Adaptation of the Nagra Activation Analysis Methodology to Serpent
|
|
- Thomasina Horton
- 5 years ago
- Views:
Transcription
1 Adaptation of the Nagra Activation Analysis Methodology to Serpent Valentyn Bykov May 31, 2018, Serpent UGM 2018
2 Nuclear Power in Switzerland Reactor Type Net First MWe power Beznau 1 PWR Beznau 2 PWR Gösgen PWR Mühleberg BWR Leibstadt BWR Closure scheduled for 2019 (Nagra supports decommissioning planning activities)
3 3D MCNP Models Made mostly manually Cart. & Hex. Lattice Nodal core definition (240x25) - Void (mat. card) - Power (sdef probability) Explicit, detailed reactor internals
4 ORIGEN-based Activation Calculations Hybrid VR ADVANTG ww MCNP run multi-group flux information (fmesh) Operating History User input Material definition Ray-trace geometry with LAVA Activation ORIGEN.csv.csv.csv.csv.silo Based on ORNL work for ITER
5 Flux Information (MCNP) Run Mesh tally for each component - Cartesian - Millions of cells (generally 5cm res.) - SCALE-56 energy group structure (can use any other) Very difficult transport problem: Accelerated with Hybrid variance reduction
6 Variance Reduction (ADVANTG) Hybrid VR with ADVANTG User defines (structured) grid - only non-rotated Cartesian :-( LAVA ray-traces MCNP model, assigns materials to each cell DENOVO solves adjoint flux (CADIS and FW-CADIS) WW file generated Separate optimization for each component run Makes calculations affordable for single CPU workstations (but sometimes there are issues)
7 Validation Phase 1: Transport - Foil Activation Phase 2: Activation - Direct spectroscopy of modeled components
8 Material Definition (LAVA) Use LAVA (from ADVANTG) to ray-trace the model, assign materials for each cell of mesh tally Can use modified MCNP input removing surroundings of the component (water, other components) Impurities defined here, taken from: - from elementary analysis of the actual component (best) - from elementary analysis of similar components, other plants (ok) - from literature (risky!)
9 Operation History Operation History provider by the user in the input file Describe full NPP life + decay until reference date Two vectors: time (in sec) and power (value by which to multiply the MCNP result particles / s) Merely scales MCNP result no changes in void profile, power distribution / leakage - Generally: pic one point in time as representative, accept effects - Can split calculation into two separate ones, e.g.: 50yr full power = (25yr full power + 25yr decay) core1 + (25yr full power) core
10 Activation Calculations Using ORIGEN (like SCALE 6.2 but modified to have a C++ API) - Based on CRAM = very fast (1-2 sec per cell with full NPP history) Information provided over API MPI Parallelization Results output as both text (CSV) and graphic (SILO) - User defines nuclides of interest - For each nuclide, mesh tally sized table with values in Bq or Bq/g - Can post-process CSV (bash scripts, spreadsheet operations) - Can post-process SILO (VisIt)
11
12 Application to Packaging Planning Detailed activation distribution used to plan segmentation, packaging ILW cask (with various shielding) LLW cask (cheaper, bigger) Iterative process with the NPP Number of subcomponents ILW casks LLW casks
13 Application to Packaging Planning 13 Datum/Autorenkürzel Filename
14 Application to Decommissioning Planning Dose Rate fields repeatedly requested Currently no automated way, but link planned ORNL accelerates calculations with MS-CADIS VR
15 Current Methodology Under development, but already used for production runs Done for Swiss NPPs (all by 2019), but also commercially - Gundremmingen B&C (Germany s two biggest BWRs) - Basel Research Reactor Nuclear Eng. MSc students involved - 7 MSc Theses But there is room for improvement for VR, post-processing Could the same (and more) be done with Serpent?
16 Transition Checklist N/A Models with large, complex geometry Powerful variance reduction Flux mesh tallies or Direct Component Activation (given known operation history) Possibly define resulting gamma source term (for dose rate calculations) Tools for results post-processing 16 Datum/Autorenkürzel Filename
State of the art of Monte Carlo technics for reliable activated waste evaluations
State of the art of Monte Carlo technics for reliable activated waste evaluations Matthieu CULIOLI a*, Nicolas CHAPOUTIER a, Samuel BARBIER a, Sylvain JANSKI b a AREVA NP, 10-12 rue Juliette Récamier,
More informationClick to edit Master title style
New features in Serpent 2 for fusion neutronics 5th International Serpent UGM, Knoxville, TN, Oct. 13-16, 2015 Jaakko Leppänen VTT Technical Research Center of Finland Click to edit Master title Outline
More informationRadiological Characterization and Decommissioning of Research and Power Reactors 15602
Radiological Characterization and Decommissioning of Research and Power Reactors 15602 INTRODUCTION Faezeh Abbasi *, Bruno Thomauske *, Rahim Nabbi * RWTH University Aachen The production of the detailed
More information1 st International Serpent User Group Meeting in Dresden, Germany, September 15 16, 2011
1 st International Serpent User Group Meeting in Dresden, Germany, September 15 16, 2011 Discussion notes The first international Serpent user group meeting was held at the Helmholtz Zentrum Dresden Rossendorf
More informationAutomated ADVANTG Variance Reduction in a Proton Driven System. Kenneth A. Van Riper1 and Robert L. Metzger2
Automated ADVANTG Variance Reduction in a Proton Driven System Kenneth A. Van Riper1 and Robert L. Metzger2 1 White Rock Science, P. O. Box 4729, White Rock, NM 87547, kvr@rt66.com Radiation Safety Engineering,
More informationPSG2 / Serpent a Monte Carlo Reactor Physics Burnup Calculation Code. Jaakko Leppänen
PSG2 / Serpent a Monte Carlo Reactor Physics Burnup Calculation Code Jaakko Leppänen Outline Background History The Serpent code: Neutron tracking Physics and interaction data Burnup calculation Output
More informationBreaking Through the Barriers to GPU Accelerated Monte Carlo Particle Transport
Breaking Through the Barriers to GPU Accelerated Monte Carlo Particle Transport GTC 2018 Jeremy Sweezy Scientist Monte Carlo Methods, Codes and Applications Group 3/28/2018 Operated by Los Alamos National
More informationDevelopment of a Variance Reduction Scheme in the Serpent 2 Monte Carlo Code Jaakko Leppänen, Tuomas Viitanen, Olli Hyvönen
Development of a Variance Reduction Scheme in the Serpent 2 Monte Carlo Code Jaakko Leppänen, Tuomas Viitanen, Olli Hyvönen VTT Technical Research Centre of Finland, Ltd., P.O Box 1000, FI-02044 VTT, Finland
More informationOutline. Monte Carlo Radiation Transport Modeling Overview (MCNP5/6) Monte Carlo technique: Example. Monte Carlo technique: Introduction
Monte Carlo Radiation Transport Modeling Overview () Lecture 7 Special Topics: Device Modeling Outline Principles of Monte Carlo modeling Radiation transport modeling with Utilizing Visual Editor (VisEd)
More informationHPC Particle Transport Methodologies for Simulation of Nuclear Systems
HPC Particle Transport Methodologies for Simulation of Nuclear Systems Prof. Alireza Haghighat Virginia Tech Virginia Tech Transport Theory Group (VT 3 G) Director of Nuclear Engineering and Science Lab
More informationPopulation Control Variance Reduction in MCNP
Population Control Variance Reduction in MCNP One of the oldest, but simplest and often effective, variance reduction techniques in MCNP is to control the population of MCNP particles passing through a
More informationParticle track plotting in Visual MCNP6 Randy Schwarz 1,*
Particle track plotting in Visual MCNP6 Randy Schwarz 1,* 1 Visual Editor Consultants, PO Box 1308, Richland, WA 99352, USA Abstract. A visual interface for MCNP6 has been created to allow the plotting
More informationGraphical User Interface for High Energy Multi-Particle Transport
Graphical User Interface for High Energy Multi-Particle Transport Phase I Final Report PREPARED BY: P.O. Box 1308 Richland, WA 99352-1308 PHONE: (509) 539-8621 FAX: (509) 946-2001 Email: randyschwarz@mcnpvised.com
More informationAutomatic Mesh Adaptivity for Hybrid Monte Carlo/Deterministic Neutronics Modeling of Difficult Shielding Problems. Ahmad Ibrahim
Automatic Mesh Adaptivity for Hybrid Monte Carlo/Deterministic Neutronics Modeling of Difficult Shielding Problems by Ahmad Ibrahim A dissertation submitted in partial fulfillment of the requirement for
More informationChallenges and developments in fusion neutronics a CCFE perspective
Challenges and developments in fusion neutronics a CCFE perspective A. Turner Applied Radiation Physics group SERPENT fusion neutronics workshop, Cambridge, June 2015 CCFE is the fusion research arm of
More informationDevelopment of a Radiation Shielding Monte Carlo Code: RShieldMC
Development of a Radiation Shielding Monte Carlo Code: RShieldMC Shenshen GAO 1,2, Zhen WU 1,3, Xin WANG 1,2, Rui QIU 1,2, Chunyan LI 1,3, Wei LU 1,2, Junli LI 1,2*, 1.Department of Physics Engineering,
More informationBEAVRS benchmark calculations with Serpent-ARES code sequence
BEAVRS benchmark calculations with Serpent-ARES code sequence Jaakko Leppänen rd International Serpent User Group Meeting Berkeley, CA, Nov. 6-8, Outline Goal of the study The ARES nodal diffusion code
More informationCALCULATION OF THE ACTIVITY INVENTORY FOR THE TRIGA REACTOR AT THE MEDICAL UNIVERSITY OF HANNOVER (MHH) IN PREPARATION FOR DISMANTLING THE FACILITY
CALCULATION OF THE ACTIVITY INVENTORY FOR THE TRIGA REACTOR AT THE MEDICAL UNIVERSITY OF HANNOVER (MHH) IN PREPARATION FOR DISMANTLING THE FACILITY Gabriele Hampel, Friedemann Scheller, Medical University
More informationClick to edit Master title style
Fun stuff with the built-in response matrix solver 7th International Serpent UGM, Gainesville, FL, Nov. 6 9, 2017 Jaakko Leppänen VTT Technical Research Center of Finland Click to edit Master title Outline
More informationIMPROVING COMPUTATIONAL EFFICIENCY OF MONTE-CARLO SIMULATIONS WITH VARIANCE REDUCTION
International Conference on Mathematics and Computational Methods Applied to Nuclear Science & Engineering (M&C 2013), Sun Valley, Idaho, USA, May 5-9, 2013, on CD-ROM, American Nuclear Society, LaGrange
More informationAttila4MC. Software for Simplifying Monte Carlo. For more info contact or
Attila4MC Software for Simplifying Monte Carlo For more info contact attila@varian.com or Gregory.Failla@varian.com MCNP and MCNP6 are trademarks of Los Alamos National Security, LLC, Los Alamos National
More informationModeling Radiation Transport Using MCNP6 and Abaqus/CAE Chelsea A. D Angelo, Steven S. McCready, Karen C. Kelley Los Alamos National Laboratory
Modeling Radiation Transport Using MCNP6 and Abaqus/CAE Chelsea A. D Angelo, Steven S. McCready, Karen C. Kelley Los Alamos National Laboratory Abstract: Los Alamos National Laboratory (LANL) has released
More informationApplication of MCNP Code in Shielding Design for Radioactive Sources
Application of MCNP Code in Shielding Design for Radioactive Sources Ibrahim A. Alrammah Abstract This paper presents three tasks: Task 1 explores: the detected number of as a function of polythene moderator
More informationClick to edit Master title style
Introduction to Serpent Code Fusion neutronics workshop, Cambridge, UK, June 11-12, 2015 Jaakko Leppänen VTT Technical Research Center of Finland Click to edit Master title Outline style Serpent overview
More informationClick to edit Master title style
Greetings from the Serpent Developer Team 7th International Serpent UGM, Gainesville, FL, Nov. 6 9, 2017 Jaakko Leppänen VTT Technical Research Center of Finland Source code development: Click Jaakko to
More informationMesh Human Phantoms with MCNP
LAUR-12-01659 Mesh Human Phantoms with MCNP Casey Anderson (casey_a@lanl.gov) Karen Kelley, Tim Goorley Los Alamos National Laboratory U N C L A S S I F I E D Slide 1 Summary Monte Carlo for Radiation
More informationSERPENT Cross Section Generation for the RBWR
SERPENT Cross Section Generation for the RBWR Andrew Hall Thomas Downar 9/19/2012 Outline RBWR Motivation and Design Why use Serpent Cross Sections? Modeling the RBWR Generating an Equilibrium Cycle RBWR
More informationDETERMINISTIC 3D RADIATION TRANSPORT SIMULATION FOR DOSE DISTRIBUTION AND ORGAN DOSE EVALUATION IN DIAGNOSTIC CT
DETERMINISTIC 3D RADIATION TRANSPORT SIMULATION FOR DOSE DISTRIBUTION AND ORGAN DOSE EVALUATION IN DIAGNOSTIC CT Monica Ghita,, Glenn Sjoden, Manuel Arreola, Ahmad Al-Basheer Basheer, Choonsik Lee, Wesley
More informationParallel PENTRAN Applications. G. E. Sjoden and A. Haghighat Nuclear and Radiological Engineering University of Florida
Parallel PENTRAN Applications G. E. Sjoden and A. Haghighat Nuclear and Radiological Engineering University of Florida Overview Introduction Parallel Computing & MPI Boltzmann & Transport PENTRAN TM Code
More informationVerification of the Hexagonal Ray Tracing Module and the CMFD Acceleration in ntracer
KNS 2017 Autumn Gyeongju Verification of the Hexagonal Ray Tracing Module and the CMFD Acceleration in ntracer October 27, 2017 Seongchan Kim, Changhyun Lim, Young Suk Ban and Han Gyu Joo * Reactor Physics
More informationA premilinary study of the OECD/NEA 3D transport problem using the lattice code DRAGON
A premilinary study of the OECD/NEA 3D transport problem using the lattice code DRAGON Nicolas Martin, Guy Marleau, Alain Hébert Institut de Génie Nucléaire École Polytechnique de Montréal 28 CNS Symposium
More informationModeling the ORTEC EX-100 Detector using MCNP
Modeling the ORTEC EX-100 Detector using MCNP MCNP is a general-purpose Monte Carlo radiation transport code for modeling the interaction of radiation with materials based on composition and density. MCNP
More informationDirect Use of CAD Geometry in Monte Carlo Radiation Transport. Paul Wilson CNERG/FTI Neutronics Team U. Wisconsin-Madison
Direct Use of CAD Geometry in Monte Carlo Radiation Transport Paul Wilson CNERG/FTI Neutronics Team U. Wisconsin-Madison CNERG/FTI Neutronics Team 8 Research Staff 1 Visiting Scientist T. Bohm Nuclear
More informationVisual MCNP Editor Lore
Visual MCNP Editor Lore 1. The Visual MCNP Editor can not run my input file. 2. I do not use the Visual Editor it dies all the time. 3. Vised died and I lost my input file. 4. Vised messes up my input
More informationDosimetry Simulations with the UF-B Series Phantoms using the PENTRAN-MP Code System
Dosimetry Simulations with the UF-B Series Phantoms using the PENTRAN-MP Code System A. Al-Basheer, M. Ghita, G. Sjoden, W. Bolch, C. Lee, and the ALRADS Group Computational Medical Physics Team Nuclear
More informationELECTRON DOSE KERNELS TO ACCOUNT FOR SECONDARY PARTICLE TRANSPORT IN DETERMINISTIC SIMULATIONS
Computational Medical Physics Working Group Workshop II, Sep 30 Oct 3, 2007 University of Florida (UF), Gainesville, Florida USA on CD-ROM, American Nuclear Society, LaGrange Park, IL (2007) ELECTRON DOSE
More informationInvestigations into Alternative Radiation Transport Codes for ITER Neutronics Analysis
CCFE-PR(17)10 Andrew Turner Investigations into Alternative Radiation Transport Codes for ITER Neutronics Analysis Enquiries about copyright and reproduction should in the first instance be addressed to
More informationComparison of Shutdown Dose Rate Results using MCNP6 Activation Capability and MCR2S
APPLIED RADIATION PHYSICS GROUP TECHNICAL NOTE ARP-097 July 2014 Comparison of Shutdown Dose Rate Results using MCNP6 Activation Capability and MCR2S A. Turner 1, Z. Ghani 1, J. Shimwell 2 1: CCFE, Culham
More informationIMPROVEMENTS TO MONK & MCBEND ENABLING COUPLING & THE USE OF MONK CALCULATED ISOTOPIC COMPOSITIONS IN SHIELDING & CRITICALITY
IMPROVEMENTS TO MONK & MCBEND ENABLING COUPLING & THE USE OF MONK CALCULATED ISOTOPIC COMPOSITIONS IN SHIELDING & CRITICALITY N. Davies, M.J. Armishaw, S.D. Richards and G.P.Dobson Serco Technical Consulting
More informationAdvances in neutronics tools with accurate simulation of complex fusion systems
Advances in neutronics tools with accurate simulation of complex fusion systems Mohamed Sawan P. Wilson, T. Tautges(ANL), L. El-Guebaly, T. Bohm, D. Henderson, E. Marriot, B. Kiedrowski, A. Ibrahim, B.
More informationMCNP CLASS SERIES (SAMPLE MCNP INPUT) Jongsoon Kim
MCNP CLASS SERIES (SAMPLE MCNP INPUT) Jongsoon Kim Basic constants in MCNP Lengths in cm Energies in MeV Times in shakes (10-8 sec) Atomic densities in units of atoms/barn*-cm Mass densities in g/cm 3
More informationGeometric Templates for Improved Tracking Performance in Monte Carlo Codes
Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo 2013 (SNA + MC 2013) La Cité des Sciences et de l Industrie, Paris, France, October 27-31, 2013 Geometric Templates
More informationDose rate calculation at transport and storage casks for spent nuclear fuel
Dose rate calculation at transport and storage casks for spent nuclear fuel B. Gmal, U. Hesse, K. Hummelsheim, R. Kilger, M. Wagner Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbh, Forschungsinstitute,
More informationMCNP Variance Reduction technique application for the Development of the Citrusdal Irradiation Facility
IYNC 2008 Interlaken, Switzerland, 20-26 September 2008 Paper No. 376 MCNP Variance Reduction technique application for the Development of the Citrusdal Irradiation Facility R Makgae Pebble Bed Modular
More informationNeutronics analysis for ITER Diagnostic Generic Upper Port Plug
2017 ANS Annual Meeting Technical Session: Neutronics Challenges of Fusion Facilities - Neutronics Challenges of Fusion Facilities - Neutronics analysis for ITER Diagnostic Generic Upper Port Plug Arkady
More informationDesign Calculation or Analysis Cover Sheet
BSC Design Calculation or Analysis Cover Sheet Complete only applicable items. ENG.20061026.0003 QA: QA 2. Page 1 3. System DOE and Commercial Waste Package 5. Title Tabulation of Fundamental Assembly
More informationKIT Fusion Neutronics R&D Activities and Related Design Applications
1 FTP/P7-19 KIT Fusion Neutronics R&D Activities and Related Design Applications U. Fischer1), D. Große1), K. Kondo1), D. Leichtle1), M. Majerle2), P. Pereslavtsev1), A. Serikov1), S. P. Simakov1,3) 1)
More informationParallel computations for the auto-converted MCNP5 models of the ITER ECRH launcher
Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft Parallel computations for the auto-converted MCNP5 models of the ITER ECRH launcher A. Serikov, U. Fischer, R. Heidunger, L. Obholz, P. Spaeh,
More informationTREAT Modeling & Simulation Using PROTEUS
TREAT Modeling & Simulation Using PROTEUS May 24, 2016 ChanghoLee Neutronics Methods and Codes Section Nuclear Engineering Division Argonne National Laboratory Historic TREAT Experiments: Minimum Critical
More informationA PRACTICAL LOOK AT MONTE CARLO VARIANCE REDUCTION METHODS IN RADIATION SHIELDING
A PRACTICAL LOOK AT MONTE CARLO VARIANCE REDUCTION METHODS IN RADIATION SHIELDING RICHARD H. OLSHER Health Physics Measurements Group, Los Alamos National Laboratory MS J573, P.O. Box 1663, Los Alamos,
More informationResearch Article Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis
Science and Technology of Nuclear Installations Volume 2, Article ID 68347, 7 pages doi:.55/2/68347 Research Article Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2
More informationgpmc: GPU-Based Monte Carlo Dose Calculation for Proton Radiotherapy Xun Jia 8/7/2013
gpmc: GPU-Based Monte Carlo Dose Calculation for Proton Radiotherapy Xun Jia xunjia@ucsd.edu 8/7/2013 gpmc project Proton therapy dose calculation Pencil beam method Monte Carlo method gpmc project Started
More informationSteam Generator Replacement of the Belgian Doel 1 unit: follow-up and on site dosimetry
Steam Generator Replacement of the Belgian Doel 1 unit: follow-up and on site dosimetry B. Walschaerts*, J. Defloor**, R. Wyckmans** * Radwaste management Decommissioning Radiation protection, Tractebel
More informationVCell. vcell.org. To run VCell go to: modeling environment for mathematical simulation of cellular events.
VCell modeling environment for mathematical simulation of cellular events. To run VCell go to: vcell.org Virtual Cell is developed by the Center for Cell Analysis and Modeling at the University of Connecticut
More informationBWR-club Status and Program Robust Power Workshop Uppsala, Sweden May 3 and 4, OKG/Uniper
BWR-club Status and Program Robust Power Workshop Uppsala, Sweden May 3 and 4, 2017 OKG/Uniper BWR-club... provides a forum for its members (European BWR utilities) to maintain and improve plant safety
More informationSimulation of the ILL High Flux Reactor using MCNP and Tripoli
Simulation of the ILL High Flux Reactor using MCNP and Tripoli E. Farhi, S. Fuard (ILL) C. Hennane, P.A. Harraud C. Campioni (CEA/Saclay) MC PSI 2006 1 Source models not satisfactory What do we know about
More informationOF THE, AMERICAN NUCLEAR SOCIiZTY
OF THE, AMERICAN NUCLEAR SOCIiZTY June 4-8, 2000 Volume 82 San Diego Town and Country Hotel and Convention Center TANSAO 82 l-300 (2000) San Diego, California ISSN: 0003-O 18X Raymond H. Gabaldon III Technical
More informationGraphical User Interface for Simplified Neutron Transport Calculations
Graphical User Interface for Simplified Neutron Transport Calculations Phase 1 Final Report Instrument No: DE-SC0002321 July 20, 2009, through April 19, 2010 Recipient: Randolph Schwarz, Visual Editor
More informationNeutronics Analysis of TRIGA Mark II Research Reactor. R. Khan, S. Karimzadeh, H. Böck Vienna University of Technology Atominstitute
Neutronics Analysis of TRIGA Mark II Research Reactor R. Khan, S. Karimzadeh, H. Böck Vienna University of Technology Atominstitute 23-03-2010 TRIGA Mark II reactor MCNP radiation transport code MCNP model
More informationMethodology for spatial homogenization in Serpent 2
Methodology for spatial homogenization in erpent 2 Jaakko Leppänen Memo 204/05/26 Background patial homogenization has been one of the main motivations for developing erpent since the beginning of the
More informationA FLEXIBLE COUPLING SCHEME FOR MONTE CARLO AND THERMAL-HYDRAULICS CODES
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011) Rio de Janeiro, RJ, Brazil, May 8-12, 2011, on CD-ROM, Latin American Section (LAS)
More informationEVALUATION OF SPEEDUP OF MONTE CARLO CALCULATIONS OF TWO SIMPLE REACTOR PHYSICS PROBLEMS CODED FOR THE GPU/CUDA ENVIRONMENT
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2011) Rio de Janeiro, RJ, Brazil, May 8-12, 2011, on CD-ROM, Latin American Section (LAS)
More informationMichael Speiser, Ph.D.
IMPROVED CT-BASED VOXEL PHANTOM GENERATION FOR MCNP MONTE CARLO Michael Speiser, Ph.D. Department of Radiation Oncology UT Southwestern Medical Center Dallas, TX September 1 st, 2012 CMPWG Workshop Medical
More informationDRAGON SOLUTIONS FOR BENCHMARK BWR LATTICE CELL PROBLEMS
DRAGON SOLUTIONS FOR BENCHMARK BWR LATTICE CELL PROBLEMS R. Roy and G. Marleau Institut de Génie Nucléaire École Polytechnique de Montréal P.O.Box 6079, Station CV, Montreal, Canada roy@meca.polymtl.ca
More informationISOCS Characterization of Sodium Iodide Detectors for Gamma-Ray Spectrometry
ISOCS Characterization of Sodium Iodide Detectors for Gamma-Ray Spectrometry Sasha A. Philips, Frazier Bronson, Ram Venkataraman, Brian M. Young Abstract--Activity measurements require knowledge of the
More informationValidation of GEANT4 for Accurate Modeling of 111 In SPECT Acquisition
Validation of GEANT4 for Accurate Modeling of 111 In SPECT Acquisition Bernd Schweizer, Andreas Goedicke Philips Technology Research Laboratories, Aachen, Germany bernd.schweizer@philips.com Abstract.
More informationElectron Dose Kernels (EDK) for Secondary Particle Transport in Deterministic Simulations
Electron Dose Kernels (EDK) for Secondary Particle Transport in Deterministic Simulations A. Al-Basheer, G. Sjoden, M. Ghita Computational Medical Physics Team Nuclear & Radiological Engineering University
More informationMCNP Monte Carlo & Advanced Reactor Simulations. Forrest Brown. NEAMS Reactor Simulation Workshop ANL, 19 May Title: Author(s): Intended for:
LA-UR- 09-03055 Approved for public release; distribution is unlimited. Title: MCNP Monte Carlo & Advanced Reactor Simulations Author(s): Forrest Brown Intended for: NEAMS Reactor Simulation Workshop ANL,
More informationEvaluation of RayXpert for shielding design of medical facilities
Evaluation of Raypert for shielding design of medical facilities Sylvie Derreumaux 1,*, Sophie Vecchiola 1, Thomas Geoffray 2, and Cécile Etard 1 1 Institut for radiation protection and nuclear safety,
More information2-D Reflector Modelling for VENUS-2 MOX Core Benchmark
2-D Reflector Modelling for VENUS-2 MOX Core Benchmark Dušan Ćalić ZEL-EN d.o.o. Vrbina 18 8270, Krsko, Slovenia dusan.calic@zel-en.si ABSTRACT The choice of the reflector model is an important issue in
More informationStatus of the Serpent criticality safety validation package
VTT TECHNICAL RESEARCH CENTRE OF FINLAND LTD Status of the Serpent criticality safety validation package Serpent UGM 2017 Riku Tuominen and Ville Valtavirta, VTT Outline Criticality Safety Evaluation What
More informationWhite Paper 3D Geometry Visualization Capability for MCNP
White Paper 3D Geometry Visualization Capability for MCNP J. B. Spencer, J. A. Kulesza, A. Sood Los Alamos National Laboratory Monte Carlo Methods, Codes, and Applications Group June 12, 2017 1 Introduction
More informationANSYS Improvements to Engineering Productivity with HPC and GPU-Accelerated Simulation
ANSYS Improvements to Engineering Productivity with HPC and GPU-Accelerated Simulation Ray Browell nvidia Technology Theater SC12 1 2012 ANSYS, Inc. nvidia Technology Theater SC12 HPC Revolution Recent
More informationCORE MONITORING EXPERIENCE WITH GARDEL
CORE MONITORING EXPERIENCE WITH GARDEL Axel Becker, Alejandro Noël Studsvik Scandpower GmbH Studsvik Scandpower Suisse GmbH Abstract The GARDEL core surveillance and analysis system is a standard, modular
More informationStatus and development of multi-physics capabilities in Serpent 2
Status and development of multi-physics capabilities in Serpent 2 V. Valtavirta VTT Technical Research Centre of Finland ville.valtavirta@vtt.fi 2014 Serpent User Group Meeting Structure Click to of edit
More informationEvaluation of PBMR control rod worth using full three-dimensional deterministic transport methods
Available online at www.sciencedirect.com annals of NUCLEAR ENERGY Annals of Nuclear Energy 35 (28) 5 55 www.elsevier.com/locate/anucene Evaluation of PBMR control rod worth using full three-dimensional
More informationImproving Uintah s Scalability Through the Use of Portable
Improving Uintah s Scalability Through the Use of Portable Kokkos-Based Data Parallel Tasks John Holmen1, Alan Humphrey1, Daniel Sunderland2, Martin Berzins1 University of Utah1 Sandia National Laboratories2
More informationA Dosimetry and Visualization Tool for the Fukushima Daiichi Power Plant
A Dosimetry and Visualization Tool for the Fukushima Daiichi Power Plant Justin Vazquez, Aiping Ding, Tianyu Liu, Chao Liang, Benjamin Lawrence, Christopher Gaylord, and Lin Su Advisor: Dr. X. George Xu
More informationThe Monte Carlo simulation of a Package formed by the combination of three scintillators: Brillance380, Brillance350, and Prelude420.
EURONS I3 506065 JRA9 RHIB Report made during stay IEM-CSIC Madrid december 2006 MINISTERIO DE ASUNTOS EXTERIORES Y DE COOPERACIÓN AECI VICESECRETARÍA GENERAL The Monte Carlo simulation of a Package formed
More informationD.1 Radon Model Emanation Analysis
D.1 Radon Model Emanation Analysis This appendix presents the radon emanation analyses performed in support of the EE/CA for the Ross-Adams Mine Site. The radon emanation analyses were used to determine
More informationMultiphysics simulations of nuclear reactors and more
Multiphysics simulations of nuclear reactors and more Gothenburg Region OpenFOAM User Group Meeting Klas Jareteg klasjareteg@chalmersse Division of Nuclear Engineering Department of Applied Physics Chalmers
More informationRobin P. Gardner* and. Lianyan Liu. Computalog U.S.A., Inc., 501 Winscott Road, Fort Worth, Texas 76126
NUCLEAR SCIENCE AND ENGINEERING: 133, 80 91 ~1999! Monte Carlo Simulation of Neutron Porosity Oil Well Logging Tools: Combining the Geometry-Independent Fine-Mesh Importance Map and One-Dimensional Diffusion
More informationHARNESSING IRREGULAR PARALLELISM: A CASE STUDY ON UNSTRUCTURED MESHES. Cliff Woolley, NVIDIA
HARNESSING IRREGULAR PARALLELISM: A CASE STUDY ON UNSTRUCTURED MESHES Cliff Woolley, NVIDIA PREFACE This talk presents a case study of extracting parallelism in the UMT2013 benchmark for 3D unstructured-mesh
More informationNUC E 521. Chapter 6: METHOD OF CHARACTERISTICS
NUC E 521 Chapter 6: METHOD OF CHARACTERISTICS K. Ivanov 206 Reber, 865-0040, kni1@psu.edu Introduction o Spatial three-dimensional (3D) and energy dependent modeling of neutron population in a reactor
More informationDaedeok-daero, Yuseong-gu, Daejeon , Republic of Korea b Argonne National Laboratory (ANL)
MC 2-3/TWODANT/DIF3D Analysis for the ZPPR-15 10 B(n, α) Reaction Rate Measurement Min Jae Lee a*, Donny Hartanto a, Sang Ji Kim a, and Changho Lee b a Korea Atomic Energy Research Institute (KAERI) 989-111
More informationDevelopment and Use of Computational Anthropomorphic Phantoms for Medical Dosimetry Nina Petoussi-Henss
Medical Radiation Physics and Diagnostics, AMSD Development and Use of Computational Anthropomorphic Phantoms for Medical Dosimetry Nina Petoussi-Henss HMGU, HELENA Lecture series, 16.09.2015 Outline Why
More informationDose Calculations: Where and How to Calculate Dose. Allen Holder Trinity University.
Dose Calculations: Where and How to Calculate Dose Trinity University www.trinity.edu/aholder R. Acosta, W. Brick, A. Hanna, D. Lara, G. McQuilen, D. Nevin, P. Uhlig and B. Slater Dose Calculations - Why
More informationParallel computation performances of Serpent and Serpent 2 on KTH Parallel Dator Centrum
KTH ROYAL INSTITUTE OF TECHNOLOGY, SH2704, 9 MAY 2018 1 Parallel computation performances of Serpent and Serpent 2 on KTH Parallel Dator Centrum Belle Andrea, Pourcelot Gregoire Abstract The aim of this
More informationHigh Performance Parallel Monte Carlo Transport Computations for ITER Fusion Neutronics Applications
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol. 2, pp.294-300 (2011) ARTICLE High Performance Parallel Monte Carlo Transport Computations for ITER Fusion Neutronics Applications Arkady SERIKOV *, Ulrich
More informationTechnology for a better society. SINTEF ICT, Applied Mathematics, Heterogeneous Computing Group
Technology for a better society SINTEF, Applied Mathematics, Heterogeneous Computing Group Trond Hagen GPU Computing Seminar, SINTEF Oslo, October 23, 2009 1 Agenda 12:30 Introduction and welcoming Trond
More informationHitachi-GE Nuclear Energy, Ltd.(HGNE) Junichi Kawahata. Ref. CNJ-OG-G056Rev.7.3 All Rights Reserved Copyright 2009, Hitachi-GE Nuclear Energy, Ltd.
Advanced Technologies and Further Evolution Towards New Build NPP Projects International Conference on Opportunities and Challenges for Water Cooled Reactors in the 21st Century Hitachi-GE Nuclear Energy,
More informationABSTRACT. W. T. Urban', L. A. Crotzerl, K. B. Spinney', L. S. Waters', D. K. Parsons', R. J. Cacciapouti2, and R. E. Alcouffel. 1.
COMPARISON OF' THREE-DIMENSIONAL NEUTRON FLUX CALCULATIONS FOR MAINE YANKEE W. T. Urban', L. A. Crotzerl, K. B. Spinney', L. S. Waters', D. K. Parsons', R. J. Cacciapouti2, and R. E. Alcouffel ABSTRACT
More informationCHAPTER 10: TALLYING IN MCNP
_or_.e_sa_m_h_us_se_in 6:..:...7 ---"M.=o:.c.;;nte.:-C-"-=arlo Particle Transport with MCNP CHAPTER 10: TALLYING IN MCNP Tallying is the process of scoring the parameters of interest, Le. providing the
More informationComputing Acceleration for a Pin-by-Pin Core Analysis Method Using a Three-Dimensional Direct Response Matrix Method
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol., pp.4-45 (0) ARTICLE Computing Acceleration for a Pin-by-Pin Core Analysis Method Using a Three-Dimensional Direct Response Matrix Method Taeshi MITSUYASU,
More information5.3 cm. Lateral access holders for foils. Figure 1: ITER mock-up geometry
Introduction The calculations for the ITER Benchmark Experiment on Tungsten of the Shielding Integral Benchmark Archive and Database (SINBAD 2000) kept by the Radiation Safety Information Computational
More informationlecture 8 Groundwater Modelling -1
The Islamic University of Gaza Faculty of Engineering Civil Engineering Department Water Resources Msc. Groundwater Hydrology- ENGC 6301 lecture 8 Groundwater Modelling -1 Instructor: Dr. Yunes Mogheir
More informationExperience in Neutronic/Thermal-hydraulic Coupling in Ciemat
Madrid 2012 Experience in Neutronic/Thermal-hydraulic Coupling in Ciemat Miriam Vazquez (Ciemat) Francisco Martín-Fuertes (Ciemat) Aleksandar Ivanov (INR-KIT) Outline 1. Introduction 2. Coupling scheme
More informationCoupled calculations with Serpent
Coupled calculations with Serpent 2.1.29 Serpent UGM University of Florida, Gainesville, November 8, 2017 V. Valtavirta VTT Technical Research Center of Finland Background Serpent 2 has been designed for
More informationUSE OF CAD GENERATED GEOMETRY DATA IN MONTE CARLO TRANSPORT CALCULATIONS FOR ITER
USE OF CAD GENERATED GEOMETRY DATA IN MONTE CARLO TRANSPORT CALCULATIONS FOR ITER U. Fischer 1, H. Iida 2, Y. Li 3, M. Loughlin 4, S. Sato 2, A. Serikov 1, H. Tsige-Tamirat 1, T. Tautges 5, P. P. Wilson
More informationA COARSE MESH RADIATION TRANSPORT METHOD FOR PRISMATIC BLOCK THERMAL REACTORS IN TWO DIMENSIONS
A COARSE MESH RADIATION TRANSPORT METHOD FOR PRISMATIC BLOCK THERMAL REACTORS IN TWO DIMENSIONS A Thesis Presented to The Academic Faculty By Kevin John Connolly In Partial Fulfillment Of the Requirements
More information