三层熔池结构IVR分析程序开发及验证
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摘要
目前对熔融物堆内滞留(IVR)进行分析时,主要采用两层熔池模型进行点估算分析。然而随着研究的深入,已有IVR分析程序不能准确模拟三层熔池模型。为此,本文采用三层熔池模型开发了模块化IVR分析程序SPIRE,并对计算结果进行了验证。结果表明,SPIRE程序的计算结果与文献结果吻合较好,适用于IVR分析。利用SPIRE程序进行分析可知,与两层熔池相比,三层熔池结构下压力容器底部和轻金属层热流密度均会有明显增强。敏感性分析结果表明,铀氧化份额和不锈钢总质量会显著影响热流密度分布及最大临界热流密度比。
Currently when in-vessel retention(IVR)is evaluated,two-layer melt pool model is usually applied for point estimate analysis.After years of intensive study on IVR,it is found that existing code could not simulate three-layer melt pool model accurately.Thus a modular IVR analysis code using three-layer melt pool model,SPIRE,was developed and validated.The results show that the calculation results of SPIRE code have a good agreement with the data from references.Thus the SPIRE code is feasible to perform IVR analysis.Based on simulation results,the heat fluxes at the bottom of the vessel and light metallic layer notably increase in three-layer melt pool configuration compared with two-layer melt pool configuration.According to sensitive analysis,uranium oxidation rate and total mass of stainless steel have larger influence on heat flux distribution and maximum critical heat flux(CHF)ratio.
Currently when in-vessel retention(IVR)is evaluated,two-layer melt pool model is usually applied for point estimate analysis.After years of intensive study on IVR,it is found that existing code could not simulate three-layer melt pool model accurately.Thus a modular IVR analysis code using three-layer melt pool model,SPIRE,was developed and validated.The results show that the calculation results of SPIRE code have a good agreement with the data from references.Thus the SPIRE code is feasible to perform IVR analysis.Based on simulation results,the heat fluxes at the bottom of the vessel and light metallic layer notably increase in three-layer melt pool configuration compared with two-layer melt pool configuration.According to sensitive analysis,uranium oxidation rate and total mass of stainless steel have larger influence on heat flux distribution and maximum critical heat flux(CHF)ratio.
引文
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[2]徐红,周志伟.大型先进压水堆熔融物堆内滞留初步研究[J].原子能科学技术,2013,47(6):969-974.XU Hong,ZHOU Zhiwei.Preliminary study on in-vessel retention in large-scale advanced PWR[J].Atomic Energy Science and Technology,2013,47(6):969-974(in Chinese).
[3]鲍晗,金越,刘晓晶,等.大功率先进压水堆IVR有效性评价中熔池换热研究[J].原子能科学技术,2014,48(2):234-240.BAO Han,JIN Yue,LIU Xiaojing,et al.Study of nature convection heat transfer in molten pool for advanced large size PWR[J].Atomic Energy Science and Technology,2014,48(2):234-240(in Chinese).
[4]金頔,李飞,刘晓晶,等.大功率先进压水堆压力容器外部冷却能力研究[J].原子能科学技术,2014,48(2):277-284.JIN Di,LI Fei,LIU Xiaojing,et al.Study on external reactor vessel cooling capacity for advanced large size PWR[J].Atomic Energy Science and Technology,2014,48(2):277-284(in Chinese).
[5]杨磊,伊雄鹰,陈玉清.海洋核动力平台严重事故下熔融物堆内滞留分析程序开发[J].原子能科学技术,2017,51(11):1 997-2 003.YANG Lei,YI Xiongying,CHEN Yuqing.Development of IVR code for marine nuclear power plant under severe accident[J].Atomic Energy Science and Technology,2017,51(11):1 997-2 003(in Chinese).
[6]ASMOLOV V,STRIZHOV V.Overview of the progress in the OECD MASCA project[C]∥Proceedings of CSARP Meeting.Washington D.C.:[s.n.],2004.
[7]曹克美,许以全,史国宝,等.三层熔池结构情况下反应堆压力容器外水冷有效性分析[J].核动力工程,2013,34(4):20-23.CAO Kemei,XU Yiquan,SHI Guobao,et al.ERVC effectiveness analysis for three-layer configuration molten pool[J].Nuclear Power Engineering,2013,34(4):20-23(in Chinese).
[8]文青龙,陈军,卢冬华,等.严重事故条件下压力容器完整性评价的研究进展[J].核科学与工程,2011,31(3):222-229.WEN Qinglong,CHEN Jun,LU Donghua,et al.Research progress on assessment of reactor vessel integrity under severe accident conditions[J].Chinese Journal of Nuclear Science and Engineering,2011,31(3):222-229(in Chinese).
[9]THEOFANOUS T G,LIU C,ADDITON S,et al.In-vessel cool-ability and retention of a core melt,DOE/ID-10460[R].[S.l.]:[s.n.],1996.
[10]KYMLINEN O,TUOMISTO H,HONGISTO O,et al.Heat flux distribution from a volumetrically heated pool with high Rayleigh number[J].Nuclear Engineering and Design,1994,149(1-3):401-408.
[11]TRAN C T.The effective convectively model for simulation and analysis of melt pool heat transfer in a light water reactor pressure vessel lower head[D].Sweden:KTH,2009.
[12]GABOR J D,ELLSION P G,CASSULO J C.Heat transfer from internally heated hemispherical pools[R].USA:Argonne National Laboratory,1980.
[13]ZHANG L,ZHANG Y,ZHAO B,et al.COPRA:A large scale experiment on natural convection heat transfer in corium pools with internal heating[J].Progress in Nuclear Energy,2016,86:132-140.
[14]ZHANG Y P,QIU S Z,SU G H,et al.Analysis of safety margin of in-vessel retention for AP1000[J].Nuclear Engineering and Design,2010,240(8):2 023-2 033.
[15]PARK J W,BAE J,SONG H J.Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling[J].Annals of Nuclear Energy,2016,88:57-67.
[16]向清安,关仲华,邓纯锐,等.AP1000IVR三层熔池结构评价分析[J].核动力工程,2013,34(6):83-87.XIANG Qingan,GUAN Zhonghua,DENG Chunrui,et al.An assessment methodology of thee-layers melt configuration during IVR for AP1000[J].Nuclear Power Engineering,2013,34(6):83-87(in Chinese).
[17]PARK R J,KANG K H,SONG S W,et al.Detailed evaluation of melt pool configuration in the low plenum of the APR1400reactor vessel during severe accidents[J].Annals of Nuclear Energy,2015,75:476-482.
[18]REMPE J L,KNUDSON D L.Margin for in-vessel retention in the APR1400-VESTA and SCDAP/RELAP5-3D[R].USA:Idaho National Engineering and Environmental Laboratory,2004.