Evaluation of {sup 28,29,30}Si Neutron Induced Cross Sections for ENDF/B-VI.

Evaluation of {sup 28,29,30}Si Neutron Induced Cross Sections for ENDF/B-VI.
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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.

Evaluation Of[sup 28,29,30]i Neutron Induced Cross Sections for ENDF/B-VI.

Evaluation Of[sup 28,29,30]i Neutron Induced Cross Sections for ENDF/B-VI.
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Total Pages:
Release: 2001
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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.

ENDF-6 Compatible Evaluation of Neutron Induced Reaction Cross Sections for [sup]106,108,110,111,112,113,114,116Cd

ENDF-6 Compatible Evaluation of Neutron Induced Reaction Cross Sections for [sup]106,108,110,111,112,113,114,116Cd
Author: Ivan Sirakov
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Total Pages: 8
Release: 2013
Genre:
ISBN: 9789279284212
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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.

ENDF-6 Compatible Evaluation of Neutron Induced Reaction Cross Sections for 182,183,184,186W.

ENDF-6 Compatible Evaluation of Neutron Induced Reaction Cross Sections for 182,183,184,186W.
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Total Pages: 10
Release: 2013
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ISBN: 9789279285394
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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.

Calculated Cross Sections for Neutron Induced Reactions on Sup 19 F and Uncertainties of Parameters

Calculated Cross Sections for Neutron Induced Reactions on Sup 19 F and Uncertainties of Parameters
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Total Pages: 73
Release: 1990
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Nuclear model codes were used to calculate cross sections for neutron-induced reactions on 19F for incident energies from 2 to 20 MeV. The model parameters in the codes were adjusted to best reproduce experimental data and are given in this report. The calculated results are compared to measured data and the evaluated values of ENDF/B-V. The covariance matrix for several of the most sensitive model parameters is given based on the scatter of measured data around the theoretical curves and the long-range correlation error of measured data. The results of these calculations form the basis for the new ENDF/B-VI fluorine evaluation. 44 refs., 64 figs., 14 tabs.

Description of Evaluation for /sup 63,65/Cu for ENDF/B-VI.

Description of Evaluation for /sup 63,65/Cu for ENDF/B-VI.
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Total Pages:
Release: 1987
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Isotopic evaluations for /sup 63,65/Cu performed for ENDF/B-VI are briefly reviewed. The evaluations are based on analysis of experimental data and results of model calculations which reproduce the experimental data. Evaluated data are given for neutron-induced reaction cross sections, angular and energy distributions, and for gamma-ray production cross sections associated with the reactions. File 6 formats are used to represent energy-angle correlated data and recoil spectra. Uncertainty files are included for the major cross sections. Full evaluations are given for /sup 63,65/Cu.

Evaluated Neutron-induced Cross Sections for /sup 40/Ca from 20 to 40 MeV.

Evaluated Neutron-induced Cross Sections for /sup 40/Ca from 20 to 40 MeV.
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Release: 1982
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Nuclear model codes were used to compute cross sections for neutron-induced reactions on /sup 40/Ca for incident energies from 20 to 40 MeV. The input parameters for the model codes were determined through analysis of experimental data in this energy region. Computed cross sections along with emission spectra for each product were combined into an Evaluated Nuclear Data File (ENDF) using the proposed format for charged-particle reactions. Discussion of the models used, the resulting calculations, and the final evaluated data file are presented.