ENDF-6 Format Proposals for Dosimetry Library Improvements
In support of the IRDFF-II Dosimetry Library, this proposal suggests enhancements to the ENDF-6 format. Recommendations include revising restrictions on File 10 use, formalizing MT-261 for dosimetry spectra, and addressing URR interpolation issues to improve cross section data accuracy.
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Some ENDF-6 Format Proposals in Support of the IRDFF-II Dosimetry Library Andrej Trkov, Skip Kahler
File 10 General Description ENDF-6 Manual says: File 10 is allowed only for evaluations that represent the data for single isotopes. MF10/MT5 is used for radionuclide production in IRDFF-II Dosimetry library for natural elements where the same residual is produced by different reactions from different isotopes. Example: 58Ni(n,p), 60Ni(n,nd), 60Ni(n,2np), 60Ni(n,t), 62Ni(n,2nt), 62Ni(n,4np), 62Ni(n,3nd) all lead to 60Co Above 20 MeV the 60Ni(n,2np) reaction represents the dominant contribution to 58Co production. Proposal: Remove the last sentence in the first paragraph of Section 10.1 mentioned above, which administratively excludes the use of MF=10 for elemental data.
Define MT-261 for dosimetry spectra Standard and reference neutron fields are essential for cross section validation Operations on spectra are similar to those on cross sections (group-averaging, plotting, covariances) Non-cross section data exist in MF=3 (e.g. 251, 252, 253 ) Proposal Formalize and legalize the use of MT=261 to represent neutron fields in special purpose libraries
URR Interpolation Currently in the ENDF-6 Manual: In such cases, if two adjacent grid points differ by more than a factor of three, the processing code should add additional intermediate energy points at a spacing of approximately ten-per-decade and compute the cross sections at the intermediate points using parameter interpolation. This is a contradiction and nonsense that leads to ambiguous cross section reconstruction Proposal: Admit the wrong decision and prescribe interpolation of resonance parameters
