Neutron-photon Multigroup Cross Sections for Neutron Energies Less Than Or Equal to 400 MeV.


Book Description

Multigroup cross sections (66 neutron groups and 21 gamma ray groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, /sup 10/B, /sup 11/B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten and lead. These cross sections are available from the Radiation Shielding Information Center of the Oak Ridge National Laboratory. 19 references.




Neutron-photon Multigroup Cross Sections for Neutron Energies Less Than Or Equal To400 MeV. Revision 1


Book Description

Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, /sup 10/B, /sup 11/B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs.




Neutron-photon Multigroup Cross Sections for Neutron Energies [60 MeV.


Book Description

Multigroup cross sections (47 n-groups, 21 .gamma.-groups) in ANISN format for neutron energies from thermal to 60 MeV and for the elements H, /sup 10/B, /sup 11/B, C, O, Si, Ca, Cr, Fe, and Ni are described. A P/sub 5/-Legendre expansion is used at energies equal to or greater than 14.9 MeV and a P/sub 3/-Legendre expansion is used at energies equal to or less than 14.9 MeV. Calculated results of the dose equivalent vs. depth in the shield from a point isotropic source at the center of a spherical shell (366 cm thick) heavy-concrete (density = 3.6 g cm/sup -3/) shield are presented. The energy distribution of the source neutrons is approximately that from a Li(D, n) neutron radiation damage facility.










Nuclear Cross Sections for 95-Mev Neutrons


Book Description

The total cross sections of twelve different elements were measured using the neutron beam from the 184-in. cyclotron, operating with deuterons. Bismuth fission ionization chambers were employed as both monitor and detector in conventional 'good geometry' attenuation measurements in the neutron flux emerging from the 3-in. diameter collimating port in the 10-ft-thick concrete shielding. The mean energy of detection of the neutrons in this experiment is estimated to be 95 Mev. Measurements were also made with a monitor and detector placed inside the concrete shielding where an intense neutron flux over a large area could be obtained. Attenuators of four different elements were placed in front of the detector in a 'poor geometry' arrangement so that attenuation was due essentially to inelastic collisions which degrade the neutron energy below the fission threshold. A second detector was placed outside the concrete shielding In the collimated neutron beam in line with the neutron source, absorber, and first detector. Attenuation in it is caused by both inelastic and elastic scattering. By this arrangement the ratio of inelastic to total cross section can be determined directly in one experiment. The nuclear radii as calculated from the observed cross section, using the theory of the transparent nucleus, vary as 1.38 x 10(exp-13) A(exp(1/3)) cm. In this energy range the ratios of the inelastic to total cross sections are all less than one-half.