Analysis of Thermal Reactor Benchmarks with Design Codes Based on ENDF/B-V Data

Thermal reactor benchmark calculations have been performed with the “design codes” EPRI-CELL and PDQ using ENDF/B-V cross-section data. The purpose of these calculations is to determine the quality of ENDF/B-V data for predicting reactor parameters when used with methods typically employed for power reactor analysis. This verification is essential if ENDF/B-V cross sections are to be used widely by the nuclear industry for reactor design, core reload, and core-follow studies. It appears that ENDF/B-V, when used in typical reactor design codes, is an accurate data set for light water reactor analysis. Computed resonance integrals and reaction ratios for ²³⁸U seem to be slightly high but are within the uncertainty. The average k_eff obtained for a diverse set of 27 UO₂ and MO₂ critical configurations is 1.002 ± 0.002. Critical UO₂ eigenvalues are consistently slightly overestimated, on the average by 0.2%. The average eigenvalue obtained for the mixed-oxide lattices is 1.0007 with a standard deviation of 0.0023. Plutonium isotopic ratios generally show good agreement with measured values obtained from burned power reactor pins.