TY - GEN
T1 - Transient analysis of cp300 based on relap5/mod3.4
AU - Jianchang, Liu
AU - Yue, Nina
AU - Xu, Xie
AU - Sun, Ducheng
AU - Wenxi, Tian
AU - Su, Guanghui
AU - Qiu, Suizheng
N1 - Publisher Copyright:
Copyright © 2014 by ASME.
PY - 2014
Y1 - 2014
N2 - The thermal-hydraulic characteristics of Nuclear Power Plant (NPP) during Steam Generator Tube Rupture (SGTR) accident are of great concern. In this paper, the thermalhydraulic characteristics of CP300 during SGTR accident with no operator actionsduring the first 30 min are investigated using the best estimate RELAP5/MOD3.4. Modeling and nodalization of CP300 was executed, including the vessel, pumps, pressurizer, steam generations and necessary auxiliary systems. Some main transient parameters were obtained, such as Reactor Coolant Pump (RCP) coolant temperature and pressure, steam generator flow rate and pressure. The calculation results gives the sequences of the NPP during the SGTR as described below. As the tube rupture occurs, the primary pressure drops and secondary pressure increases. When the primary pressure drops down to the setpoint of the scram, the control rods drop down and the power of the NPP begins to decrease, causing the primary coolant temperature to decrease. The primary pressure continues to drop. When it drops down to 10.78MPa, the High pressure Safety Injection System is put into operation and the accident is mitigated, even without operator's actions. In conclusion, the calculated results indicate that the key thermal hydraulic parameters of CP300 during the SGTR accident are in acceptable ranges and the accident is effectively controlled..
AB - The thermal-hydraulic characteristics of Nuclear Power Plant (NPP) during Steam Generator Tube Rupture (SGTR) accident are of great concern. In this paper, the thermalhydraulic characteristics of CP300 during SGTR accident with no operator actionsduring the first 30 min are investigated using the best estimate RELAP5/MOD3.4. Modeling and nodalization of CP300 was executed, including the vessel, pumps, pressurizer, steam generations and necessary auxiliary systems. Some main transient parameters were obtained, such as Reactor Coolant Pump (RCP) coolant temperature and pressure, steam generator flow rate and pressure. The calculation results gives the sequences of the NPP during the SGTR as described below. As the tube rupture occurs, the primary pressure drops and secondary pressure increases. When the primary pressure drops down to the setpoint of the scram, the control rods drop down and the power of the NPP begins to decrease, causing the primary coolant temperature to decrease. The primary pressure continues to drop. When it drops down to 10.78MPa, the High pressure Safety Injection System is put into operation and the accident is mitigated, even without operator's actions. In conclusion, the calculated results indicate that the key thermal hydraulic parameters of CP300 during the SGTR accident are in acceptable ranges and the accident is effectively controlled..
UR - https://www.scopus.com/pages/publications/84911870574
U2 - 10.1115/ICONE22-30440
DO - 10.1115/ICONE22-30440
M3 - 会议稿件
AN - SCOPUS:84911870574
T3 - International Conference on Nuclear Engineering, Proceedings, ICONE
BT - Thermal Hydraulics
PB - American Society of Mechanical Engineers (ASME)
T2 - 2014 22nd International Conference on Nuclear Engineering, ICONE 2014
Y2 - 7 July 2014 through 11 July 2014
ER -