monte carlo with fuel burnup method for the enhs benchmark calculations

Abstract

Estimates of the uncertainties arising from approximations in the methods used in different nuclear data processing and neutron transport codes are usually obtained by inter-comparing calculations made using different code systems. This paper gives details of an investigation of differences between results obtained by using different codes for a single zone model of the Encapsulated Nuclear Heat Source (ENHS) benchmark core fuelled with metallic alloy of Pu, U, and Zr. The ENHS is a new lead-bismuth or lead cooled novel reactor concept for 20 effective full power years without refuelling and with very small reactivity swing. The computational tools benchmarked include MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluation; KENO-V.a/ORIGEN2.1 code system, recently developed by authors of this paper, with the ENDFB-V based 238 group library; the design-oriented procedure based on the simplified one-dimensional (1D) geometry model and SAS2H control module; and the well-established fast reactor neutronics design tools in use at Argonne National Laboratory. Calculations made for the ENHS benchmark have shown that the differences between the results obtained when using different code schemes are quite significant and should be taken into account in assessing the quality of the nuclear data library.