RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

被引:15
|
作者
Trivedi, A. K. [1 ]
Allison, C. [2 ]
Khanna, A. [1 ]
Munshi, P. [1 ]
机构
[1] Indian Inst Technol, Nucl Engn & Technol Program, Kanpur 208016, Uttar Pradesh, India
[2] Innovat Syst Software, Idaho Falls, ID 83406 USA
关键词
SAFETY; CODE;
D O I
10.1016/j.nucengdes.2016.05.018
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
The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA-TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved all steady state thermal hydraulic conditions as reported in the DCD. The analysis has been performed for the primary safety criteria, the peak clad temperature (PCT) as it is well established for a PWR that oxidation and hydrogen generation do not violate the safety criteria as long as PCT is under the safe limit. Results from this study show that the calculated value for the PCT is 1229 K well below the acceptance criteria of 1477 K, lower than the DCD value of 1311 K and higher than the TRACE value of 1186 K. (C) 2016 Elsevier B.V. All rights reserved.
引用
收藏
页码:222 / 229
页数:8
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