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| Author | : M. Divadeenam |
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| Total Pages | : 96 |
| Release | : 1979 |
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| Release | : 1979 |
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A completely new evaluation of the nickel neutron induced reaction cross sections was undertaken as a part of the ENDF/B-V effort. (n, xy) reactions and capture reaction time from threshold to 20 MeV were considered for 58 6° 61 62 64Ni isotopes to construct the corresponding reaction cross section for natural nickel. Both experimental and theoretical calculated results were used in evaluating different partial cross sections. Precompound effects were included in calculating (n, xy) reaction cross sections. Experimentally measured total section data extending from 0.7 MeV to 20 MeV were used to generate smooth cross section. Below 0.7 to MeV elastic and capture cross sections are represented by resonance parameters. Inelastic angular distributions to the discrete isotopic levels and elemental elastic angular distributions are included in the evaluated data file. Gamma production cross sections and energy distribution due to capture and the (n, xy) reactions were evaluated from experimental data. Finally, error files are constructed for all partial cross sections.
| Author | : M. Divadeenam |
| Publisher | : |
| Total Pages | : 96 |
| Release | : 1979 |
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| Total Pages | : 135 |
| Release | : 1997 |
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Separate evaluations have been done for the three stable isotopes of silicon for ENDF/B-VI. The evaluations are based on analysis of experimental data, supplemented by results of nuclear model calculations. The computational methods and the parameters required as input to the nuclear model codes are reviewed. Discussion of the evaluated data given for resonance parameters, neutron induced reaction cross sections, associated angular and energy distributions, and gamma-ray production cross sections is included. Extensive comparisons of the evaluated cross sections to measured data are shown in this report. The evaluations include all necessary data to allow KERMA (Kinetic Energy Released in MAterials) and displacement cross sections to be calculated directly. These quantities are fundamental to studies of neutron heating and radiation damage.
| Author | : Ivan Sirakov |
| Publisher | : |
| Total Pages | : 8 |
| Release | : 2013 |
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| ISBN | : 9789279284212 |
An ENDF-6 compatible evaluation for neutron induced reactions in the resonance region has been completed for 106,108,110,111,112,113,114,116Cd. The parameters are the result of an analysis of experimental data available in the literature together with a parameter adjustment on transmission and capture data obtained at the time-of-flight facility GELINA. Complete evaluated data files in ENDF-6 format have been produced by joining the evaluations in the resonance region with corresponding files from the JEFF-3.1.2 nuclear data library (or with the JEFF-Beta-CAD proposed evaluation in case of 113Cd). These files were produced for use in the JEFF32T2 library. For neutron induced reactions in the unresolved resonance region the JENDL 4.0 evaluation for 111Cd and 113Cd was adopted. The evaluated files have been processed with the latest updates of NJOY. 99 to test their format and application consistency as well as to produce a continuous-energy data library in ACE format for use in Monte Carlo codes. The ACE files have been utilized to study the effect of the evaluated resonance parameters on results of integral experiments. The evaluated files will be implemented in the next release of the JEFF-3 library which is maintained by the Nuclear Energy Agency of the OECD.
| Author | : Ivan Sirakov |
| Publisher | : |
| Total Pages | : 9 |
| Release | : 2013 |
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| ISBN | : 9789279283192 |
A new evaluation of neutron induced reactions on 197Au nucleus in the energy regions below 500 eV and from 4 keV to 100 keV is presented. Complete evaluated data files in ENDF-6 format have been produced by joining the evaluation with corresponding files from the ENDF/B-VII. 1 library. The evaluation in the unresolved resonance region between 4 keV and 100 keV is based on a generalized single-level representation compatible with the energy-dependent option of the ENDF-6 format. The average partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser-Feshbach statistical reaction theory including width fluctuations. The transmission coefficients have been obtained from a combined analysis of the capture cross section resulting from the cross section standards evaluation project and theoretical nonfluctuating cross sections derived from a dispersive coupled channel optical model. The evaluated cross sections have been validated by a comparison with transmission and capture data obtained at the time-of-flight facility GELINA. The evaluated files have been processed with the latest updates of NJOY. 99 to test their format and application consistency as well as to produce a continuous-energy data library in ACE format for use in Monte Carlo codes. The ACE files have been utilized to study the effect of the evaluated resonance parameters on results of lead slowing-down experiments. The evaluated files will be implemented in the next release of the JEFF-3 library which is maintained by the Nuclear Energy Agency of the OECD.
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| Release | : 1987 |
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Nuclear model codes were used to compute cross sections for neutron-induced reactions on both /sup 58/Ni and /sup 60/Ni for incident energies from 1 to 20 MeV. The input parameters for the model codes were determined through analysis of experimental data in this energy region. Discussion of the models used, the input data, the resulting calculations, extensive comparisons to measured data, and comparisons to the Evaluated Nuclear Data File (ENDF/B-V) for Ni (MAT 1328) are included in this report. 118 refs., 101 figs., 19 tabs.
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| Total Pages | : 31 |
| Release | : 2000 |
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Using existing experimental data for neutron-induced total, elastic, inelastic, reaction and fission cross sections, as well as results from nuclear model calculations and evaluations from nuclear reaction data libraries, we derived an estimate for the cross sections for the 235U(n,2n) and 239Pu(n,2n) reactions for the neutron energy range from threshold to approximately 12 MeV. In effect, our approach is based on subtracting the fission and inelastic cross sections from the total reaction cross section where the difference is expected to yield the (n,2n) cross section. In addition to this subtraction approach, a ratio method and a differential method have also been explored. For 235U(n,2n), as a test case, we arrive at a cross section consistent with previous measurements, and for 239Pu(n,2n) we obtain a peak value of 400 " 60 mb for the incident neutron energy range of 10 (less-than or equal to) E{sub n} (less-than or equal to) 12 MeV.
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| Total Pages | : 10 |
| Release | : 2013 |
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| ISBN | : 9789279285394 |
An ENDF-6 compatible evaluation for neutron induced reactions in the resonance region has been completed for 182,183,184,186W. The parameters are the result of an analysis of experimental data available in the literature together with a parameter adjustment on transmission and capture data obtained at the time-of-flight facility GELINA. Complete evaluated data files in ENDF-6 format have been produced by joining the evaluations in the resonance region with corresponding files from the JEFF- 32T1 and ENDF/B-VII. 1 library. The evaluated files have been processed with the latest updates of NJOY. 99 to test their format and application consistency as well as to produce a continuous-energy data library in ACE format for use in Monte Carlo codes. The evaluated files will be implemented in the next release of the JEFF-3 library which is maintained by the Nuclear Energy Agency of the OECD.
| Author | : David T. Goldman |
| Publisher | : |
| Total Pages | : 736 |
| Release | : 1968 |
| Genre | : Neutron cross sections |
| ISBN | : |
