Loss of coolant accident analysis under restriction of reverse flow

被引:8
作者
Radaideh, Majdi, I [1 ]
Kozlowski, Tomasz [1 ]
Farawila, Yousef M. [2 ]
机构
[1] Univ Illinois, Dept Nucl Plasma & Radiol Engn, Talbot Lab, 104 South Wright St, Urbana, IL 61801 USA
[2] Farawila Etal Inc, 306 Rockwood, Richland, WA 99352 USA
关键词
LOCA; Reverse flow; TRACE; BWR; Thermal-hydraulic; TRACE; LOCA; QUANTIFICATION; WILKS;
D O I
10.1016/j.net.2019.04.016
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
This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD. (C) 2019 Korean Nuclear Society, Published by Elsevier Korea LLC.
引用
收藏
页码:1532 / 1539
页数:8
相关论文
共 15 条
[1]  
Bajorek S., 2008, TRACE V5. 0 Theory manual, field equations, solution methods and physical models
[2]   Stochastic uncertainty quantification for safety verification applications in nuclear power plants [J].
Boafo, Emmanuel ;
Gabbar, Hossam A. .
ANNALS OF NUCLEAR ENERGY, 2018, 113 :399-408
[3]   The alternate mitigation strategies on the extreme event of the LOCA and the SBO with the TRACE Chinshan BWR4 model [J].
Chen, Chun-Yu ;
Shih, Chunkuan ;
Wang, Jong-Rong .
NUCLEAR ENGINEERING AND DESIGN, 2013, 256 :332-340
[4]   Quantification of LOCA core damage frequency based on thermal-hydraulics analysis [J].
Cho, Jaehyun ;
Park, Jin Hee ;
Kim, Dong-San ;
Lim, Ho-Gon .
NUCLEAR ENGINEERING AND DESIGN, 2017, 315 :77-92
[5]  
Farawila Y. M., 2017, P557, Patent No. [14/997,2017, 14997]
[6]  
Farawila Y. M., 2019, U. S. Patent Application, Patent No. [10/176,897, 10176897]
[7]  
Farawila Y. M., 2015, P 16 INT TOP M NUCL
[8]   AP1000® SBLOCA simulations with TRACE code [J].
Montero-Mayorga, J. ;
Queral, C. ;
Gonzalez-Cadelo, J. .
ANNALS OF NUCLEAR ENERGY, 2015, 75 :87-100
[9]   Confirmation of Wilks' method applied to TRACE model of boiling water reactor spray cooling experiment [J].
Mui, Travis ;
Kozlowski, Tomasz .
ANNALS OF NUCLEAR ENERGY, 2018, 117 :53-59
[10]  
Pettersson K., 2009, NUCL FUEL BEHAV LOSS