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COSINE程序再淹没模型验证及参数敏感性分析 被引量:1

Verification of COSINE Reflooding Model and Sensitivity Analysis of Parameters
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摘要 再淹没是压水堆大破口失水事故后的重要阶段,为评估系统程序在该阶段的计算能力,需要选择多种传热模型对失水事故进行复现并分析参数的敏感性响应。本文对压水堆失水事故实验(LOFT)台架进行建模,将COSINE程序中不同传热模型的计算结果与实验数据比较,验证了传热模型精确度;同时进行再淹没阶段的参数敏感性计算,识别出了对第二包壳峰值温度(PCT)影响最大的参数。计算表明:COSINE程序的传热模型能较好地预测再淹没现象;对计算结果影响较大的敏感性参数包括:UO_(2)体积热容、液滴直径、液滴相间传热系数和膜态沸腾壁面对汽相的传热系数。 Reflooding is an important stage after the PWR large break accident.In order to evaluate the calculation capability of the system program at this stage,it is necessary to select a variety of heat transfer models to reproduce the accident and analyze the sensitivity response of the parameters.In this paper,the PWR Loss of Fluid Test(LOFT)bench is modeled,and the calculation results of different heat transfer models in the COSINE program are compared with the experimental data to verify the accuracy of the model;At the same time,the parameter sensitivity calculation in the reflooding stage is carried out,and the parameters which have the greatest influence on the second peak cladding temperature(PCT)are identified.The calculation shows that the heat transfer model of COSINE program can well predict the reflooding phenomenon,and the sensitivity parameters that have great influence on the calculation results include UO_(2) volume heat capacity,droplet diameter,droplet interphase heat transfer coefficient and heat transfer coefficient of film boiling wall against vapor phase.
作者 李雪琳 张昊 杨燕华 Li Xuelin;Zhang Hao;Yang Yanhua(School of Nuclear Science and Engineering,Shanghai Jiaotong University,Shanghai,200240,China)
出处 《核动力工程》 EI CAS CSCD 北大核心 2022年第6期73-78,共6页 Nuclear Power Engineering
关键词 再淹没 敏感性分析 COSINE LOFT实验 Reflooding Sensitivity analysis COSINE LOFT experiment
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  • 1Tong Soo Choi,Hee Cheon NO.An improved RELAP5/MOD3.3 reflood model considering the effect of spacer grids[J].Nuclear Engineering and Design.2012
  • 2Tong Soo Choi,Hee Cheon NO.Improvement of the reflood model of RELAP5/MOD3.3 based on the assessments against FLECHT-SEASET tests[J].Nuclear Engineering and Design.2009(4)

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