Abstract :
The paper presents the methods developed and calculations carried out, using the SCALE4.3 code system to analyse the possibility of safe storage of the Romanian VVR-S reactor spent fuel load inside the CASTOR MTR2 cask. A method has been developed to derive specific problem-dependent nuclear libraries, defined for groups of fuel assemblies. The SAS2H calculation module has been used to simulate different irradiation histories of all spent fuel assemblies and to process the resulting libraries for input to ORIGEN-S code depletion-decay calculations. Using this procedure the characteristics of the spent fuel assemblies for six cooling times have been calculated. Obtained results are presented and discussed in the paper. The CSAS6 module with KENO-VI criticality code based on the 44-group energy cross-sections library has been used to solve criticality aspects. Some relevant problems to justify the calculation model have been studied and are presented in the paper. The sequence SAS4 with MORSE, a Monte Carlo code based on the (27n–18g)-coupled group energy cross-sections, has been used for solving the shielding problem. Gamma and neutron dose rates estimated on the surface of the container, and respectively at 1 m, 2 m, and 3 m in both radial and axial directions for normal loading, and for audit conditions are presented and discussed.
