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CMRR堆内高温高压辐照考验回路典型事故分析

胡泊 郭斯茂 王冠博 钱达志 郭玉川 余恒

胡泊, 郭斯茂, 王冠博, 等. CMRR堆内高温高压辐照考验回路典型事故分析[J]. 强激光与粒子束, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023
引用本文: 胡泊, 郭斯茂, 王冠博, 等. CMRR堆内高温高压辐照考验回路典型事故分析[J]. 强激光与粒子束, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023
Hu Bo, Guo Simao, Wang Guanbo, et al. Simulation of typical accidents in fuel test loop of CMRR[J]. High Power Laser and Particle Beams, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023
Citation: Hu Bo, Guo Simao, Wang Guanbo, et al. Simulation of typical accidents in fuel test loop of CMRR[J]. High Power Laser and Particle Beams, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023

CMRR堆内高温高压辐照考验回路典型事故分析

doi: 10.11884/HPLPB201931.190023
基金项目: 

国家自然科学基金青年科学项目基金项目 11605173

详细信息
    作者简介:

    胡泊(1992—),男,硕士,研究方向为反应堆热工计算及安全分析; 350516326@qq.com

    通讯作者:

    钱达志(1968—),男,研究员,研究方向为核能科学与工程; qdz1968@vip.sina.com

  • 中图分类号: TL364+.4

Simulation of typical accidents in fuel test loop of CMRR

  • 摘要: 基于中国绵阳研究堆(CMRR)高温高压辐照考验回路初步设计方案,就回路失水事故(LOCA)及失流事故(LOFA)两类典型事故进行分析。结果表明:回路在冷管段及热管段失水事故下包壳热点温度最高为880.6 ℃及367.6 ℃,均远低于1204 ℃;全部失流事故下最小偏离泡核沸腾比(MDNBR)大于1.5,不会发生偏离泡核沸腾;卡轴事故中包壳最高温度为734.1 ℃,低于1482 ℃。上述结果均满足验收准则,符合安全法规要求。
  • 图  1  高温高压试验回路设计方案

    Figure  1.  Outline of the high-temperature high-press fuel test loop

    图  2  辐照装置俯视图

    Figure  2.  Top view of irradiation facility

    图  3  试验回路RELAP5节点图

    Figure  3.  RELAP5 nodalization of fuel test loop (FTL) main loop

    图  4  失水事故破口流量

    Figure  4.  Flow rate through the break during loss-of-coolant accident

    图  5  失水事故稳压器压力

    Figure  5.  Pressure in the main loop during loss-of-coolant accident

    图  6  失水事故辐照装置水位

    Figure  6.  Coolant level in the irradiation facility during loss-of-coolant accident

    图  7  失水事故包壳热点温度

    Figure  7.  Cladding temperature during loss-of-coolant accident

    图  8  失水事故氩气间隙压力

    Figure  8.  Pressure in the argon gap during loss-of-coolant accident

    图  9  失水事故CMRR冷却水温

    Figure  9.  Temperature of China's Mianyang Research Reactor coolant during loss-of-coolant accident

    图  10  卡轴事故与失流事故主回路冷却剂流量

    Figure  10.  Flow rate through the irradiation facility during loss-of-flow accidents (LOFAs)

    图  11  卡轴事故与失流事故回路功率

    Figure  11.  Power during clamp shaft accident and loss-of-flow accidents

    图  12  卡轴事故与失流事故主回路压力

    Figure  12.  Pressure in the main loop during clamp shaft accident and loss-of-flow accidents

    图  13  失流事故的偏离泡核沸腾比

    Figure  13.  DNBR during the total LOFAs

    图  14  卡轴事故包壳热点温度

    Figure  14.  Cladding temperature during the clamp shaft accident

    表  1  主回路主要设计参数

    Table  1.   Designed value of key parameters of primary loop

    power/kW primary loop pressure/MPa argon gap pressure/MPa coolant flow rate/(kg·s-1) inlet coolant temperature/K outlet coolant temperature/K
    60 15.5 0.165 3.5 553 583
    下载: 导出CSV

    表  2  系统稳态参数

    Table  2.   Key parameters in steady-state

    power/ kW pressure/ MPa inlet coolant temperature/℃ outlet coolant temperature/℃ main loop coolant flow rate/(kg·s-1) argon gap pressure/ MPa cladding temperature/ K
    design value 600 15.500 280.00 310.00 3.500 0.165 630.00
    RELAP5 calculation value 600 15.556 278.89 310.41 3.476 0.175 640.63
    下载: 导出CSV
  • [1] 程作用, 张劲松, 李忠宪, 等. 500 kW考验回路水中F-和Cl-的低压离子色谱分析[J]. 核动力工程, 1996, 17(5): 477-480. https://www.cnki.com.cn/Article/CJFDTOTAL-HDLG605.016.htm

    Cheng Zuoyong, Zhang Jingsong, Li Zhongxian, et al. Analysis of F- and Cl- ions in the primary circuit water of 500 kW test loop by low pressure ion chromatography. Nuclear Power Engineering, 1996, 17(5): 477-480 https://www.cnki.com.cn/Article/CJFDTOTAL-HDLG605.016.htm
    [2] Hadjam A, Souidi F, Loubar A, et al. Simulation of a LBLOCA in the CALLISTO test facility using the best estimate computer code RELAP5/SCDAP3.2[J]. Nuclear Engineering and Design, 2013, 262: 153-167. doi: 10.1016/j.nucengdes.2013.03.052
    [3] Choo K N, Cho M S, Kim B G, et al. Material irradiation at Hanaro, Korea[C]//Research Reactor application for Materials under High Neutron Fluence. 2011: 50-51.
    [4] 张毅. CARR高温高压试验回路事故分析[D]. 北京: 中国原子能科学研究院, 2007: 23-28.

    Zhang Yi. Accidents analysis of high temperature and high pressure test loop of CARR. Beijing: China Institute of Atomic Energy, 2007: 23-28
    [5] Bae H, Kim D E, Yi S J, et al. Comparison of three SBLOCA tests with different break locations using the SMART-ITL facility to estimate the safety of the SMART design[J]. Nuclear Engineering and Technology, 2017, 49(5): 968-978. doi: 10.1016/j.net.2017.04.006
    [6] 于平安, 朱瑞安, 喻真烷, 等. 核反应堆热工分析[M]. 西安: 西安交通大学出版社, 1979: 268-269.

    Yu Ping'an, Zhu Ruian, Yu Zhenwan, et al. Thermal analysis of nuclear reactors. Xi'an: Xi'an Jiaotong University Press, 1979: 268-269
    [7] 张学学. 热工基础[M]. 北京: 高等教育出版社, 2006.

    Zhang Xuexue. Thermal engineering foundation. Beijing: Higher Education Press, 2006
    [8] The RELAP5-3D Code Development Team. RELAP5-3D Code Manual Volume1-5[M]. Idaho: INEEL, 2005.
    [9] 朱继洲, 奚树人, 单建强, 等. 核反应堆安全分析[M]. 西安: 西安交通大学出版社, 2004: 96-102.

    Zhu Jizhou, Xi Shuren, Shan Jianqiang, et al. Safety analysis of nuclear reactors. Xi'an: Xi'an Jiaotong University Press, 2004: 96-102
    [10] 余冀阳, 余尔俊. 核电厂事故分析[M]. 北京: 清华大学出版社, 2012: 23-24.

    Yu Jiyang, Yu Erjun. Accidents analysis of nuclear plants. Beijing: Tsinghua University Press, 2012: 23-24
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出版历程
  • 收稿日期:  2019-01-24
  • 修回日期:  2019-05-24
  • 刊出日期:  2019-09-15

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