Characterization of Fast Neutron Transmission Through an Iron Shield

In this paper we give an analysis of the neutron transmission through iron sphere using Monte Carlo and transport theory methods based on ENDF/B-VII.1 general purpose library. The motivation for investigation comes from a well-known deficiency in inelastic data older library evaluation (ENDF/B-V), giving concern fast flux underestimation within reactor pressure vessels. order to benchmark next-generation ENDF/B-VI data, U.S. Nuclear Regulatory Commission former Czechoslovakian National Research Institute have jointly preformed several experiments 1990s, addressing leakage spectra obtained 252Cf fission source centre sphere. It was shown that cross section, containing improvements over previous evaluations, will not entirely resolve spectrum discrepancies observed at high energies. Since safety analyses vessel embrittlement are often calculations specific multigroup section libraries, simulation performed hybrid shielding methodology ADVANTG3.0.3 MCNP6.1.1b codes. Comparison calculated referenced dosimeter activation rates presented "standard" nuclear reactions, used dosimetry. For purpose, new IRDFF-II special IAEA Data Services as reference dosimetry sections. code calculation reaction rates, which were also compared with IRDFF-1.05