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R-Matrix Analysis of 235U Neutron Transmission and Cross-Section Measurements in the 0- to 2.25-keV Energy Range

L. C. Leal, H. Derrien, N. M. Larson, R. Q. Wright

Nuclear Science and Engineering

Volume 131 / Number 2 / February 1999 / Pages 230-253


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A new R-matrix analysis of the 235U cross-section data in the 0- to 2250-eV energy region is presented. The analysis was performed with the SAMMY computer code that has recently been updated to permit, for the first time, inclusion of both differential and integral data within the analysis process. Fourteen differential data sets and six integral quantities were used in this evaluation: two measurements of fission plus capture, one of fission plus absorption, six of fission alone, two of transmission, and one of eta, plus standard values of thermal cross sections for fission and capture, and of K1 and the Westcott g factors for both fission and absorption. An excellent representation was obtained for the high-resolution transmission, fission, and capture cross-section data as well for the integral quantities. The result is a single set of resonance parameters spanning the entire range up to 2250 eV, a decided improvement over the present ENDF/B-VI evaluation, in which 11 discrete resonance parameter sets are required to cover that same energy range. This new evaluation is expected to greatly improve predictability of the criticality safety margins for nuclear systems in which 235U is present.

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