Titre :	Study and modeling of a Heat pipe Microreactor core using OpenMC code
Type de document :	document électronique
Auteurs :	Anfel Ounis, Auteur ; Salah-Eddine Bentridi, Directeur de thèse
Editeur :	Setif:UFA
Année de publication :	2024
Importance :	1 vol (46 f.)
Format :	29 cm
Langues :	Anglais (eng)
Catégories :	Thèses & Mémoires:Physique
Mots-clés :	Physique
Index. décimale :	530 - Physique
Résumé :	In this study, a microreactor model for space missions is studied. The microreactor, made from a solid monobloc moderator including fuel rods and heat-pipes and utilizes the OpenMC Monte- Carlo code for neutronic study and geometrical modelling. Characteristics and features of the model are based on recent works from an Argentinean compact microreactor project.
Note de contenu :	Sommaire Introduction 06 Chapter 1: Overview of Nuclear Microreactors 07 I.1. Introduction 07 I.2. Microreactors designs and technological features 08 I.2.1. Enhancement of reactivity control 09 I.2.2. The coolant choice for efficient heat transfer 09 I.2.3. Better Confinement of radioactive materials 11 I.3. The most relevant advantages of microreactors 11 I.4. Relevant challenges of Microreactors 14 Chapter 02: Modeling and simulation of a Heat-pipe Microreactor with OpenMC code 16 II.1. The chosen heat-pipe microreactor model 16 II.2. Design aspects and requirements 17 II.2.1. Design aspects 17 II.2.2. Design requirements 17 II.3. Characteristics and features of the chosen microreactor model 18 II.4. The modelling of the Microreactor Core by using OpenMC 18 II.4.1. Physical model: Definition of materials 19 II.4.2. Geometrical model: Definition of surfaces and cells 21 II.4.3. Definition of simulation Settings: 25 Chapter 3: Results and Discussion of OpenMC simulations 27 III.1. Definition of Tallies (Recording functions) 27 III.1.1. Spectrum Tally 27 III.1.2. Mesh tally 28 III.2. Criticality and neutron flux spectrum 29 III.2.1. Criticality results 29 III.2.2. Neutron flux spectrum 29 III.3. Neutron flux distribution and profile 30 III.3.1 Neutron flux XY-distribution 31 III.3.2. Neutron flux YZ-distribution 33 III.4. Power density distribution and power profile 35 III.5. Reflector reactivity worth 36 III.5.1. The effect of reflector extraction on the reactor criticality 36 III.5.2. Integral reactivity worth of the lateral reflector 36 III.5.3. Differential reactivity worth of the lateral reflector 36 III.5.4. The deformation of flux and power density distribution 38 Conclusion 41 Bibliography
Côte titre :	MAPH/0658

Study and modeling of a Heat pipe Microreactor core using OpenMC code [document électronique] / Anfel Ounis, Auteur ; Salah-Eddine Bentridi, Directeur de thèse . - [S.l.] : Setif:UFA, 2024 . - 1 vol (46 f.) ; 29 cm.
Langues : Anglais (eng)

Catégories :	Thèses & Mémoires:Physique
Mots-clés :	Physique
Index. décimale :	530 - Physique
Résumé :	In this study, a microreactor model for space missions is studied. The microreactor, made from a solid monobloc moderator including fuel rods and heat-pipes and utilizes the OpenMC Monte- Carlo code for neutronic study and geometrical modelling. Characteristics and features of the model are based on recent works from an Argentinean compact microreactor project.
Note de contenu :	Sommaire Introduction 06 Chapter 1: Overview of Nuclear Microreactors 07 I.1. Introduction 07 I.2. Microreactors designs and technological features 08 I.2.1. Enhancement of reactivity control 09 I.2.2. The coolant choice for efficient heat transfer 09 I.2.3. Better Confinement of radioactive materials 11 I.3. The most relevant advantages of microreactors 11 I.4. Relevant challenges of Microreactors 14 Chapter 02: Modeling and simulation of a Heat-pipe Microreactor with OpenMC code 16 II.1. The chosen heat-pipe microreactor model 16 II.2. Design aspects and requirements 17 II.2.1. Design aspects 17 II.2.2. Design requirements 17 II.3. Characteristics and features of the chosen microreactor model 18 II.4. The modelling of the Microreactor Core by using OpenMC 18 II.4.1. Physical model: Definition of materials 19 II.4.2. Geometrical model: Definition of surfaces and cells 21 II.4.3. Definition of simulation Settings: 25 Chapter 3: Results and Discussion of OpenMC simulations 27 III.1. Definition of Tallies (Recording functions) 27 III.1.1. Spectrum Tally 27 III.1.2. Mesh tally 28 III.2. Criticality and neutron flux spectrum 29 III.2.1. Criticality results 29 III.2.2. Neutron flux spectrum 29 III.3. Neutron flux distribution and profile 30 III.3.1 Neutron flux XY-distribution 31 III.3.2. Neutron flux YZ-distribution 33 III.4. Power density distribution and power profile 35 III.5. Reflector reactivity worth 36 III.5.1. The effect of reflector extraction on the reactor criticality 36 III.5.2. Integral reactivity worth of the lateral reflector 36 III.5.3. Differential reactivity worth of the lateral reflector 36 III.5.4. The deformation of flux and power density distribution 38 Conclusion 41 Bibliography
Côte titre :	MAPH/0658