Micro cross-section parameterization based on RMC code

Feng Zhiyuan; Li Kaiwen; Luo Hao; Wang Kan

doi:10.11884/HPLPB202234.210309

Volume 34 Issue 2

Jan. 2022

Turn off MathJax

Article Contents

Article Navigation > High Power Laser and Particle Beams > 2022 > 34(2): 026006

Feng Zhiyuan, Li Kaiwen, Luo Hao, et al. Micro cross-section parameterization based on RMC code[J]. High Power Laser and Particle Beams, 2022, 34: 026006. doi: 10.11884/HPLPB202234.210309

Citation:

Feng Zhiyuan, Li Kaiwen, Luo Hao, et al. Micro cross-section parameterization based on RMC code[J]. High Power Laser and Particle Beams, 2022, 34: 026006. doi: 10.11884/HPLPB202234.210309

Feng Zhiyuan, Li Kaiwen, Luo Hao, et al. Micro cross-section parameterization based on RMC code[J]. High Power Laser and Particle Beams, 2022, 34: 026006. doi: 10.11884/HPLPB202234.210309

Citation:

Feng Zhiyuan, Li Kaiwen, Luo Hao, et al. Micro cross-section parameterization based on RMC code[J]. High Power Laser and Particle Beams, 2022, 34: 026006. doi: 10.11884/HPLPB202234.210309

PDF( 822 KB)

Micro cross-section parameterization based on RMC code

doi: 10.11884/HPLPB202234.210309

Department of Engineering Physics, Tsinghua University, Beijing 100084, China

Received Date: 2021-07-22
Rev Recd Date: 2021-09-06

Available Online: 2021-10-14

Publish Date: 2022-01-11

Abstract

Abstract

To perform realistic core calculations, few-group neutron cross-sections library by functions of burn-up and thermal hydraulics parameters should be prepared in advance. Traditional deterministic parameterization process is based on the macro cross-section. However, this method should consider the historical effect of some physical states, which increases the number of calculation branches. Thus, this paper proposes a new parameterization process based on the micro cross-section of nuclide. This method effectively eliminates the historical impact. Therefore, only burn-up and material temperature need to be considered. The calculation process is performed by RMC code. Firstly, all of material nuclides micro cross-section and nuclides density is calculated. Then, macro cross-section is obtained by the micro cross-section lib. To test the method accuracy, a self-design pressurized water reactor model is built. The test results agree well with the reference results calculated by RMC full core calculation.
- cross-section parameterization,
- RMC,
- few-group micro cross-section,
- burn-up,
- effective multiplication factor

FullText(HTML)

References(8)

References

[1]	Weston M. S. Nuclear reactor physics[M]. America: Wiley Interscience, 2001.
[2]	Wang Y, Bangerth W, Ragusa J. Three-dimensional h-adaptivity for the multi-group neutron diffusion equations[J]. Progress in Nuclear Energy, 2009, 51(3): 543-555. doi: 10.1016/j.pnucene.2008.11.005
[3]	Sugimura N, Yamamoto A. Resonance treatment based on ultra-fine-group spectrum calculation in the AEGIS code[J]. Journal of Nuclear Science and Technology, 2007, 44: 958-966. doi: 10.1080/18811248.2007.9711335
[4]	Fridman E, Leppanen J. On the use of the Serpent Monte Carlo code for few-group cross section generation[J]. Annals of Nuclear Energy, 2011, 38(6): 1399-1405. doi: 10.1016/j.anucene.2011.01.032
[5]	Liu Z, Smith K, Forget B, et al. Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo[J]. Annals of Nuclear Energy, 2018, 112: 507-516. doi: 10.1016/j.anucene.2017.10.039
[6]	Shim H, Cho Y, Song S, et al. Generation of few group diffusion theory constants by Monte Carlo code[J]. Transactions of the American Nuclear Society, 2008, 172: 66-77.
[7]	Shinzaburo M. MVP/GMVP II: General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods[M]. Japan: Japan Atomic Energy Research Institute, 2005.
[8]	Nikitin E, Fridman E, Mikityuk K. On the use of the SPH method in nodal diffusion analyses of SFR cores[J]. Annals of Nuclear Energy, 2015, 85: 544-551. doi: 10.1016/j.anucene.2015.06.007