Research On Transient Mechanism Of Reactor Coolant Pump And Reactor Coolant System Piping Under Shaft Stuck Accident Condition

Ensuring safety is a critical requirement for stable operation of nuclear power plants,and this paper focuses on studying the dynamic response of the reactor coolant pump and its coolant system under specific accident conditions.In the event of a system blockage,the rotor of the reactor coolant pump can rapidly stall,leading to a sudden decrease in coolant flow in the affected loop.This results in a rapid change in fluid pressure and density at the pump inlet,causing a water hammer pressure wave to propagate rapidly through the pipeline.In this paper,the actual transient flow process and the real-time change law of the parameters inside the system under the accident conditions are accurately reproduced for the key technical problems of the Hualong-1reactor coolant pump and its reactor coolant system.Firstly,the hydraulic model of the Hualong-1 reactor coolant pump was selected and the technical specifications of the calculated model were verified by test methods.To simplify the steam generator tube bundle area and pressure vessel core area,the equal flow section method was applied.Additionally,the energy balance principle was utilized to regulate the system’s operating conditions by controlling the flow radius of resistance elements and matching the resistance of pipeline overflow components.The results demonstrate that the relative error between theoretical and non-steady-state computational resistance values of each component is less than 5%.Furthermore,the reactor coolant pump’s flow deviation from the rated value is less than 2%during normal system operation.Based on these considerations,a simplified three-dimensional model of the three-loop reactor coolant system was established.Furthermore,the secondary development technology of the user-defined functions in CFX software was utilized to achieve the dynamic response of fluid medium density and the linear regulation of reactor coolant pump speed during shaft stuck accidents,which provided essential input for numerical calculations.Secondly,the transient external characteristics of the reactor coolant pump in each loop of the system were compared,and entropy production theory was used to analyze the energy loss characteristics of reactor coolant pumpⅠin the accident loop.The results showed that during the shaft stuck accident,there was an abrupt change in the flow rate,torque,and head of reactor coolant pumpⅠwithin a short period of time.Backflow in loopⅠbegan at about 1 second,and the system reached dynamic equilibrium again at 3 seconds.During this period,reactor coolant pumpⅠunderwent a variety of changes in working conditions.The inertial energy of the unit and the kinetic energy of the fluid were transformed into hydraulic work and unit friction work.Additionally,there was a large velocity gradient inside and at the boundary of unstable flow regions,such as impact,vortex,and deliquescence.This was an important factor that triggered irreversible energy loss in the pump.In addition,the pressure fluctuation characteristics of the reactor coolant pumpⅠin the accident loop were analyzed using the wavelet transform method.The findings indicate that a large vortex at the inlet of the impeller working surface of reactor coolant pumpⅠblocked the flow channel after the shaft stuck accident.The high pressure area resulting from the impact caused a sudden change in local fluid density,triggering the water hammer phenomenon in the system pipeline.As the impeller speed decreases,the high frequency band wavelet response caused by the interference between the static and dynamic rotors in each flow channel of reactor coolant pumpⅠis gradually destroyed,and the low frequency band wavelet response caused by the water hammer effect dominates.As the flow rate further decreases,the wavelet response in the low frequency band in the same flow channel weakens and tends to be consistent,and when the system reaches dynamic equilibrium again,the frequency band centered on the rotor pass frequency becomes the main disturbance frequency.Finally,an accident safety assessment method for the reactor coolant system under shaft stuck conditions was developed,and the transient mechanism of the piping in each loop was analyzed.The results showed that,following a shaft stuck accident,the local sudden pressure change in the system pipeline was transmitted as acoustic waves,and induced sudden hydraulic loads acted on the wall surface of each elbow.The maximum load force peak was found to be at the W1_Ⅰsub-wall surface at the first elbow of the transition section within loopⅠ,with a value of 3.152×10~6N.In the accident loop,the peak pressure wave gradually decreased along the direction of pressure wave transmission on the steam generator heat transfer tube elbow with the weakest wall,and the energy of the pressure wave was gradually dissipated during the transmission process.The maximum transient pressure peak was located at the monitoring position out4 of the heat transfer tube re56-3,with a value of 15.9928 MPa.