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E 1.99: endf-273
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Cross sections for the Cu(n,xn) and Cu(n,x. gamma. ) reactions between 1 and 20 MeV |
1 |
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E 1.99: endf-274
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VITAMIN-E an ENDF/B-V multigroup cross-section library for LMFBR core and shield, LWR shield, dosimetry and fusion blanket technology. |
1 |
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E 1.99: endf-277
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/sup 238/U neutron-induced fission cross section for incident neutron energies between 5 eV and 3. 5 MeV |
1 |
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E 1.99: endf-278
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Summary of fission spectrum workshop held at the National Neutron Cross Section Center, Brookhaven National Laboratory, October 23, 1978. [Cross sections, review, neutron spectra] |
1 |
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E 1.99: endf-280
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Status of ENDF/B-V neutron emission spectra induced by 14-MeV neutrons. [Partial and total spectra, cross sections, graphs] |
1 |
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E 1.99: endf-283
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Evaluated data for n + ⁹Be reactions. [10⁻⁵ eV to 20 MeV] |
1 |
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E 1.99: endf-286
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Evaluation of natural chromium neutron cross sections for ENDF/B-V |
1 |
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E 1.99: endf-287
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FCXSEC multigroup cross-section libraries for nuclear fuel cycle shielding calculations. |
1 |
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E 1.99: endf-288
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Comparison of photon-production processing codes LAPHNGAS, MACK-IV, and NJOY |
1 |
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E 1.99: endf-289
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Measurement of the average number of prompt neutrons emitted per fission of ²³⁵U relative to ²⁵²Cf for the energy region 500 eV to 10 MeV |
1 |
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E 1.99: endf-290
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Preliminary study of pseudorandom binary sequence pulsing of ORELA |
1 |
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E 1.99: endf-291
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COVERX service module of the FORSS system. [LMFBR] |
1 |
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E 1.99: endf-294
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Ni elemental neutron induced reaction cross-section evaluation |
1 |
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E 1.99: endf-295
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Processing ENDF/B-V uncertainty data into multigroup covariance matrices |
1 |
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E 1.99: endf-297
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User's guide for SAMMY a computer model for multilevel r-matrix fits to neutron data using Bayes' equations. |
1 |
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E 1.99: endf-298
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Evaluation of resonance parameters for neutron interaction with iron isotopes for energies up to 400 keV |
1 |
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E 1.99: endf-299
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Evaluation of cross sections for neutron-induced reactions in sodium. [10⁻⁵ eV to 20 MeV] |
1 |
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E 1.99: endf-300
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Standard reference and other important nuclear data by the Cross Section Evaluation Working Group |
1 |
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E 1.99: endf-303
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GLUCS a generalized least-squares program for updating cross section evaluations with correlated data sets. [In FORTRAN IV for PDP-10] |
1 |
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E 1.99: endf-308
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Calculation of neutron and gamma-ray production cross sections for calcium from 8 to 20 MeV |
1 |
