Serpent code
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This paper presents a model development of the Dalat nuclear research reactor (DNRR) with the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code.
9p vicedric 08-02-2023 3 3 Download
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The ADS (accelerator driven system) is recognized as a promising system to annihilate the radioactivity of nuclear waste with its inherent safety feature and waste transmutation potential. Thus, conceptual designs of ADS are widely carrying out.
10p visnape 30-01-2023 7 4 Download
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In this study, the SCALE/TRITON code (based on deterministic method) and the Serpent 2 code (based on Monte Carlo method) were utilized to prepare the group constants of the pressurized water reactor (PWR) mixed-oxide (MOX) fuel assemblies for transient analyses of PWR MOX fueled cores in normal operation and control rod ejection accident condition with 3D reactor kinetics codes.
10p meyerowitz 25-12-2021 7 0 Download
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The purpose of this study is to verify the accuracy of innovative ADS core modeling by using simulation codes. The reactivity calculations of CERMET loaded fuel ADS was conducted using two Monte Carlo codes, Serpent [8] and MCNP6 [9] with ENDF/B-VII.0 library [10]. The comparison of results obtained from the two codes is analyzed and discussed in this study.
6p nguaconbaynhay11 07-04-2021 12 1 Download
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This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR) loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code.
9p trinhthamhodang1218 14-03-2021 5 1 Download
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In nuclear safety research, the quality of the results of simulation codes is widely determined by the reactor design and safe operation, and the description of neutron transport in the reactor core is a feature of particular importance.
14p christabelhuynh 30-05-2020 37 0 Download
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The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The “Fast Total Monte Carlo” method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on k∞, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.
10p minhxaminhyeu5 30-06-2019 12 0 Download
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In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1) and 233U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode.
6p minhxaminhyeu5 30-06-2019 32 0 Download
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This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to 235U enrichment ratio compared to the UO2/Zr fuel.
12p minhxaminhyeu3 12-06-2019 16 0 Download