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A method is presented for calculating thermal neutron fluxes in the primary shields of reactor systems which eliminates reliance on mock-up experimental data. A multigroup P/sub 1/ approach is ernployed with the spatial dependence of the neutron sttenuation adjusted through use of a point source attenuation kernel for a homogeneous hydrogenous medium. Comparison of calculation with experiment is presentad. (auth).
A method is presented which has been used to calculate "thermal" neutron fluxes in reactor shields. Application of the method to the calculation of neutron spectra and other shielding quantities is discussed.
Included are 687 selected references to unclassified reports and scientific journal literature on radiation shields and shielding. Author, report number, and subject indexes are also included.