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Numerical Study of Thermal hydraulics Characteristics of 900 MW PWR
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[1]WANG Wei-shu,GUO Hui-jun,LIANG Cheng-sheng,et al.Numerical Study of Thermal hydraulics Characteristics of 900 MW PWR[J].Journal of Zhengzhou University (Engineering Science),2015,36(01):129-.[doi:10.3969/ j. issn.1671 -6833.2015.01.011]
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Journal of Zhengzhou University (Engineering Science)[ISSN 1671-6833/CN 41-1339/T] Volume: 36卷 Number of periods: 2015 01 Page number: 129- Column: Public date: 2015-01-10
Title:
Numerical Study of Thermal hydraulics Characteristics of 900 MW PWR
Author(s):
WANG Wei-shu1; GUO Hui-jun1; LIANG Cheng-sheng1; 2; XU Wei-hui1
1.Thermal Engineering Research Center，North China University of Water Resources and Electric Power, Zhengzhou 450011,China;2. Hebei Huare Engineering Design co. Ltd.,Shijiazhuang 050000，China
Keywords:
nuclear reactor; subchannel analysis;  heat transmission in the core; thermal hydraulic
CLC:
TL331
DOI:
10.3969/ j. issn.1671 -6833.2015.01.011
Abstract:
The steady-state thermal analysis model of reactor core was established for a 900MW pressurizedwater reactor. The steady - state thermal-hydraulic of reactor core was calculated and analyzed with COBRA-IV. The temperature of fuel element,coolant flow distribution and temperature and the departure from nucleateboiling ratio ( DNBR ) of the reactor core were obtained. The results show that the coolant in the core exits lat-eral flow from the center to the around. The maximum temperature of coolant in the core outlet is up to 338.2℃. The maximum temperature of fuel in the core is up to 1 350 ℃. The maximum temperature of claddingsurface and fuel pellet appears above the center. The DNBR near the inlet is much higher than near the outlet,and the minimum DNBR appears near the center.
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