Criticality and rod worth analysis of the DNRR research reactor using the SRAC and MCNP5 codes

Binh, Quang; Phan, Giang T.T.; Nguyen, Kien-Cuong; Ngo, Quang Huy; Tran, Hoai-Nam

doi:10.1016/j.nucengdes.2019.01.011

Cited by 17 publications

(7 citation statements)

References 16 publications

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“…e core and the graphite reflector are placed in the reactor pool. More detailed description of the DNRR reactor can be seen in [8,9]. Normally, the DNRR is operated continuously for a period of about 130 hours per month.…”

Section: The Dnrr and Calculation Methodsmentioning

confidence: 99%

Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

Nguyen

Huynh

et al. 2020

Science and Technology of Nuclear Installations

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Radiation safety analysis of a new interim storage of the Dalat Nuclear Research Reactor (DNRR) for keeping spent high enriched uranium (HEU) fuel bundles during the core conversion to low enriched uranium (LEU) fuel had been performed and presented. The photon source and decay heat of the spent HEU fuel bundles were calculated using the ORIGEN2.1 code. Gamma dose rates of the spent fuel interim storage were evaluated using the MCNP5 code with various scenarios of water levels in the reactor tank and cooling time. The radiation safety analysis shows that the retention of 106 spent HEU fuel bundles at the interim storage together with a core of 92 LEU fuel bundles meets the requirements of radiation safety. The results indicate that in the most severe case, i.e., the complete loss of water in the reactor tank, the operators still can access the reactor hall to mitigate the accident within a limited time. Particularly, in the control room, the dose rate of about 1.4 μ Sv / h is small enough for people to work normally.

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Section: The Dnrr and Calculation Methodsmentioning

confidence: 99%

Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

Nguyen

Huynh

et al. 2020

Science and Technology of Nuclear Installations

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“…To control the operation of the core, the DNRR has 2 safety rods, 4 shim rods and 1 automatic regulating rod. Safety rods and shim rods are made of B4C, automatic regulating rod is made of stainless steel [1][2].…”

Section: Description For the Dnrr Fuel Assembly Configurationsmentioning

confidence: 99%

“…To evaluate the operation of the reactor core, researchers use simulation models. The physical characteristics of the DNRR core have been calculated using computational programs such as MCNP, WIMSD/CITATION, SRAC codes [2] and the reliability of the result depends directly on the accuracy of the cross section libraries. Thus, it is essential to do sensitivity and uncertainty analysis to evaluate the impact of nuclear data error on the calculation result.…”

Section: Introductionmentioning

confidence: 99%

Sensitivity and Uncertainty Analysis on Reactivity for HEU and LEU Fuel Assemblies of Dalat Nuclear Research Reactor using Monte Carlo Code and ENDF/B-VII.0 and ENDF/B-VII.1 Nuclear Libraries

Thoi Nam,

Tran Quang Linh,

Hartanto

et al. 2022

MaP

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This paper presents the analysis of sensitivity and uncertainty for the infinite multiplication factor (kinf) for the VVR-M2 typed HEU and LEU fuel assemblies of the Dalat nuclear research reactor (DNRR) using the MCNP6.1-Whisper1.1 code. Sensitivity calculations were performed for the ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. In Whisper, the keff uncertainty due to nuclear data was evaluated by the uncertainty propagation law using the sensitivities obtained by MCNP and the available covariance matrix. The most significant sensitivity coefficients in positive contribution are the coefficient total n and fission reaction of U-235, elastic scattering reaction of H-1 and inelastic scattering of thermal neutrons and in negative contribution are capture reactions of U-235, H-1, Al-27, U-238 and U-234 isotopes. The large discrepancy between sensitivities of elastic scattering reaction cross section of H-1 and inelastic scattering of thermal neutrons between two libraries are found because of the change in neutron spectra of HEU and LEU fuel assemblies using the two library versions. The uncertainty of the kinf from the ENDF/B-VII.0 nuclear library error for all isotopes was found to be significant (about 0.45% of reactivity effect) with the largest contribution from isotope U-235 contributions (about 0.3% to 0.40%).

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“…For neutronics design calculation of the LEU working core, the MCNP-REBUS linkage system code was used to estimate the shutdown margin, the excess reactivity, the control rod's worth, the kinetics parameters, the burn-up distribution, and the power density distribution [4][5][6]. e power peaking factors with different positions of four shim rods inside the reactor core were also determined.…”

Section: Introductionmentioning

confidence: 99%

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

Nguyen

Huynh

et al. 2021

Science and Technology of Nuclear Installations

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This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

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Criticality and rod worth analysis of the DNRR research reactor using the SRAC and MCNP5 codes

Cited by 17 publications

References 16 publications

Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

Sensitivity and Uncertainty Analysis on Reactivity for HEU and LEU Fuel Assemblies of Dalat Nuclear Research Reactor using Monte Carlo Code and ENDF/B-VII.0 and ENDF/B-VII.1 Nuclear Libraries

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

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