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 keff 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.