Application of code scaling, applicability and uncertainty methodology to large break LOCA analysis of two loop PWR

In NED 119, No. 1 (May 1990) a series of six papers published by a Technical Program Group presented a new methodology for the safety evaluation of emergency core cooling systems in nuclear power plants. This paper describes the application of that new methodology to the LB LOCA analysis of the two loop Westinghouse power plant. Results of the original work were used wherever possible, so that the analysis was finished in less than one man year of work. Steam generator plugging level and safety injection flow rate were used as additional uncertainty parameters, which had not been used in the original work. The computer code RELAP5/MOD2 was used. Response surface was generated by the regression analysis and by the artificial neural network like Optimal Statistical Estimator method. Results were compared also to the analytical calculation.