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Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.Nghiên cứu tính toán vật lý nơtron, thủy nhiệt và quản lý vùng hoạt để vận hành an toàn và sử dụng hiệu quả lò phản ứng hạt nhân Đà Lạt.

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 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 DOCTORAL THESIS Subject: Atomic and Nuclear Physics Code number: 9-44-01-06 Supervisor: Assoc Prof PhD NGUYỄN NHỊ ĐIỀN Hà Nội – 2023 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 theconfidence and motivation tocompletethe 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 INTRODUCTION 10 CHAPTER GENERAL OVERWIEW 20 1.1 Motivation of the thesis 20 1.2 General introduction about the DNRR 21 1.2.1 History, structure and reactor core arrangement 21 1.2.2 Fuel of VVR-M2 HEU and LEU 28 1.2.3 Neutronics characteristics of the DNRR 30 1.2.4 Thermal hydraulics characteristics of the DNRR 32 1.3 The development of computer codes for reactor calculation in the world 33 1.4 The research situation about research reactor in Vietnam 35 1.5 Reactor kinetics in three dimensions 38 1.6 Burn-up calculation for core and fuel management 39 CHAPTER CALCULATION MODELS FOR THE DALAT NUCLEAR RESEARCH REACTOR USING LEU FUEL 41 2.1 Neutronics calculation models 41 2.1.1 Deterministic code 41 2.1.2.1 Lattice cell model 41 2.1.2.2 Whole core model 46 2.1.2 Calculation model for computer codes using Monte Carlo method 48 2.2 Thermal hydraulics calculation for the DNRR 50 2.3 Reactor kinetics application for the DNRR 53 2.3.1 Preparation group constants for the PARCS code 53 2.3.2 Calculation model for the DNRR using the PARCS code 54 2.4 Burn-up calculation for the DNRR 54 2.4.1 Development MCDL computer code 54 2.4.2 Application of MCDL for burn-up and refueling calculation for LEU core 61 2.5 Summary of Chapter 62 CHAPTER RESULTS AND DISSCUSSIONS 64 3.1 Neutronics and thermal hydraulics for LEU core 64 3.1.1 Neutronics calculation results 64 3.1.1.1 Neutronics characteristics of the HEU and LEU VVR-M2 fuel types 64 3.1.1.2 Criticality, reactivity, control rod worths 67 3.1.1.3 Excess reactivity, control rod worth, beryllium rods 72 3.1.1.4 Neutron flux distribution 74 3.1.1.5 Power peaking factor 78 3.1.1.6 Feedback reactivity temperature coefficients and void factor 80 3.1.1.7 Kinetics parameters 81 3.1.2 Thermal hydraulics calculation results 81 3.1.2.1 Validation of the PLTEMP code 81 3.1.2.2 Steady state of PLTEMP code without hot channel factors 85 3.1.2.3 Steady state of PLTEMP code with hot channel factors 87 3.2 Kinetics calculation results for LEU core 88 3.2.1 Calculation results from the Serpent and PARCS codes at steady state condition 88 3.2.2 Calculation and experiment results during increasing power of the DNRR from 80% to 100% 90 3.2.3 Simulation of the accident when uncontrolled withdrawal of one control rod at nominal power 93 3.2.4 Simulation of the changing of power when inserting positive reactivity smaller than 10 cents 95 3.3 Burn-up calculation results 97 3.3.1 Validation of the MCDL code 97 3.3.2 Calculation results of the HEU core 102 3.3.3 Calculation results of the LEU core 106 3.4 Summary of Chapter 113 CONCLUSIONS 113 NEW CONTRIBUTIONS OF THE THESIS 117 LIST OF PUBLICATIONS 117 REFERENCES 119 LIST OF FIGURES Fig 1.1 Cross sections of the DNRR in axial and radial directions 22 Fig 1.2 Annual operation time of the DNRR from 1984 to 2011 (the core configuration using HEU fuel and mixed cores) 27 Fig 1.3 Annual operation time of the DNRR from 2012 to 2022 (using LEU fuel) 27 Fig 1.4 Specific dimensions and geometry of the VVR-M2 fuel type 29 Fig 2.1 Calculation model for VVR-M2 FA of the DNRR 42 Fig 2.2 Calculation model for group constants of the neutron trap 44 Fig 2.3 Calculation model for lattice cells in side the DNRR core 45 Fig 2.4 Calculation model for lattice cells outside the DNRR core 46 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) 47 Fig 2.6 Calculation model in axial direction using the CITATION code 48 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 c-example of detailed power peaking factor calculation inside FA) 49 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 50 Fig 2.9 Calculation model for full core of the DNRR using the MCNP code 50 Fig 2.10 The DNRR model calculation for the PLTEMP/ANL code and LEU core with 92 FAs 51 Fig 2.11 Super-cell model in the Serpent code to create group constants of ShR 54 Fig 2.12 Calculation model of the DNRR using the PARCS code 54 Fig 2.13 General structure of the MCDL code 58 Fig 2.14 Burn-up chain model of actinide and fission products isotopes in the MCDL code 60 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) 66 Fig 3.2 Neutron spectrum of LEU VVR-M2 fuel type with different calculation libraries 67 Fig 3.3 Critical core configuration with a) 72 FAs and b) working core with 92 FAs 70 Fig 3.4 Thermal neutron flux distribution in the radial direction (unit ×1012 n/cm2.s) 77 Fig 3.5 Relative power distribution in axial direction and depending on control rod positions of working core of 92 LEU FAs 79 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) 80 Fig 3.7 Comparison of the measured cladding and coolant temperatures of the HEU core to validate the PLTEMP/ANL code 82 Fig 3.8 The HEU VVR-M2 IFA of the DNRR 83 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 84 Fig 3.10 Calculation results at nominal power without errors and uncertainties in the hottest FA 86 Fig 3.11 Comparison of the fuel cladding and coolant temperatures at different reactor power levels 87 Fig 3.12 Calculation results at nominal power with systematic errors 87 Fig 3.13 Calculation results at nominal power with systematic and random errors 88 Fig 3.14 Calculation results of the increasing power of the DNRR from 80 to 100% 91 Fig 3.15 Experimental data of the changing position of ReR (TD) when increasing reactor power from 80% to 100% 91 Fig 3.16 Position of ReR (TD) and power when increasing reactor power from 80% to 100% 92 Fig 3.17 Reactivity and reactor power when increasing reactor power from 80 to 100% 92 Fig 3.18 Calculation results the changing position of ReR (TD) when increasing power from 80 to 100% 93 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 94 Fig 3.20 Reactor power transient of one ShR withdrawal from operating power 100% 95 Fig 3.21 Reactor power and reactivity transient of ShR number uncontrolled withdrawal from operating power 100% 95 Fig 3.22 Power and reactivity changing when inserting 10 cents reactivity 96 Fig 3.23 The changing of the ReR (TD) when inserting 10 cents 97 Fig 3.24 Experimental data of power (D1) and ReR (TD) position when inserting 10 cents 97 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 U-235) 99 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) 99 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) 101 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) 102 Fig 3.29 The changing of heavy isotopes of a) HEU and b) LEU fuels of the DNRR 102 Fig 3.30 Difference (%) of calculation results and experimental data (using Cs-137 isotope) for 106 burnt HEU FAs 104 Fig 3.31 Atomic number density of Li-6 and He-3 for 240 nodes in calculation model for LEU core 105 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) 106 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) 107 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%) 108 Fig 3.35 Fuel burn-up distribution of LEU core with 98 FAs (under values) 110 Fig 3.36 Core configuration of 98 FAs loaded fresh FAs and discharged burnt FAs having burn-up about 27% (under values) 111 Fig 3.37 Fuel burn-up distribution of the last cycle using 10 fresh LEU FAs 112 LIST OF TABLES Table 1.1 Material in structure of the DNRR 23 Table 1.2 The parameters of VVR-M2 HEU and LEU fuel 28 Table 2.1 Length and material of LEU FA in axial direction 43 Table 3.1 Infinite multiplication factor of the VVR-M2 HEU and LEU fuels 64 Table 3.2 Calculation results of infinite multiplication factors with different calculation libraries 65 Table 3.3 The critical core configurations established during physical start-up 69 Table 3.4 Multiplication factor of the critical cores using LEU fuel 71 Table 3.5 The effective control rods worth of the working core using 92 LEU FAs 72 Table 3.6 The effective reactivity of LEU FAs 73 Table 3.7 Effective reactivity of beryllium rods 73 Table 3.8 The calculation results and experimental data of relative thermal neutron flux in radial direction 74 Table 3.9 The calculation results and experimental data of relative thermal neutron flux in axial direction 75 Table 3.10 Calculation results and experimental data of thermal neutron flux at the neutron trap of LEU working core 76 Table 3.11 Neutron flux distribution at irradiation positions of LEU working core 77 Table 3.12 Power peaking factor of working core using 92 LEU FAs 78 Table 3.13 The calculation results of feedback reactivity coefficients of LEU core 80 Table 3.14 Calculation results and experimental data of kinetics parameters of LEU core 81 Table 3.15 Hot channel factors in thermal hydraulic analysis of the DNRR 84 Table 3.16 Calculation results and experimental data for decay constant 88 Table 3.17 Calculation results and experimental data of delayed neutron fraction 88 Table 3.18 Calculation results of multiplication factors from the Serpent and PARCS codes 89 Table 3.19 Infinite multiplication factors of HEU and LEU FAs depending on burn-up (% mass of U-235) 98 Table 3.20 Operation time and excess reactivity of the HEU cores and mixed-core 103 Table 3.21 The calculation results and experimental data of negative effective reactivity of beryllium rods 105 Fig 3.36 Core configuration of98 FAs loaded 4fresh FAsanddischarged4 burnt FAs havingburn-up about 27% (undervalues) The multiplication factors of the core configuration evaluated with and without samples around 1.024 and 1.034 respectively With these multiplication factor values, the core excess reactivity was estimated about 0.66 $ for fuel burn-up The reactor operating time was estimated about 2640 hours, corresponding to more than 110 days and nights Thus, it can be seen that with the using of 10 new FAs loaded for the 92 LEU FAs working core, the total operating time under this option is up to 17,640 hours If operating time about 2000 hours to 3000 hours each year, thetime to use of 10FAs can be archived fromnearly6 yearsto more than years The fuel burn-up distribution at the end of this cycle has an average fuel burn-up of 18.46%, with a multiplication factor of 1,034 and reactor operating times of this configuration to 110 days and nights, corresponding to 2640 hoursand a massof U- 235burn-up about100g 111 Fig 3.37 Fuelburn-updistributionofthelastcycleusing 10fresh LEUFAs (undervalues areburn-uppercentof U-235) In theoretical calculations, the physical parameters of both refueling core configurations were mainly calculated usingthe MCNP code withthe ENDF/BVII.1 library The simulated calculation model was quite similar to reality, including details of additional irradiation positions in the core when performing fuel burn-up calculations The thermal hydraulics parameters were also calculated mainly by the PLTEMP4.2 code The calculation of thermal hydraulics parameters mainly focus on core configurations that were recommended to consider in the process of refueling Experiments can be carried out before and after the refueling proceduresin order to have reliable data, and the experimental data willbe used for theprocess ofre-evaluatingthe reliabilityofthe calculation programs When calculating thermal hydraulics safety analysis, it was found that the core configuration with 98 FAs had the maximum temperature of the fuel cladding of the hottest FA in the core around 90.6 0C, slightly higher than those in the initial core of 92 FAs However, this temperature still completely meets the manufacturer's requirements with the fuel cladding temperaturelimit of 103 0C With the core configuration loading FAs loading, the neutron flux is reduced almost 9% at the trap and 8% at the new irradiation channels without loading samples target TeO2 In this working core, the average I-131 activity of container of TeO2 sample target was only about 3.94 Ci after to 10 hours of cooling Neutron fluxes at the trap and the two new irradiation channels decreased between 19.5% and 10.7% with loaded samples target However, the neutron flux distributionsof both core optionsat the irradiation positionssuchas drychannels 7-1, 13-2 or wet channel 1-4 and the rotary specimen on the graphite reflector only decreased average about 5% and was still fully met requirements for basic applications such as neutron activation analysis or basics research on neutron source fromhorizontal beamtubes 112 3.4 Summary of Chapter The chapter presents calculation results of neutronics, thermal hydraulics, primary RIA analysis, and fuel burn-up for core and fuel management for the DNRR using fully LEU fuels The obtained results were compared with experimental data to confirm about the high fidelity and good calculation model that have been built for the DNRR In neutronics, the characteristics of the DNRR using LEU fuel were investigated to determine parameters related to reactivity, control rod worths, neutron flux distribution, power peaking factor, feedback coefficients and kinetics These obtained data were used for core management and assured ability for safe operation as well as effective utilization Then, MCNP5 or codes and ENDF/BVII.1 library were used as official codes for core management of the DNRR inneutronicscalculation The thermal hydraulics analsysis was carried out to estimate safety factors at steady state operation of 500 kW Under normal operation condition with inlet temperature 32 0C, the maximum fuel cladding was under 91 0C far below limitation of 103 0C When adding hotpot enginerring data in thermal hydraulics analysis, the cladding temperature was stillunder melting temperature of aluminumandthe fuel claddingintegritywas ensured The primary safety analysis for RIA of the DNRR using LEU fuel was implemented using the 3-D kinetics PARCS code The calculation results showed that under the accident condition when inserting positive reactivity, the DNRR was stillinsafestatus and the fuel cladding was notdamaged The MCDL code was applied to calculate for burn-up as well as refueling to the DNRR using effectively 10 fresh remained LEU FAs After establishing98 LEU FAs core configuration, the abilityto extendoperation time ofthe DNRRup to2030ifadding28 LEUFAs with morethan 32000hours runningatfull powerof500kW The burn-up and refueling calculations for the DNRR were conducted with abilityto enhance radioisotope production of I-131 isotope When adding two new irradiation channels at cell 5-6 and 9-6 together with accumulated irradiation of target containers by moving position from rotary specimen to inside the core at each operating cycle in week, the activity of each irradiated container after cycles can be reached more than Ci and each month the DNRR can produce morethan 100Ci of I-131 isotope CONCLUSIONS 113 In the dissertation, three major problems were solved including: neutronics and thermal hydraulics calculation and analysis for the LEU core with 92 FAs; using the PARCS code to apply for the DNRR in preliminary with 3-D kinetics calculation; and developing the MCDL code for burn-up calculation integrated beryllium poisoning to investigate the HEU and LEU cores Using the system computer codes including MCNP code for neutronics calculation, the PLTEMP4.2 code for thermal hydraulics analysis, MCDL code combined with optimization code LPO_V (GA and SA methods) to determine the refueling loading patterns for using 10 remained fresh LEU fuels and enhancement of radioisotope production for I-131 isotope on the DNRR In the first purposerelatedto neutronics andthermal hydraulics calculationand analysis,some mainresultsobtained as follow: The detailed investigation of VVR-M2 LEU and HEU fuels was conducted with a focus on multiplication factors,spectrum, anddeterminingthe most suitable libraryfor DNRR as ENDF/BVII Using the MCNP code and other computer codes such as SRAC2006, WIMS-ANL and REBUS-PC to evaluate all neutronics characteristics of the LEU core with 92 LEU FAs The obtained results were compared to experimental data or calculation results from other codes and the discrepancy between calculation results and experimental data was found to be a maximum of 10% The effective multiplication factors of twenty-five distinct core configurations were computed, and the average values from three codes and experimentaldata were within20pcmof one another In comparison toexperimentaldata,the effective reactivityof FAs, berylliumrods,and the control rods' valuefor LEUcore were estimated witha maximum 10%variance Neutron flux distribution at each FA in the core and irradiation positions were also calculated for radioisotope production and other applications The power peaking factor was an essential value for the thermal hydraulics analysis and was determined following the positions of control rods Other parameters including feedback temperature coefficients and kinetics parameters for the LEU core were evaluated and usingforthermal hydraulics andsafety analysis The validation of the PLTEMP/ANL4.2 code was carried out by comparing the calculation results of the LEU core to experimental data of the HEU core of the DNRR with very good consistency The LEU working core with 92 FAs and 12 beryllium rods was evaluated and subjected to steady-state thermal hydraulic analysis without hot channel factors at the nominal thermal power of 500 kW The obtained results show that at the hottest FA, the fuel cladding temperature was only90.4 oC, which was far below the limitation of the VVR-M2 fuel cladding temperature of 103 oC, the ONB temperature was about 115.6 oC, according to the Forster-Greif correlation, and the minimum DNBR value was about 32.0 When the systematic uncertainties were taken into account, the maximum fuel cladding temperature was predicted to 114 be 98.4 oC, which again was well below the limit value of 103 oC, the DNBR value was 17.79, and the coolant temperature at the outlet of the hottest FA was very low compared with the saturation temperature of 107 oC This means that the DNRR met the thermal hydraulics safety at steady-state condition with application of the global hot channel factors When all systematic as well as random uncertainties were applied to a limiting calculation, the maximum fuel cladding temperature obtained was 114.3 oC, which was several degrees below the ONB point of 116.2 oC The minimum DNBR value, according to Shah’s correlation, was estimated as 15.2, which was much higher than the acceptable criterion of 1.5 for the DNRR To the second purpose, the PARCS code was used for preliminary calculations in kinetic and transient analysis of LEU core with 92 FAs and calculation results were compared with experimental data or calculation results from the RELAP5 code The Serpent code was used to generate multi-group cross-sections for 3-D kinetics calculations of the DNRR in PARCS code model The DNRR calculation model, completely specified with the Serpent code, was also utilized to validate the steady-state and using in transient/accident investigation The calculation results showed the limitation of applying the PARCS code to the research reactor so to have a tool for 3D kinetics coupling with system code as the RELAP5 code The main reason of disadvantage of the PARCS code inthermal hydraulics isthat safetyanalysis factors are defaultparameters for PWR type The final objective of this research was to develop the MCDL code to calculate burn-up for the DNRR using HEU and LEUfuels The specific outcomes are as follows: The MCDL code was validated for HEU and LEU fuels, as well as the initial HEU and LEU core with 89 and 92 FAs, respectively The generated calculation results were compared with the REBUSMCNP code, and theburn-uppercentagedifference betweenthe two codes waslessthan5% Detailed examination for HEU and mixed cores including fuel burn-up and beryllium poisoning estimation to isotopes He-3, H-3, Li-6 The burn-up distribution of 106 HEU FAs were calculated by the MCDLcode and compared to experimental data using gamma-scanning method The maximum difference between calculation results and experimental data was under 18% The poisoned beryllium blocksandrods were updatedfor using indesign calculationfor LEUcore inneutronics For the LEU core, fuel burn-up calculation was also carried out for the fuel and core management purposes at the beginning of the core life in 2012 Before carrying out refueling, the estimation of burnup distribution of LEU core was performed The MCDL code was also applied for burn-up 115 calculation of fuel loading pattern core configurations in refueling scheme for abilityto use 10 remained fresh LEUFAs The MCDL code and other codes such as MCNP, PLEMP4.2 were used to compute the utilization of effective LEU VVR-M2 fuel for the DNRR by refueling at the end of the cycle of 92 LEU FAs configuration and creating a refueling strategy to use 10 existing LEU FAs The fuel loading configurations of the DNRR were found with the aid of a fuel loading patterns optimization code using a simulated annealing optimization algorithm and a 2D triangle map diffusion code The primary requirements of the fuel-loading pattern calculation process are operating limit circumstances, increasing fuel burn-up, prolonging operation time, and radioisotope production capability Each fuel-loading pattern was evaluated in detailed neutronics, fuel burn-up distribution and thermal hydraulics analysis The enhancement of radioisotope production was also studied in accordance with operational safety requirements, with the radioisotope I-131 produced on the DNRR increasing from 30 Ci to 40 Ci with 100 hours of reactor operation per week and the accumulation irradiation method by moving the TeO2 containers from the rotary specimen to the core during each operation cycle was applied By building extra irradiation channels at cells 5-6 and 9-6, the capacity to load TeO2 targets into the reactor's core was increased, resulting in a reduction in operation time per week while maintaining enoughI-131 production Basing on the results from the dissertation, research direction in the future was proposed especially to meet thepurposesin safetyoperation,effectiveutilization anddeveloping ofthe DNRRincluding: - Measuring burn-up of LEU fuels byusing method as reactivity, gamma scanning to obtain the data needed to validateburn-upcomputer codes and enhancethe DNRR's core and fuel managementcapabilities - Coupling computation in thermal hydraulics and safety analysis for the DNRR using the PARCS and the RELAP5 codes to improve the accuracy and fidelity of results obtained under transient or accident scenarios - Coupling CFD codes (such as OpenFOAM or ANSYS) and the PARCS code with Serpent code to considerin detailed multi-physics for the DNRR as well asfor new research reactor The experiment can be carried out by using the IFA loaded at the neutron trap for measuring neutronics and thermal hydraulics data in natural or forced convection - Application of artificial intelligence (AI) to optimization in refueling and safety analysis for the DNRR by using operational data and calculation results In order to identify the optimal fuel-loading pattern, the detailed power peaking factor, thermal hydraulics, and fuel burn-up data are calculated in a short amount of time and serve as the constraintconditionofthe functiontarget 116 - Conceptual design of the DNRR and increasing power through using new fuel type as TRIGA, HANARO or MTR reactors with higher uranium density (>3.0 g/cm3) in order to improve the DNRR's applicationandutilization capabilities NEW CONTRIBUTIONS OFTHE THESIS - Calculating in physics and thermal hydraulics applied to the core and fuel management of the DNRR in order to ensure safe operation and effective utilization Specifically, the evaluation and determination of specific physical parameters ofthe core afterlong-termoperation, as well as preparation for refueling - Using the 3-D kinetics program PARCS to investigate and assess the safety of the DNRR in the transition condition and reactivity insertion accident The obtained result is a prerequisite for coupling with the RELAP5 code, allowingfor more detailed futuresafetystudies for the DNRR andthe newresearch reactor - Developing a fuel burn-up calculation code that was coupled between the MCNP code and the burn-up module with beryllium poisoning computation, and that can update the MCNP code or the calculation library to applyto the DNRR in determining the distribution of 3-D fuel burn-up The primary function of the MCDL code is to provide fuel burn-up distribution for determining fuel loading patterns in refueling for the DNRR and other research reactors containing beryllium materialinthe core LIST OFPUBLICATIONS [1]Kien-Cuong Nguyen, Vinh-Vinh Le, Ton-Nghiem Huynh, Ba-Vien Luong, Nhi-Dien Nguyen, Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor, Science and Technology of Nuclear Installations, vol.2021, Article ID6673162,10 pages https://doi.org/10.1155/2021/6673162 117 [2] Giang Phan, Hoai-Nam Tran, Kien-Cuong Nguyen, Viet-Phu Tran, Van-Khanh Hoang, Pham Nhu Viet Ha, Hoang Anh Tuan Kiet, Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5, Science and Technology of Nuclear Installations, vol 2017, Article ID 2615409, 10 pages, 2017.https://doi.org/10.1155/2017/2615409 [3] Tran, VP., Nguyen, KC., Hartanto, D et al Development of a PARCS/Serpent model for neutronics analysis ofthe Dalat nuclear researchreactor NUCLSCI TECH 32,15,2021, https://doi.org/10.1007/s41365-021-00855-5 [4] Nhi-Dien Nguyen, Kien-Cuong Nguyen, Ton-Nghiem Huynh, Doan-Hai-Dang Vo, Hoai-Nam Tran, Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel, Science and Technology of Nuclear Installations,vol.2020, Article ID7972827,11 pages https://doi.org/10.1155/2020/7972827 [5] Kien-Cuong Nguyen, Vinh-Vinh Le, Ton-Nghiem Huynh, Ba-Vien Luong, Nhi-Dien Nguyen, Hoai-Nam Tran, "InterimStorage of the Dalat Nuclear Research Reactor:RadiationSafetyAnalysis", Science and Technology of Nuclear Installations,vol.2020, Article ID7327045,10 pages,2020 https://doi.org/10.1155/2020/7327045 [6] Nguyen Kien, C., Nguyen Thi, D., Tran Viet, P., Nguyen Huu, T., and Pham Nhu Viet, H., Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent Monte Carlo code, Nuclear Science and Technology, 9(3),21-29,2019,https://doi.org/10.53747/jnst.v9i3.41 [7] Nguyen Kien, C., Huynh Ton, N., Le Vinh, V., Luong Ba, V., Pham Quang, H., Tran Quoc, D., and Bui Van, C., CalculationResultsfor EnhancingAbilityof I-131RadioisotopeProduction Using TelluriumDioxide Target on the Dalat Nuclear Research Reactor, Nuclear Science and Technology, 9(3), 2021, https://doi.org/10.53747/jnst.v9i3.39 Conferences [1] Kien Cuong NGUYEN, et.al, The development depletion code couped with Monte Carlo computer code, VINANTSXIConference, Danang,Vietnam,2015 [2] Kien Cuong NGUYEN, et.al, Fuel burn-up calculation for the Dalat Nuclear Research Reactor by using Serpent and MCNP6 computercodes,VINANTSXIVConference,Dalat,Vietnam,2021 118 REFERENCES Vietnamese Báo cáo kết khởi động vật lý khởi động lượng để chuyển đởi tồn vùng hoạt Lò Phản ứng hạt nhân Đà Lạt sang nhiên liệu độ giàu thấp, Viện Nghiên cứu hạt nhân, Viện Năng lượng nguyên tử Việt nam, 2012,ĐàLạt Nguyễn Minh Tuân, Xây dựng hệ thiết bị đo độ cháy bó nhiên liệu qua sử dụng Lò phản ứng hạt nhân Đà Lạt kỹ thuật 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