Download or read book EVALUATION OF 238U NEUTRON CROSS SECTIONS FOR THE ENDF B VERSION II FILE written by and published by . This book was released on 1971 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Program RECENT version 79 1 written by Dermott Edward Cullen and published by . This book was released on 1979 with total page 44 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Evaluation of Fission Cross Sections and Covariances 233U 235U 238U 239Pu 240Pu and 241Pu written by and published by . This book was released on 2000 with total page 88 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Evaluation of the 238U Neutron Cross Sections for Incident Neutron Energies Up to 4 KeV written by and published by . This book was released on 1978 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report describes an evaluation of the 238U cross sections below 4 keV. Recent measurements and reanalyses of older data are discussed. Evaluated resonance parameters are obtained for 164 s-wave and 280 p-wave levels. The capture widths of the first three s-wave levels are significantly lower than in the ENDF/B-IV evaluation. The s-wave strength function above 1.5 keV is sytematically larger than in ENDF/B-IV. Statistical and systematic uncertainties are evaluated for the resonance parameters and for the smooth backgrounds. The statistical distributions of the resonance parameters are compared with theoretically expected distributions.

Download or read book EVALUATED NEUTRON CROSS SECTIONS OF sup 23 Na FOR THE ENDF B FILE written by and published by . This book was released on 1968 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Nuclear Data for Science and Technology written by Syed M. Qaim and published by Springer Science & Business Media. This book was released on 2012-12-06 with total page 1041 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book describes the Proceedings of the International Conference on Nuclear Data for Science and Technology held at Jillich in May 1991. The conference was in a series of application oriented nuclear data conferences organized in the past under the auspices of the Nuclear Energy Agency-Nuclear Data Committee (NEANDC) and with the support of the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP). It was the fIrst international conference on nuclear data held in Germany, with the scientific responsibility entrusted to the Institute of Nuclear Chemistry of the Research Centre Jillich. The scientific programme was established by the International Programme Committee in consultation with the International Advisers, and the NEA and IAEA cooperated in the organization. A total of 328 persons from 37 countries and fIve international organizations participated. The scope of these Proceedings extends to a wide range of interdisciplinary topics dealing with measu rement, calculation, evaluation and application of nuclear data, with a major emphasis on numerical data. Both energy and non-energy related applications are considered and due attention is given to some fundamental aspects relevant to the understanding of nuclear data.

Download or read book EVALUATED NEUTRON CROSS SECTIONS OF sup 240 Pu FOR THE ENDF B FILE written by and published by . This book was released on 1968 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Proceedings written by and published by . This book was released on 1971 with total page 502 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book ENDF 6 Compatible Evaluation of Neutron Induced Reaction Cross Sections for 182 183 184 186W written by and published by . This book was released on 2013 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.

Download or read book Program GROUPLE version 79 1 written by Dermott Edward Cullen and published by . This book was released on 1980 with total page 76 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book ENDF 6 Compatible Evaluation of Neutron Induced Reaction Cross Sections for sup 106 108 110 111 112 113 114 116Cd written by Ivan Sirakov and published by . This book was released on 2013 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.

Download or read book CROSS SECTION EVALUATIONS FOR ENDF B VII written by M. HERMAN and published by . This book was released on 2006 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: This is the final report of the work performed under the LANL contract on neutron cross section evaluations for ENDF/B-VII (April 2005-May 2006). The purpose of the contract was to ensure seamless integration of the LANL neutron cross section evaluations in the new ENDF/B-VII library. The following work was performed: (1) LANL evaluated data files submitted for inclusion in ENDF/B-VII were checked and, when necessary, formal formatting errors were corrected. As a consequence, ENDF checking codes, run on all LANL files, do not report any errors that would rise concern. (2) LANL dosimetry evaluations for {sup 191}Ir and {sup 193}Ir were completed to match ENDF requirements for the general purpose library suitable for transport calculations. A set of covariances for both isotopes is included in the ENDF files. (3) Library of fission products was assembled and successfully tested with ENDF checking codes, processed with NJOY-99.125 and simple MCNP calculations. (4) KALMAN code has been integrated with the EMPIRE system to allow estimation of covariances based on the combination of measurements and model calculations. Covariances were produced for 155,157-Gd and also for 6 remaining isotopes of Gd.

Download or read book Experimental and Evaluated Nuclear Plus Interference Cross Sections for Light Charged Particles written by S. T. Perkins and published by . This book was released on 1980 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1976 with total page 1118 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Evaluation of the Cross Sections of Iron Endf b Mat 1101 written by D. C. Irving and published by . This book was released on 1970 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: The neutron cross sections for iron have been evaluated and placed in the ENDF/B format with the material (MAT) number 1101. The total cross section was evaluated from recent experimental measurements and is compatible with measurements from very thick sample penetration made at the Tower Shielding Facility. Other data were taken from evaluations by J.J. Schmidt as contained in the UKAEA Nuclear Data File, DFN 64. (Author).

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1974-03 with total page 1100 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Evaluation Of sup 28 29 30 i Neutron Induced Cross Sections for ENDF B VI written by and published by . This book was released on 2001 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.