Analysis of uncontrolled withdrawal of control rods group accident based on three-dimensional neutronics/thermal-hydraulic coupling method

BackgroundStrong interaction between neutronics and thermal-hydraulic phenomena exists in the core under reactivity initiated accidents for VVER (Vodo-Vodyanoy Energetic Reactor) units. Three-dimensional simulation considering coupling effect is particularly important in the accident analysis.PurposeThis study aims to simulate the process and consequences of reactivity initiated accidents in detail and accurately analyze uncontrolled withdrawal of control rods group accident for VVER based on three-dimensional neutronics/thermal-hydraulic coupling method.MethodsThree-dimensional neutronics/thermal-hydraulic coupling method including internal coupling type and loosely-coupled data transfer scheme was adopted in this study. Firstly, three-dimensional neutronics code combined with interface functions was compiled into dynamic link library (DLL), and quasi three-dimensional thermal-hydraulic system code was taken as the leading role to implement the coupling of neutronics and thermal-hydraulic codes. Then, the reliability of the coupled code was validated using the trip of one out of four reactor coolant pump sets test conducted in the VVER-1000 power plant, and calculation and analysis on the uncontrolled withdrawal of control rods group accident during minimal controlled power level were performed based on the coupled code. Finally, the core neutronics model and system thermal-hydraulic model were built with consideration of the uncertainty of neutronics parameters, code-to-code comparison was carried out to validate these models.ResultsCalculation results indicate that sequence of events and extreme values of safety-relevant important parameters are close to those of FSAR (Final Safety Analysis Report). Response processes of major thermal-hydraulic parameters in the transient generally agree well with those of FSAR. Acceptance criteria under the considered accident condition are satisfied. The protective automatic actions of safety systems bring the reactor plant into a safe state. The axial relative power peak in the core hot channel assembly is 3.07. The minimum DNBR (Departure from Nucleate Boiling Ratio) in the core is 2.34, and the maximum temperature of the fuel rods is 1 061.3 ℃, both of which remain a large margin compared with the limited values.ConclusionsResults of this study demonstrate that three-dimensional neutronics/thermal-hydraulic coupling method is of great significance to the accurate and reliable simulation on the reactivity initiated accidents for VVER units.