Advance Search
Volume 44 Issue 2
Apr.  2023
Turn off MathJax
Article Contents
Zhang Tengfei, Yin Han, Sun Qizheng, Xiao Wei. Application Research on VITAS—a General-purpose Neutron Transport Code[J]. Nuclear Power Engineering, 2023, 44(2): 15-23. doi: 10.13832/j.jnpe.2023.02.0015
Citation: Zhang Tengfei, Yin Han, Sun Qizheng, Xiao Wei. Application Research on VITAS—a General-purpose Neutron Transport Code[J]. Nuclear Power Engineering, 2023, 44(2): 15-23. doi: 10.13832/j.jnpe.2023.02.0015

Application Research on VITAS—a General-purpose Neutron Transport Code

doi: 10.13832/j.jnpe.2023.02.0015
  • Received Date: 2022-06-02
  • Rev Recd Date: 2023-02-14
  • Publish Date: 2023-04-15
  • In order to improve the applicability of deterministic whole-core neutron transport code, a general-purpose neutron transport code VITAS is developed. The verification results of TAKEDA3 benchmark problem (rectangular assembly), the TAKEDA4 benchmark problem (hexagonal assembly), the Dodds benchmark problem (R-Z geometry) and the C5G7-TD5 benchmark problem (PWR high fidelity calculation) show that higher-order spatial and angular basis functions can make the results converge asymptotically and steadily to the reference solution, reaching the calculation accuracy level equivalent to that of multi-group Monte Carlo method. Compared with the reference solution, the deviations are less than 60pcm (1pcm = 10−5) in effective multiplication coefficient (keff), −3pcm in control rod worth and 1.03% in neutron flux distribution root mean square (RMS) for TAKEDA3 benchmark problem; less than 20pcm in keff, 32pcm in control rod worth and 0.70% in neutron flux distribution RMS for TAKEDA4 benchmark problem; less than 1% (maximum) in power for Dodds benchmark problem; and less than 0.9% in power for C5G7-TD5 benchmark problem. The research in this paper shows that VITAS has the potential to become a general-purpose calculation tool for accurately solving the neutron transport problems.

     

  • loading
  • [1]
    曹良志, 谢仲生, 李云召. 近代核反应堆物理分析[M]. 北京: 中国原子能出版社, 2017:4-7.
    [2]
    DOWNAR T, STAUDENMIER J. Theory Manual for the PARCS Neutronics core simulator: PARCS v2.6 U. S. NRC Core Neutronics Simulator, theory manual[Z]. Indiana, United States: Purdue University, 2004.
    [3]
    LAWRENCE R D. The DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry: ANL-83-1[R]. Argonne, Illinois, United States: Argonne National Lab, 1983.
    [4]
    VER PLANCK D M, COBB W R, BORLAND R S, et al. SIMULATE-E: a nodal core-analysis program for light-water reactors. Computer code user's manual: EPRI-NP-2792-CCM[R]. Framingham: Yankee Atomic Electric Co, 1983.
    [5]
    YANG W, WU H C, LI Y Z, et al. Development and verification of PWR-core fuel management calculation code system NECP-Bamboo: part Ⅱ Bamboo-Core[J]. Nuclear Engineering and Design, 2018, 337: 279-290. doi: 10.1016/j.nucengdes.2018.07.017
    [6]
    ZHANG T F, LI Z P. Variational nodal methods for neutron transport: 40 years in review[J]. Nuclear Engineering and Technology, 2022, 54(9): 3181-3204. doi: 10.1016/j.net.2022.04.012
    [7]
    ZHANG T F, WU H C, CAO L Z, et al. An improved variational nodal method for the solution of the three- dimensional steady-state multi-group neutron transport equation[J]. Nuclear Engineering and Design, 2018, 337: 419-427. doi: 10.1016/j.nucengdes.2018.07.009
    [8]
    ZHANG T F, XIONG J B, LIU L G, et al. Development and implementation of an integral variational nodal method to the hexagonal geometry nuclear reactors[J]. Annals of Nuclear Energy, 2019, 131: 210-220. doi: 10.1016/j.anucene.2019.03.031
    [9]
    ZHUANG K, SHANG W, LI T, et al. Variational nodal method for three-dimensional multigroup neutron diffusion equation based on arbitrary triangular prism[J]. Annals of Nuclear Energy, 2021, 158: 108285. doi: 10.1016/j.anucene.2021.108285
    [10]
    ZHANG T F, LEWIS E E, SMITH M A, et al. A variational nodal approach to 2D/1D pin resolved neutron transport for pressurized water reactors[J]. Nuclear Science and Engineering, 2017, 186(2): 120-133. doi: 10.1080/00295639.2016.1273023
    [11]
    ZHANG T F, WANG Y P, LEWIS E E, et al. A three-dimensional variational nodal method for pin-resolved neutron transport analysis of pressurized water reactors[J]. Nuclear Science and Engineering, 2017, 188(2): 160-174. doi: 10.1080/00295639.2017.1350002
    [12]
    XIAO W, LI X Y, LI P J, et al. High-fidelity multi-physics coupling study on advanced heat pipe reactor[J]. Computer Physics Communications, 2022, 270: 108152. doi: 10.1016/j.cpc.2021.108152
    [13]
    GEUZAINE C, REMACLE J F. Gmsh: a 3-D finite element mesh generator with built-in pre- and post-processing facilities[J]. International Journal for Numerical Methods in Engineering, 2009, 79(11): 1309-1331. doi: 10.1002/nme.2579
    [14]
    LOELIGER J, MCCULLOUGH M. Version control with Git: powerful tools and techniques for collaborative software development[M]. 2nd ed. 1005 Gravenstein Highway North, Sebastopol: O'Reilly Media, 2012:336.
    [15]
    BONNEY G E, KISSLING G E. Gram‐Schmidt orthogonalization of multinormal variates: applications in genetics[J]. Biometrical Journal, 1986, 28(4): 417-425. doi: 10.1002/bimj.4710280407
    [16]
    TAKEDA T, IKEDA H. 3-D neutron transport benchmarks[J]. Journal of Nuclear Science and Technology, 1991, 28(7): 656-669. doi: 10.1080/18811248.1991.9731408
    [17]
    ROMANO P K, FORGET B. The OpenMC Monte Carlo particle transport code[J]. Annals of Nuclear Energy, 2013, 51: 274-281. doi: 10.1016/j.anucene.2012.06.040
    [18]
    SEUBERT A, SUREDA A, BADER J, et al. The 3-D time-dependent transport code TORT-TD and its coupling with the 3D thermal-hydraulic code ATTICA3D for HTGR applications[J]. Nuclear Engineering and Design, 2012, 251: 173-180. doi: 10.1016/j.nucengdes.2011.09.067
    [19]
    GOLUOGLU S, BENTLEY C, DEMEGLIO R, et al. A deterministic method for transient, three-dimensional neutron transport: MOL. 19980504.0231[R]. Las Vegas: Yucca Mountain Project, 1998.
    [20]
    Argonne Code Center. Argonne Code Center: benchmark problem book: ANL-7416[R]. Argonne, Illinois, United States: American Nuclear Society, 1977.
    [21]
    HOU J, IVANOV K N, BOYARINOV V F, et al. OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization[J]. Nuclear Engineering and Design, 2017, 317: 177-189. doi: 10.1016/j.nucengdes.2017.02.008
    [22]
    SHEN Q C, WANG Y R, JABAAY D, et al. Transient analysis of C5G7-TD benchmark with MPACT[J]. Annals of Nuclear Energy, 2019, 125: 107-120. doi: 10.1016/j.anucene.2018.10.049
  • 加载中

Catalog

    通讯作者: 陈斌, bchen63@163.com
    • 1. 

      沈阳化工大学材料科学与工程学院 沈阳 110142

    1. 本站搜索
    2. 百度学术搜索
    3. 万方数据库搜索
    4. CNKI搜索

    Figures(8)  / Tables(4)

    Article Metrics

    Article views (1541) PDF downloads(73) Cited by()
    Proportional views
    Related

    /

    DownLoad:  Full-Size Img  PowerPoint
    Return
    Return