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A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d + 6,7 Li cross section data. A new code McDeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The McDeLicious code was checked against available experimental data and calculation results of McDeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that McDeLicious along with newly evaluated d + 6,7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.

The MARS15(2012) is the latest version of a multi-purpose Monte-Carlo code developed since 1974 for detailed simulation of hadronic and electromagnetic cascades in an arbitrary 3-D geometry of shielding, accelerator, detector and spacecraft components with energy ranging from a fraction of an electronvolt to 100 TeV. Driven by needs of the intensity frontier projects with their Megawatt beams, e.g., ESS, FAIR and Project X, the code has been recently substantially improved and extended. These include inclusive and exclusive particle event generators in the 0.7 to 12 GeV energy range, proton inelastic interaction modeling below 20 MeV, implementation of the EGS5 code for electromagnetic shower simulation at energies from 1 keV to 20 MeV, stopping power description in compound materials, new module for DPA calculations for neutrons from a fraction of eV to 20-150 MeV, user-friendly DeTra-based method to calculate nuclide inventories, and new ROOT-based geometry., Comment: 7 p

The CASCADE/INPE code system, developed jointly by the Joint Institute of Nuclear Search and IATE, is intended for calculating the transport of charged particles and neutrons in the energy range from several GeV down to thermal energy in media with a complicated geometry and composition. It includes as individual blocks the CASCADE and MCNP-4B programs and an interface-program. The CASCADE block is used to model the interaction of charged particles with target nuclei and particle transport in the medium at energies greater than a fixed cutoff energy (20 MeV for neutrons). Transport of neutrons with energy below 20 MeV is modeled using the MCNP-4B block. The interface program is intended for processing computational results using the CASCADE block and preparation of initial data for the MCNP-4B block. An example of the calculation of the characteristics of a subcritical reactor with an external neutron source is presented, 1 figure, 6 references.