R-matrix analysis of {sup 235}U neutron transmission and cross sections in the energy range 0 to 2.25 keV
This document describes a new R-matrix analysis of {sup 235}U cross section data in the energy range from 0 to 2,250 eV. The analysis was performed with the computer code SAMMY, 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, capture, and scattering, 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 as for the integral quantities. The result is a single set of resonance parameters spanning the entire range up to 2,250 eV, a decided improvement over the present ENDF/VI evaluation, in which eleven 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 {sup 235}U is present.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE, Washington, DC (United States); USDOE Office of Environmental Restoration and Waste Management, Washington, DC (United States)
- DOE Contract Number:
- AC05-96OR22464
- OSTI ID:
- 631239
- Report Number(s):
- ORNL/TM-13516; ON: DE98007037; TRN: 98:006994
- Resource Relation:
- Other Information: PBD: Nov 1997
- Country of Publication:
- United States
- Language:
- English
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