Capture and fission cross sections for /sup 235/U in the ''unresolved resonance'' energy region were evaluated to permit determination of local-average resonance parameters for the ENDF/B-IV cross section file. Microscopic data were examined for infinitely dilute average fission and capture cross sections and also for intermediate structure unlikely to be reproduced by statistical fluctuations of resonance widths and spacings within known laws. Evaluated cross sections, averaged over lethargy intervals greater than 0.1, were obtained as an average over selected data sets after appropriate renormalization. Estimated uncertainties are given for these evaluated average cross sections. The ''intermediate'' structure fluctuations common to a few independent data sets were approximated by straight lines joining successive cross sections at 120 selected energy points; the cross sections at the vertices were adjusted to reproduce the evaluated average cross sections over the broad energy regions. Data sources and methods are reviewed, output values are tabulated, and some modified procedures are suggested for future evaluations. Evaluated fission and capture integrals for the resolved resonance region are also tabulated. These are not in agreement with integrals based on the resonance parameters of ENDF/B versions III and IV. 8 tables, 5 figures.
Peelle, R.W.. Evaluation for ENDF/B-IV of the neutron cross sections for /sup 235/U from 82 eV to 25 keV. United States: N. p., 1976.
Web. doi:10.2172/7189870.
Peelle, R.W.. Evaluation for ENDF/B-IV of the neutron cross sections for /sup 235/U from 82 eV to 25 keV. United States. doi:10.2172/7189870.
Peelle, R.W.. Sat .
"Evaluation for ENDF/B-IV of the neutron cross sections for /sup 235/U from 82 eV to 25 keV". United States.
doi:10.2172/7189870. https://www.osti.gov/servlets/purl/7189870.
@article{osti_7189870,
title = {Evaluation for ENDF/B-IV of the neutron cross sections for /sup 235/U from 82 eV to 25 keV},
author = {Peelle, R.W.},
abstractNote = {Capture and fission cross sections for /sup 235/U in the ''unresolved resonance'' energy region were evaluated to permit determination of local-average resonance parameters for the ENDF/B-IV cross section file. Microscopic data were examined for infinitely dilute average fission and capture cross sections and also for intermediate structure unlikely to be reproduced by statistical fluctuations of resonance widths and spacings within known laws. Evaluated cross sections, averaged over lethargy intervals greater than 0.1, were obtained as an average over selected data sets after appropriate renormalization. Estimated uncertainties are given for these evaluated average cross sections. The ''intermediate'' structure fluctuations common to a few independent data sets were approximated by straight lines joining successive cross sections at 120 selected energy points; the cross sections at the vertices were adjusted to reproduce the evaluated average cross sections over the broad energy regions. Data sources and methods are reviewed, output values are tabulated, and some modified procedures are suggested for future evaluations. Evaluated fission and capture integrals for the resolved resonance region are also tabulated. These are not in agreement with integrals based on the resonance parameters of ENDF/B versions III and IV. 8 tables, 5 figures.},
doi = {10.2172/7189870},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sat May 01 00:00:00 EDT 1976},
month = {Sat May 01 00:00:00 EDT 1976}
}