Coupled Neutron and Gamma Heating Calculation Based on VARIANT Transport Solutions

Abstract

For accurate assessment of nuclear heating in fast reactors, a new coupled neutron and gamma heating calculation scheme has been developed based on VARIANT nodal transport solutions of neutron and gamma flux distributions. The MC²-3 code was extended to generate multigroup neutron and gamma cross sections and kinetic energy release in materials (KERMA) factors, and a utility program CURVE was developed to reconstruct detailed pin and duct wall powers from VARIANT output files. The improved heating calculation scheme has been verified against MCNP6 Monte Carlo reference solutions for the Advanced Burner Test Reactor (ABTR) and Experimental Breeder Reactor II (EBR-II) benchmark problems. Compared to the existing coupled heating calculation method based on DIF3D diffusion theory solutions, the new heating calculation scheme utilizes more accurate gamma cross sections and KERMA factors, accounts for the transport effects, and eliminates the approximations in the existing pin power reconstruction scheme. As a result, it produces more accurate assembly and pin power distributions. For both the ABTR and EBR-II problems, the maximum assembly power error was ~1% in fuel assemblies and ~2% in instrumented structure assemblies, and the maximum error in pin segment powers in an axial node of fuel assembly was ~4%. In the blankets of the EBR-II problem, the maximum error in pin segment powers was increased to ~8%, mainly due to the lower power level and the relatively large error in the nodal power of the VARIANT solution.

Keywords:

• Sodium-cooled fast reactor
• coupled neutron and gamma heating
• variational nodal transport
• pin power reconstruction

Acknowledgments

This research was performed using funding received from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy University Program (DE-NE0008426).