Development of constitutive equations in reactor safety analysis code with data-based modeling using artificial neural network

In a nuclear reactor safety evaluation process, it is too costly to experiment in the same scale with a commercial nuclear power plant. Therefore, the safety evaluation of a nuclear reactor relies on a safety analysis computer code substantially, whose accuracy directly affects the nuclear safety. The reactor safety analysis code is consisted of governing equations and constitutive equations. The constitutive equations in a reactor safety analysis code has high accuracy for simulating a separate effect test (SET). They are typically a result of experimental data regression with a mathematically limited form. Furthermore, SET can be deliberately used for improving constitutive relations’ accuracy. The code validation process also includes comparison of the code result with an integral effect test. If there is a mismatch between experiment results and simulation results, quantifying the cause and using the information to improve constitutive relations are not straightforward. Therefore, if a methodology which the accuracy of the constitutive relations is improved as the number of experimental data increases is developed, one can expect that the safety analysis code’s accuracy will automatically improved as more data is accumulated. This methodology can be developed using an artificial neural network that enables data-driven modeling and has less mathematical limitations. In the previous studies [1, 2], artificial neural networks were applied to replace the wall heat transfer coefficient, and wall friction coefficients in thermal hydraulic (TH) conditions. In this study, artificial neural networks (ANN) that substitute constitutive equations including interfacial heat transfer, interfacial friction are trained on the range that can cover wider TH conditions for analyzing design basis accidents. Methodology for the training data generation is developed to capture the twophase flow characteristics as much as possible. Also, the methodology for increasing the model accuracy is newly tested for wall heat transfer. The reference nuclear safety analysis code used in this study is MARS-KS.