Aiming at providing test problems that may be used to verify an adequate performance of the current version of a neutron and 7-ray transport computer code used for reactor core or shield calculations, we summarize the input data and the calculated results for three benchmark problems. The 1st problem deals with a 1-dimensional small spherical reactor for use to test 1-dimensional SN codes, DTF-lV and ANISN, and possibly the MORSE Monte Carlo code. The 2nd problem concerned with 2-dimensional (2, y) neutron propagation through an absorbing medium gives a severe test for 2-dimensional S,V codes, TWOTRAN-GG, TWOTRAN-IT, DOT-3 and TRIPLET. The last problem dealing with 2-dimensional (r, z) radiation streaming is to test also the finite difference SN codes, TWOTRAN-IT and DOT-3, and the finite element S.V code FEMRZ. The present article summarizes also the general tendency of the effect of parameters used for these calculations on the numerical results and computation time.
Neutron nuclear data of 233 U have been evaluated in the energy range from 10-5 eV to 20 MeV. Evaluated quantities are the total, fission, capture, elastic and inelastic scattering, (n, 2n) and (n, 3n) reaction cross sections, and the average numbers of prompt and delayed neutrons emitted per fission. The thermal and resonance cross sections have been evaluated on the basis of the measured data. The resolved resonance parameters are given up to 100eV and the unresolved resonance parameters between 100eV and 30keV. The total and fission cross sections have been evaluated in the higher energy region on the basis of the recently measured data, while the theoretical calculation with the optical, statistical and evaporation models has been used for evaluation of the other cross sections. The presently adopted optical potential parameters have reproduced well the experimental total cross section in the entire energy range as well as the measured data of the s-wa ve strength function. The structure observed in the vp values below 1 MeV is reproduced by the semi-empirical formula based on the fission fragment kinematics. The presently evaluated fission cross section is considerably lower than that of ENDF /8-IV between 10 and 50 keY. This low fission cross section is expected to resolve the kef! discrepancy pointed out from the benchmark tests in 233 U critical assemblies.
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