Development Of Accident Analysis Model For Nuclear Power Plant Based On RELAP5 And GOTHIC

In the research of nuclear power design and general design review(GDA),evaluating the mass energy release of the main system and the response of pressure and temperature in the containment under the breach accident are one of the core tasks of nuclear power plant safety analysis.At present,studying the response of the main system’s mass energy release and containment pressure and temperature during a nuclear power plant breach accident is calculated by separating the main system analysis program(RELAP5)from the containment system analysis program(GOTHIC).This process requires Repeated iterative calculations are tedious and time-consuming,and the working conditions are often set to the most serious accidents during the calculation.The calculated results are too conservative and do not belong to the best assessment.The designed nuclear power plant has low economic operation.In order to make the calculation results closer to the real situation,this article aims to establish a coupling interface between RELAP5 and GOTHIC,so that the two programs can transfer data to each other in real time,and jointly calculate,so as to ease the single program during the transient and accident analysis of the nuclear power system.Conservative calculations and improved calculation accuracy.In this paper,the link between RELAP5 and GOTHIC program is realized by using static link library as a data exchange bridge.After analyzing the programming structure and data storage mode of RELAP5,the RELAP5 program data output and input model is established.Then,the user interface and modeling principles of GOTHIC are studied,and the data output and input models of GOTHIC are determined.Because the target coupling program should have complete universality,it should not only be able to simulate the one-way transmission of the vapor-water mixture between the two programs,but also be able to simulate the reverse flow of the coolant,and it can simulate the two-way transmission of a small amount of non-condensable gas in the fluid.The data that needs to be exchanged between the two programs.In addition,for the case where different physical property reference points may be used for the same substance in different programs,the internal calculations of the programs are also used to reconcile.For the difficulty that the time steps used in different programs are generally not synchronized,a linear interpolation method is used for processing..In order to verify the correctness of the coupled program in data transfer,coupling calculations were performed on several typical examples,mainly to check the correctness of the data exchange between the two programs.The calculation results show that the coupled nodes in the two programs are calculated The properties and composition parameters of the produced fluid are completely consistent,which verifies the correctness of the data transmission of the coupling program.Finally,this coupling program was used to simulate the accident conditions of a small marine reactor(ACPR50S)of China General Nuclear Power Group(CGN),and the calculation results were compared with the results of uncoupled calculations.The comparison results showed that the coupling program calculations The results are more reasonable,release the safety margin,and provide a reference for the subsequent design of nuclear power safety systems.