分类： 核科学技术 >> 裂变堆工程技术 提交时间： 2021-12-31
摘要： Nuclear data are the cornerstones of reactor physics and shielding calculations. Recently, China released CENDL-3.2 in 2020, and the United States released ENDF/B-VIII.0 in 2018. Therefore, it is necessary to comprehensively evaluate the criticality computing performance of these newly released evaluated nuclear libraries. In this study, we used the NJOY2016 code to generate ACE format libraries based on the latest neutron data libraries (including CENDL-3.2, JEFF3.3, ENDF/B-VIII.0, and JENDL4.0). The MCNP code was used to conduct a detailed analysis of fission nuclides, including 235U, 233U, and 239Pu, in different evaluated nuclear data libraries based on 100 benchmarks. The criticality calculation performance of each library was evaluated using three statistical parameters: , , and . Analysis of the parameter showed that CENDL-3.1 and JENDL-4.0 both had >10 benchmarks that exceeded 3σ, whereas CENDL-3.2, ENDFB-VIII.0, and JEFF-3.3 had, 7, 5, and 4 benchmarks, respectively, exceeding 3σ. The ENDF/B-VII.1 library performed best, with only two benchmarks exceeding 3σ. Compared with CENDL-3.1, CENDL-3.2 offers an improvement in criticality calculations. Compared with the JEFF-3.3 and ENDF/B-VIII.0 libraries, CENDL3.2 performs better in the calculation of the 233U assemblies, but it performs poorly in the pusl11 series case calculation of the 239Pu assemblies, and thus further improvement is needed.
分类： 核科学技术 >> 裂变堆工程技术 提交时间： 2021-12-31
摘要： The transport cross-section based on inflow transport approximation can significantly improve the accuracy of light water reactor (LWR) analysis, especially for the treatment of the anisotropic scattering effect. The previous inflow transport approximation is based on the moderator cross-section and normalized fission source, which is approximated using transport theory. Although the accuracy of reactivity is increased, the P0 flux moment has a large error in the Monte Carlo code. In this study, an improved inflow transport approximation was introduced with homogenization techniques, applying the homogenized cross-section and accurate fission source. The numerical results indicated that the improved inflow transport approximation can increase the P0 flux moment accuracy and maintain the reactivity calculation precision with the previous inflow transport approximation in typical LWR cases. In addition to this investigation, the improved inflow transport approximation is related to the temperature factors. The improved inflow transport approximation is flexible and accurate in the treatment of the anisotropic scattering effect, which can be directly used in the temperature-dependent nuclear data library.
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