Parameter Estimation in a Reduced Nuclear Reactor Model
Marzio Marseguerra, Enrico Zio and Raffaele Canetta

 
276
Moosung Jae
Hanyang University
A Reliability Assessment Method Using System Dynamics and Application
Kyung Min Kang, Moosung Jae, Sangman Kwak

 
23
Stochastic Transport
 
051
Imre Pazsit
Chalmers University of Technology
Neutron fluctuations in a medium randomly varying in time
L. Pal, I. Pazsit

 
052
Philippe Humbert
CEA
Time Dependent and Asymptotic Neutron Number Probability Distribution Calculation Using Discrete Fourier Transform
P. Humbert

 
101
José L. Mu–oz-Cobo
Polytechnic University of Valencia
Stochastic transport theory of the deterministic and stochastic pulsing Feynman-alpha methods: Validation with MUSE experiments
D. Ballester, J.L. Mu–oz-Cobo, J.L. Kloosterman

 
131
Anil K. Prinja
University of New Mexico
Atomic Mix Synthetic Acceleration of Dose Computations in Binary Statistical Media
A.K. Prinja

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195
Rémi Sentis
CEA
On the slowing down of particles in a heterogeneous medium
R. Sentis

286
Frank Graziani
Lawrence Livermore National Laboratory
Effective Mean Free Paths in Absorbing and Scattering Random Media: Direct Numerical Simulation
F. Graziani, D. Miller, D. Slone

288
Edward Larsen
University of Michigan
An Asymptotic Correction to the Atomic Mix Approximation
E.W. Larsen, L.M. Davis, T.S. Palmer

 
289
Edward Larsen
University of Michigan
The 1-D Atomic Mix Diffusion Limit
E.W. Larsen, R. Vasques, M.T. Vihlena

 
324
Ian Davis
Oregon State University
Two-Group k-Eigenvalue Benchmark Calculations for Planar Geometry Transport in a Binary Stochastic Medium
I.M. Davis, T.S. Palmer

 
325
Todd Palmer
Oregon State University
A Coupled Diffusion Synthetic Acceleration for Binary Stochastic Mixture Transport Iterations in Slab Geometry
T.S. Palmer

 
335
Richard Sanchez
CEA Saclay
A variance model for Markov statistics
R. Sanchez

 
343
Ziya Akcasu
University of Michigan
A new approach to stochastic transport via the functional Volterra expansion
A.Z. Akcasu

 
24
Recent Developments in Stochastic and Deterministic Methods for Charged Particle Transport
 
156
Anil K. Prinja
University of New Mexico
A Moment-Preserving Energy-Loss Decomposition Scheme for Electron Transport
A.K. Prinja

 
169
Brian Franke
Sandia National Laboratories
Monte Carlo Electron Transport Using Generalized Boltzmann Fokker-Planck Scattering Models
B. Franke

 
184
Clif Drumm
Sandia National Laboratories
Delta-Function Down Scatter and Extended First-Collision Source for Electron Beam Transport
C. Drumm

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281
H. Grady Hughes
Los Alamos National Laboratory
Improved Logic for Sampling Landau Straggling in MCNP5
H. Grady Hughes

25
Transport calculation methods for unstructured meshes
053
Alexander Elchine
Alexandrov Research Institute of Technology
The Evolution of the Surface Harmonic Method to Derive Equations for a Distribution of Neutrons and Their Importance in a Heterogeneous Reactor
A.V. Elshin, FSUE "Alexandrov NITI", Sosnovy Bor

 
054
Teruhiko Kugo
Japan Atomic Energy Research Institute
Benchmark Solution for Unstructured Geometry PWR Problem by Method of Characteristics using Combinatorial Geometry
T. Kugo

 
055
Victor F. Boyarinov
RRC Kurchatov Institute
Generation of Multigroup Cross Sections from Microgroup Ones in Code System SUHAM-U Used for VVER-1000 Reactor Core Calculations with MOX Loading
V.F. Boyarinov, V.D. Davidenko, A.A. Polismakov, V.F. Tsybulsky

 
144
Igor Suslov
IPPE
Method of the Characteristics for Calculation of VVER without Homogenization
I. Suslov

 
157
Robert Roy
Ecole Polytechnique de Montreal
Large-scale 3D characteristics solver: can the dream live on?
R. Roy

 
234
Shawn Pautz
Sandia National Laboratories
Manufactured Solution Verification of the Ceptre Code
S. Pautz

 
272
Jim Morel
LANL
A lumped discontinuous finite-element spatial discretization for triangular-mesh Sn calculations in r-z geometry
J.E. Morel, A. Gonzalez-Aller, J.S. Warsa

 
329
Robert E. Grove
KAPL

The Slice Balance Approach (SBA): A Characteristic-Based Multiple Balance SN Approach on Unstructured Polyhedral Meshes
R.E. Grove

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330
Christopher Pain
Imperial College
A High-Order Petrov-Galerkin Method for the Boltzmann Transport Equation
C.C. Pain, A.S. Candy, M.D. Piggott, C.J. Cotter, G.J. Gorman, A.J.H. Goddard, C.R.E. de Oliveira

26
High Temperature Reactors Physics and Methods
063
J.L. Kloosterman
Delft University of Technology
Spatial Effects in Dancoff Factors for Pebble-Bed HTRs
J.L. Kloosterman

 
064
Victoria Kulik
CEA
Core Homogenization Method for Pebble Bed Modular Reactors
V. Kulik, R. Sanchez

 
122
J.C. Kuijper
NRG
HTR core physics and transient analyses by the PANTHERMIX code system
J.C. Kuijper

 
220
Rian Prinsloo
NECSA
Application of SP2 and the Spatially continuous Sp2-P1 equations to the PBMR reactor
R. Prinsloo, D.I.Tomasevic, F. Albornoz

 
221
Frederik Reitsma
PBMR Pty Ltd
Investigation of the power peaking in the PBMR pebble-bed reactor
Frederik. Reitsma and Wessel R Joubert

 
222
Frederik Reitsma
PBMR Pty Ltd
An Overview of the FZJ-tools for HTR core design and reactor dynamics, the past, present and future
Frederik. Reitsma, H. J. Rźtten, W Scherer

 
239
Rian Prinsloo
NECSA
Development of the Analytic Nodal Method in Cylindrical Geometry
R. Prinsloo

 
249
Nathanael Hudson
Georgia Institute of Technology
Spectral History Correction of Microscopic Cross Sections For the PBR Using the Slowing Down Balance
N. Hudson

 
254
Ayman I. Hawari Investigation of the Effect of Carbon Interstitials on the Thermal Neutron Scattering Cross Sections of Graphite
A.I. Hawari, I.I. Al-Qasir, A.M. Ougouag

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313
Abderrafi M. Ougouag
Idaho National Laboratory
PEBBED Analysis of Hot Spots in Pebble-Bed Reactors
A. M. Ougouag, H. D. Gougar, W.K. Terry

314
Abderrafi M. Ougouag
Idaho National Laboratory
Methods for Modeling the Packing of Fuel Elements in Pebble Bed Reactors
A. M. Ougouag, J.J. Cogliati

333
Bryan Miles
Imperial College
Linear-Solution-Preservation and Diffusive Solutions in Multidimensional S_N Calculations
B. Miles, C.C. Pain, M.D. Eaton, K.A. Ziver, A.H.H. Goddard

 
27
Multi-physics coupled code systems
for nuclear reactor design and safety
 
056
Valery Malofeev
RRC Kurchatov Institute
Application of the BARS Code for Coupled Neutronics and Thermal-Hydraulics Calculations
A. Avvakumov, V. Malofeev, V. Sidorov

 
057
Werner Maschek
Forschungszentrum Karlsruhe /IKET
SIMMER-III and SIMMER-IV Safety Code Development for Reactors with Transmutation Capability
W. Maschek, A. Rineiski, T. Suzuki, Mg. Mori, S. Wang, E. Wiegner, K. Morita, T. Cadiou, P. Coste, S. Pigny, Sa. Kondo, Y. Tobita, H. Yamano, S. Fujita

 
076
Paul Coddington
Paul Scherrer Institut
Development and application of the coupled code system FAST for the analysis of fast reactor transients
S.Pelloni, E. Bubelis, K. Mikityuk, P. Coddington

 
091
Antti Daavittila
VTT - Technical Recearch Centre of Finland
Transient and fuel performance analysis with VTT's coupled code system
A. Daavittila, A.HŠmŠlŠinen, J. Miettinen, and H. RŠty

 
093
Nikolay Kolev
Framatome-ANP
Is it possible to design universal multi-phase flow analyzer?
N. I. Kolev

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096
Christine Poinot
CEA
Optimization of a Coupling Calculation for a BWR Core Depletion using CRONOS2/FLICA4 Coupled System and Kalif Supervision Tool
N. Todorova, E. Rigautr, C. Poinot, E. Royer, K. Ivanov

163
M. Langenbuch
GRS
Overview on the Development and Application of the Coupled Code System ATHLET-QUABOX/CUBBOX
M. Langenbuch

171
Erwin Mźller
Westinghouse Electric Sweden
Studies on Consistent Generation of Neutron Kinetics Data for Coupled Three-Dimensional Two-Group Transient Applications
E. Mźller

 
208
Kostadin N. Ivanov
Pennsylvania State University
Methodologies for BWR Stability Analysis with TRACE/PARCS
Y. Xu, T. Downar , A. Petruzzi, K. Ivanov, F. Maggini, R. Miro, J. Staudenmeier

 
212
Dr. Dobromir Panayotov
Westinghouse Electric Sweden AB
POLCA-T - A Coupled Multi-Physics Tool for Design and Safety Analyses
Dobromir Panayotov, Ulf Bredolt, Henric Lindgren

 
226
Yoshiro Asahi
JAERI
Analysis of Loss of On-site Power without Scram for the High Temperature Engineering Test Reactor
Y. Asahi

 
235
Soeren Kliem
Forschungszentrum Rossendorf
Uncertainty and sensitivity analysis of the Kozloduy pump trip test by coupled RELAP5/3.3-PARCS/2.6 code
S. Kliem

 
237
Rafael Mir—
Universidad Politecnica de Valencia
Assessment of the coupled code TRACE/PARCS to study BWR out-of-phase instabilities: Applications to Ringhals-1 NPP
R. Mir—

 
240
Quan Zhou
University of Illinois at Urbana-Champaign
Stability Analyses of an Integrated Neutronics-Thermal Hydraulics Model of Natural Circulation BWR under Low Pressure
Q. Zhou

 
287
Eija Karita Puska
VTT Technical Research Centre of Finland

APROS Couplings from Core to Containment
E. K. Puska and J. Ylijoki

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295
Diana Cuervo
Polytechnic University of Madrid
Implementation and performance of Krylov methods for the solution of two fluid hydrodynamics equations in the COBRA TF code
D. Cuervo

309
Ana Mar’a S‡nchez Hern‡ndez
Chemical and Nuclear Engineering
Sensitivity Analysis of Cofrentes NPP SCRAM-61 to Different Thermalhydraulic-Neutronic Core Mappings with TRAC-BF1/VALKIN
A.M. S‡nchez Hern‡ndez

28
Iterative methods for transport problems
 
058
Gabriele Grassi
CEA
A non-linear spatial multi-grid acceleration for the Method of Characteristics in unstructured meshes
G. Grassi

 
070
Andrei M. Voloschenko
Keldysh Institute of Applied Mathematics
Consistent P1 Synthetic Acceleration Scheme for Transport Equation in 3D Geometries
A.M. Voloschenko

 
071
Andrei M. Voloschenko
Keldysh Institute of Applied Mathematics
Experience in the Use of the Consistent P1SA Scheme for Accelerating Upscattering Iterations Convergence in Solving Thermalization and Subcritical Fission Problems in 1D/2D/3D Geometries
A.M. Voloschenko

 
080
Jae Chang
Los Alamos National Laboratory
Effectiveness of Various Transport Synthetic Acceleration Methods With and Without GMRES
J. Chang, M.L. Adams

 
103
Simone Santandrea
CEA
A macro domain DPn technique as an acceleration tool for the method of characteristics
S. Santandrea

 
228
James Warsa
LANL
Preconditioning a Parallel, Inexact Block-Jacobi Splitting of the SN Algorithm
J. Warsa

 
317
D.Y. Anistratov
NC State University

Nonlinear Weighted Flux Methods for Particle Transport Problems
L. Roberts, D.Y. Anistratov

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319
Marvin L. Adams
Texas A&M University
A New Approach to the Iterative Solution of Transport Problems
A. Maslowski, M.L. Adams

334
Cassiano R. E. de Oliveira
Georgia Tech
A Space-Angle Algebraic Multigrid Preconditioner for the Finite Element-Spherical Harmonics Method
H.K. Park, C.R.E. de Oliveira

345
Brian D. Lansrud
Los Alamos National Laboratory
A Spatial Multigrid Iterative Method for One-Dimensional Discrete-Ordinates Transport Problems
B.D. Lansrud, M.L. Adams

 
29
Monte Carlo Simulation: recent trends
in modeling and applications
 
110
A. Dubi
Ben Gurion University of the Negev
Resources optimization in system Engineering. A hybrid Monte-Carlo/ Analytic General solution of the problem
A. Dubi

 
132
Enrico Zio
Politecnico di Milano
Monte Carlo simulation of radioactive contaminant transport in unsaturated porous media
E. Zio

 
135
Jean-Marc Palau
CEA
3D Monte-Carlo Transport Calculations of Whole Slab Reactor Cores : Validation of Deterministic Neutronic Calculation Route
J.M. Palau

 
161
Francesca Giacobbo
Politecnico di Milano
Monte Carlo simulation of radioactive contaminant transport in unsaturated porous media
F. Giacobbo

 
164
Marek Flaska
EC-JRC-IRMM
Modeling of an improved GELINA target for the production of neutrons
M. Flaska

 
225
Yann Richet
IRSN
Initialization Bias Suppression in Iterative Monte Carlo Simulations: Benchmarks on Criticality Monte Carlo Calculations
Y. Richet

 
242
Sara Pozzi
Oak Ridge National Laboratory
Nuclear Materials Identification by Photon Interrogation
S. Pozzi

 
243
Bernard Rottner Is the Overall Activity Stored ina Final Repository Systematically Underestimated?
B. Rottner

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248
Nicolas Freud
CNDRI
Fast simulation of X-ray imaging: implementation of deterministic forced detection in Geant4
N. Freud

271
Joachim Miss
IRSN
The MORET 4B Monte Carlo code - New features to treat complex criticality systems
J. Miss, O. Jacquet, L. Heulers, F. Fernex, Y. Richet

336
Michael Loughlin
UKAEA Fusion
Comparison of AttilaTM and MCNPTM for Fusion Applications
M. Loughlin, T. Wareing, A. Barnett, G.Failla, J. McGhee

 
30
Resonance self-shielding models and algorithms
for reactor calculations
 
059
Masayuki Tohjoh
Chuden CTI Co., Ltd.,
Effects of the spatial discretization and temperature approximation on the BWR assembly calculations
M. Tohjoh, M. Watanabe, A. Yamamoto

 
060
Dave Knott
GE Nuclear Energy / Global Nuclear Fuel - Americas
Modelling Resonance Interference Effects in the GE Lattice Physics Code LANCER02
E. Wehlage, D. Knott, V.W. Mills

 
095
Alain Hebert
Ecole Polytechnique de Montreal
A Review of Legacy and Advanced Self-Shielding Models for Lattice Calculations
A. Hebert

 
108
Joo G. Han
Seoul National University
Implementation of Subgroup Method in Direct Whole Core Transport Calculation involving Nonuniform Temperature Distribution
H. G. Joo, B. S. Han, C. H. Kim, K. S. Kim

 
190
Mireille Coste-Delclaux
CEA
New Developments in Resonant Mixture Self-Shielding Treatment with APOLLO2 code and Application to Jules Horowitz Reactor Core Calculation
M. Coste-Delclaux

 
31
The Changing R™le of Analytical Transport Methods
in Nuclear Science and Engineering and Related Fields
 
061
Barry Ganapol
University of Arizona
Multigroup 1D Homogeneous Medium Benchmark using the Green's Function Method
B. Ganapol

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062
Tamara S. Poveschenko
RRC Kurchatov Institute
Some Aspects of Boundary Condition Formulation on Reactor Cell Surface
T.S. Poveschenko, N.I. Laletin

337
Barry D. Ganapol
U. of Arizona
Guidance of Monte Carlo calculations via analytical benchmarks
B.D. Ganapol, S. Woolf, C. de Oliveira

338
Pierre Benoist
CEA retiree
The CN Method and the Time-Dependent Half-Space Albedo Problem
P. Benoist

 
32
Numerical modelling of underground radioactive waste repository: a focus on radionuclides transfer through geological medium
 
197
Peter Frolkovic
University of Heidelberg
R3T - software package for numerical simulations of radioactive contaminant transport in groundwater
P. Frolkovic

 
274
Jean Roberts
INRIA-Rocquencourt
Multidomain Modeling for the Simulation of Flow and Transport in a Neighborhood of an Underground Nuclear Waste Repository
F. Clément, J. Jaffré, M. Kern, V. Martin, J.E. Roberts, A. Sboui

 
305
Alain Bourgeat
UCB-Lyon1
Derivation of a Global Model for an Underground High-level, long-lived Nuclear Waste Repository
A. Bourgeat

 
33
OECD/NEA benchmark on 3-D deterministic transport calculations without spatial homogenization (3-D C5G7 MOX)
 
077
Jon A Dahl
Los Alamos National Laboratory
3-D Extension C5G7 MOX Benchmark Results Using PARTISN
J.A. Dahl

 
099
A. Seubert
Gesellschaft fźr Anlagen- und Reaktorsicherheit (GRS) mbH
Solving the C5G7 3-D Extension Benchmark Problem with the SN Code TORT
A. Seubert, S. Langenbuch, W. Zwermann

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104
Nam Zin Cho
Korea Advanced Institute of Science and Technology
Three-Dimensional Whole-Core Transport Calculations of the OECD Benchmark Problem C5G7 MOX by the CRX Code
G.S. Lee, N.Z. Cho

120
Philippe Humbert
CEA
Results for the C5G7 3-D Extension Benchmark Using the Discrete Ordinates Code PANDA
P. Humbert

133
Micheal A Smith
Argonne National Laboratory
Argonne National Laboratory Results for the OECD 3-D MOX fuel assembly benchmark
M.A. Smith

 
147
Shinya Kosaka
TEPSYS
CHAPLET-3D Results for the 3-D Extension C5G7 MOX Benchmarks
S. Kosaka

 
200
Igor Suslov
IPPE
Construction of Extrapolated Solution of the C5G7MOX Benchmark with the MCCG3D Characteristics code
I. Suslov

 
205
Hong-Chul KIM
Hanyang University
3-D Extension C5G7 MOX Benchmark Calculation Using THREEDANT Code
Hong-Chul KIM, Chi Young HAN, Jong Kyung KIM, and Byung-Chan NA

 
331
Byung-Chan Na
OECD, NEA
Benchmark on Deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization
M. A. Smith, E. E. Lewis,
Byung-Chan Na


 
341
Todd A. Wareign
Transpire Inc
Attila calculations for the 3-D-C5G7 MOX Benchmark Extension
T.A. Wareign, J.M. McGhee, D.A. Barnett, G.A. Failla