Analysis on LOCA for CSR1000

被引：6

作者：

Chen, Juan ^{[1
,2
]}

Zhou, Tao ^{[1
,2
]}

Liu, Liang ^{[3
]}

Fang, Xiaolu ^{[4
]}

机构：

[1] North China Elect Power Univ, Sch Nucl Sci & Engn, Beijing 102206, Peoples R China

[2] North China Elect Power Univ, Nucl Safety & Thermal Power Standardizat Inst, Beijing 102206, Peoples R China

[3] Beijing Guodian Zhishen Control Technol Co Ltd, Beijing 102200, Peoples R China

[4] China Nucl Power Engn Co Ltd, Beijing 100840, Peoples R China

来源：

ANNALS OF NUCLEAR ENERGY | 2017年 / 110卷

关键词：

CSR1000; LOCA; Break flow; Safety system;

D O I：

10.1016/j.anucene.2017.07.026

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

In CSR1000 (Chinese Supercritical Water Cooled Reactor 1000), based on the existing SCAC-CSR1000 code, by incorporating break flow model and passive safety system, CSR1000-DP02 code for LOCA analysis is developed. 50% hot leg break loss of coolant accident (LOCA) is analyzed. The results show that, when LOCA happens, the reactor scrams immediately. The main feed water mass flow, and coolant mass flow of the two fuel channels overall both increase in LOCA. In the flow of passive containment cooling system curve, there is a peak from 2 s to 4 s, and there is a critical flow phenomenon from 5.7 s to 12 s. When LOCA occurs, the cladding temperature gradually decreases, the maximum cladding temperature is below the safety criterion of 1260 degrees C. (C) 2017 Elsevier Ltd. All rights reserved.

引用

页码：903 / 908

页数：6

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