MONTE CARLO POWER ITERATION: ENTROPY AND SPATIAL CORRELATIONS
|
|
- Ami Lynch
- 6 years ago
- Views:
Transcription
1 MONTE CARLO POWER ITERATION: ENTROPY AND SPATIAL CORRELATIONS ANDREA ZOIA, M. NOWAK (CEA/SACLAY) E. DUMONTEIL, A. ONILLON (IRSN) J. MIAO, B. FORGET, K. S. SMITH (MIT) NEA EGAMCT meeting Andrea ZOIA DEN/DANS/DM2S/SERMA/LTSD July 6th JUILLET 2016 PAGE 1
2 OUTLINE Power iteration with Monte Carlo The impact of correlations in criticality simulations Entropy and convergence Spatial moments and correlations Relation to neutron clustering theory Perspectives CEA July 6th 2016 PAGE 2
3 THE CRITICAL BOLTZMANN EQUATION We would like to determine the fundamental mode j 1 and the associated fundamental eigenvalue k 1 of Critical Boltzmann equation for the neutron flux j Net disappearance operator L Creation (fission) operator F CEA July 6th 2016 PAGE 3
4 POWER ITERATION A generalized eigenvalue equation Power iteration algorithm: Guess solution Iterate Hypothesis Convergence CEA July 6th 2016 PAGE 4
5 Source MONTE CARLO APPROACH: CRITICALITY SIMULATION 1 st gen. 2 nd gen. 3 rd gen. 4 th gen. 5 th gen. Fission chain t Source 1 st gen. 2 nd gen. 3 rd gen. N 0 particles Fission generations CEA July 6th 2016 PAGE 5
6 POWER ITERATION: THE STANDARD (?) TOOL Source 1 st gen. G th gen. (G+1) th gen. M th gen. N 0 particles Convergence to the fundamental mode j 1 (statistical equilibrium) Stationarity: sample j 1 = <j 1 (g)> Hypothesis: (I)ID replicas What about correlations? CEA July 6th 2016 PAGE 6
7 A TOY MODEL OF A NUCLEAR REACTOR Neutrons in a box Scattering Capture Fission Descendants per fission Reflecting boundary conditions Assumption: the reactor is critical Expected fundamental mode j 1 : spatially uniform over the box CEA July 6th 2016 PAGE 7
8 IMPACT OF SYSTEM SIZE L ON POWER ITERATION Generations Delta-like source at the center of the box Initial number of neutrons per generation N = 10 4 CEA July 6th 2016 PAGE 8
9 IMPACT OF SYSTEM SIZE L ON POWER ITERATION Neutrons per generation N = 10 4 Neutron clustering CEA July 6th 2016 PAGE 9
10 IMPACT OF POPULATION SIZE N ON POWER ITERATION Generations Delta-like source at the center of the box System size L = 400 cm CEA July 6th 2016 PAGE 10
11 System size L = 400 cm IMPACT OF POPULATION SIZE N ON POWER ITERATION N Neutron clustering CEA July 6th 2016 PAGE 11
12 THE ENTROPY FUNCTION: CONVERGENCE ANALYSIS Shannon entropy: m m Generations to convergence: CEA July 6th 2016 PAGE 12
13 THE EFFECTS OF CLUSTERING ON THE ENTROPY FUNCTION Shannon entropy: Theoretical expected value for independent replicas Measured value Impact of correlations between generations CEA July 6th 2016 PAGE 13
14 ANALYSIS OF SPATIAL MOMENTS: THE CENTER OF MASS CEA July 6th 2016 PAGE 14
15 A STATISTICAL MECHANICS DESCRIPTION Neutrons as a collection of N stochastic particles: {x 1, x 2, x i, x N } A remarkable identity for the spatial moments: Square COM Mean square displacement: Mean square pair distance: CEA July 6th 2016 PAGE 15
16 A STATISTICAL MECHANICS DESCRIPTION Neutrons as a collection of N stochastic particles: {x 1, x 2, x i, x N } A remarkable identity for the spatial moments: Average particle density y Square COM Mean square displacement: Pair correlation function h Mean square pair distance: CEA July 6th 2016 PAGE 16
17 RECIPROCITY OF RANDOM WALKS Forward time flow Backward time flow Measure in z z t Source Measure in z z t Source CEA July 6th 2016 PAGE 17
18 THE AVERAGE NEUTRON DENSITY z Measurement z t Source z 0 CEA July 6th 2016 PAGE 18
19 THE PAIR CORRELATION FUNCTION Correlated measurements z 1 z 2 z 1 z z 2 z z z 0 z t Source z 0 CEA July 6th 2016 PAGE 19
20 THE HOMOGENEOUS CUBE REACTOR Average neutron density: Pair correlation function: System size L Single dimensionless parameter Population size N Migration area M 2 CEA July 6th 2016 PAGE 20
21 POWER ITERATION AS A FUNCTION OF c Uniform initial condition CEA July 6th 2016 PAGE 21
22 POWER ITERATION AS A FUNCTION OF c Uniform initial condition CEA July 6th 2016 PAGE 22
23 THE HOMOGENEOUS CUBE REACTOR: SPATIAL MOMENTS Mean square displacement Mean square pair distance CEA July 6th 2016 PAGE 23
24 THE HOMOGENEOUS CUBE REACTOR: SPATIAL MOMENTS Mean square displacement Mean square pair distance Square COM Fluctuations of COM CEA July 6th 2016 PAGE 24
25 SPATIAL MOMENTS: STATISTICAL ANALYSIS Theory MC Theory MC CEA July 6th 2016 PAGE 25
26 THE HOOGENBOOM-MARTIN PWR BENCHMARK Spatial moments Power iteration CEA July 6th 2016 PAGE 26
27 CONCLUSIONS Statistical mechanics approach to power iteration Neutron clustering can be suppressed by acting on c Applicability to real-world (heterogeneous) systems? CEA July 6th 2016 PAGE 27
28 Thanks for your attention E. Dumonteil et al., Annals of Nuclear Energy 63, (2014). A. Zoia, E. Dumonteil, A. Mazzolo, C. de Mulatier, A. Rosso, Phys. Rev. E 90, (2014). C. de Mulatier, E. Dumonteil, A. Rosso, A. Zoia, J. Stat. Mech. P08021 (2015). B. Houchmandzadeh, E. Dumonteil, A. Mazzolo, A. Zoia, Phys. Rev. E 92, (2015). M. Nowak et al., Annals of Nuclear Energy 94, 856 (2016). CEA July 6th 2016 PAGE 28
29 PAGE 29 CEA 10 AVRIL JUILLET 2016 Commissariat à l énergie atomique et aux énergies alternatives Centre de Saclay Gif-sur-Yvette Cedex T. +33 (0) Secr :+33 (0) Etablissement public à caractère industriel et commercial RCS Paris B DEN/DANS DM2S SERMA
30 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Pure diffusion (ideal gas) Initial condition: N 0 particles with uniform density CEA March 22nd 2016 PAGE 30
31 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Fluctuations: Initial condition: N 0 particles with uniform density CEA March 22nd 2016 PAGE 31
32 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Fluctuations: Initial condition: N 0 particles with uniform density CEA March 22nd 2016 PAGE 32
33 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Diffusion + branching + capture (critical gas) Initial condition: N 0 particles with uniform density CEA March 22nd 2016 PAGE 33
34 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Fluctuations: Initial condition: N 0 particles with uniform density CEA March 22nd 2016 PAGE 34
35 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Mixing time Initial condition: N 0 particles with uniform density Clustering Capture Fission Diffusion CEA March 22nd 2016 PAGE 35
36 SPATIAL BEHAVIOUR OF THE NEUTRON DENSITY Mixing time Renewal time? Initial condition: N 0 particles with uniform density Clustering Capture Fission Diffusion CEA March 22nd 2016 PAGE 36
37 A STATISTICAL MECHANICS DESCRIPTION Neutrons as a collection of N particles Spatial moments: Mean square displacement: Mean square pair distance: Center of mass: CEA April 11th 2016 PAGE 37
THE ROLE OF CORRELATIONS IN MONTE CARLO CRITICALITY SIMULATIONS
THE ROLE OF CORRELATIONS IN MONTE CARLO CRITICALITY SIMULATIONS ANDREA ZOIA CEA/SACLAY Séminaire MANON Andrea ZOIA DEN/DANS/DM2S/SERMA/LTSD March 22nd 2016 APRILE 13, 2016 PAGE 1 OUTLINE q Power iteragon
More informationMechanical modelling of SiC/SiC composites and design criteria
Mechanical modelling of SiC/SiC composites and design criteria F. Bernachy CEA, DEN/DMN/SRMA/LC2M, Gif-sur-Yvette, France L. Gélébart CEA, DEN/DMN/SRMA/LC2M, Gif-sur-Yvette, France J. Crépin Centre des
More informationModelling of Specimen Interaction with Ferrite Cored Coils by Coupling Semi-Analytical and Numerical Techniques
Modelling of Specimen Interaction with Ferrite Cored Coils by Coupling Semi-Analytical and Numerical Techniques A. Skarlatos, E. Demaldent, A. Vigneron and C. Reboud 25 th 28 th of June, Bratislava, Slovak
More informationQUALIFICATION OF A CFD CODE FOR REACTOR APPLICATIONS
QUALIFICATION OF A CFD CODE FOR REACTOR APPLICATIONS Ulrich BIEDER whole TrioCFD Team DEN-STMF, CEA, UNIVERSITÉ PARIS-SACLAY www.cea.fr SÉMINAIRE ARISTOTE, NOVEMBER 8, 2016 PAGE 1 Outline Obective: analysis
More informationNew methods implemented in TRIPOLI-4. New methods implemented in TRIPOLI-4. J. Eduard Hoogenboom Delft University of Technology
New methods implemented in TRIPOLI-4 New methods implemented in TRIPOLI-4 J. Eduard Hoogenboom Delft University of Technology on behalf of Cheikh Diop (WP1.1 leader) and all other contributors to WP1.1
More informationTHERMOHYDRAULIC TRANSIENTS IN BOILING HELIUM NATURAL CIRCULATION LOOPS
THERMOHYDRAULIC TRANSIENTS IN BOILING HELIUM NATURAL CIRCULATION LOOPS Hernán FURCI Director: Chantal MEURIS Supervisor: Bertrand BAUDOUY Laboratoire de Cryogénie et Station d Essais SACM/IRFU/CEA de Saclay
More informationULTRASONIC WAVE PROPAGATION IN DISSIMILAR METAL WELDS APPLICATION OF A RAY-BASED MODEL AND COMPARISON WITH EXPERIMENTAL RESULTS
ULTRASONIC WAVE PROPAGATION IN DISSIMILAR METAL WELDS APPLICATION OF A RAY-BASED MODEL AND COMPARISON WITH EXPERIMENTAL RESULTS Audrey GARDAHAUT 1, Hugues LOURME 1, Frédéric JENSON 1, Shan LIN 2, Masaki
More informationQuality control of neutron-absorber materials for the nuclear fuel cycle, Principle of the JEN-3 neutron Backscattering gauge
Quality control of neutron-absorber materials for the nuclear fuel cycle, Principle of the JEN-3 neutron Backscattering gauge Hamid MAKIL (CEADRT/LIST/LCAE) Patrick BRISSET (IAEA) ICARST 2017, 24 28 April
More informationMAGNETIC ANISOTROPY IN TIGHT-BINDING
MAGNETIC ANISOTROPY IN TIGHT-BINDING Cyrille Barreteau Magnetic Tight-Binding Workshop, London10-11 Sept. 2012 9 SEPTEMBRE 2012 CEA 10 AVRIL 2012 PAGE 1 SOMMAIRE TB Model TB 0 : Mehl & Papaconstantopoulos
More informationInvestigation of Prompt Fission Neutron and Gamma Spectra with their covariance matrices. Application to 239 Pu+n th, 238 U+n 1.
Investigation of Prompt Fission Neutron and Gamma Spectra with their covariance matrices. Application to 239 Pu+n th, 238 U+n 1.8MeV, 235 U+n th O. Litaize, L. Berge, D. Regnier, O. Serot, Y. Peneliau,
More informationCode Strategy for Simulating Severe Accident Scenario
Code Strategy for Simulating Severe Accident Scenario C. SUTEAU, F. SERRE, J.-M; RUGGIERI, F. BERTRAND -CEA- March 4-7, 2013, Paris, France christophe.suteau@cea.fr OULINES INTRODUCTION AND CONTEXT REFERENCE
More informationA DFA ON AES BASED ON THE ENTROPY OF ERROR DISTRIBUTIONS
A DFA ON AES BASED ON THE ENTROPY OF ERROR DISTRIBUTIONS FDTC2012 Ronan Lashermes, Guillaume Reymond, Jean-Max Dutertre, Jacques Fournier, Bruno Robisson and Assia Tria 9 SEPTEMBER 2012 INTRODUCTION Introduction
More informationN/réf. AP/cp Votre N/réf. AP/cp: Ecublens, April 30, 2014
INSTITUT DE PHYSIQUE DE L'ENERGIE ET DES PARTICULES (IPEP) Laboratoire de physique des réacteurs et de comportement des systèmes (LRS) Université Paris Sud Centre d Orsay Service des Etudes Doctorales
More informationSpatio-temporal correlations in fuel pin simulation : prediction of true uncertainties on local neutron flux (preliminary results)
Spatio-temporal correlations in fuel pin simulation : prediction of true uncertainties on local neutron flux (preliminary results) Anthony Onillon Neutronics and Criticality Safety Assessment Department
More information1 Introduction. É É â. EPJ Web of Conferences 42, (2013) C Owned by the authors, published by EDP Sciences, 2013
É É EPJ Web of Conferences 42, 05004 (2013) DOI: 10.1051/ epjconf/ 20134205004 C Owned by the authors, published by EDP Sciences, 2013 Reactivity effect breakdown calculations with deterministic and stochastic
More informationDiffusion coefficients and critical spectrum methods in Serpent
Diffusion coefficients and critical spectrum methods in Serpent Serpent User Group Meeting 2018 May 30, 2018 Espoo, Finland A. Rintala VTT Technical Research Centre of Finland Ltd Overview Some diffusion
More informationRecent Developments in the TRIPOLI-4 Monte-Carlo Code for Shielding and Radiation Protection Applications
Recent Developments in the TRIPOLI-4 Monte-Carlo Code for Shielding and Radiation Protection Applications Yi-Kang LEE, Fadhel MALOUCH, and the TRIPOLI-4 Team CEA-Saclay France Journées Codes de calcul
More informationSupplementary Information:
Supplementary Information: Atomistic Simulation of Solubilization of Polycyclic Aromatic Hydrocarbons in a Sodium Dodecyl Sulfate Micelle Xujun Liang 1,2,3, Massimo Marchi 2,3, Chuling Guo 1,4, Zhi Dang
More informationPrototypes and fuel cycle options including transmutation
A S T R I D Prototypes and fuel cycle options including transmutation General introduction, GEN IV fast reactors Transmutation demonstration Fuel cycle Conclusions www.cea.fr DEN/CAD/DER/CPA Jean-Paul
More informationMass transfer kinetics of uranium(vi) and plutonium(iv) extracted by N,Ndialkylamides. Comparison of different techniques
Mass transfer kinetics of uranium(vi) and plutonium(iv) extracted by N,Ndialkylamides Comparison of different techniques R. Berlemont 1, A. Lélias 1, M. Miguirditchian 1, J.-P. Simonin 2 1- CEA Marcoule,
More informationOn the Use of Shannon Entropy of the Fission Distribution for Assessing Convergence of Monte Carlo Criticality Calculations.
Organized and hosted by the Canadian Nuclear Society. Vancouver, BC, Canada. 2006 September 10-14 On the Use of Shannon Entropy of the Fission Distribution for Assessing Convergence of Monte Carlo Criticality
More informationRecent developments in the TRIPOLI-4 Monte-Carlo code for shielding and radiation protection applications
Recent developments in the TRIPOLI-4 Monte-Carlo code for shielding and radiation protection applications Fadhel Malouch 1,a, Emeric Brun 1, Cheikh Diop 1, François-Xavier Hugot 1, Cédric Jouanne 1, Yi-Kang
More informationNEW DEVELOPMENTS OF AUTORADIOGRAPHY TECHNIQUE TO IMPROVE ALPHA AND BETA MEASUREMENTS FOR DECOMMISSIONING FACILITIES
NEW DEVELOPMENTS OF AUTORADIOGRAPHY TECHNIQUE TO IMPROVE ALPHA AND BETA MEASUREMENTS FOR DECOMMISSIONING FACILITIES PASCAL FICHET, C MOUGEL, Y DESNOYERS, P SARDINI, H WORD DEN SERVICE D ETUDES ANALYTIQUES
More informationPS1-10 COMPARISON OF SIZE-DETERMINING TECHNIQUES FOR NANOPARTICLES IN SUSPENSION: APPLICATION TO Ag NPs.
PS1-10 COMPARISON OF SIZE-DETERMINING TECHNIQUES FOR NANOPARTICLES IN SUSPENSION: APPLICATION TO Ag NPs Sylvie Motellier, Nathalie Pélissier, Jean-Gabriel Mattei, Olivier Sicardy 8 November 2016 Univ.
More informationECT Lecture 2. - Reactor Antineutrino Detection - The Discovery of Neutrinos. Thierry Lasserre (Saclay)
ECT Lecture 2 - Reactor Antineutrino Detection - The Discovery of Neutrinos Thierry Lasserre (Saclay) Reactor Neutrino Detection Inverse Beta Decay p + anti-v e à e + + n cross section @2 MeV : 5 10-43
More informationImpact of Photon Transport on Power Distribution
Impact of Photon Transport on Power Distribution LIEGEARD Clément, CALLOO Ansar, MARLEAU Guy, GIRARDI Enrico Électricité de France, R&D, Simulation neutronique techniques de l information et calcul scientifique
More informationCross Section Generation Strategy for High Conversion Light Water Reactors Bryan Herman and Eugene Shwageraus
Cross Section Generation Strategy for High Conversion Light Water Reactors Bryan Herman and Eugene Shwageraus 1 Department of Nuclear Science and Engineering Massachusetts Institute of Technology 77 Massachusetts
More informationComparison of Monte Carlo methods for adjoint neutron transport
Eur. Phys. J. Plus (2018) 133: 317 DOI 10.1140/epjp/i2018-12132-9 Regular Article THE EUROPEAN PHYSICAL JOURNAL PLUS Comparison of Monte Carlo methods for adjoint neutron transport Vito Vitali 1, Sandra
More informationPreventing xenon oscillations in Monte Carlo burnup calculations by forcing equilibrium
Preventing xenon oscillations in Monte Carlo burnup calculations by forcing equilibrium Aarno Isotaloa), Jaakko Leppänenb), Jan Dufekcc) a) Aalto University, Finland b) VTT Technical Research Centrte of
More informationBEAM DYNAMICS STUDIES FOR HILUMI LHC
BEAM DYNAMICS STUDIES FOR HILUMI LHC BARBARA DALENA IN COLLABORATION WITH: J. PAYET, A. CHANCÉ, O. GABOUEV The HiLumi LHC Design Study is included in the High Luminosity LHC project and is partly funded
More informationReactor radiation skyshine calculations with TRIPOLI-4 code for Baikal-1 experiments
DOI: 10.15669/pnst.4.303 Progress in Nuclear Science and Technology Volume 4 (2014) pp. 303-307 ARTICLE Reactor radiation skyshine calculations with code for Baikal-1 experiments Yi-Kang Lee * Commissariat
More informationMonte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW
Monte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW M. Knebel (Presented by V. Valtavirta) Institute for Neutron Physics and Reactor Technology (INR) Reactor Physics
More informationSolving Bateman Equation for Xenon Transient Analysis Using Numerical Methods
Solving Bateman Equation for Xenon Transient Analysis Using Numerical Methods Zechuan Ding Illume Research, 405 Xintianshiji Business Center, 5 Shixia Road, Shenzhen, China Abstract. After a nuclear reactor
More informationDemonstration of Full PWR Core Coupled Monte Carlo Neutron Transport and Thermal-Hydraulic Simulations Using Serpent 2/ SUBCHANFLOW
Demonstration of Full PWR Core Coupled Monte Carlo Neutron Transport and Thermal-Hydraulic Simulations Using Serpent 2/ SUBCHANFLOW M. Daeubler Institute for Neutron Physics and Reactor Technology (INR)
More informationLecture 20 Reactor Theory-V
Objectives In this lecture you will learn the following We will discuss the criticality condition and then introduce the concept of k eff.. We then will introduce the four factor formula and two group
More informationCore Physics Second Part How We Calculate LWRs
Core Physics Second Part How We Calculate LWRs Dr. E. E. Pilat MIT NSED CANES Center for Advanced Nuclear Energy Systems Method of Attack Important nuclides Course of calc Point calc(pd + N) ϕ dn/dt N
More informationHigh performance computing for neutron diffusion and transport equations
High performance computing for neutron diffusion and transport equations Horizon Maths 2012 Fondation Science Mathématiques de Paris A.-M. Baudron, C. Calvin, J. Dubois, E. Jamelot, J.-J. Lautard, O. Mula-Hernandez
More informationGEM TOMOGRAPHIC MEASUREMENTS FOR WEST AND VALIDATION STRATEGIES
GEM TOMOGRAPHIC MEASUREMENTS FOR WEST AND VALIDATION STRATEGIES 2 nd IAEA TM FDPVA 30 May-2 June 2017 Boston (USA) D.Mazon Many thanks to A. Jardin PAGE 1 GENERAL INTRODUCTION The tomographic SXR project
More informationStudy of X-ray diagnostics for corium-sodium interaction during severe accident scenario
Study of X-ray diagnostics for corium-sodium interaction during severe accident scenario PHENIICS Fest Orsay, 31 st May, 2017 Shifali SINGH DEN/DTN/SMTA/LPMA CEA, Cadarache Christophe Journeau Thesis Director
More informationSENSITIVITY ANALYSIS OF ALLEGRO MOX CORE. Bratislava, Iľkovičova 3, Bratislava, Slovakia
SENSITIVITY ANALYSIS OF ALLEGRO MOX CORE Jakub Lüley 1, Ján Haščík 1, Vladimír Slugeň 1, Vladimír Nečas 1 1 Institute of Nuclear and Physical Engineering, Slovak University of Technology in Bratislava,
More informationESNII + WP7 Fuel Safety
ESNII + WP7 Fuel Safety Nathalie CHAUVIN CEA Cadarache Fuel Studies Department CEA/DEN/Cad ESNII + 24 MARS 2015 17-19 March, 2015 PAGE 1 WP7 Fuel Safety Fuel characteristics of 3 prototypes (starting cores)
More informationMODELLING OF HTRs WITH MONTE CARLO: FROM A HOMOGENEOUS TO AN EXACT HETEROGENEOUS CORE WITH MICROPARTICLES
MODELLING OF HTRs WITH MONTE CARLO: FROM A HOMOGENEOUS TO AN EXACT HETEROGENEOUS CORE WITH MICROPARTICLES Rita PLUKIENE a,b and Danas RIDIKAS a 1 a) DSM/DAPNIA/SPhN, CEA Saclay, F-91191 Gif-sur-Yvette,
More informationSerpent Monte Carlo Neutron Transport Code
Serpent Monte Carlo Neutron Transport Code NEA Expert Group on Advanced Monte Carlo Techniques, Meeting September 17 2012 Jaakko Leppänen / Tuomas Viitanen VTT Technical Research Centre of Finland Outline
More informationA Hybrid Deterministic / Stochastic Calculation Model for Transient Analysis
A Hybrid Deterministic / Stochastic Calculation Model for Transient Analysis A. Aures 1,2, A. Pautz 2, K. Velkov 1, W. Zwermann 1 1 Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Boltzmannstraße
More informationPrompt Alpha Calculation With Monte Carlo Code JMCT
Prompt Alpha Calculation With Monte Carlo Code JMCT Rui LI 1,2, Tao YE 1, Gang LI 1,2 *, Yun BAI 1, Li DENG 1,2 1 Institute of Applied Physics and Computational Mathematics(IAPCM): FengHao East Road No.2,
More informationImprovements of Isotopic Ratios Prediction through Takahama-3 Chemical Assays with the JEFF3.0 Nuclear Data Library
PHYSOR 2004 -The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments Chicago, Illinois, April 25-29, 2004, on CD-ROM, American Nuclear Society, Lagrange Park, IL. (2004) Improvements
More informationNonlinear Iterative Solution of the Neutron Transport Equation
Nonlinear Iterative Solution of the Neutron Transport Equation Emiliano Masiello Commissariat à l Energie Atomique de Saclay /DANS//SERMA/LTSD emiliano.masiello@cea.fr 1/37 Outline - motivations and framework
More informationModeling and Simulation of Dispersion Particle Fuels in Monte Carlo Neutron Transport Calculation
18 th IGORR Conference 2017 Modeling and Simulation of Dispersion Particle Fuels in Monte Carlo Neutron Transport Calculation Zhenping Chen School of Nuclear Science and Technology Email: chzping@yeah.net
More informationFIFRELIN TRIPOLI-4 coupling for Monte Carlo simulations with a fission model. Application to shielding calculations
FIFRELIN TRIPOLI-4 coupling for Monte Carlo simulations with a fission model. Application to shielding calculations Odile Petit 1,*, Cédric Jouanne 1, Olivier Litaize 2, Olivier Serot 2, Abdelhazize Chebboubi
More informationDIRECT EXPERIMENTAL TESTS AND COMPARISON BETWEEN SUB-MINIATURE FISSION CHAMBERS AND SPND FOR FIXED IN-CORE INSTRUMENTATION OF LWR
DIRECT EXPERIMENTAL TESTS AND COMPARISON BETWEEN SUB-MINIATURE FISSION CHAMBERS AND SPND FOR FIXED IN-CORE INSTRUMENTATION OF LWR G. Bignan, J.C. Guyard Commisariat à l Energie Atomique CE CADARACHE DRN/DER/SSAE
More informationKriging models with Gaussian processes - covariance function estimation and impact of spatial sampling
Kriging models with Gaussian processes - covariance function estimation and impact of spatial sampling François Bachoc former PhD advisor: Josselin Garnier former CEA advisor: Jean-Marc Martinez Department
More informationDetermination of the boron content in polyethylene samples using the reactor Orphée
Determination of the boron content in polyethylene samples using the reactor Orphée F. Gunsing, A. Menelle CEA Saclay, F-91191 Gif-sur-Yvette, France O. Aberle European Organization for Nuclear Research
More informationFigure 1. Layout of fuel assemblies, taken from BEAVRS full core benchmark, to demonstrate the modeling capabilities of OpenMC.
Treatment of Neutron Resonance Elastic Scattering Using Multipole Representation of Cross Sections in Monte Carlo Simulations Vivian Y. Tran Benoit Forget Abstract Predictive modeling and simulation plays
More informationR&D ON FUTURE CIRCULAR COLLIDERS
R&D ON FUTURE CIRCULAR COLLIDERS Double Chooz ALICE Edelweiss HESS Herschel CMS Detecting radiations from the Universe. Conseil Scientifique de l Institut 2015 Antoine Chance and Maria Durante MOTIVATIONS
More informationSolving the neutron slowing down equation
Solving the neutron slowing down equation Bertrand Mercier, Jinghan Peng To cite this version: Bertrand Mercier, Jinghan Peng. Solving the neutron slowing down equation. 2014. HAL Id: hal-01081772
More informationLE JOURNAL DE PHYSIQUE - LETTRES
Nous We Tome 46? 21 1 er NOVEMBRE 1985 LE JOURNAL DE PHYSIQUE - LETTRES J. Physique Lett. 46 (1985) L-985 - L-989 1 er NOVEMBRE 1985, L-985 Classification Physics Abstracts 64.70P - 05.40-75.40 Diffusion
More information2013 EDDY CURRENT BENCHMARK PROBLEM: SOLUTION VIA A COUPLED INTEGRAL APPROACH
2013 EDDY CURRENT BENCHMARK PROBLEM: SOLUTION VIA A COUPLED INTEGRAL APPROACH R. MIORELLI 1, C. REBOUD 1 AND T. THEODOULIDIS 2 1 CEA, LIST, Centre de Saclay, Gif-sur-Yvette, France 2 Department of Mechanical
More informationMONTE CARLO SIMULATION OF VHTR PARTICLE FUEL WITH CHORD LENGTH SAMPLING
Joint International Topical Meeting on Mathematics & Computation and Supercomputing in Nuclear Applications (M&C + SNA 2007) Monterey, California, April 5-9, 2007, on CD-ROM, American Nuclear Society,
More informationFission Yield CEA-Cadarache
Fission Yield Activities @ CEA-Cadarache Olivier SEROT CEA-Cadarache DEN/DER/SPRC/LEPh 13108 Saint Paul lez Durance France PAGE 1 Content Fission Yield Measurements on Lohengrin @ Institut Laue Langevin
More informationThe Boltzmann Equation and Its Applications
Carlo Cercignani The Boltzmann Equation and Its Applications With 42 Illustrations Springer-Verlag New York Berlin Heidelberg London Paris Tokyo CONTENTS PREFACE vii I. BASIC PRINCIPLES OF THE KINETIC
More informationMCNP CALCULATION OF NEUTRON SHIELDING FOR RBMK-1500 SPENT NUCLEAR FUEL CONTAINERS SAFETY ASSESMENT
MCNP CALCULATION OF NEUTRON SHIELDING FOR RBMK-15 SPENT NUCLEAR FUEL CONTAINERS SAFETY ASSESMENT R. Plukienė 1), A. Plukis 1), V. Remeikis 1) and D. Ridikas 2) 1) Institute of Physics, Savanorių 231, LT-23
More informationChapter V: Interactions of neutrons with matter
Chapter V: Interactions of neutrons with matter 1 Content of the chapter Introduction Interaction processes Interaction cross sections Moderation and neutrons path For more details see «Physique des Réacteurs
More informationA Hybrid Stochastic Deterministic Approach for Full Core Neutronics Seyed Rida Housseiny Milany, Guy Marleau
A Hybrid Stochastic Deterministic Approach for Full Core Neutronics Seyed Rida Housseiny Milany, Guy Marleau Institute of Nuclear Engineering, Ecole Polytechnique de Montreal, C.P. 6079 succ Centre-Ville,
More informationNanocomposite for building constructions and civil infrastructures: European network pilot production line to promote industrial application cases.
PROJECT NANOLEAP (646397)- H2020-NMP-PILOTS-2014 NMP1-2014: Open access pilot plants for cost effective nanocomposites Nanocomposite for building constructions and civil infrastructures: European network
More informationOn the use of SERPENT code for few-group XS generation for Sodium Fast Reactors
On the use of SERPENT code for few-group XS generation for Sodium Fast Reactors Raquel Ochoa Nuclear Engineering Department UPM CONTENTS: 1. Introduction 2. Comparison with ERANOS 3. Parameters required
More informationIntroduction. Problem Summary. Attila Code-to-Code Comparison Subcritical Spent Nuclear Fuel Canister with Primary Neutron and Gamma Sources
Introduction An example spent nuclear fuel canister calculation was performed with Attila, with results compared to the MCNPX Monte Carlo code. The canister contained representative subcritical neutron
More information2. The Steady State and the Diffusion Equation
2. The Steady State and the Diffusion Equation The Neutron Field Basic field quantity in reactor physics is the neutron angular flux density distribution: Φ( r r, E, r Ω,t) = v(e)n( r r, E, r Ω,t) -- distribution
More informationFuel BurnupCalculations and Uncertainties
Fuel BurnupCalculations and Uncertainties Outline Review lattice physics methods Different approaches to burnup predictions Linkage to fuel safety criteria Sources of uncertainty Survey of available codes
More informationLesson 6: Diffusion Theory (cf. Transport), Applications
Lesson 6: Diffusion Theory (cf. Transport), Applications Transport Equation Diffusion Theory as Special Case Multi-zone Problems (Passive Media) Self-shielding Effects Diffusion Kernels Typical Values
More informationCost-accuracy analysis of a variational nodal 2D/1D approach to pin resolved neutron transport
Cost-accuracy analysis of a variational nodal 2D/1D approach to pin resolved neutron transport ZHANG Tengfei 1, WU Hongchun 1, CAO Liangzhi 1, LEWIS Elmer-E. 2, SMITH Micheal-A. 3, and YANG Won-sik 4 1.
More informationTime-Dependent Statistical Mechanics 1. Introduction
Time-Dependent Statistical Mechanics 1. Introduction c Hans C. Andersen Announcements September 24, 2009 Lecture 1 9/22/09 1 Topics of concern in the course We shall be concerned with the time dependent
More informationLesson 9: Multiplying Media (Reactors)
Lesson 9: Multiplying Media (Reactors) Laboratory for Reactor Physics and Systems Behaviour Multiplication Factors Reactor Equation for a Bare, Homogeneous Reactor Geometrical, Material Buckling Spherical,
More informationMarkov Chain Monte Carlo The Metropolis-Hastings Algorithm
Markov Chain Monte Carlo The Metropolis-Hastings Algorithm Anthony Trubiano April 11th, 2018 1 Introduction Markov Chain Monte Carlo (MCMC) methods are a class of algorithms for sampling from a probability
More informationProcessing of Basic Nuclear Data for Criticality Coefficients Calculations of Fast Homogeneous Systems
Processing of Basic Nuclear Data for Criticality Coefficients Calculations of Fast Homogeneous Systems Jefferson Neves Pereira jeffersonnevespereira@gmail.com Instituto Militar de Engenharia Sergio de
More informationCalculation of a Reactivity Initiated Accident with a 3D Cell-by-Cell Method: Application of the SAPHYR System to a Rod Ejection Accident in TMI1
Calculation of a Reactivity Initiated Accident with a 3D Cell-by-Cell Method: Application of the SAPHYR System to a Rod Ejection Accident in TMI1 S. Aniel-Buchheit 1, E. Royer 2, P. Ferraresi 3 1 S. Aniel
More informationSimple benchmark for evaluating self-shielding models
Simple benchmark for evaluating self-shielding models The MIT Faculty has made this article openly available. Please share how this access benefits you. Your story matters. Citation As Published Publisher
More informationNonclassical Particle Transport in Heterogeneous Materials
Nonclassical Particle Transport in Heterogeneous Materials Thomas Camminady, Martin Frank and Edward W. Larsen Center for Computational Engineering Science, RWTH Aachen University, Schinkelstrasse 2, 5262
More informationA Deterministic against Monte-Carlo Depletion Calculation Benchmark for JHR Core Configurations. A. Chambon, P. Vinai, C.
A Deterministic against Monte-Carlo Depletion Calculation Benchmark for JHR Core Configurations A. Chambon, P. Vinai, C. Demazière Chalmers University of Technology, Department of Physics, SE-412 96 Gothenburg,
More informationStatus of MORET5 source convergence improvements and benchmark proposal for Monte Carlo depletion calculations
Status of MORET5 source convergence improvements and benchmark proposal for Monte Carlo depletion calculations Y. Richet ; W. Haeck ; J. Miss Criticality analysis department Study, Research, Codes Development
More informationVladimir Sobes 2, Luiz Leal 3, Andrej Trkov 4 and Matt Falk 5
A Study of the Required Fidelity for the Representation of Angular Distributions of Elastic Scattering in the Resolved Resonance Region for Nuclear Criticality Safety Applications 1 Vladimir Sobes 2, Luiz
More informationSENSITIVITY AND PERTURBATION THEORY IN FAST REACTOR CORE DESIGN
Journal of ELECTRICAL ENGINEERING, VOL. 65, NO. 7s, 214, 25 29 SENSITIVITY AND PERTURBATION THEORY IN FAST REACTOR CORE DESIGN Jakub Lüley Branislav Vrban Štefan Čerba Ján Haščík Vladimír Nečas Sang-Ji
More informationNeutron capture and fission reactions on. sections, -ratios and prompt -ray emission from fission. 1 Introduction and Motivation
EPJ Web of Conferences 42, 01002 (2013) DOI: 10.1051/ epjconf/ 20134201002 C Owned by the authors, published by EDP Sciences, 2013 Neutron capture and fission reactions on 235 U : cross sections, -ratios
More informationPhD Qualifying Exam Nuclear Engineering Program. Part 1 Core Courses
PhD Qualifying Exam Nuclear Engineering Program Part 1 Core Courses 9:00 am 12:00 noon, November 19, 2016 (1) Nuclear Reactor Analysis During the startup of a one-region, homogeneous slab reactor of size
More informationORIGAMIX, A CDTE-BASED SPECTRO-IMAGER DEVELOPMENT FOR NUCLEAR APPLICATIONS
ORIGAMIX, A CDTE-BASED SPECTRO-IMAGER DEVELOPMENT FOR NUCLEAR APPLICATIONS Sébastien Dubos 1 Hermine Lemaire 2 Frédérick Carrel 2 Olivier Limousin 1 Aline Meuris 1 Stéphane Schanne 1 Vincent Schoepff 2
More informationLE JOURNAL DE PHYSIQUE - LETTRES
05.50 On We Tome 45 No 14 15 JUILLET 1984 LE JOURNAL DE PHYSIQUE - LETTRES J. Physique Lett. 45 (1984) L-701 - L-706 15 JUILLET 1984, L-701 Classification Physics Abstracts - - 05.20 64.60 On a model of
More informationLecture 7 Problem Set-2
Objectives In this lecture you will learn the following In this lecture we shall practice solving problems. We will solve 5 out of 10 problems in Assignment-2. Background Information Mole Molecular weight
More informationComparison of the Monte Carlo Adjoint-Weighted and Differential Operator Perturbation Methods
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol., pp.836-841 (011) ARTICLE Comparison of the Monte Carlo Adjoint-Weighted and Differential Operator Perturbation Methods Brian C. KIEDROWSKI * and Forrest
More informationPART I INTRODUCTION The meaning of probability Basic definitions for frequentist statistics and Bayesian inference Bayesian inference Combinatorics
Table of Preface page xi PART I INTRODUCTION 1 1 The meaning of probability 3 1.1 Classical definition of probability 3 1.2 Statistical definition of probability 9 1.3 Bayesian understanding of probability
More informationHomogenization Methods for Full Core Solution of the Pn Transport Equations with 3-D Cross Sections. Andrew Hall October 16, 2015
Homogenization Methods for Full Core Solution of the Pn Transport Equations with 3-D Cross Sections Andrew Hall October 16, 2015 Outline Resource-Renewable Boiling Water Reactor (RBWR) Current Neutron
More informationTwo simple lattice models of the equilibrium shape and the surface morphology of supported 3D crystallites
Bull. Nov. Comp. Center, Comp. Science, 27 (2008), 63 69 c 2008 NCC Publisher Two simple lattice models of the equilibrium shape and the surface morphology of supported 3D crystallites Michael P. Krasilnikov
More informationConvergence Analysis and Criterion for Data Assimilation with Sensitivities from Monte Carlo Neutron Transport Codes
PHYSOR 2018: Reactor Physics paving the way towards more efficient systems Cancun, Mexico, April 22-26, 2018 Convergence Analysis and Criterion for Data Assimilation with Sensitivities from Monte Carlo
More informationFirst-Passage Kinetic Monte Carlo Algorithm for Complex Reaction-Diffusion Systems
First-Passage Kinetic Monte Carlo Algorithm for Complex Reaction-Diffusion Systems Aleksandar Donev 1 Lawrence Postdoctoral Fellow Lawrence Livermore National Laboratory In collaboration with: Vasily V.
More informationMonte Caro simulations
Monte Caro simulations Monte Carlo methods - based on random numbers Stanislav Ulam s terminology - his uncle frequented the Casino in Monte Carlo Random (pseudo random) number generator on the computer
More informationPerturbation/sensitivity calculations with Serpent
SERPENT workshop Cambridge, 17-19 September 2014 Perturbation/sensitivity calculations with Serpent Manuele Auero, Adrien Bidaud, Pablo Rubiolo LPSC/CNRS Grenoble Calculating the complete β e...coupling
More informationCovariance Generation using CONRAD and SAMMY Computer Codes
Covariance Generation using CONRAD and SAMMY Computer Codes L. Leal a, C. De Saint Jean b, H. Derrien a, G. Noguere b, B. Habert b, and J. M. Ruggieri b a Oak Ridge National Laboratory b CEA, DEN, Cadarache
More informationVVER-1000 Coolant Transient Benchmark - Overview and Status of Phase 2
International Conference Nuclear Energy for New Europe 2005 Bled, Slovenia, September 5-8, 2005 VVER-1000 Coolant Transient Benchmark - Overview and Status of Phase 2 Nikola Kolev, Nikolay Petrov Institute
More informationA PERTURBATION ANALYSIS SCHEME IN WIMS USING TRANSPORT THEORY FLUX SOLUTIONS
A PERTURBATION ANALYSIS SCHEME IN WIMS USING TRANSPORT THEORY FLUX SOLUTIONS J G Hosking, T D Newton, B A Lindley, P J Smith and R P Hiles Amec Foster Wheeler Dorchester, Dorset, UK glynn.hosking@amecfw.com
More informationK-effective of the World and Other Concerns for Monte Carlo Eigenvalue Calculations
Progress in NULER SIENE and TEHNOLOGY, Vol. 2, pp.738-742 (2011) RTILE K-effective of the World and Other oncerns for Monte arlo Eigenvalue alculations Forrest. ROWN Los lamos National Laboratory, Los
More informationFuel cycle studies on minor actinide transmutation in Generation IV fast reactors
Fuel cycle studies on minor actinide transmutation in Generation IV fast reactors M. Halász, M. Szieberth, S. Fehér Budapest University of Technology and Economics, Institute of Nuclear Techniques Contents
More informationHT HT HOMOGENIZATION OF A CONDUCTIVE AND RADIATIVE HEAT TRANSFER PROBLEM, SIMULATION WITH CAST3M
Proceedings of HT2005 2005 ASME Summer Heat Transfer Conference July 17-22, 2005, San Francisco, California, USA HT2005-72193 HT2005-72193 HOMOGENIZATION OF A CONDUCTIVE AND RADIATIVE HEAT TRANSFER PROBLEM,
More information