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Techniques of fast neutron assessment will be described which are applicable when no assumption is made as to the flux distribution; as for example, assuming a fission spectrum. These techniques have been utilized to program the IBM 1620 Digital Computer. The cross section data as a function of energy and a first estimate of flux distribution is utilized in this program, as opposed to the customary technique of utilizing effective cross sections based upon a particularly assumed flux distribution. The computer program is iterative, taking the flux distribution calculated in each pass through the analysis and comparing it to the distribution assumed at the beginning of the calculations for that particular pass. If the differences are not sufficiently small, the resultant calculation of flux distribution is taken as a new estimate of the flux distribution and the calculations repeated for another pass. The output of the program is an estimate of the differential neutron flux spectrum measured in neutrons/cm2/sec/ev/watt. The energy range of the response of each foil selected for the experimental phase of this work is determined by the computer program based upon that foil’s cross section with respect to energy and the estimated flux distribution. The first few percent of the foil’s response, both at the lower and higher energy portions, are discarded; thus, for each successive iteration the total energy range of the foil’s response is truncated. This is done for three successive iterations. Experience has shown that the computer program converges adequately after a few iterations.
Neutron, flux measurement, foil detectors, duct shielding
Gardner, L. B.
Research Physicist, Physics and Mathematics Div., U. S. Naval Civil Engineering Laboratory, Port Hueneme, Calif.
Shoemaker, N. F.
Mathematician, Physics and Mathematics Div., U. S. Naval Civil Engineering Laboratory, Port Hueneme, Calif