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 + ^6,7Li 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.

OpenMC是社区开发的蒙特卡洛中子和光子传输模拟代码。它能够基于连续能核横截面数据执行固定源粒子迁移模拟。在这项工作中，OpenMC已针对基于加速器的中子源（例如国际聚变材料辐照设施-面向DEMO的中子源设施（IFMIF-DONES））的应用进行了基准测试和验证。为此，McDeLicious代码是MCNP Monte Carlo代码的扩展，具有在评估的d + ^6,7的基础上模拟氘锂中子源的能力。Li截面数据，已迁移到OpenMC代码。通过与通过MCNP6-McDeLicious获得的结果进行比较，已对OpenMC-McDeLicious进行了计算，以计算非碰撞中子和光子通量及光谱。辛巴达实验基准和d-Li实验已用于验证OpenMC。这些测试和验证表明OpenMC和MCNP代码以及实验数据之间具有良好的一致性。此外，OpenMC-McDeLicious用于在IFMIF-DONES中子源的辐射测试模块中获得核响应，例如，每原子排量（DPA），产气，核加热。大多数OpenMC-McDeLicious结果显示与5％的MCNP6-McDeLcious达成了良好的协议。这表明适用于基于加速器的中子源设施的OpenMC代码的成功基准测试和验证。