Preliminary thermal-hydraulic safety analysis of Tehran research reactor during fuel irradiation experiment

摘要

Following domestic fabrication of nuclear fuels in Iran, it is necessary to investigate fuel material behavior, fission gas release, fuel swelling, cladding material behavior and fuel integrity of domestic fuels at different burnup in a research reactor during irradiation. Currently, Tehran research reactor is the sole operating research reactor which can be used for fuel irradiation experiments in the country. In this regard, standard codes as well as developed complementary computer programs are applied to verify thermal-hydraulic performance of irradiating a domestic rod-type fuel assembly of natural UO_2 pellets in Tehran research reactor core, which itself contains 20% enriched plate-type U_3O_8-Al fuels. Maximum temperatures of fuel, clad and coolant, onset of nucleate boiling, onset of flow instability and departure from nucleate boiling during irradiation experiment are investigated by subchannel analysis as indicators to verify the reactor core safe operation during the experiment. The results give the confidence that during this irradiation experiment, thermal-hydraulic steady state safety criteria of the mixed-core are satisfied and the fuel irradiation experiment does not induce any significant operational change.