MCNPX–BUCAL1 code to code verification through burnup analysis

Abstract

The availability of accurate burnup data is an essential first step in any systematic approach to enhancement of economics, safety and performance of a research reactor. This first step requires the utilization of a well verified burnup code system. In this work a newly home-developed burnup code called BUCAL1 is presented. The code provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP (version 5c). BUCAL1 has the capability of using several depletion calculation schemes that do not exist in several other burnup code systems such as: shuffling, refueling and multicycles burnup calculation, in an automatic way.The accuracy and precision of BUCAL1 were tested for U-Zrh fuels, by a code to code verification with MCNPX2.7, by predicting the burnup parameters of the 2MW TRIGA Mark II Moroccan research reactor. Continuous energy cross section data from the more recent nuclear data evaluation ENDF/B-VII.0 as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. Analysis of the verification results shows that BUCAL1 is enough accurate to be used in burnup calculations.