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SNAP程序在核电厂安全分析中的应用 被引量：3

The Application of SNAP to NPP Safety Anlysis
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摘要 SNAP程序是由NRC资助开发的用于简化分析过程的软件,本文对SNAP程序进行了详细的介绍,并利用SNAP程序与RELAP5/MOD3.3程序对某典型四环路压水堆进行模拟,描述了SNAP程序在核电厂安全分析中应用的特点,并对关键现象进行分析。研究表明,SNAP程序的应用可以大大简化程序建模和数据处理过程,并能直观实时的观测计算结果,在核电厂安全分析中应用的前景广泛。 Symbolic Nuclear Analysis Package （SNAP） consists of a suite of integrated applications designed to simplify the process of performing engineering analysis.It is developed by The Nuclear Regulatory Commission（NP, C）. In this paper, SNAP cede is introduced in details, a typical four loops PWR with SNAP and RELAP5/MOD3.3 is simulated, the application specialty of SNAP Code in safety analysis is discussed, and typical phenomena is analysed. Research shows, SNAP has its application prospect, as by the use of it, the procedure of model building and data processing would be simplified greatly, and the results could be observed directly in real time.
作者 孙微 靖剑平 付陟玮 刘巧凤 潘昕怿
机构地区 中华人民共和国环境保护部核与辐射安全中心
出处 《中国科技信息》 2012年第18期78-79,共2页 China Science and Technology Information
关键词 SNAP程序 核电厂安全分析 图形化分析程序 SNAP Code NPP Safety anlysis graphics analysiscode
分类号 TM623.8 [电气工程—电力系统及自动化]
  • 相关文献

参考文献3

  • 1Symbolic Nuclear Analysis Package (SNAP) User,s Manual. Applied Programming Technology, Inc.
  • 2RHLAP5/M0D3.3 CODE MANUAL.
  • 3Jayanti S, Hewin G F. Prediction of the Slug-Chum Flow Transition in Vertical 丁wo-Phase Flow,1992,18(6):847,860.

同被引文献27

  • 1I.H. Bokhari, T. Mahmood. Analysis of loss of flowaccident at Pakistan research reactor-1 [J]. Annals of Nuclear Energy, 2005, 32(18): 1963-1968.
  • 2A. Hainoun, N. Ghazi, B. Mansour Abdul-Moaiz. Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT [J]. Nuclear Engineering and Design, 2010, 240 (5) : 1132-1138.
  • 3Zhi hongXu, Dong Hou, Sheng weiFu, et al. Loss of flow accident and its mitigation measures for nuclear systems with SCWR-M [J]. Annals of Nuclear Energy, 2011, 38(12): 2634-2644.
  • 4Edward L. Carlin, Peter A. Hilton and Yixing Sung. margin assessment of AP1000 loss of flow transient [C]. // International Conference on Nuclear Engineering, Miami: ASME, 2006, 1-9.
  • 5上海核工程研究设计院.三门核电一期工程1&2号机组最终安全分析报告[R].上海:国家核电上海核工程研究设计院,2012.
  • 6NRC.10CFR 50.46, Acceptance criteria for emergency core cooling systems (ECCS) in light water nuclear reactors, Appendix K to Part 50 " ECCS Evaluation Models" [S]. Washington DC: NRC, 1977.
  • 7D'AURIA F, CAMARGO C, MAZZANTINI O. The Best Estimate Plus Uncertainty ( BEPU ) approach in licensing of current nuclear reactors [J]. Nuclear Engineering & De- sign, 2012, 248 (1): 317-328.
  • 8D'Auria F, Petruzzi A, Muellner N, et al. The BEPU Chal- lenge in Current Licensing of Nuclear Reactors [C]//ASME 2010 3rd Joint US-European Fluids Engineering Summer Meeting collocated with 8th International Conference on Nan- ochannels, Microchannels, and Minichannels. Monitreal, 2010: 1511-1517.
  • 9Hallee B' T. Feed-and-bleed transient analysis of OSU APEX facility using the modem Code Scaling, Applicability, and Uncertainty method [D] .Oregon: Oregon State University, 2013.
  • 10MARTIN R P, O'DELL L D. AREVA's realistic large break LOCA analysis methodology [J]. Nuclear Engineering & De- sign, 2005, 235 (16): 1713-1725.

引证文献3

二级引证文献7

中国科技信息
2012年 第18期

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