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Volume 43 Issue 3
Jun.  2022
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Zu Tiejun, Xu Ning, Cao Liangzhi, Wu Hongchun. Development and Verification of Shielding Database Generation Module in NECP-Atlas[J]. Nuclear Power Engineering, 2022, 43(3): 15-20. doi: 10.13832/j.jnpe.2022.03.0015
Citation: Zu Tiejun, Xu Ning, Cao Liangzhi, Wu Hongchun. Development and Verification of Shielding Database Generation Module in NECP-Atlas[J]. Nuclear Power Engineering, 2022, 43(3): 15-20. doi: 10.13832/j.jnpe.2022.03.0015

Development and Verification of Shielding Database Generation Module in NECP-Atlas

doi: 10.13832/j.jnpe.2022.03.0015
  • Received Date: 2021-03-09
  • Accepted Date: 2021-09-07
  • Rev Recd Date: 2021-05-26
  • Publish Date: 2022-06-07
  • The shielding database generation module Shield_calc is developed in the nuclear data processing program NECP-Atlas. This module firstly uses NECP-Atlas to generate a database of fine-group neutron and photon cross sections in MATXS format which is irrelevant to the problem; then uses the ultrafine group method and Bondarenko iterative method to perform resonance self-shielding calculation to obtain effective self-shielding cross-sections; Finally, based on the one-dimensional reactor model, NECP-Hydraa is used for transportation calculation to obtain the typical weight spectrum of the applied reactor type, and the fine-group shielding database is merged into the wide-group shielding database NECL-SHILED. Using Shield_calc module, based on the same evaluation nuclear database ENDF/B-Ⅶ.0 as BUGLE-B7, NECL-SHILED with 47 groups of neutrons and 20 groups of photons is generated and compared with BUGLE-B7. The numerical results show that the calculation results of NECL-SHILD and BUGLE-B7 are in good agreement, which verifies that the Shield_calc module has high accuracy.

     

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  • [1]
    RISNER J M, WIARDA D, DUNN M E, et al. Production and testing of the VITAMIN-B7 fine-group and BUGLE-B7 broad-group coupled neutron/gamma cross-section libraries derived from ENDF/B-Ⅶ. 0 nuclear data: ORNL/TM-2011/12, NUREG/CR-7045[R]. Oak Ridge: Oak Ridge National Lab., 2011.
    [2]
    CHADWICK M B, OBLOŽINSKÝ P, HERMAN M, et al. ENDF/B-Ⅶ. 0: next generation evaluated nuclear data library for nuclear science and technology[J]. Nuclear Data Sheets, 2006, 107(12): 2931-3060. doi: 10.1016/j.nds.2006.11.001
    [3]
    MACFARLANE R, MUIR D W, BOICOURT R M et al. The NJOY nuclear data processing system, version 2016: LA-UR-17-20093[R]. Los Alamos, NM, United States: Los Alamos National Laboratory, 2017.
    [4]
    MACFARLANE R E. TRANSX 2: a code for interfacing MATXS cross-section: LA-12312-MS[R]. Los Alamos NM, United States: Los Alamos National Laboratory, 1992.
    [5]
    陈义学,陈朝斌,吴军,等. 基于ENDF/B-Ⅶ. 0评价库的多群参数库MUSE1.0的开发与初步验证[J]. 核动力工程,2010, 31(2): 6-10,15.
    [6]
    WIARDA D, DUNN M. PUFF-Ⅳ: code system to generate multigroup covariance matrices from ENDF/B-Ⅵ uncertainty files: PSR-534[R]. Oak Ridge TN, United States: Oak Ridge National Laboratory, 2006
    [7]
    SCALE. A comprehensive modeling and simulation suite for nuclear safety analysis and design: ORNL/TM-2005/39[R]. Version 6.1. Oak Ridge, TN (United States): Oak Ridge National Laboratory, 2011
    [8]
    BROWN D A, CHADWICK M B, CAPOTE R, et al. ENDF/B-Ⅷ. 0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data[J]. Nuclear Data Sheets, 2018, 148: 1-142. doi: 10.1016/j.nds.2018.02.001
    [9]
    彭超,丁谦学,梅其良,等. 压水堆离散纵标屏蔽设计多群参数库的开发与初步验证[J]. 核动力工程,2020, 41(3): 19-23.
    [10]
    ZU T J, XU J L, TANG Y Q, et al. NECP-Atlas: a new nuclear data processing code[J]. Annals of Nuclear Energy, 2019, 123: 153-161. doi: 10.1016/j.anucene.2018.09.016
    [11]
    毕沪超,祖铁军,徐嘉隆,等. NECP-Atlas不可辨共振能区概率表模块的开发和验证[J]. 核动力工程,2020, 41(3): 8-13.
    [12]
    徐嘉隆,祖铁军,曹良志,等. NECP-Atlas中多群常数处理模块的开发与验证[J]. 核动力工程,2019, 40(1): 12-17.
    [13]
    谢仲生. 压水堆核电厂堆芯燃料管理计算及优化[M]. 北京: 原子能出版社, 2001: 27-36.
    [14]
    ZU T J, YIN W, HE Q M, et al. Application of the hyperfine group self-shielding calculation method to the lattice and whole-core physics calculation[J]. Annals of Nuclear Energy, 2020, 136: 107045. doi: 10.1016/j.anucene.2019.107045
    [15]
    XU L F, CAO L Z, ZHENG Y Q, et al. Development of a new parallel SN code for neutron-photon transport calculation in 3-D cylindrical geometry[J]. Progress in Nuclear Energy, 2017, 94: 1-21. doi: 10.1016/j.pnucene.2016.09.005
    [16]
    OECD-NEA. International handbook of evaluated criticality safety benchmark experiments: NEA-7231[R]. Paris, France: Nuclear Energy Agency, Organization for Economic Cooperation and Development, 2006.
    [17]
    PESCARINI M, ORSI R, FRISONI M. PCA-Replica (H2O/Fe) neutron shielding benchmark experiment – deterministic analysis in cartesian (X,Y,Z) geometry using the TORT-3.2 3D transport code and the BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-96 Cross Section Libraries: UTFISSM-P9H6-009[R]. Bologna: ENEA-Bologna Research Centre, 2014
    [18]
    REMEC I, KAM F B K. H. B. robinson-2 pressure vessel benchmark: NUREG/CR-6453, ORNL/TM-13204[R]. Washington: U. S. Nuclear Regulatory Commission, 1998.
    [19]
    LEENDERS L. LWR-PVS benchmark experiment VENUS-3: FCP/VEN/01[R]. Mol, Belgium: CEN/SCK, 1988.
    [20]
    ORSI R. H. B. Robinson-2 pressure vessel dosimetry benchmark: deterministic three-dimensional analysis with the TORT transport code[J]. Nuclear Engineering and Technology, 2020, 52(2): 448-455. doi: 10.1016/j.net.2019.07.025
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