核电厂停堆工矿下事故及其处置研究
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摘要
在核电厂的安全研究中，通常着重研究功率运行下的瞬态和事故，而对停堆工况下
的安全问题研究较少。近年来，核电厂停堆工况下发生了儿起典型事故，这些事故具有
可能导致堆芯损坏的严重后果的特性，这引起了人们对此类事故的重视和研究。
本文主要围绕核电厂停堆工况下的安全问题开展研究，采用确定论的方法对我国秦
山核电厂停堆工况下的事故及其处置进行了分析研究，并提出了事故处置措施。分析了
两类事故：余热排出系统失效事故和意外硼稀释事故。
选定秦山核电厂为计算对象，假定电厂处于环路半充水的冷停堆工况，采用热工水
力系统瞬态分析程序RELAPS/MOD2，对失去余热排出系统后电厂的热工水力响应进行
模拟计算（主要是堆芯和一回路的响应）。计算结果表明，在丧失余热排出系统冷却后大
约1040秒，堆芯上部即开始裸露，大约3380秒堆芯温度快速上升，如果没有缓解措施，
将导致堆芯损坏；在执行缓解措施的情况下，假定在30分钟启动投入安注，此后堆芯水
位能及时得到恢复，堆芯温度陡升的现象不再发生，从而可避免堆芯的损坏。即操纵员
可以有30分钟的时间米采取措施缓解事故。
对于意外硼稀释事故，应用RELAPS/MOD2程序计算了热停堆工况下的快速硼稀释
事故。计算结果表明，该事故进程非常迅速，事故后果也很严重。在低硼浓度水团己经
形成的情况下，主泵再启动后大约10秒左右，该水团将穿过堆芯，使堆芯反应性迅速增
加，如果堆芯达到临界，堆芯功率将快速增长，并可能造成燃料元件的损坏。根据不同
的硼浓度，计算得到的临界水团体积从2m～3到5m～3不等。
对上述两种典型事故的分析结果及结论可实际应用于秦山核电厂停堆工况下事故管
理中，同时对我国其他核电厂的停堆工况下事故管理也具有指导意义。
The general concern for the safety analysis of PWR power plant is mainly on the study of the transients and accidents at power operation, but the safety issue of plant at shutdown condition was seldom studied. In recent years, several accidents occurred when the PWR power plant was operated at shutdown conditions, and some of these accidents had the potential consequences to cause the core damage, and this has attracted people's attention.
This thesis initiates a study focused on safety issues of nuclear power plant at shutdown conditions. And a deterministic methodology is used to perform a study on shutdown accident of Qinshan Nuclear Power Plant (QNPP). The details of this research include safety anaIysis about two accidents: Loss of Residual Heat Removal Accidents, and Inadvertent Boron Dilution Accidents.
Selecting the Qinshan Nuclear Power plant as the reference plant, assuming that the pIant operates at cold shutdown and mid-loop operation mode, the thermal-hydraulic response of the plant after the RHR system failed has been evaluated with the thermal-hydraulic transient analysis code RELAP5/MOD2. The evaluating results show that the up part of the core will start to be uncovered at l7 minutes after lost of RHR, then at about 1 hr the temperature of the core will rise rapidly. If no measures have been taken to mitigate the accident, the core would be damaged. If safety injection into the cold loop is activated at l800 seconds following' the accident, the core water level can be restored, and the rapid rising of the core temperature would not occur, so the core damage can be avoided. This gives the plant operator about 30 minutes to take action to mitigate the accident.
RELAP5/MOD2 code also has been used to simu1ate the inadvertent boron dilution accident at hot shutdown operation. The results show that the accident's process is very fast and. it's sequence is hazardous. If the low concentration boron water package has formed, the water package will go through the core after l0 seconds of the primary pump restart. This will make the core reactivity increase abruptly, and the core power will increase rapidly too. lf core reach criticism, it would have damages to the fuel elements. The critical volume of low concentration boron water package varies from 2 to 5 stere according to different boron concentration.
The results and conclusions from this research can be applied to the accident management at shutdown mode for QNPP, and they are also of significance for the accident management at shutdown mode for other domestic NPPs.
的安全问题研究较少。近年来，核电厂停堆工况下发生了儿起典型事故，这些事故具有
可能导致堆芯损坏的严重后果的特性，这引起了人们对此类事故的重视和研究。
本文主要围绕核电厂停堆工况下的安全问题开展研究，采用确定论的方法对我国秦
山核电厂停堆工况下的事故及其处置进行了分析研究，并提出了事故处置措施。分析了
两类事故：余热排出系统失效事故和意外硼稀释事故。
选定秦山核电厂为计算对象，假定电厂处于环路半充水的冷停堆工况，采用热工水
力系统瞬态分析程序RELAPS/MOD2，对失去余热排出系统后电厂的热工水力响应进行
模拟计算（主要是堆芯和一回路的响应）。计算结果表明，在丧失余热排出系统冷却后大
约1040秒，堆芯上部即开始裸露，大约3380秒堆芯温度快速上升，如果没有缓解措施，
将导致堆芯损坏；在执行缓解措施的情况下，假定在30分钟启动投入安注，此后堆芯水
位能及时得到恢复，堆芯温度陡升的现象不再发生，从而可避免堆芯的损坏。即操纵员
可以有30分钟的时间米采取措施缓解事故。
对于意外硼稀释事故，应用RELAPS/MOD2程序计算了热停堆工况下的快速硼稀释
事故。计算结果表明，该事故进程非常迅速，事故后果也很严重。在低硼浓度水团己经
形成的情况下，主泵再启动后大约10秒左右，该水团将穿过堆芯，使堆芯反应性迅速增
加，如果堆芯达到临界，堆芯功率将快速增长，并可能造成燃料元件的损坏。根据不同
的硼浓度，计算得到的临界水团体积从2m～3到5m～3不等。
对上述两种典型事故的分析结果及结论可实际应用于秦山核电厂停堆工况下事故管
理中，同时对我国其他核电厂的停堆工况下事故管理也具有指导意义。
The general concern for the safety analysis of PWR power plant is mainly on the study of the transients and accidents at power operation, but the safety issue of plant at shutdown condition was seldom studied. In recent years, several accidents occurred when the PWR power plant was operated at shutdown conditions, and some of these accidents had the potential consequences to cause the core damage, and this has attracted people's attention.
This thesis initiates a study focused on safety issues of nuclear power plant at shutdown conditions. And a deterministic methodology is used to perform a study on shutdown accident of Qinshan Nuclear Power Plant (QNPP). The details of this research include safety anaIysis about two accidents: Loss of Residual Heat Removal Accidents, and Inadvertent Boron Dilution Accidents.
Selecting the Qinshan Nuclear Power plant as the reference plant, assuming that the pIant operates at cold shutdown and mid-loop operation mode, the thermal-hydraulic response of the plant after the RHR system failed has been evaluated with the thermal-hydraulic transient analysis code RELAP5/MOD2. The evaluating results show that the up part of the core will start to be uncovered at l7 minutes after lost of RHR, then at about 1 hr the temperature of the core will rise rapidly. If no measures have been taken to mitigate the accident, the core would be damaged. If safety injection into the cold loop is activated at l800 seconds following' the accident, the core water level can be restored, and the rapid rising of the core temperature would not occur, so the core damage can be avoided. This gives the plant operator about 30 minutes to take action to mitigate the accident.
RELAP5/MOD2 code also has been used to simu1ate the inadvertent boron dilution accident at hot shutdown operation. The results show that the accident's process is very fast and. it's sequence is hazardous. If the low concentration boron water package has formed, the water package will go through the core after l0 seconds of the primary pump restart. This will make the core reactivity increase abruptly, and the core power will increase rapidly too. lf core reach criticism, it would have damages to the fuel elements. The critical volume of low concentration boron water package varies from 2 to 5 stere according to different boron concentration.
The results and conclusions from this research can be applied to the accident management at shutdown mode for QNPP, and they are also of significance for the accident management at shutdown mode for other domestic NPPs.
引文
[1] T-L Chu, W.T.Pratt, "Risk Contribution from Low Power and Shutdown of a Pressurized Water Reactor", The Fifth International Topical Meeting on Nuclear Thermal-Hydraulic, Operation and Safety(NUTHOS-5) , April 14-18,1997, Beijing, China
[2] International Atomic Energy Agency, Summary Report on the Post-Accident Review, Meeting on the Chernobyl Accident, Safety Series No. 75-INSAG-1, Vienna(1986)
[3] 熊本和，“核电厂停堆工况下的安全问题”，核科学与工程，1995,15(2) :97-100
[4] U.S.NRC, "Loss of Residual Heat Removal System, Diablo Cayon, Unit 2, April 10,1987", NUREG-1269, June 1987
[5] CEA, IPSN, "A Probabilistic Safety Assessment of the Standard French 900Mwe Pressurized Water Reactor", EPS900, April 1990
[6] L.W. Ward, et al., "Consequences of the Loss of the Residual Heat Removal System in Pressurized Water Reactors," NUREG/CR-5820, Brookhaven National Laboratory, May 1992
[7] U.S. Nuclear Regulatory Commission, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, NUREG-1449 (Feb. 1992)
[8] U.S.NRC, "Loss of Vital AC Power and the Residual heat Removal System During Mid-loop Operation at Vogtle Unit 1 on March 20,1990," NUREG-1410, June 1990
[9] D.J. Diamond, et al., "Probability and Consequences of Rapid Boron Dilution in Pressurized Water Reactors," NUREG/CR-5 819, Brookhaven National Laboratory
[10] LEONARD W WARD, "Evaluation of the Loss of Residual Heat Removal Systems in PWRs with U-tube Steam Generators," Nuclear Technology, V.100, P25-38, OCT 1992
[11] Y. Narumiya, et al., "Thermal hydraulic analysis of RHR cooling during shutdown operation in PWR", Eighth International Topic Meeting on Nuclear Reactor Thermal-hydraulic, Kyoto, Japan, Sep-Oct, Proceedings Volume 3
[12] V. Hoensch, "Points of view on shutdown cooling events among the members of the WANO-Paries Centre," Proceedings of a Technical Committee meeting held in Stockholm, 30 Nov.-3 Dec., 1992:PSA for the shutdown mode for nuclear power plants, IAEA-TECDOC-751 (June 1994)
[13] T.S. Andreychek, et al., "Loss of RHRS Cooling While the RCS is Partially Filled," WCAP-11916, Westinghouse Electric Corporation July 1998
[14] Miraglia, F.J., "Loss of Residual Heat Removal(RHR) while the Reactor Coolant System(RCS) is Partially Filled," Generic Letter 87-12, Letter to all licensees of operating PWRs and holders of construction permits for PWRs, July 9,1987．
[15] "Bethsy Isp-38 Parametric Analysis and Influences of the Time Step", Transactions of the ANS, Volume 77, p446,1997
[16] Z.MUSICKI, et al., "Evaluation of Potential Severe Accidents During low Power and Shutdown Operations at Surry, Unit 1, Analysis of Core Damage Frequency from Internal Fires During Mid-Loop Operations", NUREG/CR-6144, Volume 3, July, 1994
[17] Donald T. Chung, James F. Meyer, "Consideration of Accident Management During Shutdown", Safety of Thermal Reactors Proceedings, July 1991, T1431． 064-53 I61
[18] A.A. Troshko, Y.A. Hassan, "Simulation of the Loss of Residual Heat Removal System of an Integral Test Facility Using French Computer Code CATHARE", The Fifth International Topical Meeting on Nuclear Thermal-Hydraulic, Operation and Safety(NUTHOS-5) , April 14-18,1997, Beijing, China
[19] 熊本和，“核电厂硼酸水失控稀释事故研究”，国家核安全局
[20] T. Chatain, et al., "Code Analysis of Multidimensional Phenomena in a ROSA-IV/LSTF Experiment Simulating Loss of Residual Heat Removal Event During PWR Mid-loop Operation", ASME/JSME Nuclear Engineering Conference-Volume 1 ASME1993
[21] Y. Hansan, et al., "Relap5/Mod3 Simulation of the Loss of the RHRS During a Mid-Loop Operation Experiment Conducted at the/Rosa-Ⅳ Large Scale Test Facility, Nuclear Technology, V103, p310, Sep, 1993
[22] “核电站停堆工况PSA研究”，柯国土，博士论文，中国原子能科学研究院，1998年
[23] “停堆期间环路半充水(MID-LOOP)运行丧失余热排出热工水力分析”，阚强生等，核动力传热和流体力学，中国核能动力学会反应堆热工流体专业委员会，1997年
[24] "Effect of Flushing Boron During a Boiling Water Reactor Anticipated Transient without Scram", Nuclear Technology, V95, p162, Aug, 1991
[25] U.S.NRC, "Inadvertent Boron Dilution Events", Generic Letter 85-05
[26] U.S,NRC, "Boron Dilution of Reactor Coolant During Steam Generator Decontamination", IE Information Notice 80-34
[27] U.S.NRC, "Foreign Experience Regarding Boron Dilution", Information Notice 91-54
[28] DRAGAN MIRKOVIC and DAVID J. DIAMOND, "Effect of Flushing Boron During a Boiling Water Reactor Without Scram," Nuclear Technology, V.9.5, P162-174, AUG 1991
[29] "Loss of Decay Heat Removal Due to Loss of Fluid Levels in Reactor Coolant System." NRC Information Notice No. 86-101
[30] RELAP5/MOD2 Code Manual
[31] “秦山核电厂SGTR事故及其处置研究”，李吉根，博士论文，中国原子能科学研究院，1991
[32] “秦山核电厂30万千瓦机组参数汇总”，上海核工程研究设计院，1987年9月
[33] “秦山核电厂30万千瓦机组系统说明”，上海核工程研究设计院，1987年9月
[34] “秦山核电厂最终安全分析报告”，第15章，事故分析，上海核工程研究设计院
[35] “秦山核电厂最终安全分析报告”，第16章，技术规格书，上海核工程研究设计院
[36] “秦山核电厂事故分析数据汇编”，郝老迷，中国原子能科学研究院内部资料，1988年
[37] “压水堆核电厂系统与设备”，浦胜娣，核工业研究生部
[38] “RELAP5／MOD2反应性反馈计算模型的改进”，孙礼亚，RELAP5程序的开发研究专题报告，中国原子能科学研究院，1990年6月
[39] “RELAP5／MOD2的稳态初始化方法”，李吉根，RELAP5程序的开发研究专题报告，中国原子能科学研究院，1990年6月
[40] “核电厂事故分析”，俞尔俊，核工业研究生部，1991年6月
[41] “压水堆核电厂安全与事故对策”，濮继龙，原子能出版社，1995年
[42] “核反应堆安全分析”，朱继洲主编，原子能出版社，1988年
[43] “核反应堆热工分析”，于平安等编著，原子能出版社，1986年
[2] International Atomic Energy Agency, Summary Report on the Post-Accident Review, Meeting on the Chernobyl Accident, Safety Series No. 75-INSAG-1, Vienna(1986)
[3] 熊本和，“核电厂停堆工况下的安全问题”，核科学与工程，1995,15(2) :97-100
[4] U.S.NRC, "Loss of Residual Heat Removal System, Diablo Cayon, Unit 2, April 10,1987", NUREG-1269, June 1987
[5] CEA, IPSN, "A Probabilistic Safety Assessment of the Standard French 900Mwe Pressurized Water Reactor", EPS900, April 1990
[6] L.W. Ward, et al., "Consequences of the Loss of the Residual Heat Removal System in Pressurized Water Reactors," NUREG/CR-5820, Brookhaven National Laboratory, May 1992
[7] U.S. Nuclear Regulatory Commission, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, NUREG-1449 (Feb. 1992)
[8] U.S.NRC, "Loss of Vital AC Power and the Residual heat Removal System During Mid-loop Operation at Vogtle Unit 1 on March 20,1990," NUREG-1410, June 1990
[9] D.J. Diamond, et al., "Probability and Consequences of Rapid Boron Dilution in Pressurized Water Reactors," NUREG/CR-5 819, Brookhaven National Laboratory
[10] LEONARD W WARD, "Evaluation of the Loss of Residual Heat Removal Systems in PWRs with U-tube Steam Generators," Nuclear Technology, V.100, P25-38, OCT 1992
[11] Y. Narumiya, et al., "Thermal hydraulic analysis of RHR cooling during shutdown operation in PWR", Eighth International Topic Meeting on Nuclear Reactor Thermal-hydraulic, Kyoto, Japan, Sep-Oct, Proceedings Volume 3
[12] V. Hoensch, "Points of view on shutdown cooling events among the members of the WANO-Paries Centre," Proceedings of a Technical Committee meeting held in Stockholm, 30 Nov.-3 Dec., 1992:PSA for the shutdown mode for nuclear power plants, IAEA-TECDOC-751 (June 1994)
[13] T.S. Andreychek, et al., "Loss of RHRS Cooling While the RCS is Partially Filled," WCAP-11916, Westinghouse Electric Corporation July 1998
[14] Miraglia, F.J., "Loss of Residual Heat Removal(RHR) while the Reactor Coolant System(RCS) is Partially Filled," Generic Letter 87-12, Letter to all licensees of operating PWRs and holders of construction permits for PWRs, July 9,1987．
[15] "Bethsy Isp-38 Parametric Analysis and Influences of the Time Step", Transactions of the ANS, Volume 77, p446,1997
[16] Z.MUSICKI, et al., "Evaluation of Potential Severe Accidents During low Power and Shutdown Operations at Surry, Unit 1, Analysis of Core Damage Frequency from Internal Fires During Mid-Loop Operations", NUREG/CR-6144, Volume 3, July, 1994
[17] Donald T. Chung, James F. Meyer, "Consideration of Accident Management During Shutdown", Safety of Thermal Reactors Proceedings, July 1991, T1431． 064-53 I61
[18] A.A. Troshko, Y.A. Hassan, "Simulation of the Loss of Residual Heat Removal System of an Integral Test Facility Using French Computer Code CATHARE", The Fifth International Topical Meeting on Nuclear Thermal-Hydraulic, Operation and Safety(NUTHOS-5) , April 14-18,1997, Beijing, China
[19] 熊本和，“核电厂硼酸水失控稀释事故研究”，国家核安全局
[20] T. Chatain, et al., "Code Analysis of Multidimensional Phenomena in a ROSA-IV/LSTF Experiment Simulating Loss of Residual Heat Removal Event During PWR Mid-loop Operation", ASME/JSME Nuclear Engineering Conference-Volume 1 ASME1993
[21] Y. Hansan, et al., "Relap5/Mod3 Simulation of the Loss of the RHRS During a Mid-Loop Operation Experiment Conducted at the/Rosa-Ⅳ Large Scale Test Facility, Nuclear Technology, V103, p310, Sep, 1993
[22] “核电站停堆工况PSA研究”，柯国土，博士论文，中国原子能科学研究院，1998年
[23] “停堆期间环路半充水(MID-LOOP)运行丧失余热排出热工水力分析”，阚强生等，核动力传热和流体力学，中国核能动力学会反应堆热工流体专业委员会，1997年
[24] "Effect of Flushing Boron During a Boiling Water Reactor Anticipated Transient without Scram", Nuclear Technology, V95, p162, Aug, 1991
[25] U.S.NRC, "Inadvertent Boron Dilution Events", Generic Letter 85-05
[26] U.S,NRC, "Boron Dilution of Reactor Coolant During Steam Generator Decontamination", IE Information Notice 80-34
[27] U.S.NRC, "Foreign Experience Regarding Boron Dilution", Information Notice 91-54
[28] DRAGAN MIRKOVIC and DAVID J. DIAMOND, "Effect of Flushing Boron During a Boiling Water Reactor Without Scram," Nuclear Technology, V.9.5, P162-174, AUG 1991
[29] "Loss of Decay Heat Removal Due to Loss of Fluid Levels in Reactor Coolant System." NRC Information Notice No. 86-101
[30] RELAP5/MOD2 Code Manual
[31] “秦山核电厂SGTR事故及其处置研究”，李吉根，博士论文，中国原子能科学研究院，1991
[32] “秦山核电厂30万千瓦机组参数汇总”，上海核工程研究设计院，1987年9月
[33] “秦山核电厂30万千瓦机组系统说明”，上海核工程研究设计院，1987年9月
[34] “秦山核电厂最终安全分析报告”，第15章，事故分析，上海核工程研究设计院
[35] “秦山核电厂最终安全分析报告”，第16章，技术规格书，上海核工程研究设计院
[36] “秦山核电厂事故分析数据汇编”，郝老迷，中国原子能科学研究院内部资料，1988年
[37] “压水堆核电厂系统与设备”，浦胜娣，核工业研究生部
[38] “RELAP5／MOD2反应性反馈计算模型的改进”，孙礼亚，RELAP5程序的开发研究专题报告，中国原子能科学研究院，1990年6月
[39] “RELAP5／MOD2的稳态初始化方法”，李吉根，RELAP5程序的开发研究专题报告，中国原子能科学研究院，1990年6月
[40] “核电厂事故分析”，俞尔俊，核工业研究生部，1991年6月
[41] “压水堆核电厂安全与事故对策”，濮继龙，原子能出版社，1995年
[42] “核反应堆安全分析”，朱继洲主编，原子能出版社，1988年
[43] “核反应堆热工分析”，于平安等编著，原子能出版社，1986年