Study on neutronics, thermal hydraulics and core management for safe operation and effective utilization of the Dalat Nuclear Research Reactor

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 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 co-authored 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. LIST OF ABBREVIATIONS BOC Beginning of Cycle BWR Boiling Water Reactor CFD Computational Fluid Dynamics CHF Critical Heat Flux CITATION Nuclear Reactor Core Analysis Code DNBR Departure Nucleate Boiling Ratio DNRI Dalat Nuclear Research Institute DNRR Dalat Nuclear Research Reactor ENDF Evaluated Nuclear Data File FA Fuel Assembly FIR Flow Instability Ratio FPD Full Power Day GA Genetic Algorithm HANARO Korean Research Reactor HEU Highly Enriched Uranium IAEA International Atomic Energy Agency IFA Instrumental Fuel Assembly JEFF Joint Evaluated Fission and Fusion (European Evaluated) Nuclear Data Library JENDL Japanese Evaluated Nuclear Data Library LEU Low Enriched Uranium MCNP Monte Carlo N-Particle Computer Code MPI Message Passing Interface MTR Material Testing Reactor NPP Nuclear Power Plant ONBR Onset Nucleate Boiling Ratio PVM Parallel Virtual Machine PWR Pressurized Water Reactor ReR Automatic Regulating Rod RERTR Reduced Enrichment of Research and Test Reactor RIA Reactivity Insertion Accident SA Simulated Annealing SaR Safety Rod ShR Shim Rod SRAC Standard Reactor Analysis Code TRIGA Training, Research, Isotope Production of General Atomics VVER Water-Water Energetic Reactor WIMSD Winfrith Improved Multigroup Scheme 1-D Unidimensional 2-D Bidimensional 3-D Tridimensional CONTENT LIST OF FIGURES Fig. 1.1. Cross sections of the DNRR in axial and radial directions 23 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) 30 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 49 Fig. 2.2. Calculation model for group constants of the neutron trap 51 Fig. 2.3. Calculation model for lattice cells in side the DNRR core 52 Fig. 2.4. Calculation model for lattice cells outside the DNRR core 53 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) 54 Fig. 2.6. Calculation model in axial direction using the CITATION code 55 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) 56 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 57 Fig. 2.9. Calculation model for full core of the DNRR using the MCNP code 58 Fig. 2.10. The DNRR model calculation for the PLTEMP/ANL code and LEU core with 92 FAs 59 Fig. 2.11. Super-cell model in the Serpent code to create group constants of ShR 62 Fig. 2.12. Calculation model of the DNRR using the PARCS code 63 Fig. 2.13. General structure of the MCDL code 68 Fig. 2.14. Burn-up chain model of actinide and fission products isotopes in the MCDL code 70 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) 77 Fig. 3.2. Neutron spectrum of LEU VVR-M2 fuel type with different calculation libraries 78 Fig. 3.3. Critical core configuration with a) 72 FAs and b) working core with 92 FAs 82 Fig. 3.4. Thermal neutron flux distribution in the radial direction (unit ×1012 n/cm2.s) 90 Fig. 3.5. Relative power distribution in axial direction and depending on control rod positions of working core of 92 LEU FAs 92 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) 93 Fig. 3.7. Comparison of the measured cladding and coolant temperatures of the HEU core to validate the PLTEMP/ANL code 96 Fig. 3.8. The HEU VVR-M2 IFA of the DNRR 97 Fig. 3.9. Axial power distribution of the hottest FA (at cell 10-5) calculated by the MCNP code for 25-cm insertion of 4 ShRs 98 Fig. 3.10. Calculation results at nominal power without errors and uncertainties in the hottest FA 100 Fig. 3.11. Comparison of the fuel cladding and coolant temperatures at different reactor power levels 101 Fig. 3.12. Calculation results at nominal power with systematic errors 102 Fig. 3.13. Calculation results at nominal power with systematic and random errors 103 Fig. 3.14. Calculation results of the increasing power of the DNRR from 80 to 100% 106 Fig. 3.15. Experimental data of the changing position of ReR (TD) when increasing reactor power from 80% to 100% 107 Fig. 3.16. Position of ReR (TD) and power when increasing reactor power from 80% to 100% 107 Fig. 3.17. Reactivity and reactor power when increasing reactor power from 80 to 100% 108 Fig. 3.18. Calculation results the changing position of ReR (TD) when increasing power from 80 to 100% 108 Fig. 3.19. Power and reactivity in the accident of uncontrolled withdrawal of the ShR number 1 with and without feedback reactivity temperature coefficients of water and fuel 110 Fig. 3.20. Reactor power transient of one ShR withdrawal from operating power 100% 111 Fig. 3.21. Reactor power and reactivity transient of ShR number 1 uncontrolled withdrawal from operating power 100% 111 Fig. 3.22. Power and reactivity changing when inserting 10 cents reactivity 112 Fig. 3.23. The changing of the ReR (TD) when inserting 10 cents 113 Fig. 3.24. Experimental data of power (D1) and ReR (TD) position when inserting 10 cents 113 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) 115 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) 115 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) 117 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) 118 Fig. 3.29. The changing of heavy isotopes of a) HEU and b) LEU fuels of the DNRR 119 Fig. 3.30. Difference (%) of calculation results and experimental data (using Cs-137 isotope) for 106 burnt HEU FAs 121 Fig. 3.31. Atomic number density of Li-6 and He-3 for 240 nodes in calculation model for LEU core 123 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) 124 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) 125 Fig. 3.34. The procedure to carrying out refueling 6 FAs of the LEU working core with 92 FAs (upper values are order number of FAs, lower numbers are BU%) 126 Fig. 3.35. Fuel burn-up distribution of LEU core with 98 FAs (under values) 128 Fig. 3.36. Core configuration of 98 FAs loaded 4 fresh FAs and discharged 4 burnt FAs having burn-up about 27% (under values) 129 Fig. 3.37. Fuel burn-up distribution of the last cycle using 10 fresh LEU FAs 130 LIST OF TABLES Table 1.1. Material in structure of the DNRR 25 Table 1.2. The parameters of VVR-M2 HEU and LEU fuel 31 Table 2.1. Length and material of LEU FA in axial direction 50 Table 3.1. Infinite multiplication factor of the VVR-M2 HEU and LEU fuels 75 Table 3.2. Calculation results of infinite multiplication factors with different calculation libraries 76 Table 3.3. The critical core configurations established during physical start-up 81 Table 3.4. Multiplication factor of the critical cores using LEU fuel 83 Table 3.5. The effective control rods worth of the working core using 92 LEU FAs 84 Table 3.6. The effective reactivity of LEU FAs 85 Table 3.7. Effective reactivity of beryllium rods 86 Table 3.8. The calculation results and experimental data of relative thermal neutron flux in radial direction 87 Table 3.9. The calculation results and experimental data of relative thermal neutron flux in axial direction 88 Table 3.10. Calculation results and experimental data of thermal neutron flux at the neutron trap of LEU working core 89 Table 3.11. Neutron flux distribution at irradiation positions of LEU working core 91 Table 3.12. Power peaking factor of working core using 92 LEU FAs 91 Table 3.13. The calculation results of feedback reactivity coefficients of LEU core 94 Table 3.14. Calculation results and experimental data of kinetics parameters of LEU core 95 Table 3.15. Hot channel factors in thermal hydraulic analysis of the DNRR 99 Table 3.16. Calculation results and experimental data for decay constant 103 Table 3.17. Calculation results and experimental data of delayed neutron fraction. 104 Table 3.18. Calculation results of multiplication factors from the Serpent and PARCS codes 104 Table 3.19. Infinite multiplication factors of HEU and LEU FAs depending on burn-up (% mass of U-235) 114 Table 3.20. Operation time and excess reactivity of the HEU cores and mixed-core 120 Table 3.21. The calculation results and experimental data of negative effective reactivity of beryllium rods 122 INTRODUCTION Research reactors are essential to the implementation of a nation's nuclear program, and can be used for research and training, material testing, neutron activation analysis, production of radioisotopes for medicine or industry, and other purposes. More than 220 research reactors with different power, fuel type, and neutron energy are currently operating in 53 countries [11]. The structure and power of all research reactors are quite simple with low power, temperature, and pressure when compared to nuclear power plants. Under Reduced Enrichment Research and Test Reactor (RERTR) program [12], almost all operational research or test reactors were converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel but they still kept the purpose in utilizations and applications. Three main factors related to the existence of the research reactor are management, operation, and utilization. From a management point of view, the reactor must be in good condition and operating staff or managers can know clearly or deeply about the reactor in practice and parameters as well. In terms of safety operation, the reactor must meet or exceed the design requirements for safety in physics, thermal hydraulics, and adequate operation. The reactor also has a design to meet for safe operation even in abnormal, transients, and accident conditions. Depending on the characteristics of the reactor, the utilization can be exploited as much as possible. The purposes of application of the reactor must be explicitly defined before building and operating. In general, reactor physics can be divided into two problems: statics and dynamics, along with reactor kinetics and burn-up. In statics calculations, the time variable in transport or diffusion neutron equations is ignored. Multiplication factor or reactivity and neutron flux distributions or power distribution are the most important characteristics derived from static neutronics calculations. For thermal hydraulics, the safety parameters need to be evaluated including fuel, fuel cladding, and coolant temperatures, other safety parameters (ONBR or DNBR) under maximum nominal power, and inlet coolant temperature condition. Reactor kinetics describe the behavior of a reactor based on the insertion or withdrawal of reactivity in reactor core at time step intervals. Three-dimensional (3-D) reactor kinetics is crucial and must be considered for any reactor in normal and transient/accident conditions. During the simulation's subsequent time steps, the power distribution of the hottest fuel assembly (FA) in radial and axial directions within the reactor core can be determined by using 3-D reactor kinetics computer code. Fuel burn-up is also a very important process that directly influences the properties and safety of a reactor. Changing fuel compositions such as the production of actinide isotopes and fission products, and reducing the reactor’s excess reactivity or core lifetime are the two most significant factors affecting reactor characteristics. The burn-up process of a reactor occurred according to operation time in day, month, or even yearly timescale. The burn-up distribution of FAs, excess reactivity, and other parameters are important for core and fuel management in addition to enhancing safe operation and effective utilization as well. In order to conduct experiments on a reactor, it is possible to obtain accurate, reliable data, but careful preparation in terms of equipment and other resources is required, whereas the modern calculating method is more straightforward, economical, and practical due to advanced capabilities of computer nowadays. During the full core conversion of the Dalat Nuclear Research Reactor (DNRR) [1] at the end of 2011, numerous experiments were conducted to determine characteristics of neutronics parameters at the start-up and working LEU core with varying LEU FAs loadings from 72 to 92 FAs. These experimental data are extremely valuable for validating computer codes used into design and management of the DNRR. In addition, the burn-up distribution of 106 burnt HEU FAs was also measured using the gamma scanning method to estimate the burn-up percentage of U-235 [2], and these data can be used to validate a self-developed burn-up computer code MCDL (Monte Carlo Depletion Light water reactor). Neutronics computer codes for core and fuel management of the DNRR use both deterministic and Monte Carlo methods for theoretical calculations, especially after the complete full core conversion in 2012. Popular deterministic codes include cell code, whole core code, and burn-up code; for example, the SRAC2006 (PIJ, CITATION, COREBN) system code [17, 18] or the WIMS-ANL [19] and the REBUS-PC [20] codes or the REBUS-MCNP linkage code [21]. For the LEU core, the MCNP5 [22] or the MCNP6 [23] codes are primarily used for design and neutronics calculations, as well as burn-up. Because of complicated geometry, the DNRR is suitable with neutronics codes employing the Monte Carlo method, such as the MCNP, MVP [24], and Serpent 2 [25] codes. The PLTEMP4.2 [26] thermal hydraulics code agrees very well with fuel, core models, fuel correlation, and using natural convection to remove produced heat from the reactor core of the DNRR. All computer codes used in design of the LEU core were validated by comparing their obtained results with experimental data or other results from different computer codes. The REBUS-MCNP linking codes were applied to calculate burn-up and burn-up distribution of the HEU or LEU cores. The MCNP code in the system code has a role in determining neutron flux, reaction rates while the REBUS code calculates burn-up and updates atomic number density of depletion regions in each FAs. In reactor kinetics calculation, the PARCS code [27, 28] was used to evaluate the power, reactivity insertion, and power distribution in radial and axial directions. The DNRR was reconstructed from the former TRIGA Mark II, which was built in the early 1960s, operated at 250 kW from 1963 to 1968, and extended shutdown until March 1975. All TRIGA fuels were unloaded at the end of March 1975 and then shipped back to the United States of America. The reactor reconstruction project began in 1982 and criticality was reached on November 1st, 1983. In February 1984, a nominal power of 500 kW was obtained. The initial fresh working core was loaded with 88 VVR-M2 fuel assemblies that were 36% HEU. The DNRR was granted permission to partial conversion from HEU core to a mixed HEU-LEU core beginning in 2006, and the first six LEU FAs (19.75% enrichment) were installed in September 2007. In 2009, the DNRR established a mixed core with 92 HEU and 12 LEU FAs. In December 2011, the DNRR completed full core conversion and established the initial critical core loaded 72 LEU FAs with a neutron trap. The working core was created with 12 slightly burnt LEU FAs and 80 fresh LEU FAs and 12 beryllium rods located around the neutron trap. In September 2019, two irradiation channels 5-6 and 9-6 were installed in the reactor core by replacing two beryllium rods to increase the amount of I-131 radioisotope production. In April 2021, the reactor implemented refueling by replacing two beryllium rods with two new LEU fuel assemblies, and in May 2022, the reactor continued refueling by replacing two other beryllium rods with two new LEU fuel assemblies. The reactor was refueled in May 2023 to attain 98 FAs in the working core [7, 8]. Even operating with low power and neutron flux, the DNRR has contributed significantly to the social-economic development of Vietnam. Various nuclear engineering and radioisotope applications for medical, agricultural, industrial, geological, hydrological, and environmental purposes have been implemented on the DNRR in order to promote economic growth and protect public health. In addition, the fundamental research on reactor engineering, nuclear physics, and other applied research conducted on the D