Radioactive inventory, dose rate and thermal power evaluation by calculus for spent fuel in the WWR-S research reactor

From July 1957 to July 1997, in the IFIN-HH research reactor 152 fuel assemblies of EK-10 type and 68 fuel assemblies of S-36 type were burned. Lately, an increased water radioactivity was found in storage ponds due of fission products and structural materials corrosion. For this reason and taking into account the necessity of the reactor decommissioning in the near future, a solution is required for the long term storage of the spent fuel, probably by dry method. The project for dry storage of the spent fuel requires the knowledge of the radioactive inventory, gamma dose rate and thermal power for each fuel assembly, at different moments from the shut-down of the reactor. To this end, it was necessary to gather and process all information concerning the spent fuel in the reactor: fuel assembly inventory, constructive details, material contents, irradiation history. The created data base is updated continuously. The calculation methodology for the radioactive inventory, thermal power and gamma dose rate allows the easy determination of these quantities for new situations in the future. For this data base the ORIGEN code was assimilated and used. Gamma dose rates for different locations around the fuel assemblies, in the air, were calculated using a point Kernel method via MERCURE code. The spectral indices REZ and FAST were calculated using the WIMS code. The results concerning the radioactive inventory and thermal power were obtained separately for structural materials, actinides and fission products. The main contribution to radioactive inventory is due to fission products, 102 Ci/fuel assembly, while that of structural materials is 10-2 to 10-3 Ci/fuel assembly and that of actinides is 1 to 5 Ci/fuel assembly. The same trend is also found in case of thermal power, i.e., 0.1 to 1 W/fuel assembly, 10-3 to 10-2 W/fuel assembly and 10-2 W/fuel assembly, respectively. Dose rates arising from the fuel assembly were calculated for both situations, in the air and inside the original lead flask. The modeling used for this calculation followed as close as possible the real geometry of fuel assembly and flask. The results obtained were validated by comparison with the values obtained during the transportation of the fuel elements towards the external ponds. For C-52 fuel assembly, which is a more intense gamma source, the dose rates 0.3 · 106 mSv/h, at the top of fuel assembly, to 0.3 · 10-2 mSv/h, outside the container wall, were determined. (authors)