Neutron Transport Simulations of RBMK Fuel Assembly Using Multigroup and Continuous Energy Data Libraries within the SCALE Code

The neutron transport simulations of RBMK-1500 fuel assembly were performed using both multigroup and continuous energy data libraries available within the SCALE code system in order to validate its suitability for the estimation of RBMK neutronic characteristics. The resonance processing of cross section, involved in the preparation of the multigroup data library, has a significant impact on neutron transport calculations. Standard Dancoff factors (DFs) used for the heterogeneous geometry of RBMK fuel assembly are insufficient for the accurate estimation of resonance self-shielding. Thus, the SCALE module MCDancoff was used in this study to determine location-specific DFs. The results of RBMK-1500 fuel assembly simulations using standard and user-defined DFs were compared. In addition, the continuous energy (CE) cross-section data library was applied for the benchmark calculations. The impact of different nuclear data libraries on neutron transport simulations was tested as well. It was found out that the usage of the multigroup data libraries generates some deviation from the reference simulations obtained with CE libraries. The CE library based on the estimated ENDF/B-VII.1 data proved to be the best alternative for neutron transport simulations of RBMK fuel assembly.

Modelling of RBMK Spent Nuclear Fuel Characteristics

After the final shutdown of Ignalina NPP, total amount of spent nuclear fuel is approximately 22 thousands of fuel assemblies. Radionuclide content and its characteristics in spent fuel are initial data for analysis of various safety related areas such as shielding, thermal analysis, radioactive releases and other processes. Experimental investigations of radionuclide content and characteristics in spent nuclear fuel are complicated and expensive, therefore numerical evaluation methods are widely used. Numerical modelling of spent RBMK fuel characteristics was performed using TRITON code from SCALE 6.1 system. Activities of fission products and actinides, gamma and neutron sources, decay heat obtained with TRITON code are compared with previous modelling results obtained using SAS2H sequence from the former SCALE 4.3 version. Some evaluated parameters are compared with published experimental data for RBMK spent nuclear fuel.

RBMK Fuel Rod Modelling and Investigation of Process During Operation Cycle

This paper presents the analysis determining the status of fuel rods after whole normal operation. The FEMAXI–6 code was selected for such analysis. Evaluating the specifics of RBMK fuel rods, the adaptation of code was provided. After the adaptation of FEMAXI-6 code, the single fuel rod model of RBMK-1500 was developed and the processes, which occur during whole life of fuel rods, were analyzed. For this analysis the fuel rod from fuel channel with average initial power (2.5 MW) was selected. After (normal) operation the fuel rods from the reactor are transferred to the spent fuel pool and the state of the fuel rods (intactness of cladding, residual stresses in the cladding and fuel pellets, gap between cladding and pellets and etc.) is very important, because fuel rod cladding is one of the safety barriers. In this paper the stresses in cladding, plastic deformation of cladding and other parameters were calculated using FEMAXI-6 and method of final elements. The performed analysis demonstrates possibility to identify state of fuel rods after normal operation that is necessary for long-term fuel storage in spent fuel pools.