摘要
将堆芯子通道热工水力分析程序ＣＯＢＲＡⅢＣ／ＭＩＴ－２的水物性、临界热流关系式、泡核沸腾起始点判断公式等加以修正或扩充，使之能用于低温低压下研究堆或实验堆的分析。利用改进的ＣＯＢＲＡⅢＣ／ＭＩＴ－２，对日本板状元件高通量研究堆ＪＲＲ－３Ｍ在不同基准流速下以及不同流道阻塞率下的热工水力特性进行了分析计算，所得结果与日本原子能研究院开发的热工水力分析软件ＣＯＯＬＯＤ的相应预测结果符合良好。
Some improvements have been made on the core subchannel analysis code COBRA Ⅲ C/MIT-2,i.e.the correction of water thermal property at low pressure and low temperature conditions,the addition of some equations for Onset of Nucleate Boiling(ONB) temperature,and the addition of correlations for DNB heat flux.The steady-state thermal-hydraulic performance of JRR-3M under different core flow velocities and different channel blockage rates are analyzed using the improved COBRA Ⅲ C/MIT-2.The results are in good accordance with those predicted by COOLOD,a thermal and hydraulic analysis code used in JRR-3M.
出处
《核科学与工程》
CSCD
北大核心
1996年第1期35-41,共7页
Nuclear Science and Engineering
