高级检索

留言板

尊敬的读者、作者、审稿人, 关于本刊的投稿、审稿、编辑和出版的任何问题, 您可以本页添加留言。我们将尽快给您答复。谢谢您的支持!

姓名
邮箱
手机号码
标题
留言内容
验证码

含强吸收体堆芯的各向异性SP3两步法计算研究

李云召 秦浚玮 夏凡 吴宏春 曹良志

李云召, 秦浚玮, 夏凡, 吴宏春, 曹良志. 含强吸收体堆芯的各向异性SP3两步法计算研究[J]. 核动力工程, 2023, 44(6): 9-15. doi: 10.13832/j.jnpe.2023.06.0009
引用本文: 李云召, 秦浚玮, 夏凡, 吴宏春, 曹良志. 含强吸收体堆芯的各向异性SP3两步法计算研究[J]. 核动力工程, 2023, 44(6): 9-15. doi: 10.13832/j.jnpe.2023.06.0009
Li Yunzhao, Qin Junwei, Xia Fan, Wu Hongchun, Cao Liangzhi. Anisotropic SP3 Two-Step Method for Reactor Cores with Strong Absorbers[J]. Nuclear Power Engineering, 2023, 44(6): 9-15. doi: 10.13832/j.jnpe.2023.06.0009
Citation: Li Yunzhao, Qin Junwei, Xia Fan, Wu Hongchun, Cao Liangzhi. Anisotropic SP3 Two-Step Method for Reactor Cores with Strong Absorbers[J]. Nuclear Power Engineering, 2023, 44(6): 9-15. doi: 10.13832/j.jnpe.2023.06.0009

含强吸收体堆芯的各向异性SP3两步法计算研究

doi: 10.13832/j.jnpe.2023.06.0009
基金项目: 国家自然科学基金(11975181)
详细信息
    作者简介:

    李云召(1984—),男,教授,现主要从事核反应堆物理相关的科研与教学工作,E-mail: yunzhao@xjtu.edu.cn

  • 中图分类号: TL324

Anisotropic SP3 Two-Step Method for Reactor Cores with Strong Absorbers

  • 摘要: 核反应堆堆芯中出现的强吸收体,尤其是可移动的控制棒,会显著增强堆芯内中子角注量率的角度各向异性效应,传统的各向同性SP3两步法不能进行有效刻画,需要构建各向异性SP3两步法。本文首先从2个方面研究各向异性效应;推导了各向异性SP3方程,建立了适用于各向异性SP3方程的均匀化模型。通过材料试验堆(MTR)对本文方法进行验证与分析。结果表明,本文研究的各向异性SP3方程及其均匀化方法较之传统计算方法偏差更小,对于有效增殖系数(keff)和功率分布改善较为明显。因此,本文研究的各向异性SP3两步法能够有效处理含强吸收体堆芯问题。

     

  • 图  1  MTR 多元件问题示意图

    Figure  1.  Configuration of MTR Multi-Assembly Problem

    图  2  新型均匀化模型示意图

    Figure  2.  Schematic Diagram of New Homogenization Model

    图  3  不同计算方法所得功率分布偏差示意图

    Figure  3.  Schematic Diagram of Power Distribution Deviation Obtained by Different Calculation Methods

    图  4  二维MTR问题堆芯及反射层模型示意图

    Figure  4.  Schematic Diagram of 2D MTR Core and Reflector Model

    图  5  二维MTR问题燃料元件模型示意图

    Figure  5.  Fuel Element Model of 2D MTR Problem

    图  6  二维MTR问题功率分布偏差示意图

    Figure  6.  Power Distribution Deviation of the 2D MTR Problem     

    表  1  不同散射阶数keff计算结果及偏差

    Table  1.   keff Results and Deviations of Different Scattering Orders

    计算方式不插控制棒Ag-In-Cd控制棒Hf控制棒B4C控制棒
    keff偏差/pcmkeff偏差/pcmkeff偏差/pcmkeff偏差/pcm
    参考解1.527850.847980.838300.72975
    Inflow1.52755–300.85518+7200.84516+6860.74195+1220
    P11.52829+440.84343–4550.83226–6040.72414–561
    P21.52834+490.84806+80.83732–980.73081+106
    P31.52832+470.84789–90.83719–1110.73087+112
    下载: 导出CSV

    表  2  MTR中2种燃料元件的几何与材料信息

    Table  2.   Geometric and Material Parameters of 2 Fuel Elements in MTR

    参数名称标准燃料元件控制燃料元件
    组件尺寸/cm7.7×8.17.7×8.1
    燃料材料UAlx-AlUAlx-Al
    燃料
    富集度/%
    LEU2020
    MEU4545
    HEU9393
    燃料板数2317
    燃料板厚度/cm0.0510.051
    燃料板长度/cm6.36.3
    包壳材料AlAl
    包壳厚度/cm0.0380.038
    板间厚度/cm0.2230.223
    控制棒材料Ag-In-Cd
      LEU—低浓铀;MEU—中浓铀;HEU—高浓铀
    下载: 导出CSV

    表  3  新型均匀化模型计算结果

    Table  3.   Calculation Results of New Homogenization Model

    计算方式参考解keffkeff偏差/
    pcm
    均方根
    偏差/%
    最大
    偏差/%
    各向异性SP3(P2)1.059311.05979+480.472.10
    各向同性SP3(P2)1.07745+18146.92−11.46
    各向同性SP3(Inflow)1.05702−2290.662.64
    下载: 导出CSV

    表  4  二维MTR堆芯问题计算结果

    Table  4.   Results of 2D MTR Core Problem

    keff功率分布偏差/%
    参考值计算值偏差/pcm均方根最大值
    1.015451.01616+710.973.72
    下载: 导出CSV
  • [1] PALMTAG S P. Advanced nodal methods for MOX fuel analysis[D]. Cambridge: Massachusetts Institute of Technology, 1997.
    [2] TATSUMI M, YAMAMOTO A. Advanced PWR core calculation based on multi-group nodal-transport method in three-dimensional pin-by-pin geometry[J]. Journal of Nuclear Science and Technology, 2003, 40(6): 376-387. doi: 10.1080/18811248.2003.9715369
    [3] YANG W, ZHENG Y Q, WU H C, et al. High-performance whole core pin-by-pin calculation based on EFEN-SP3 method[J]. Nuclear Power Engineering, 2014, 35(5): 164-167.
    [4] TATSUMI M, OHOKA Y, ENDO T, et al. Verification and validation of AEGIS/SCOPE2: the state-of-the-art in-core fuel management system for PWRs[C]//18th International Conference on Nuclear Engineering. Xi’an: American Society of Mechanical Engineers, 2010,doi: 10.1115/ICONE18-29154.
    [5] GRUNDMANN U, MITTAG S. Super-homogenisation factors in pinwise calculations by the reactor dynamics code DYN3D[J]. Annals of Nuclear Energy, 2011, 38(10): 2111-2119. doi: 10.1016/j.anucene.2011.06.030
    [6] LIU S C, WANG G B, LIANG J G, et al. Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods[J]. Annals of Nuclear Energy, 2015, 85: 830-836. doi: 10.1016/j.anucene.2015.06.041
    [7] MARLEAU G, HÉBERT A, ROY R. A user guide for DRAGON version DRAGON_000331 Realease 3.04, IGE-174[Z]. 2000
    [8] HÉBERT A, SEKKI D, CHAMBON R. A user guide for DONJON version4: IGE-300[R]. Montréal: École Polytechnique de Montréal, 2013.
    [9] MA J M, WANG G B, YUAN S, et al. An improved assembly homogenization approach for plate-type research reactor[J]. Annals of Nuclear Energy, 2015, 85: 1003-1013. doi: 10.1016/j.anucene.2015.07.018
    [10] 汤春桃. 中子输运方程特征线解法及嵌入式组件均匀化方法的研究[D]. 上海: 上海交通大学, 2009.
    [11] CHAUDRI K S, MIRZA S M. Burnup dependent Monte Carlo neutron physics calculations of IAEA MTR benchmark[J]. Progress in Nuclear Energy, 2015, 81: 43-52. doi: 10.1016/j.pnucene.2014.12.018
    [12] SHAABAN I, ALBARHOUM M. Performance of the MTR core with MOX fuel using the MCNP4C2 code[J]. Applied Radiation and Isotopes, 2016, 114: 92-103. doi: 10.1016/j.apradiso.2016.05.009
    [13] ALAWNEH L M, PARK C J, JARADAT M K, et al. Burnup estimation for plate type fuel assembly of research reactors through the least square fitting method[J]. Annals of Nuclear Energy, 2014, 71: 37-45. doi: 10.1016/j.anucene.2014.03.029
    [14] MARGULIS M, GILAD E. Monte Carlo and nodal neutron physics calculations of the IAEA MTR benchmark using Serpent/DYN3D code system[J]. Progress in Nuclear Energy, 2016, 88: 118-133. doi: 10.1016/j.pnucene.2015.12.008
    [15] YAMAMOTO A, KITAMURA Y, YAMANE Y. Simplified treatments of anisotropic scattering in LWR core calculations[J]. Journal of Nuclear Science and Technology, 2008, 45(3): 217-229. doi: 10.1080/18811248.2008.9711430
    [16] XIA F, ZU T J, WU H C. Effect and treatment of angular dependency of multi-group total cross section and anisotropic scattering in fine-mesh transport calculation[J]. Annals of Nuclear Energy, 2018, 114: 110-121. doi: 10.1016/j.anucene.2017.11.049
    [17] USHIO T, TAKEDA T, MORI M. Neutron anisotropic scattering effect in heterogeneous cell calculations of light water reactors[J]. Journal of Nuclear Science and Technology, 2003, 40(7): 464-480. doi: 10.1080/18811248.2003.9715381
    [18] GELBARD E M. Application of spherical harmonics method to reactor problems: WAPD-T-1182[R]. West Mifflin: Bettis Atomic Power Laboratory, 1960
    [19] NELSON A. Improved convergence rate of multi-group scattering moment tallies for Monte Carlo neutron transport codes[D]. Michigan: University of Michigan, 2014.
    [20] WANG S C, CAO L Z, LI Y Z, et al. An energy-group structure optimization from seven to four for PWR-core pin-by-pin calculation[J]. Nuclear Engineering and Design, 2023, 402: 112115. doi: 10.1016/j.nucengdes.2022.112115
    [21] IAEA. Research reactor core conversion from the use of highly enriched uranium fuels: guidebook[M]. Vienna: International Atomic Energy Agency, 1980: 39-45.
    [22] LI Y Z, ZHANG B, HE Q M, et al. Development and verification of PWR-core fuel management calculation code system NECP-bamboo: part I bamboo-lattice[J]. Nuclear Engineering and Design, 2018, 335: 432-440. doi: 10.1016/j.nucengdes.2018.05.030
  • 加载中
图(6) / 表(4)
计量
  • 文章访问数:  91
  • HTML全文浏览量:  30
  • PDF下载量:  29
  • 被引次数: 0
出版历程
  • 收稿日期:  2022-12-15
  • 修回日期:  2023-02-01
  • 网络出版日期:  2023-12-11
  • 刊出日期:  2023-12-15

目录

    /

    返回文章
    返回