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MINISTRY OF EDUCATION AND TRAINING MINISTRY OF SCIENCE AND TECHNOLOGY VIETNAM ATOMIC ENERGY INSTITUTE - NGUYỄN KIÊN CƯỜNG Study on neutronics, thermal hydraulics and core management for safe operation and effective utilization of the Dalat Nuclear Research Reactor A thesis submitted in fulfillment of the requirements for the degree of Doctor of Philosophy HÀ NỘI – 2023 MINISTRY OF EDUCATION AND TRAINING TECHNOLOGY MINISTRY OF SCIENCE AND VIETNAM ATOMIC ENERGY INSTITUTE - NGUYỄN KIÊN CƯỜNG Study on neutronics, thermal hydraulics and core management for safe operation and effective utilization of the Dalat Nuclear Research Reactor DOCTORAL THESIS Subject: Atomic and Nuclear Physics Code number: 9-44-01-06 Supervisor Assoc Prof PhD NGUYỄN NHỊ ĐIỀN : Hà Nội – 2023 DECLARATION OF AUTHORSHIP I, Nguyễn Kiên Cường, declare that this thesis titled, ―Study on neutronics, thermal hydraulics and core management for safe operation and effective utilization of the Dalat Nuclear Research Reactor‖ is my own work, conducted under the supervision of Assoc Prof PhD Nguyễn Nhị Điền, and has not been published in any other works or articles The results were coauthored with other authors after having permission to use for the thesis Author Nguyễn Kiên Cường ACKNOWLEDGEMENTS I would like to express my deep gratitude to my supervisor Assoc Prof PhD Nguyễn Nhị Điền for his valuable assistance and encouragement throughout my research works I would like to thank all my colleagues at the Reactor Center, particularly Mr Huỳnh Tôn Nghiêm, Mr Lê Vĩnh Vinh, Mr Lương Bá Viên and Mr Nguyễn Minh Tuân for their generous support, insightful discussions and contributions to my study I appreciate the staff of the Dalat Nuclear Research Institute, Nuclear Training Center, and Vietnam Atomic Energy Institute for their kind assistance during my research I want to send my appreciation to Assoc Prof PhD Vương Hữu Tấn, Assoc Prof PhD Phạm Đình Khang, Assoc Prof PhD Nguyễn Xuân Hải, PhD Trần Chí Thành, Assoc Prof PhD Nguyễn Tuấn Khải, Assoc Prof PhD Trịnh Anh Đức for their encouragement for my research works I am grateful to Ms Nguyễn Thúy Hằng for all of her assistance with the administrative aspects to my research Finally, I would like to express my gratitude to my parents, my wife, my son and my daughter, who always provided me with the confidence and motivation to complete the thesis BOC BWR CFD CHF CITATION DNBR DNRI DNRR ENDF FA FIR FPD GA HANARO HEU IAEA IFA JEFF JENDL LEU MCNP MPI MTR NPP ONBR PVM PWR ReR RERTR RIA SA SaR ShR SRAC TRIGA VVER WIMSD 1-D 2-D 3-D LIST OF ABBREVIATIONS Beginning of Cycle Boiling Water Reactor Computational Fluid Dynamics Critical Heat Flux Nuclear Reactor Core Analysis Code Departure Nucleate Boiling Ratio Dalat Nuclear Research Institute Dalat Nuclear Research Reactor Evaluated Nuclear Data File Fuel Assembly Flow Instability Ratio Full Power Day Genetic Algorithm Korean Research Reactor Highly Enriched Uranium International Atomic Energy Agency Instrumental Fuel Assembly Joint Evaluated Fission and Fusion (European Evaluated) Nuclear Data Library Japanese Evaluated Nuclear Data Library Low Enriched Uranium Monte Carlo N-Particle Computer Code Message Passing Interface Material Testing Reactor Nuclear Power Plant Onset Nucleate Boiling Ratio Parallel Virtual Machine Pressurized Water Reactor Automatic Regulating Rod Reduced Enrichment of Research and Test Reactor Reactivity Insertion Accident Simulated Annealing Safety Rod Shim Rod Standard Reactor Analysis Code Training, Research, Isotope Production of General Atomics Water-Water Energetic Reactor Winfrith Improved Multigroup Scheme Unidimensional Bidimensional Tridimensional CONTENT ACKNOWLEDGEMENTS LIST OF ABBREVIATIONS CONTENT LIST OF TABLES 10 INTRODUCTION 11 CHAPTER GENERAL OVERWIEW 21 1.1 Motivation of the thesis 21 1.2 General introduction about the DNRR 23 1.2.1 History, structure and reactor core arrangement 23 1.2.2 Fuel of VVR-M2 HEU and LEU 31 1.2.3 Neutronics characteristics of the DNRR 34 1.2.4 Thermal hydraulics characteristics of the DNRR 36 1.3 The development of computer codes for reactor calculation in the world 38 1.4 The research situation about research reactor in Vietnam 41 1.5 Reactor kinetics in three dimensions 44 1.6 Burn-up calculation for core and fuel management 46 CHAPTER CALCULATION MODELS FOR THE DALAT NUCLEAR RESEARCH REACTOR USING LEU FUEL 48 2.1 Neutronics calculation models 48 2.1.1 Deterministic code 48 2.1.2.1 Lattice cell model 48 2.1.2.2 Whole core model 54 2.1.2 Calculation model for computer codes using Monte Carlo method 57 2.2 Thermal hydraulics calculation for the DNRR 59 2.3 Reactor kinetics application for the DNRR 62 2.3.1 Preparation group constants for the PARCS code 62 2.3.2 Calculation model for the DNRR using the PARCS code 63 2.4 Burn-up calculation for the DNRR 64 2.4.1 Development MCDL computer code 64 2.4.2 Application of MCDL for burn-up and refueling calculation for LEU core 73 2.5 Summary of Chapter 74 CHAPTER RESULTS AND DISSCUSSIONS 76 3.1 Neutronics and thermal hydraulics for LEU core 76 3.1.1 Neutronics calculation results 76 3.1.1.1 Neutronics characteristics of the HEU and LEU VVR-M2 fuel types 76 3.1.1.2 Criticality, reactivity, control rod worths 80 3.1.1.3 Excess reactivity, control rod worth, beryllium rods 85 3.1.1.4 Neutron flux distribution 88 3.1.1.5 Power peaking factor 92 3.1.1.6 Feedback reactivity temperature coefficients and void factor 95 3.1.1.7 Kinetics parameters 96 3.1.2 Thermal hydraulics calculation results .97 3.1.2.1 Validation of the PLTEMP code 97 3.1.2.2 Steady state of PLTEMP code without hot channel factors .101 3.1.2.3 Steady state of PLTEMP code with hot channel factors 103 3.2 Kinetics calculation results for LEU core 105 3.2.1 Calculation results from the Serpent and PARCS codes at steady state condition 105 3.2.2 Calculation and experiment results during increasing power of the DNRR from 80% to 100% 107 3.2.3 Simulation of the accident when uncontrolled withdrawal of one control rod at nominal power .110 3.2.4 Simulation of the changing of power when inserting positive reactivity smaller than 10 cents 113 3.3 Burn-up calculation results 115 3.3.1 Validation of the MCDL code 115 3.3.2 Calculation results of the HEU core .121 3.3.3 Calculation results of the LEU core .125 3.4 Summary of Chapter .133 CONCLUSIONS 135 NEW CONTRIBUTIONS OF THE THESIS .139 LIST OF PUBLICATIONS 140 REFERENCES 142 LIST OF FIGURES Fig 1.1 Cross sections of the DNRR in axial and radial directions 24 Fig 1.2 Annual operation time of the DNRR from 1984 to 2011 (the core configuration using HEU fuel and mixed cores) 30 Fig 1.3 Annual operation time of the DNRR from 2012 to 2022 (using LEU fuel) 31 Fig 1.4 Specific dimensions and geometry of the VVR-M2 fuel type 33 Fig 2.1 Calculation model for VVR-M2 FA of the DNRR 50 Fig 2.2 Calculation model for group constants of the neutron trap 52 Fig 2.3 Calculation model for lattice cells in side the DNRR core 53 Fig 2.4 Calculation model for lattice cells outside the DNRR core 54 Fig 2.5 Calculation model for the DNRR using the REBUS-PC and CITATION codes (a- model in REBUS-PC and b- model in CITATION) 55 Fig 2.6 Calculation model in axial direction using the CITATION code 56 Fig 2.7 Model cross section of VVR-M2 FA in MCNP code for normal calculation and power peaking factor calculation (a- FA; b- FA model in the MCNP code and cexample of detailed power peaking factor calculation inside FA) 57 Fig 2.8 Calculation model for a) SaRs or ShRs, b) ReR, c) beryllium rod, d) aluminum chock rod, e) wet or dry irradiation channels, f) the neutron trap 58 Fig 2.9 Calculation model for full core of the DNRR using the MCNP code 59 Fig 2.10 The DNRR model calculation for the PLTEMP/ANL code and LEU core with 92 FAs 60 Fig 2.11 Super-cell model in the Serpent code to create group constants of ShR 63 Fig 2.12 Calculation model of the DNRR using the PARCS code 64 Fig 2.13 General structure of the MCDL code 69 Fig 2.14 Burn-up chain model of actinide and fission products isotopes in the MCDL code 71 Fig 3.1 Neutron spectrum of HEU and LEU VVR-M2 fuels with 108 neutron energy groups in average power (89 HEU FAs core and 92 LEU FAs core) 78 Fig 3.2 Neutron spectrum of LEU VVR-M2 fuel type with different calculation libraries 79 Fig 3.3 Critical core configuration with a) 72 FAs and b) working core with 92 FAs 84 Fig 3.4 Thermal neutron flux distribution in the radial direction (unit ×1012 n/cm2.s) 92 Fig 3.5 Relative power distribution in axial direction and depending on control rod positions of working core of 92 LEU FAs 94 Fig 3.6 Calculation results of relative power distribution in the working core using 92 LEU FAs (upper value from MCNP and below value from REBUS) 95 Fig 3.7 Comparison of the measured cladding and coolant temperatures of the HEU core to validate the PLTEMP/ANL code 98 Fig 3.8 The HEU VVR-M2 IFA of the DNRR 99 Fig 3.9 Axial power distribution of the hottest FA (at cell 10-5) calculated by the MCNP code for 25-cm insertion of ShRs 100 Fig 3.10 Calculation results at nominal power without errors and uncertainties in the hottest FA 102 Fig 3.11 Comparison of the fuel cladding and coolant temperatures at different reactor power levels 103 Fig 3.12 Calculation results at nominal power with systematic errors 104 Fig 3.13 Calculation results at nominal power with systematic and random errors 105 Fig 3.14 Calculation results of the increasing power of the DNRR from 80 to 100% 108 Fig 3.15 Experimental data of the changing position of ReR (TD) when increasing reactor power from 80% to 100% 109 Fig 3.16 Position of ReR (TD) and power when increasing reactor power from 80% to 100% 109 Fig 3.17 Reactivity and reactor power when increasing reactor power from 80 to 100% 110 Fig 3.18 Calculation results the changing position of ReR (TD) when increasing power from 80 to 100% 110 Fig 3.19 Power and reactivity in the accident of uncontrolled withdrawal of the ShR number with and without feedback reactivity temperature coefficients of water and fuel 112 Fig 3.20 Reactor power transient of one ShR withdrawal from operating power 100% 113 Fig 3.21 Reactor power and reactivity transient of ShR number uncontrolled withdrawal from operating power 100% 113 Fig 3.22 Power and reactivity changing when inserting 10 cents reactivity 114 Fig 3.23 The changing of the ReR (TD) when inserting 10 cents 115 Fig 3.24 Experimental data of power (D1) and ReR (TD) position when inserting 10 cents 115 Fig 3.25 a) Infinite multiplication factor of HEU fuel depending on burn-up steps and b) atom density of actinide isotopes at the end of burn-up step (~ 36% burn up of 117 U-235) Fig 3.26 a) Infinite multiplication factor of LEU fuel depending on burn-up steps and b) atom density of actinide isotopes at the end of burn-up step (~ 29% burn-up of U-235) .117 Fig 3.27 Burn-up (% U-235) distribution of fresh HEU core after 538 FPDs operation (REBUS-MCNP system code at upper values and MCDL code at lower values) .119 Fig 3.28 Burn-up (%U-235) distribution of fresh LEU core after 600 FPDs operation (MCNP_REBUS code at upper values and MCDL code at lower values) 120 Fig 3.29 The changing of heavy isotopes of a) HEU and b) LEU fuels of the DNRR 121 Fig 3.30 Difference (%) of calculation results and experimental data (using Cs-137 isotope) for 106 burnt HEU FAs .123 Fig 3.31 Atomic number density of Li-6 and He-3 for 240 nodes in calculation model for LEU core 125 Fig 3.32 Core configuration of the LEU core with 92 fresh FAs including 12 burnt LEU FAs (red and blue color numbers are BU% U-235 of LEU FAs slightly burn-up, black values are identification number of fresh LEU FAs) 126 Fig 3.33 Fuel burn-up distribution of the LEU core with 92 FAs in March, 2021 (upper values are order number, under values are burn-up percent of U-235) .127 Fig 3.34 The procedure to carrying out refueling FAs of the LEU working core with 92 FAs (upper values are order number of FAs, lower numbers are BU%) 128 Fig 3.35 Fuel burn-up distribution of LEU core with 98 FAs (under values) 130 Fig 3.36 Core configuration of 98 FAs loaded fresh FAs and discharged burnt FAs having burn-up about 27% (under values) 131 Fig 3.37 Fuel burn-up distribution of the last cycle using 10 fresh LEU FAs 132