Cross-Section Evaluations to 150 MeV for Accelerator-Driven Systems and Implementation in MCNPX

Abstract

New accelerator-driven technologies that utilize spallation neutrons, such as the production of tritium and the transmutation of radioactive waste, require accurate nuclear data to model the performance of the target/blanket assembly and to predict neutron production, activation, heating, shielding requirements, and material damage. To meet these needs, nuclear-data evaluations and libraries up to 150 MeV have been developed for use in transport calculations to guide engineering design. By using advanced nuclear models that account for details of nuclear structure and the quantum nature of the nuclear scattering, significant gains in accuracy can be achieved below 150 MeV, where intranuclear cascade calculations become less accurate. Evaluations are in ENDF-6 format for important target/blanket and shielding materials (isotopes of H, C, N, O, Al, Si, P, Ca, Cr, Fe, Ni, Cu, Nb, W, Hg, and Pb) for both incident neutrons and incident protons. The evaluations are based on measured data as well as predictions from the GNASH nuclear model code, which calculates cross sections using Hauser-Feshbach, exciton, and Feshbach-Kerman-Koonin preequilibrium models. Elastic scattering distributions and direct reactions are calculated from the optical model. All evaluations specify production cross sections and energy-angle correlated spectra of secondary light particles as well as production cross sections and energy distributions of heavy recoils and gamma rays. A formalism developed to calculate recoil energy distributions is presented. The use of these nuclear data in the MCNPX radiation transport code is also briefly described. This code merges essential elements of the LAHET and MCNP codes and uses these new data below 150 MeV and intranuclear cascade collision physics at higher energies. Extensive comparisons are shown between the evaluated results and experimental cross-section data to benchmark and validate the evaluated library. In addition, integral benchmarks of calculated and measured kerma coefficients for neutron energy deposition and neutron transmission through an iron slab compared with MCNPX calculations are provided. These evaluations have been accepted into the ENDF/B-VI library as Release 6.