临界热通量下反应堆压力容器的极限承载能力研究
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摘要
熔融物堆内滞留(In-vessel retention,IVR)已成为核电厂处理堆芯熔融严重事故的一种有效管理策略。为使IVR成功,既要满足热失效准则,保证反应堆压力容器(Reactor pressure vessel,RPV)的局部热通量低于堆腔内冷却剂的临界热通量(Critical heat flux,CHF),也要确保RPV的压力边界完整性,避免发生结构失效。为此,需要对CHF下RPV的结构完整性进行分析。对CHF下某堆型的RPV进行热分析,得到了RPV器壁局部熔化后的有效几何模型和沿壁厚的温度分布。进而,考虑热载荷和压力载荷作用,对该RPV模型进行极限载荷分析和IVR 72 h蠕变分析,确定RPV的极限承载能力。结果表明,IVR初始时刻4.9 MPa内压作用下该RPV最薄器壁径向截面全面屈服,达到对应的极限条件;72 h后3.6 MPa内压下的最大局部蠕变应变为4.1%,而3.7 MPa下则高达42.5%。因此,可将3.6 MPa视为该堆型RPV在CHF工况下72 h内的极限载荷。
The In-vessel retention(IVR) of molten core has been part of the effective management strategies of the severe accidents for nuclear power plant. To make IVR succeed, the thermal failure criterion must be met, so the local heat flux of reactor pressure vessel(RPV) wall should be lower than the critical heat flux(CHF) of the external coolant. Besides, the pressure boundary integrity of the RPV should also be ensured to avoid the structural failure. Accordingly, it's urgently necessary to analyze the structural integrity of the RPV under CHF. Firstly, the thermal analysis of a RPV under CHF is carried out to obtain effective geometric model with some of the RPV wall molten and temperature distribution along the wall. Then the limit load and IVR 72 hour creep behavior are analyzed for this model subjected to the thermal load and internal pressure, and the ultimate load capacity of the RPV is subsequently determined. The results show that the whole wall along the thinnest thickness of the RPV yields and the RPV reaches the ultimate load capacity under internal pressure of 4.9 MPa at the beginning of the IVR. The maximum local creep strain of 4.1% is reached under 3.6 MPa within 72 hours, while the internal pressure of 3.7 MPa results in a significant increase of maximum creep strain up to 42.5%. Therefore, the ultimate load capacity of this RPV under CHF in 72 hours is determined to be 3.6 MPa.
The In-vessel retention(IVR) of molten core has been part of the effective management strategies of the severe accidents for nuclear power plant. To make IVR succeed, the thermal failure criterion must be met, so the local heat flux of reactor pressure vessel(RPV) wall should be lower than the critical heat flux(CHF) of the external coolant. Besides, the pressure boundary integrity of the RPV should also be ensured to avoid the structural failure. Accordingly, it's urgently necessary to analyze the structural integrity of the RPV under CHF. Firstly, the thermal analysis of a RPV under CHF is carried out to obtain effective geometric model with some of the RPV wall molten and temperature distribution along the wall. Then the limit load and IVR 72 hour creep behavior are analyzed for this model subjected to the thermal load and internal pressure, and the ultimate load capacity of the RPV is subsequently determined. The results show that the whole wall along the thinnest thickness of the RPV yields and the RPV reaches the ultimate load capacity under internal pressure of 4.9 MPa at the beginning of the IVR. The maximum local creep strain of 4.1% is reached under 3.6 MPa within 72 hours, while the internal pressure of 3.7 MPa results in a significant increase of maximum creep strain up to 42.5%. Therefore, the ultimate load capacity of this RPV under CHF in 72 hours is determined to be 3.6 MPa.
引文
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[19]MAO Jianfeng,ZHANG Junhui,WANG Weizhe.Comparative study of flange to seal contact couplings with bolt relaxation under creep condition[J].ASME Journal of Engineering for Gas Turbine and Power,2014,136(7):072503-1-8.
[20]DINH T N,TU J P,SALMASSI T,et al.Limits of coolability in the AP1000-related ULPU-2400configuration V facility[R].Santa Barbara:CRSS Technical Report 03/06,University of California,2003.
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[23]KIM H,NAMGUNG I.Thermo-mechanical analysis of RVLH for APR1400 during a severe accident[C/CD]//23rd Conference on Structural Mechanics in Reactor Technology,Manchester,UK,August 10-14,2015.
[2]THEOFANOUS T G,LIU C,ADDITON S,et al.In-vessel coolability and retention of a core melt[J].Nuclear Engineering and Design,1997,169(1):1-48.
[3]THEOFANOUS T G,OH S J,SCOBEL J H.In-vessel retention technology development and use for advanced PWR designs in the USA and Korea[R].K-NERI Final Report,2002-022-K(I),2004.
[4]姚彦贵,施杨,蒋兴,等.严重事故IVR下反应堆压力容器耦合传热数值模拟分析[J].原子能科学技术,2014,48(8):1473-1478.YAO Yangui,SHI Yang,JIANG Xing,et al.Simulation on coupled heat transfer for reactor vessel in severe accident IVR[J].Atomic Energy Science and Technology,2014,48(8):1473-1478.
[5]SEHGAL B R.Nuclear safety in light water reactors:severe accident phenomenology[M].Oxford:Academic Press,Elsevier Inc,2012.
[6]ESMAILI H,KHATIB-RAHBAR M.Analysis of in-vessel retention and ex-vessel fuel coolant interaction for AP1000[R].ERI/NRC04-201,US Nuclear Regulatory Commission,2004.
[7]林诚格,郁祖盛,欧阳予.非能动安全先进核电厂AP1000[M].北京:原子能出版社,2008.LIN Chengge,YU Zusheng,OU Yangyu.An advanced passive plant AP1000[M].Beijing:Atomic Energy Press,2008.
[8]Nuclear Emergency Response Headquarters.Report of the Japanese government to the IAEA ministerial conference on nuclear safety-the accident at TEPCO’s Fukushima nuclear power stations[R].Tokyo:Government of Japan,2011.
[9]DEVOS J,SAINTE C C,POETTE C,et al.CEA program to model the failure of the lower head in severe accidents[J].Nuclear Engineering and Design,1999,191(1):3-15.
[10]CHU T Y,PILCH M M,BENTZ J H,et al.Reactor vessel lower head failure experiments and analyses[C]//Proceedings of the Twenty-Seven Water Reactor Safety Information Meeting.Maryland:1999,153-168.
[11]SEHGAL B R,NOURGALIEV R R,DING T N,et al.FOREVER experimental program in reactor pressure vessel creep behavior and core debris retention[C]//Proceedings of the 15th International Conference on Structural Mechanics in Reactor Technology(Smi RT-15).Seoul:1999,XI37-XI48.
[12]SEHGAL B R,KARBOJIANA A,GIRI A,et al.Assessment of reactor vessel integrity(ARVI)[J].Nuclear Engineering and Design,2005,235(2):213-232.
[13]Westinghouse LLC.Westinghouse AP1000 design control document[R].Pittsburgh:Westinghouse Electirc Company LLC,2008.
[14]高增梁,李曰兵,雷月葆.承压热冲击下反应堆压力容器的概率评定进展与案例分析[J].机械工程学报,2015,51(20):67-78.GAO Zengliang,LI Yuebing,LEI Yuebao.Progress and case study on probabilistic assessment of reactor pressure vessels under pressurized thermal shock[J].Journal of Mechanical Engineering,2015,51(20):67-78.
[15]李昌义,刘正东,林肇杰.核电站反应堆压力容器用钢的研究与应用[J].特殊钢,2010,31(4):14-17.LI Changyi,LIU Zhengdong,LIN Zhaojie.Research and application of steels for reactor pressure vessel of nuclear power station[J].Special Steel,2010,31(4):14-17.
[16]WILLSCHUETZ H G,ALTSTADT E.Generation of a high temperature material data base and its application to creep tests with French or German RPV-steel,FZR-353[R].Forschungszentrum Rossendorf,Dresden,Germany,2002.
[17]李曰兵,金伟娅,高增梁,等.核压力容器缺陷验收确定性准则的失效概率分析[J].机械工程学报,2015,51(6):27-35.LI Yuebing,JIN Weiya,GAO Zengliang,et al.Failure probability analysis of a reactor pressure vessel using a deterministic flaw acceptance criterion[J].Journal of Mechanical Engineering,2015,51(6):27-35.
[18]ALTSTADT E,MOESSNER T.Extension of the ANSYS@creep and damage simulation capabilities,FZR-296[R].Dresden:Forschungszentrum Rossendorf,2000.
[19]MAO Jianfeng,ZHANG Junhui,WANG Weizhe.Comparative study of flange to seal contact couplings with bolt relaxation under creep condition[J].ASME Journal of Engineering for Gas Turbine and Power,2014,136(7):072503-1-8.
[20]DINH T N,TU J P,SALMASSI T,et al.Limits of coolability in the AP1000-related ULPU-2400configuration V facility[R].Santa Barbara:CRSS Technical Report 03/06,University of California,2003.
[21]American Society of Mechanical Engineers Boiler and Pressure Vessel Committee on Materials.ASME boiler&pressure vessel code section II.part d-properties(metric)[S].New York:ASME,2010.
[22]LI Yuebing,ZHU Jianwei,BAO Shiyi,et al.Analysis on ultimate load capacity of cylinder with local discontinuity under high temperature gradient[C]//Proceedings of the ASME 2014 Pressure Vessels&Piping Conference.American Society of Mechanical Engineers,2014:V003T03A107-V003T03A107.
[23]KIM H,NAMGUNG I.Thermo-mechanical analysis of RVLH for APR1400 during a severe accident[C/CD]//23rd Conference on Structural Mechanics in Reactor Technology,Manchester,UK,August 10-14,2015.