Verification and validation of the fast reactor multi-group cross section processing code MGGC2.0

Ma Xubo; Ma Longxiao; Ma Xudong; Zhang Teng; Chen Xiang

doi:10.11884/HPLPB202537.240397

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Ma Xubo, Ma Longxiao, Ma Xudong, et al. Verification and validation of the fast reactor multi-group cross section processing code MGGC2.0[J]. High Power Laser and Particle Beams. doi: 10.11884/HPLPB202537.240397

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Ma Xubo, Ma Longxiao, Ma Xudong, et al. Verification and validation of the fast reactor multi-group cross section processing code MGGC2.0[J]. High Power Laser and Particle Beams. doi: 10.11884/HPLPB202537.240397

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Verification and validation of the fast reactor multi-group cross section processing code MGGC2.0

doi: 10.11884/HPLPB202537.240397

North China Electric Power University, School of Nuclear Science and Engineering, Beijing, 102206, China

Received Date: 2024-11-15
Accepted Date: 2025-04-07
Rev Recd Date: 2025-04-15

Available Online: 2025-05-20

Abstract

Abstract

The deterministic calculation method based on multi-group cross section has always been an important approach in the design of nuclear reactors. The accuracy of multi-group cross section directly affects the precision of nuclear reactor physics calculations. In order to generate high-precision cross section data for fast reactors, North China Electric Power University has developed the high-precision cross section processing code MGGC2.0. This paper conducts benchmark verification and validation of the code. The infinite homogeneous mixed media UO2, MOX, and U-TRU-Zr fuels are calculated based on the ENDF/B-VII.1 library, and the macroscopic cross section generated by MGGC2.0 are compared with those produced by MCNP to verify the accuracy of the program in generating multi-group cross section. The relative deviation of the macroscopic multi-group total cross section from the reference solution of MCNP is generally within 5%. Subsequently, calculations are performed for the Russian fast reactor experiment BFS97-1, and a homogenization method of the fuel few-group cross section for various fuel arrangement forms is proposed. The collision probability method in MGGC2.0 was used to calculate the few-group cross section data for the fuel, and DIF3D program was employed for core calculations. Additionally, this study compared the results obtained using different cross section homogenization methods. The research findings indicate that for BFS97-1, if the cross section generated directly by critical search are used, the absolute deviation of the keff calculated by DIF3D from the keff calculated by MCNP is 2.541×10⁻². This paper improves the calculation method of axial fuel inhomogeneity, reducing the deviation to below 5.0×10⁻⁴. The deviations between the calculated results for BFS97-1, BFS97-2, BFS97-5, and BFS97-6 and the MCNP results are all within3.0×10⁻³, validating the high accuracy of the code in generating multi-group and few-group cross section, which meets the requirements of engineering design.
- multi-group cross section,
- MGGC2.0,
- one-dimensional homogenization method,
- BFS benchmark experiment,
- fast reactor

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References(17)

References

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