Modification And Verification Assessment Of Related Models In Licensing LOCA Analysis Program Platform Development

In order to develop licensing LOCA analysis program platform for nuclear power plant (NPP), by comparing two different licensing LOCA analysis approaches, a relatively realistic technique have been adopted in this thesis, which is to satisfy the requirements of conservative evaluation models (EM) in related regulation (10CFR50, Appendix K) by modifying related models or correlations of the best estimate (BE) program RELAP5/MOD3, in an effort to form a licensing LOCA analysis tool. By using this"advanced program platform + conservative EM"technique, the author attempts to make LOCA analysis on the NPP, of which LOCA used to be conservatively evaluated, so that further potential of thermal safety margins should be achieved.By comparing study with Appendix K, ten models or correlations in original RELAP5/MOD3 program are identified which need to be modified, invoked, evaluated or verified. Modifications or invoking control have been separately introduced on such models or correlations as fission product decay model, correlations for critical heat flux (CHF), post-CHF heat transfer correlations, logic of prevention from returning to nuclear boiling and transition boiling heat transfer prior to reflood, discharge model, metal-water reaction model, emergency core cooling (ECC) bypass model and related models for pressurized water reactor (PWR) refill and reflood.With the accomplishment of modification or invoking the above-mentioned models, separate verifications by utilizing proper standard curve or separate test data have been carried out. The standard curve or separate tests data includes: (1) 1971 fission product decay standard curve for verifying reasonability and conservatism of fission product decay model modification; (2) related ORNL THTF test data for evaluating modified CHF correlations, post-CHF heat transfer correlations and logic of prevention from returning to nuclear boiling and transition boiling heat transfer prior to reflood; (3) Marviken test-22 for discharge model modification verification; (4) Cathcart oxidation test for verifying metal-water reaction model modification; and (5) Westinghouse FLECHT - SEASET test data for refill and reflood models modification verification, etc.Further, for the integral test LOFT L2-5, calculations with both original RELAP5/MOD3 program and the program of which some related models modified in compliance with the requirement of 10CFR50 Appendix K have been made. The results indicate that, on the one hand, ability of the original BE program simulating the accident progress is identified; and on the other hand, the modified models are confirmed to be conservative. Moreover, both individual contributions and combined effects of the model modification to conservatism of the integral analysis as a whole are observed and analyzed.To assess the effect of this"advanced BE program + conservative model"analysis method on integral LOCA analysis, calculations with the modified program for both hypothetical small break loss of coolant accident and large break loss of coolant accident of Qinshan NPP phase I are conducted. Compared with the calculated results of original RELAP5/MOD3 program and that delivered in Final Safety Analysis Report (FSAR) of Qinshan NPP, mutual and integrated effects with different model modifications are investigated. And finally, safety margin achieved by the present calculation is preliminarily discussed.