摘要
从反应堆热工水力实验只能获得和临界热流密度(CHF)有关的各平均参数.子通道分析程序提供了一种手段,把平均参数转化成CHF产生处的局部参数.从而可以整理出带局部参数条件的CHF关系式.
Critical heat flux(CHF) correlation under local parameter condition play very significant role in safety operating reactor.Only mean value concerning CHF can be obtained from reactor thermalhydraulic experiments.Subchannel analysis code has provided a kind of manner which can transfer mean value to local value at the place where CHF occurs.Furthermore,CHF correlations with local condition can be developed.The paper has presented a detailed step which can be used to derive CHF correlations with local condition by adopting FLICAⅢ M subchannel analysis code.
出处
《核动力工程》
EI
CAS
CSCD
北大核心
1997年第1期43-46,共4页
Nuclear Power Engineering
关键词
子通道分析
临界热流密度
堆芯
压水堆
Subchannel analysis Critical heat flux Reactor core Local parameter