• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3429-3439
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• 2022

The MCS code is a computer code developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation and calculation of nuclear reactor systems based on the Monte Carlo method. The code is currently used to solve two main types of reactor physics problems, namely, criticality problems and radiation shielding problems. In this paper, the radiation shielding capability of the MCS code is validated by simulating some selected SINBAD (Shielding Integral Benchmark Archive and Database) experiments. The whole validation was performed in two ways. Firstly, the functionality and computational rationality of the MCS code was verified by comparing the simulation results with those of MCNP code. Secondly, the validity and computational accuracy of the MCS code was confirmed by comparing the simulation results with the experimental results of SINBAD. The simulation results of the MCS code are highly consistent with the those of the MCNP code, and they are within the 2σ error bound of the experiment results. It shows that the calculation results of the MCS code are reliable when simulating the radiation shielding problems.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1355-1366
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• 2020

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2151-2161
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• 2021

Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic approach to ND S/U analysis. The paper includes a detailed description of each of the programs contained within ASUSD, the computational workflow and validation results. ASUSD was validated on two shielding benchmark experiments from the Shielding Integral Benchmark Archive and Database (SINBAD) - the fission relevant ASPIS Iron 88 experiment and the fusion relevant Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) mock-up experiment. The validation process was performed in two stages. Firstly, the Denovo discrete ordinates transport solver was validated as a standalone solver. Secondly, the ASUSD program package as a whole was validated as a ND S/U analysis tool. Both stages of the validation process yielded excellent results, with a maximum difference of 17% in final uncertainties due to ND between ASUSD and the stochastic ND S/U approach. Based on these results, ASUSD has proven to be a user friendly and computationally efficient tool for deterministic ND S/U analysis of shielding geometries.