Stabilization of multi-group neutron transport with transport-corrected cross-sections

被引:7
|
作者
Gunow, Geoffrey [1 ]
Forget, Benoit [2 ]
Smith, Kord [2 ]
机构
[1] Bloomberg LP, 731 Lexington Ave, New York, NY 10022 USA
[2] MIT, Dept Nucl Sci & Engn, 77 Massachusetts Ave,Bldg 24, Cambridge, MA 02139 USA
关键词
Transport correction; Convergence; Neutron transport; Method of characteristics; ANISOTROPIC SCATTERING;
D O I
10.1016/j.anucene.2018.10.036
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
Many deterministic neutron transport solvers rely on the source iteration method to solve the multi-group neutron transport equation. Often, these solvers rely on transport corrected cross-sections for accurate prediction of the neutron flux distribution. Transport corrected within-group scattering cross sections can become negative, and if these negative cross-sections are large in magnitude, source iteration can become unstable and fail to converge for certain cases. In this study, we present evidence of this convergence issue for the method of characteristics (MOC) on full-core PWR problems with common transport correction schemes. A theoretical discussion is presented to illustrate the reason for the convergence issues. Previously established stabilization methods are compared with a newly proposed stabilization method. Results show that the new stabilization method allows for faster convergence than previous techniques. In addition, the effect of Coarse Mesh Finite Difference (CMFD) acceleration on stability is analyzed, showing that CMFD acceleration with full-group structure can overcome the convergence issues, but a stabilization technique is necessary for convergence when a condensed group structure is used. (C) 2018 Published by Elsevier Ltd.
引用
收藏
页码:211 / 219
页数:9
相关论文
共 50 条
  • [41] Calculation of multi-group migration areas in deterministic transport simulations
    Liu, Zhaoyuan
    Smith, Kord
    Forget, Benoit
    ANNALS OF NUCLEAR ENERGY, 2020, 140
  • [42] Parallel Numerical Simulation for the Multi-group Particle Transport Equations
    Liu, Jie
    Chi, Lihua
    Chen, Jing
    DCABES 2008 PROCEEDINGS, VOLS I AND II, 2008, : 154 - 160
  • [43] MONTE-CARLO SIMULATION OF MULTIGROUP TRANSPORT CROSS-SECTIONS
    HARRIS, DR
    TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 1971, 14 (02): : 647 - &
  • [44] Transport and relaxation cross-sections for pure gases of linear molecules
    Heck, EL
    Dickinson, AS
    COMPUTER PHYSICS COMMUNICATIONS, 1996, 95 (2-3) : 190 - 220
  • [45] FLUCTUATING CROSS-SECTIONS IN MONTE-CARLO TRANSPORT CALCULATIONS
    TROUBETZKOY, ES
    TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 1976, 24 (NOV19): : 199 - 199
  • [46] APPROXIMATION OF GIVEN NEUTRON CROSS-SECTIONS BY LINEAR COMBINATION OF ACTIVATION CROSS-SECTIONS
    KONAKOV, SA
    NOVIKOV, VM
    CHUVILIN, DY
    KERNTECHNIK, 1989, 53 (03) : 223 - 226
  • [47] VALIDATION OF NEUTRON CROSS-SECTIONS CROSS-SECTIONS FOR SODIUM AND ADJUSTMENT TO INTEGRAL MEASUREMENTS
    PFISTER, G
    HEHN, G
    MATTES, M
    KERNTECHNIK, 1987, 51 (01) : 24 - 27
  • [48] Generating multi-group cross-sections using continuous-energy Monte Carlo method for fast reactor analysis
    Guo, Hui
    Shen, Yuyang
    Wu, Yiwei
    Song, Qufei
    Gu, Hanyang
    JOINT INTERNATIONAL CONFERENCE ON SUPERCOMPUTING IN NUCLEAR APPLICATIONS + MONTE CARLO, SNA + MC 2024, 2024, 302
  • [49] Development of neutron-photon coupling transport code IMPC-NP and optimization of method for fabrication of neutron-photon multi-group cross section
    Fang, Peng
    Wu, Xiang
    Lai, Hanghui
    Yang, Yongwei
    Yang, Lei
    Guo, Yuyao
    Zhou, Feng
    Zhu, Yanling
    PROGRESS IN NUCLEAR ENERGY, 2024, 175
  • [50] A single-step framework to generate spatially self-shielded multi-group cross sections from Monte Carlo transport simulations
    Boyd, William
    Forget, Benoit
    Smith, Kord
    ANNALS OF NUCLEAR ENERGY, 2019, 125 : 261 - 271