Analysis of Effect of SBLOCA before and behind Isolation Valve of PRHRS on Passive Reactor Core Cooling System

被引：0

作者：

Hao B. ^{[1
]}

Wang N. ^{[1
]}

Zhong J. ^{[1
]}

Shi Y. ^{[1
]}

Fang F. ^{[1
]}

机构：

[1] State Nuclear Hua Qing (Beijing) Nuclear Power Technology R&D Center Co. Ltd., Beijing

来源：

| 1600年 / Atomic Energy Press卷 / 54期

关键词：

Break accident before and behind isolation valve of PRHRS; Large passive core cooling integrated test facility; Passive core cooling system; Small break loss of coolant accident;

D O I：

10.7538/yzk.2019.youxian.0877

中图分类号：

学科分类号：

摘要：

For the test process of small break loss of coolant accident (SBLOCA) of AP type nuclear power plant, there is a more consistent understanding at home and abroad. However, the influence of the same size of the break on the test process and passive core cooling system in different locations still needs further study. In this paper, a large passive core cooling integrated test facility ACME was used to study the break accident test conditions of passive residual heat removal system (PRHRS) before and behind the isolation valve, and the bottom break test of the cold pipe of core makeup tank (CMT) was used as the contrast condition. The test results show that the accident process of PRHRS before and behind the isolation valve is in accordance with the process of SBLOCA, the core is always in a good cooling statement and the safety of passive core cooling system is effectively verified. There is a certain impact on the accident process for the same break size and different break locations, and the location of the break has a key impact on the CMT level and safety injection flow. In contrast, the heat transfer of PRHRS equipment is also quite different. The heat transfer of cold pipe break and break behind the isolation valve is greater than break before the isolation valve, however, the flow and heat transfer mechanism of PRHRS heat exchange tube needs further study. © 2020, Editorial Board of Atomic Energy Science and Technology. All right reserved.

引用

页码：2073 / 2080

页数：7

共 15 条

[1] (2010)
[2] (2012)
[3] CHEN Lian, CHANG Huajian, LI Yuquan, Analysis on working pressure selection of ACME integral test facility, Atomic Energy Science and Technology, 45, 10, pp. 1215-1220, (2011)
[4] LI Yuquan, HAO Botao, ZHONG Jia, Et al., Comparative experiment to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA, Nuclear Engineering and Technology, 49, pp. 54-70, (2017)
[5] LI Yuquan, CHANG Huajian, YE Zishen, Et al., Analyses of ACME integral test results on CAP1400 small-break loss of-coolant-accident transient, Progress in Nuclear Energy, 88, pp. 375-397, (2016)
[6] WULFF W, ROHATGI U S., System scaling for the Westinghouse AP600 pressurized water reactor and related test facilities, NUREG/CR-5541, BNL-NUREG-52550, (1998)
[7] WANG W W, SU G H, QIU S Z, Et al., Thermal hydraulic phenomena related to small break LOCAs in AP1000, Nuclear Engineering and Design, 53, pp. 407-419, (2011)
[8] ZHU Sheng, Study on code modeling and test initialization condition method of ACME facility, Atomic Energy Science and Technology, 50, 7, pp. 1179-1185, (2016)
[9] FANG Fangfang, YANG Fuming, HAO Botao, Et al., Experimental and numerical analyses of typical SBLOCAS on advanced core-cooling mechanism experiment, Atomic Energy Science and Technology, 51, 8, pp. 1393-1399, (2017)
[10] LIU Yusheng, XU Chao, FANG Fangfang, Et al., Test study of station blackout accident on advanced core-cooling mechanism experiment facility, Atomic Energy Science and Technology, 52, 8, pp. 1438-1444, (2018)

← 1 2 →