MELCOR程序在HTGR事故分析中的最新进展
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摘要
MELCOR程序是美国NRC在安全评审中使用的一体化系统分析程序,早期主要用于轻水堆严重事故分析。近年来,该程序逐渐用于高温气冷堆的石墨腐蚀、裂变产物行为和石墨粉尘等物理现象方面的研究。本文介绍了在最新版本的MELCOR 2.1程序中,针对高温气冷堆特点所进行的扩展和开发,以及MELCOR程序在高温气冷堆(HTGR)事故分析中的计算流程。
MELCOR code is an integrated system analysis code used by U.S.NRC for the nuclear safety review.In the early days,the code was mainly used for light water reactor(LWR) severe accident analysis.And in recent years,the code has been used in the research of high temperature gas-cooled reactor(HTGR) physical phenomena,including graphite corrosion,the behavior of fission products(FP) and graphite dust.Code extension and development was introduced according to the characteristics of HTGR in the latest edition of MELCOR code,MELCOR 2.1,and MELCOR code calculation process for HTGR was also introduced.
MELCOR code is an integrated system analysis code used by U.S.NRC for the nuclear safety review.In the early days,the code was mainly used for light water reactor(LWR) severe accident analysis.And in recent years,the code has been used in the research of high temperature gas-cooled reactor(HTGR) physical phenomena,including graphite corrosion,the behavior of fission products(FP) and graphite dust.Code extension and development was introduced according to the characteristics of HTGR in the latest edition of MELCOR code,MELCOR 2.1,and MELCOR code calculation process for HTGR was also introduced.
引文
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[3]刘杰,高祖瑛.高温气冷堆球床有效导热系数的计算模型及其在SANA基准试验分析中的应用[J].核科学与工程,1999,19(4):303.
[4]赵兆颐,朱瑞安.反应堆热工流体力学[M].北京:清华大学出版社,1992.
[5]Nuelear Safety Standards Commission.Reactor core design of high-temperature gas-cooled reactors part 3:Loss of pressure through friction in pebble bed cores[S].Salzgitter;Nuclear Safety Standards Commission(KTA),1981.
[6]Nuclear Safety Standards Commission.Reactor core design of high-temperature gas-cooled reactors part 2:Heat transfer in spherical fuel elements[S].Salzgitter:Nuclear Safety Standards Commission(KTA),1983.
[7]Sandia National Laboratories.MELCOR computer code manuals,vol.1:primer and users'guide[M].Washington:U.S.Nuclear Regulatory Commission,2011.
[8]Sandia National Laboratories.MELCOR computer code manuals,vol.2:reference manuals[M].Washington;U.S.Nuclear Regulatory Commission,2011.
[9]RICHARDS M B.Reaction of nuclear-grade graphite with low concentrations of steam in the helium coolant of an MHTGR[J].Energy,1990,15(9):729.
[2]陈召林,李斌,柴国旱.高温气冷堆HTR-PM安全审评中有关燃料最高温度的考虑[J].核安全,2012(2):16.
[3]刘杰,高祖瑛.高温气冷堆球床有效导热系数的计算模型及其在SANA基准试验分析中的应用[J].核科学与工程,1999,19(4):303.
[4]赵兆颐,朱瑞安.反应堆热工流体力学[M].北京:清华大学出版社,1992.
[5]Nuelear Safety Standards Commission.Reactor core design of high-temperature gas-cooled reactors part 3:Loss of pressure through friction in pebble bed cores[S].Salzgitter;Nuclear Safety Standards Commission(KTA),1981.
[6]Nuclear Safety Standards Commission.Reactor core design of high-temperature gas-cooled reactors part 2:Heat transfer in spherical fuel elements[S].Salzgitter:Nuclear Safety Standards Commission(KTA),1983.
[7]Sandia National Laboratories.MELCOR computer code manuals,vol.1:primer and users'guide[M].Washington:U.S.Nuclear Regulatory Commission,2011.
[8]Sandia National Laboratories.MELCOR computer code manuals,vol.2:reference manuals[M].Washington;U.S.Nuclear Regulatory Commission,2011.
[9]RICHARDS M B.Reaction of nuclear-grade graphite with low concentrations of steam in the helium coolant of an MHTGR[J].Energy,1990,15(9):729.