燃料棒束换热(RBHT)试验对低压下堆芯漂移流模型的评价  

The evaluation of rod bundle drift flux model using the Rod Bundle Heat Transfer(RBHT) test

作　　者：樊普[1] 曹克美[1] 徐财红[1] FAN Pu CAO Ke-mei XU Cai-hong(Shanghai Nuclear Engineering Research b. Design Institute, Shanghai 200233, Chin)

机构地区：[1]上海核工程研究设计院,上海200233

出　　处：《核科学与工程》2016年第4期548-554,共7页Nuclear Science and Engineering

摘　　要：我国目前正在发展基于非能动技术的三代核电,为评价和改进非能动核电厂小破口失水事故在低压下棒束区的漂移流模型,采用燃料棒束换热(RBHT)试验对EPRI[6]、Cunningham-Yeh[4]模型,Bestion[7]漂移流模型进行了计算分析,结果表明燃料棒束换热试验RBHT试验数据工况能涵盖非能动核电厂在低压下的参数,不需要建造针对燃料棒束的试验台架,Cunningham-Yeh[4]和Bestion[7]漂移流模型基本适用,而EPRI[6]在低压区过高预测了空泡份额,不适用非能动核电厂。Passive Nuclear Power Plant is developed in China now.To evaluate and improve the drift flux model in the rod bundle region under low pressure condition,EPRI[6],Cunningham-Yeh[4],Bestion[7]model is evaluated using the RBHT（Rod Bundle Heat Transfer）test data.And the RBHT（Rod Bundle Heat Transfer）test data ranges can include the passive nuclear power plant conditions under the low pressure phase of Small Break Loss of Coolant Accident（SBLOCA）scenario.The results indicate that the model of Cunningham-Yeh[4]and Bestion[7]which are used in SBLOCA code for passive nuclear power plant is applicable,and EPRI[6]is not applicable for SBLOCA code of passive nuclear power plant under low pressure condition.

关 键 词：漂移流模型 失水事故 非能动核电厂 

分 类 号：TL364[核科学技术—核技术及应用]