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- W3148492097 abstract "OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It is capable of performing fixed source particle transport simulation based on continuous-energy nuclear cross-section data. In this work, OpenMC has been benchmarked and verified for the application to accelerator-based neutron sources, e.g. the International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source facility (IFMIF-DONES). To this end, the McDeLicious code, which is an extension of the MCNP Monte Carlo code with the capability to simulate the deuterium-lithium neutron source on the basis of evaluated d + 6,7 Li cross section data, has been migrated to the OpenMC code. The OpenMC-McDeLicious has been tested for the calculation of uncollided neutron and photon flux and spectra by comparing to results obtained with MCNP6-McDeLicious. SINBAD experimental benchmarks and d -Li experiments have been used for validating OpenMC. These tests and verifications show good agreement between OpenMC and MCNP code, as well as the experimental data. In addition, OpenMC-McDeLicious is used for obtaining the nuclear responses, e.g. Displacement Per Atom (DPA), gas production, nuclear heating, in the irradiation test module of the IFMIF-DONES neutron source. Most OpenMC-McDeLicious results show good agreement with MCNP6-McDeLcious within 5%. This indicates the successful benchmarking and validation of OpenMC code for the applications to accelerator-based neutron source facilities." @default.
- W3148492097 created "2021-04-13" @default.
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- W3148492097 date "2021-09-01" @default.
- W3148492097 modified "2023-10-16" @default.
- W3148492097 title "Benchmarking and verification of the OpenMC code for accelerator-based neutron source analyses" @default.
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- W3148492097 doi "https://doi.org/10.1016/j.fusengdes.2021.112512" @default.
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