Publications Repository - Helmholtz-Zentrum Dresden-Rossendorf

BGCore reactor analysis system, recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge, couples the Monte Carlo neutronic code (MCNP4C) with an independently developed burnup and decay module SARAF. The BGCore utilizes multi-group approach for generation of one group cross-sections. According to this approach, only multi-group neutron spectrum is calculated by MCNP, while reaction rates are calculated in a separate subroutine using pre-generated multi-group cross-section set and the fine group neutron spectrum obtained from MCNP.
BGCore code system offers a number of advantages over similar MCNP-depletion codes. These include:

• Multi-group coupling approach significantly reduces the code execution time without compromising the accuracy of the results.
• Use of the most recent data based on JEFF-3.1 data files,
• Careful choice of about 1700 isotopes to cover all potentially significant aspects of fuel irradiation and decay. All nuclides are included in the calculation matrix with no asymptotic approximation,
• The fact that all of the isotopes, and not just the most neutronically important, are tracked in the BGCore throughout all depletion steps, allows calculations of post-irradiation fuel characteristics such as, activity, radiotoxicity, and decay heat with high degree of accuracy.

Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (T-H) feedback into the calculation scheme. Recently, a new T-H feedback module (THERMO) was developed and integrated into the BGCore system. At each computation point, the THERMO module receives as an input the power and the fuel burnup distributions in the core from the neutronic solver (MCNP) and calculates the temperatures distribution in core components and the coolant flow distribution in the core channels. The results of the THERMO calculations can be used for updating the relevant parameters in the MCNP input, such as fuel and moderator temperatures, and moderator density.
This study presents the results of coupled neutronic T-H analysis of full PWR core performed with BGCore system. The verification of BGCore system results against alternative state of the art computer code is also presented.