混合能谱超临界水堆失流事故缓解措施研究

许志红, 傅晟威, 刘晓晶, 杨燕华, 程旭

许志红, 傅晟威, 刘晓晶, 杨燕华, 程旭. 混合能谱超临界水堆失流事故缓解措施研究[J]. 原子能科学技术, 2012, 46(9): 1097-1106. DOI: 10.7538/yzk.2012.46.09.1097
引用本文: 许志红, 傅晟威, 刘晓晶, 杨燕华, 程旭. 混合能谱超临界水堆失流事故缓解措施研究[J]. 原子能科学技术, 2012, 46(9): 1097-1106. DOI: 10.7538/yzk.2012.46.09.1097
XU Zhi-hong, FU Sheng-wei, LIU Xiao-jing, YANG Yan-hua, CHENG Xu. Loss of Flow Accident Mitigation Measures of Mixed Spectrum SCWR-M[J]. Atomic Energy Science and Technology, 2012, 46(9): 1097-1106. DOI: 10.7538/yzk.2012.46.09.1097
Citation: XU Zhi-hong, FU Sheng-wei, LIU Xiao-jing, YANG Yan-hua, CHENG Xu. Loss of Flow Accident Mitigation Measures of Mixed Spectrum SCWR-M[J]. Atomic Energy Science and Technology, 2012, 46(9): 1097-1106. DOI: 10.7538/yzk.2012.46.09.1097

混合能谱超临界水堆失流事故缓解措施研究

Loss of Flow Accident Mitigation Measures of Mixed Spectrum SCWR-M

  • 摘要: 使用改进的系统程序RELAP5建立了一个混合能谱超临界水堆(SCWR-M)模型。为研究混合能谱超临界水堆失流事故特性,以获取缓解混合能谱超临界水堆失流事故的措施,选取反应堆冷却剂泵惰转时间、压力容器上部储水空间容积和安注流量作为主要参数进行分析。研究表明,混合能谱超临界水堆系统的设计是可行的。反应堆冷却剂泵惰转15 s,压力容器上部水空间容积大于27 m3,以及安注流量高于系统满功率稳态流量的5%是缓解混合能谱超临界水堆失流事故的主要措施。

     

    Abstract: Based on a revised version of RELAP5, a model of the mixed spectrum supercritical water cooled reactor (SCWR-M) system was established. Some important parameters were chosen as the main parameters to analysis the transient behaviour of SCWR-M and fix mitigation measures during loss of flow accident (LOFA). Reactor coolant pump (RCP) coast-down time of more than 15 seconds, RPV upper water volume of more than 27 m3, and safety injection of more than 5% of the system design flow are the main mitigation measures for the LOFA of SCWR-M.

     

  • [1] OKA Y. Review of high temperature water and steam cooled reactor concepts[C]∥Proc. of SCR-2000. Tokyo: [s. n.], 2000.
    [2] CHENG X, LIU X J, YANG Y H. A mixed core for supercritical water-cooled reactor[J]. Nucl Eng Technol, 2008, 40(2): 117-126.
    [3] 程旭,刘晓晶. 混合能谱超临界水堆堆芯设计分析[J]. 核科学与工程,2009,29(1):43-49.CHENG Xu, LIU Xiaojing. A mixed core for supercritical water-cooled reactors[J]. Chin J Nucl Sci Eng, 2009, 29(1): 43-49(in Chinese).
    [4] LIU X J, YANG T, CHENG X. Core and sub-channel analysis of SCWR with mixed spectrum core[J]. Annals of Nuclear Energy, 2010, 37: 1674-1682.
    [5] LIU X J, CHENG X. Steady-state thermal hydraulic analysis of SCWR assembly[J]. Frontiers of Energy and Power Engineering in China, 2008, 2(4): 475-478.
    [6] LIU X J, CHENG X. Coupled thermal-hydraulics and neutron-physics analysis of SCWR with mixed spectrum core[J]. Progress in Nuclear Energy, 2010, 52(7): 640-647.
    [7] LIU X J, CHENG X. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly[J]. Annals of Nuclear Energy, 2009, 36: 28-36.
    [8] LIU X J, CHENG X. Core and sub-channel evaluation of a thermal SCWR[J]. Nuclear Engineering and Technology, 2009, 41(5): 677-690.
    [9] ISHIWATARI Y. Safety of super LWR: Ⅱ. Safety analysis at supercritical pressure[J]. J Nucl Sci Technol, 2005, 42(11): 935-948.
    [10] MacDONALD P E. Supercritical water reactor (SCWR): Progress report for the FY-03 generation-Ⅳ R&D activities for the development of the SCWR in the U. S., INEEL/EXT-03-01210[R]. US: INEEL, 2003.
    [11] The RELAP5 Code Development Team. RELAP5/MOD3.2 code manual[M]. USA: Idaho National Engineering Laboratory, 1995.
    [12] BISHOP A A, SANDBERG R O, TONG L S. Forced convection heat transfer to water at near-critical temperatures and supercritical pressures, Report WCAP-2056 Part Ⅳ[R]. Pittsburgh: Westinghouse Electric Corporation, 1964.
    [13] PETUKHOV B S, KURGANOV V A. Heat transfer and flow resistance in the turbulent pipe flow of a fluid with near-critical state parameters[J]. Teplofizika Vysokikh Temperature, 1983, 21(1): 92-100.
    [14] 周翀,刘晓晶,杨燕华,等. ATHLET-SC程序的开发及适用性分析[J]. 原子能科学技术,2009,43(6):556-560.ZHOU Chong, LIU Xiaojing, YANG Yanhua, et al. Development and applicability analysis of ATHLET-SC code[J]. Atomic Energy Science and Technology, 2009, 43(6): 556-560(in Chinese).
    [15] UK AP1000 probabilistic risk assessment[R]. USA: Westinghouse Electric Company, 2007.
    [16] KYOO H B. Design options for the safety injection system of Korean next generation reactor[J]. Annals of Nuclear Energy, 2000, 27(11): 1011-1028.
    [17] TAHIR M. Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1[J]. Annals of Nuclear Energy, 2008, 35(11): 1986-1993.
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  • 刊出日期:  2012-09-19

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