Volume 30 Issue 5
May  2018
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Dou Haifeng, Li Rundong, Leng Jun, et al. Benchmarking verification of control rod effects on CMRR using MCNP codes throughout 3D core modeling and rod-drop experiment[J]. High Power Laser and Particle Beams, 2018, 30: 056001. doi: 10.11884/HPLPB201830.170345
Citation: Dou Haifeng, Li Rundong, Leng Jun, et al. Benchmarking verification of control rod effects on CMRR using MCNP codes throughout 3D core modeling and rod-drop experiment[J]. High Power Laser and Particle Beams, 2018, 30: 056001. doi: 10.11884/HPLPB201830.170345

Benchmarking verification of control rod effects on CMRR using MCNP codes throughout 3D core modeling and rod-drop experiment

doi: 10.11884/HPLPB201830.170345
  • Received Date: 2017-09-01
  • Rev Recd Date: 2018-01-15
  • Publish Date: 2018-05-15
  • In this research, MCNP code and ORIGEN code are used to calculate the control rod reactivity worth effects by simulating the 3D core model of CMRR reactor. The integral and differential behaviors of reactivity worth effects are measured by rod-drop experiments and digital inverse kinetic method with each other. The calculated and measured results are well accorded. The integral reactivity worth of one safety rod is about 4%Δk/k. Even in an accident when one safety rod gets stuck, the CMRR shutdown margin is still greater than 10%Δk/k, and CMRR is totally safe. So the physical design of CMRR is highly reliable and the operation could be safe.
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  • [1]
    IAEA. Utilization-related design features of research reactors: a compendium[R]. Technical Report No. 455, 2007.
    [2]
    IAEA. Safety analyses for research reactors[R]. Safety Report No. 55, 2008.
    [3]
    IAEA. Core management and fuel handling for research reactors[R]. Safety Standard No. NS-G-4.3, 2008.
    [4]
    IAEA. Operational limits and conditions and operating procedures for research reactors[R]. Safety Standard No. NS-G-4.4, 2008.
    [5]
    ORNL. WIMS-D4: Winfrith Improved Multigroup Scheme code system[R]. Code CCC-575, 1991.
    [6]
    Fowler T B, Vondy D R, Cunningham G W, et al. Nuclear reactor core analysis code: CITATION[R]. ORNL-TM-2496, 1971.
    [7]
    ORNL. Scale: A comprehensive modeling and simulation suite for nuclear safety analysis and design[R]. ORNL-TM-39, 2005.
    [8]
    Briesmeister J F. MCNP—A general Monte Carlo N-particle transport code version 4C[R]. LA-13709-M, 2000.
    [9]
    黄洪文, 叶林, 钱达志, 等. 新型铪控制棒的研制[J]. 核动力工程, 2008, 29(3): 48-51. https://www.cnki.com.cn/Article/CJFDTOTAL-HDLG200803012.htm

    Huang Hongwen, Ye Lin, Qian Dazhi, et al. Development and manufacture of new-style hafnium control rod. Nuclear Power Engineering, 2008, 29(3): 48-51 https://www.cnki.com.cn/Article/CJFDTOTAL-HDLG200803012.htm
    [10]
    MacFarlane R E. Data testing for ENDF/B-Ⅶ[R]. Los Alamos National Laboratory, 2011.
    [11]
    李润东, 代君龙, 王学杰. 300#堆周期和反应性数字化测量技术研究[C]//第五届全国核仪器及其应用学术会议. 2007: 188-191.

    Li Rundong, Dai Junlong, Wang Xuejie. Measurement digitalization for period and reactivity of SPRR-300//Proceedings of the China Conference on Nuclear Instrument Application & Nuclear Detection Technology & Nuclear Measurement Method. 2007: 188-191
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