用ICARE/CATHARE分析百万千瓦级PWR严重事故堆芯行为
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摘要
当前我国正积极发展核电，随着核电厂数量的增加，核电安全问题就显得越来越重要，而安全分析技术(DBA以及严重事故)是确保核电安全的技术基础。本文成功的在Linux系统下安装了ICARE／CATHARE程序，并建立了900MWe核电厂的计算模型。分析了该核电厂大破口失水加全厂断电事故序列和小破口失水加全厂断电事故序列的堆芯破坏过程，并研究了小破口失水加全厂断电事故序列的缓解措施及其有效性。
900MWe核电厂大破口失水加全厂断电事故序列的分析表明：500s时开始锆水反应，1700s时堆芯内的燃料开始损坏，2400s堆芯上部全部损坏、坍塌并产生了大约270kg的氢气。对破口位置的敏感性分析表明，冷端大破口失水事故要比热端大破口失水事故进展更快、后果也更严重。
小破口失水加全厂断电事故序列的分析结果表明小破口失水事故进程比大破口失水进程相对较慢。在大约2400s时开始堆芯裸露，13000s左右堆芯开始出现损坏，15000s计算结束。在小破口失水事故分析中还对几种事故缓解措施进行了分析，结果表明，如果能在事故开始后半小时开始注水，堆芯水位将很快恢复，能在堆芯建立冷却，并阻止堆芯损坏；如果在事故后堆芯出口蒸汽温度超过650℃(约2700s)时开始注水，堆芯内只有少量包壳发生氧化，但随后堆芯水位很快恢复，堆芯温度降低，也可以保持堆芯的完整性。
严重事故分析有助于更加详细地了解堆芯破坏过程中所发生的物理、化学、力学行为、热工水力行为等现象。为研究事故的预防和缓解措施，给出最佳的干预时间和干预方式提供技术基础。本文应用ICARE／CATHARE程序研究900MWe核电厂的堆芯破坏机理，该工作是国内的首次成功尝试，对今后该程序的进一步推广应用以及中法继续在严重事故领域的合作都具有重要的促进作用。
The safety of nuclear power plant (NPP) should be paid much attention due to an ambitious development policy in China. Severe accident analysis is one of the essential techniques to assure safety of NPP. The report describes an application of ICARE/CATHARE code to a 900MWe PWR, including the main aspects as follows:
Installation of ICARE/CATHARE code under Linux operating system;
Modeling of 900MWe PWR;
Analyzing accident sequences of LBLOCA and SBLOCA with station blackout, including investigation of the mitigation measures under SBLOCA.
The analysis of LBLOCA shows that oxidation of Zr begins at about 500s. The core begins to damage at 1700 seconds and the upper part of the core is observed full damage at 2400s while 270kg hydrogen is generated. A sensitivity study shows that the process of the accident with a cold leg break is faster than that with a hot leg break. The analysis shows that the progress of SBLOCA with station blackout sequence is slower than that of LBLOCA. The core uncovers at about 2400s and begins to damage at 13000s. The analysis shows that the core keeps intact if the inject system could put into operation at about half an hour after initialization of the accident. Whereas, only slight oxidation is found at upper of the core if the inject system could put into operation when the vapor temperature of the core outlet exceeds 650℃(about 2700 seconds).
900MWe核电厂大破口失水加全厂断电事故序列的分析表明：500s时开始锆水反应，1700s时堆芯内的燃料开始损坏，2400s堆芯上部全部损坏、坍塌并产生了大约270kg的氢气。对破口位置的敏感性分析表明，冷端大破口失水事故要比热端大破口失水事故进展更快、后果也更严重。
小破口失水加全厂断电事故序列的分析结果表明小破口失水事故进程比大破口失水进程相对较慢。在大约2400s时开始堆芯裸露，13000s左右堆芯开始出现损坏，15000s计算结束。在小破口失水事故分析中还对几种事故缓解措施进行了分析，结果表明，如果能在事故开始后半小时开始注水，堆芯水位将很快恢复，能在堆芯建立冷却，并阻止堆芯损坏；如果在事故后堆芯出口蒸汽温度超过650℃(约2700s)时开始注水，堆芯内只有少量包壳发生氧化，但随后堆芯水位很快恢复，堆芯温度降低，也可以保持堆芯的完整性。
严重事故分析有助于更加详细地了解堆芯破坏过程中所发生的物理、化学、力学行为、热工水力行为等现象。为研究事故的预防和缓解措施，给出最佳的干预时间和干预方式提供技术基础。本文应用ICARE／CATHARE程序研究900MWe核电厂的堆芯破坏机理，该工作是国内的首次成功尝试，对今后该程序的进一步推广应用以及中法继续在严重事故领域的合作都具有重要的促进作用。
The safety of nuclear power plant (NPP) should be paid much attention due to an ambitious development policy in China. Severe accident analysis is one of the essential techniques to assure safety of NPP. The report describes an application of ICARE/CATHARE code to a 900MWe PWR, including the main aspects as follows:
Installation of ICARE/CATHARE code under Linux operating system;
Modeling of 900MWe PWR;
Analyzing accident sequences of LBLOCA and SBLOCA with station blackout, including investigation of the mitigation measures under SBLOCA.
The analysis of LBLOCA shows that oxidation of Zr begins at about 500s. The core begins to damage at 1700 seconds and the upper part of the core is observed full damage at 2400s while 270kg hydrogen is generated. A sensitivity study shows that the process of the accident with a cold leg break is faster than that with a hot leg break. The analysis shows that the progress of SBLOCA with station blackout sequence is slower than that of LBLOCA. The core uncovers at about 2400s and begins to damage at 13000s. The analysis shows that the core keeps intact if the inject system could put into operation at about half an hour after initialization of the accident. Whereas, only slight oxidation is found at upper of the core if the inject system could put into operation when the vapor temperature of the core outlet exceeds 650℃(about 2700 seconds).
引文
[1] 濮继龙，《压水堆核电厂安全与事故对策》第六章，原子能出版社．
[2] 甘向阳 高祖瑛 张作义，“先进堆严重事故对策”，核动力工程 Vol．21．No．6，2000年12月．
[3] N. Kourti, I. Shepherd, "Modelling Intergranular Fuel Swelling in Severe accidents", Journal of Nuclear Materials 277 (2000) 37—44 , 15 June 1999.
[4] World Nuclear Association, Nuclear Power in the World Today (January 2005). http://www.world-nuclear.org/info/inf01.htm, 2005.
[5] 中华人民共和国国家核安全局，国核安发[2002]106号，新建核电厂设计中几个重要安全问题的技术政策。北京：国家核安全局，2002．
[6] IAEA, Status of Advanced Light Water Cooled Reactor Designs 1996, IAEA-TecDoc-968. IAEA, 1997.
[7] J. Qu, "The Determination of the Area Affected by Relocation after Nuclear Accidents", Nuclear Engineering and Design 223 (2003) 41-48, 18 December 2002.
[8] USNRC, Reactor Safety Study — An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants, WASH-1400 (NUREG-75/O14), U. S. Nuclear Regulatory Commission, October 1975.
[9] JAMES M., BROUGHTON, et al., "A SCENARIO OF THREE MILE ISLAND UNIT 2 ACCIDENT", NUCLEAR SAFETY vol. 87, AUG 1989.
[10] D. A. Petti, et al., Power Burst Facility (PBF) Severe Fuel Damage Test 1-4 Test Results. NUREG/CR-5163, 1989.
[11] P. von der HARDT et al., The Severe Accident Research Program PHEBUS FP. The 5th International Topical Meeting on NUTHOS (NUTHOS-5), 1997.
[12] Bernhard Kuczera, "Thirty Years of LWR Safety Research at Karlsruhe", Nuclear Engineering and Design 202 (2000) 129—142.
[13] Idaho National Engineering Laboratory, SCDAP/RELAP5/MOD3.1 Code Manual, NUREG/CR-6150, 1995.
[14] Lillington J N., "Light Water Reactor Safety: the Development of Advanced Models and Codes for Light Water Reactor Safety Analysis[M]", Amsterdam, Elsevier, 1995.
[15] H. J. Allelein, J. Bestele, K. Neu , et al., "Severe accident code ASTEC development and validation", EUROSAFE, Paris, 18-19 November 1999.
[16] Summers R M, Cole R K, Smith R C, et al., "MELCOR Computer Code Mannuals: Primer and User's Guides, Version 1. 8. 3 [M].", U. S. NRC, March 1995.
[17] Summers. R. M., Stephen W. Webb, Paul Demmie et al. , "MELCOR 1. 8. 1: Computer Code for Nuclear Reactor Severe Accident Source Term and Risk Assessment Analysis", NUREG/CR-5531. Sandia National Laboratories, 1991.
[18] 张应超 季松涛 陈彭，《国内外核电厂严重事故分析方法调研报告》，内部资料，2002．6．
[19] 季松涛，“法国严重事故程序的应用”，中国原子能科学院年报，1998．
[20] 季松涛 张应超，“秦山核电厂小破口失水加全厂断电事故序列的堆芯早期破坏过程分析”，原子能科学技术 第34卷增刊，2000年9月．
[21] R. Gonzalez, F. Jacp, et al., "ICARE2: A Tool for Making Fast Running Calculations on LWR Core Degradation", Second Workshop on Super Simulators for Nuclear power Plants, Toranomon Pastral, Tokyo, November 2 1994.
[22] P. Chatelard, F. Fichot, et al., "ICARE2 V3mod 1.2 User's Mannual"
[23] F. Jacq, S. Matteo., "TIC 98 User's Manual", drs/semar/98-47.
[24] M. Farvacqu, "User's manual of CATHARE2 V1.3E.", STR/LML/EM/91-61.
[25] M. Farvacque, C. Sarrette, "CATHARE2 version V1.3E dictionary of operators and directives", STR/LML-EM/92-124.
[26] S. Melis, P. chatelard, F. Fichot, et al., "ICARE/CATHARE Vlmod 1.0 User's Manuual and Guidelines", IPSN/DRS/SEMAR.
[27] P. Chatelard, "ICARE2 V3mod 1.4 and ICARE/CATHARE V1.4 Users' Guidelines", SEMCA-2005-068.
[28] 浦胜娣等，严重事故管理控制室导则1—初始响应导则(SACRG—1)，原子能院堆工所内部报告，2005年3月
[2] 甘向阳 高祖瑛 张作义，“先进堆严重事故对策”，核动力工程 Vol．21．No．6，2000年12月．
[3] N. Kourti, I. Shepherd, "Modelling Intergranular Fuel Swelling in Severe accidents", Journal of Nuclear Materials 277 (2000) 37—44 , 15 June 1999.
[4] World Nuclear Association, Nuclear Power in the World Today (January 2005). http://www.world-nuclear.org/info/inf01.htm, 2005.
[5] 中华人民共和国国家核安全局，国核安发[2002]106号，新建核电厂设计中几个重要安全问题的技术政策。北京：国家核安全局，2002．
[6] IAEA, Status of Advanced Light Water Cooled Reactor Designs 1996, IAEA-TecDoc-968. IAEA, 1997.
[7] J. Qu, "The Determination of the Area Affected by Relocation after Nuclear Accidents", Nuclear Engineering and Design 223 (2003) 41-48, 18 December 2002.
[8] USNRC, Reactor Safety Study — An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants, WASH-1400 (NUREG-75/O14), U. S. Nuclear Regulatory Commission, October 1975.
[9] JAMES M., BROUGHTON, et al., "A SCENARIO OF THREE MILE ISLAND UNIT 2 ACCIDENT", NUCLEAR SAFETY vol. 87, AUG 1989.
[10] D. A. Petti, et al., Power Burst Facility (PBF) Severe Fuel Damage Test 1-4 Test Results. NUREG/CR-5163, 1989.
[11] P. von der HARDT et al., The Severe Accident Research Program PHEBUS FP. The 5th International Topical Meeting on NUTHOS (NUTHOS-5), 1997.
[12] Bernhard Kuczera, "Thirty Years of LWR Safety Research at Karlsruhe", Nuclear Engineering and Design 202 (2000) 129—142.
[13] Idaho National Engineering Laboratory, SCDAP/RELAP5/MOD3.1 Code Manual, NUREG/CR-6150, 1995.
[14] Lillington J N., "Light Water Reactor Safety: the Development of Advanced Models and Codes for Light Water Reactor Safety Analysis[M]", Amsterdam, Elsevier, 1995.
[15] H. J. Allelein, J. Bestele, K. Neu , et al., "Severe accident code ASTEC development and validation", EUROSAFE, Paris, 18-19 November 1999.
[16] Summers R M, Cole R K, Smith R C, et al., "MELCOR Computer Code Mannuals: Primer and User's Guides, Version 1. 8. 3 [M].", U. S. NRC, March 1995.
[17] Summers. R. M., Stephen W. Webb, Paul Demmie et al. , "MELCOR 1. 8. 1: Computer Code for Nuclear Reactor Severe Accident Source Term and Risk Assessment Analysis", NUREG/CR-5531. Sandia National Laboratories, 1991.
[18] 张应超 季松涛 陈彭，《国内外核电厂严重事故分析方法调研报告》，内部资料，2002．6．
[19] 季松涛，“法国严重事故程序的应用”，中国原子能科学院年报，1998．
[20] 季松涛 张应超，“秦山核电厂小破口失水加全厂断电事故序列的堆芯早期破坏过程分析”，原子能科学技术 第34卷增刊，2000年9月．
[21] R. Gonzalez, F. Jacp, et al., "ICARE2: A Tool for Making Fast Running Calculations on LWR Core Degradation", Second Workshop on Super Simulators for Nuclear power Plants, Toranomon Pastral, Tokyo, November 2 1994.
[22] P. Chatelard, F. Fichot, et al., "ICARE2 V3mod 1.2 User's Mannual"
[23] F. Jacq, S. Matteo., "TIC 98 User's Manual", drs/semar/98-47.
[24] M. Farvacqu, "User's manual of CATHARE2 V1.3E.", STR/LML/EM/91-61.
[25] M. Farvacque, C. Sarrette, "CATHARE2 version V1.3E dictionary of operators and directives", STR/LML-EM/92-124.
[26] S. Melis, P. chatelard, F. Fichot, et al., "ICARE/CATHARE Vlmod 1.0 User's Manuual and Guidelines", IPSN/DRS/SEMAR.
[27] P. Chatelard, "ICARE2 V3mod 1.4 and ICARE/CATHARE V1.4 Users' Guidelines", SEMCA-2005-068.
[28] 浦胜娣等，严重事故管理控制室导则1—初始响应导则(SACRG—1)，原子能院堆工所内部报告，2005年3月