The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification

Abstract

The article presents information on the validation and verification (V&V) of the first version (V1) of the EUCLID integrated code intended for safety analysis of operating or designed liquid metal (sodium, lead, or lead–bismuth) cooled reactors under normal operation and under anticipated operational occurrences by carrying out interconnected neutronics, thermal–mechanical, and thermal–hydraulic calculations. The list of processes and phenomena that have to be modeled in the integral code for correctly describing the above-mentioned operating conditions is given. Based on this list, the most high-quality experimental data are selected for carrying out the validation. It is shown that, for sodium cooled reactors, a significant number of experiments was carried out around the world on studying individual thermal–hydraulic processes and phenomena, which made it possible to perform validation of the thermal–hydraulic module. The validation of the code—as applied to description of processes that take place in fuel rods with oxide or nitride fuel and gas gap—is carried out against the results of post-pile investigations of fuel rods irradiated in fast sodium cooled research and power-generating reactors. The obtained results opened up the possibility to determine the errors of calculating such fuel rod parameters as release of gaseous fission products from the fuel and sizes of pellet and cladding in a limited range of burnup values. To perform validation of the neutronics module as applied to calculation of such parameters as power density distribution over the core and decay heat release, a sufficient number of experiments and benchmarks were selected. The results obtained from experimental operating conditions of a BN-600 reactor and startup conditions of a BN-800 reactor made it possible to estimate how correctly the integral code performs calculations of interconnected thermal–hydraulic and neutronic processes. Only a limited set of experimental investigations is available for heavy liquid metal cooled reactors. In view of this circumstance, programs for obtaining the lacking data are developed. To estimate the quality with which the experiments are modeled by means of the EUCLID/V1 integrated code, a procedure for evaluating the errors of calculation results is developed. In accordance with this procedure, the error of calculating the parameters playing the main role in the reactor safety assessment is evaluated.