熔融物特性试验台架研究
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- 英文论文题名：Research on test facilities for studying characteristic of corium
- 论文作者：刘宇生 ; 温丽晶 ; 张盼 ; 李聪新
- 英文论文作者：LIU Yu-sheng ; WEN Li-jing ; ZHANG Pan ; LI Cong-xing ; Nuclear and Radiation Safety Center ; MEP
- 年：2015
- 作者机构：环境保护部核与辐射安全中心;
- 论文关键词：严重事故 ; 熔融物 ; 试验研究 ; 核安全
- 英文论文关键词：severe accident ; corium ; experimental study ; nuclear safety
- 会议召开时间：2015-09-21
- 会议录名称：中国核科学技术进展报告（第四卷）——中国核学会2015年学术年会论文集第10册（核安全分卷）
- 语种：中文
- 分类号：TM623
- 学会代码：EGVD
- 会议名称：中国核学会2015年学术年会
- 会议地点：中国四川绵阳
- 主办单位：中国核学会
- 学会名称：中国核学会
- 页数：6
- 文件大小：325k
- 原文格式：O
- 会议级别：全国
摘要
作为在国际上受到普遍关注的严重事故缓解措施,熔融物堆内滞留技术(1VR)在Loviisa核电厂、西屋非能动核电厂的设计中得到了应用和优化。在IVR的设计中,利用冷却水淹没反应堆堆腔,在压力容器下封头外壁和保温层之间的环形流道内形成稳定的自然循环,从而实现对压力容器下封头壁面的外部冷却,维持压力容器的完整性,避免出现威胁安全壳完整性的事故现象。其中,对于迁移到压力容器下封头的熔融物,确定其换热特性并给出可能的热流密度分布,是IVR设计中的重要问题。针对研究熔融物特性问题的试验台架,本文选取代表性的试验台架进行了研究,重点分析了这些试验台架的设计参数、试验方法、系统布置及试验结果,并进行了详细的对比分析。研究结果表明:现有熔融物特性试验瑞利数的最大值已达2×1017;试验工质包括水、氟利昂、熔盐及真实堆芯材料,工质的选择对试验测量方式及结果可用性有显著影响;结合已有的试验及本文的分析,建议新设计的试验台架采用三维试验模型,以熔盐作为模拟工质。
As one of severe accident mitigation measures gained considerable attention,design concept of retaining the corium inside the nuclear reactor pressure vessel has been successful applied to Loviisa plant and westinghouse plant with passive safety feature.The annular clearance between thermal insulation and reactor pressure vessel is utilized for cooling the vessel wall under hypothetical core melting accidents in In-Vessel Retention(IVR)design.However,the characteristic of core corium is an important problem in determination of heat flux transmitted to vessel wall.Typical test facilities for studying characteristic of corium are studied in this paper,of which the design parameters,experimental methods,system configurations and results are emphatically discussed.The results show that the Rayleigh number tested in experiments is achieved about 2×10~(17).Simulant melt material has significant effects on measure and test data.Three-dimensional test model with non-eutectic mixture of several components is recommended for new designed facility.
As one of severe accident mitigation measures gained considerable attention,design concept of retaining the corium inside the nuclear reactor pressure vessel has been successful applied to Loviisa plant and westinghouse plant with passive safety feature.The annular clearance between thermal insulation and reactor pressure vessel is utilized for cooling the vessel wall under hypothetical core melting accidents in In-Vessel Retention(IVR)design.However,the characteristic of core corium is an important problem in determination of heat flux transmitted to vessel wall.Typical test facilities for studying characteristic of corium are studied in this paper,of which the design parameters,experimental methods,system configurations and results are emphatically discussed.The results show that the Rayleigh number tested in experiments is achieved about 2×10~(17).Simulant melt material has significant effects on measure and test data.Three-dimensional test model with non-eutectic mixture of several components is recommended for new designed facility.
引文
[1]T.G.Theofanous,C.Liu,S.Additon,S.Angelini.In-vessel coolability and retention of a core melt[J].Nuclear Engineering and Design,1997,169,1-48.
[2]T.G.Theofanous,S.Angelini.Natural convection for in-vessel retention at prototypic Rayleigh numbers[J].Nuclear Engineering and Design,2000,200,1-9.
[3]Kulacki,EA.,A.A.Emara.High Rayleigh Number Convection in Enclosed Fluid Layers with Internal Heat Sources,U.S.Nuclear Regulatory Commission Report.1975,NUREG-7 5/065.
[4]T.G.Theofanous,M.Maguire,S.Angelini,T.Salmassi.The first results from the ACOPO experiment[J].Nuclear Engineering and Design,1997,169,49-57.
[5]S.V.Bechta,E.V.Krushinov,S.A.Vitol,V.B.Khabensky.Influence of corium oxidation on fission product release from molten pool[J].Nuclear Engineering and Design,2010,240,1229-1241
[6]Main Results of the First Phase of MASCA Project.OECD/NEA MASCA Project.Russian Research.November,2004.
[7]X.Gaus-Liu,A.Miassoedov,T.Cron,T.Wenz.In-vessel melt pool coolibility test-Description and results of LIVE experiments.Nuclear Engineering and Design.240(2010),3898-3903.
[2]T.G.Theofanous,S.Angelini.Natural convection for in-vessel retention at prototypic Rayleigh numbers[J].Nuclear Engineering and Design,2000,200,1-9.
[3]Kulacki,EA.,A.A.Emara.High Rayleigh Number Convection in Enclosed Fluid Layers with Internal Heat Sources,U.S.Nuclear Regulatory Commission Report.1975,NUREG-7 5/065.
[4]T.G.Theofanous,M.Maguire,S.Angelini,T.Salmassi.The first results from the ACOPO experiment[J].Nuclear Engineering and Design,1997,169,49-57.
[5]S.V.Bechta,E.V.Krushinov,S.A.Vitol,V.B.Khabensky.Influence of corium oxidation on fission product release from molten pool[J].Nuclear Engineering and Design,2010,240,1229-1241
[6]Main Results of the First Phase of MASCA Project.OECD/NEA MASCA Project.Russian Research.November,2004.
[7]X.Gaus-Liu,A.Miassoedov,T.Cron,T.Wenz.In-vessel melt pool coolibility test-Description and results of LIVE experiments.Nuclear Engineering and Design.240(2010),3898-3903.