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