Development and Verification of Thermal-hydraulic System Analysis Code Based on RELAP5 MOD3.2 for SFRs

被引：0

作者：

Song J. ^{[1
]}

Tan C. ^{[1
,2
]}

Tang S.-M. ^{[1
]}

Liu L.-M. ^{[1
]}

Tian W.-X. ^{[1
]}

Wu Y.-W. ^{[1
]}

Qiu S.-Z. ^{[1
]}

Su G.-H. ^{[1
]}

机构：

[1] Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, School of Nuclear Science and Technology, Xi'an Jiaotong University, Xi'an

[2] China Nuclear Power Operation Technology Corporation, Ltd., Wuhan

来源：

| 1600年 / Atomic Energy Press卷 / 51期

关键词：

Code development; Liquid metal physical property; RELAP5; Sodium-cooled fast reactor; Thermal-hydraulic analysis;

D O I：

10.7538/yzk.2017.51.06.0994

中图分类号：

学科分类号：

摘要：

The general-purpose reactor thermal-hydraulic analysis code RELAP5 MOD3.2 was modified for system analysis of sodium-cooled fast reactor (SFR). The thermodynamic and transport properties of sodium liquid and vapor were implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The methods of code modifications are universal for other working fluids and will not affect the code original performance. The benchmark on loss of flow accident transient of Experimental Breeder Reactor II (EBR-II) for Argonne National Laboratory (ANL) was analyzed to verify the modified code. The results show that the reactor can shut down automatically relying on inherent negative feedbacks and major thermal-hydraulic parameters are lower than the safe limits. Moreover, the calculation results are consistent with the experimental data and other SFR system analysis code results. The work in the paper can demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further. © 2017, Editorial Board of Atomic Energy Science and Technology. All right reserved.

引用

页码：994 / 1001

页数：7

共 11 条

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