Development of a dynamic stochastic neutronic code for the analysis of conventional and hybrid nuclear reactors Thalia Xenofontos To cite this version: Thalia Xenofontos. Development of a dynamic stochastic neutronic code for the analysis of conven- tional and hybrid nuclear reactors. Nuclear Experiment [nucl-ex]. Université Paris Saclay (COmUE); Université Aristote (Thessalonique, Grèce), 2018. English. NNT : 2018SACLX013. tel-01865831 HAL Id: tel-01865831 https://pastel.archives-ouvertes.fr/tel-01865831 Submitted on 2 Sep 2018 HAL is a multi-disciplinary open access L’archive ouverte pluridisciplinaire HAL, est archive for the deposit and dissemination of sci- destinée au dépôt et à la diffusion de documents entific research documents, whether they are pub- scientifiques de niveau recherche, publiés ou non, lished or not. The documents may come from émanant des établissements d’enseignement et de teaching and research institutions in France or recherche français ou étrangers, des laboratoires abroad, or from public or private research centers. publics ou privés. Développement d’un code neutronique stochastique dynamique pour l’analyse de 2018SACLX013 : réacteurs nucléaires NNT conventionnels et hybrides Thèse de doctorat de l'Université Paris-Saclay préparée à l’Ecole Polytechnique (France) et à l’Université Aristote de Thessalonique (Grèce) École doctorale n°576 Particules hadrons énergie et noyau : instrumentation, image, cosmos et simulation (Pheniics) Spécialité de doctorat: Energie Nucléaire Thèse présentée et soutenue à Thessalonique, Grèce, le 19 Janvier 2018, par Thalia A. Xenofontos Composition du Jury : Mr Nicolas Catsaros Directeur de Recherche, Centre Nationale pour la Recherche Scientifique “Demokritos”, Grèce Président Mr Evangelos Gazis Professeur, Université Nationale Polytechnique d’Athènes “Metsovio”, Grèce Rapporteur Mr Ivan Kodeli Directeur de Recherche, Institut Jozef Stefan, Slovenie Rapporteur Mme Melpomeni Varvayanni Directrice de Recherche, Centre Nationale pour la Recherche Scientifique “Demokritos”, Grèce Examinatrice Mr Constantin Meis Professeur, CEA - INSTN, Université Paris-Saclay, France Examinateur Mr Marc-Thierry Jaekel Directeur de Recherche, École Normale Supérieure de Paris, France Directeur de thèse Mr Alexandros Clouvas Professeur, Université Aristote de Thessalonique, Grèce Co-Directeur de thèse Στον πατέρα μου που οι συνθήκες δεν μου επέτρεψαν να γνωρίσω και σε αυτούς που με βοήθησαν να διακρίνω και να συνεισφέρω στην φωτεινή εξέλιξη της ανθρωπότητας. Table of Contents ABSTRACT ............................................................................................................................. 1 ACKNOWLEDGMENTS ........................................................................................................ 3 1 INTRODUCTION ............................................................................................................ 5 2 THE FISSION NUCLEAR REACTORS ....................................................................... 10 2.1 Nuclear Reactors Components ................................................................................. 10 2.2 Power Nuclear Reactors ........................................................................................... 12 3 METHODOLOGIES FOR NEUTRONIC ANALYSIS OF REACTOR CORES.........15 3.1 Neutron Transport and Criticality Equation ............................................................. 15 3.2 Methodologies for the Solution of the Neutron Transport Equation ....................... 16 3.2.1 Deterministic Approach ................................................................................... 17 3.2.2 Stochastic (Monte Carlo) Approach ................................................................. 17 3.3 Fuel Depletion Equation .......................................................................................... 19 3.4 Methodologies for the Solution of the Depletion Equation ..................................... 20 3.4.1 Transmutation Trajectory Analysis (TTA) ...................................................... 20 3.4.2 Matrix Exponential Methods ............................................................................ 21 3.5 Well Established and Under Development Monte Carlo Codes .............................. 23 3.6 State of the Art in the ADSs Simulation .................................................................. 28 3.6.1 Fuel Depletion Mechanisms ............................................................................. 28 3.6.2 Spallation Process ............................................................................................ 28 4 THE ANET CODE ......................................................................................................... 32 4.1 Criticality Calculations ............................................................................................ 34 4.1.1 Collision Estimator ........................................................................................... 34 4.1.2 Absorption Estimator ....................................................................................... 35 4.1.3 Track Length Estimator .................................................................................... 36 4.2 Flux Calculations ..................................................................................................... 36 4.3 Reaction Rates Calculations ..................................................................................... 37 4.4 Dynamic Assessment of Core Isotopic Composition .............................................. 38 5 SETUPS OF THE SIMULATED INSTALLATIONS ................................................... 49 5.1 The Training Nuclear Reactor Model 9000 of the Aristotle University of Thessaloniki (TNR-AUTh) ...................................................................................... 49 5.2 The Portuguese Research Reactor (RPI) .................................................................. 54 5.3 The VENUS Facility ................................................................................................ 58 5.4 OECD/NEA Burnup Credit Calculation Benchmark ............................................... 59 5.5 Kyoto University Critical Assembly (KUCA) ......................................................... 61 6 ANET VALIDATION & VERIFICATION STUDIES AND RESULTS ..................... 68 6.1 Criticality Assessment .............................................................................................. 68 6.2 Flux Assessment ...................................................................................................... 70 6.2.1 Measurements ................................................................................................... 70 6.2.2 Simulations ....................................................................................................... 71 6.3 Fission Rate Distribution Assessment ...................................................................... 77 6.3.1 Measurements ................................................................................................... 77 6.3.2 Simulations ....................................................................................................... 78 6.4 Time Dependent ANET Calculations ...................................................................... 86 6.5 Accelerator Driven Systems ANET Simulations ..................................................... 89 7 CONCLUSIONS ............................................................................................................. 91 8 FUTURE WORK AND PERSPECTIVES ..................................................................... 93 REFERENCES ....................................................................................................................... 95 SUMMARY IN GREEK ...................................................................................................... 104 SUMMARY IN FRENCH ................................................................................................... 109 LIST OF PUBLICATIONS ................................................................................................. 114 APPENDIX I ........................................................................................................................ 116 Fission Power Nuclear Reactor Designs .............................................................................. 116 REFERENCES ..................................................................................................................... 130 APPENDIX II ....................................................................................................................... 133 Accelerator Driven Systems (ADS) ..................................................................................... 133 REFERENCES ..................................................................................................................... 142 APPENDIX III ...................................................................................................................... 145 ABSTRACT The necessity for precise simulations of a nuclear reactor especially in case of complex core and fuel configurations has imposed the increasing use of Monte Carlο neutronics codes. Besides, a demand of additional stochastic codes’ inherent capabilities has emerged regarding mainly the simulation of the temporal variations in the core isotopic composition as well as the incorporation of the T-H feedback. In addition to the above,