Generation of few energy group cross-sections for some of generation-IV reactors

Abstract EN

Routine neutron calculations for a nuclear power reactor are usually performed using deterministic transport and diffusion codes.

Such codes need few group cross sections.

The group cross-sections are generated using transport codes.

For the global reactor calculations one needs cell parameters averaged over the fuel assembly.

The reactor assembly geometry and materials are reactor type dependent.

In the present work few group cross sections averaged over the fuel assembly are generated using MCNP5 for five types of generation IV reactors namely; HTGR, HTGR+BP, SCWR, UOX/MOX, and SFR together with PWR and VVER.

The results are compared with previous calculations.

The present results have relative statistical error in the range of 2% to 4%.

The present results are compared with published calculations and a good agreement was obtained.