Methods for processing and use of thermal neutron scattering data in OpenMC

Abstract

OpenMC, an open source Monte Carlo particle transport code, relies on its own nuclear data format that is based on the HDF5 file format. The Python API within OpenMC includes an openmc.data module, which handles generation of thermal scattering HDF5 libraries starting from either ACE or ENDF files. When starting from ENDF files, NJOY is automatically executed to produce intermediate ACE files. In most cases, the resulting HDF5 data is equivalent to that stored in an ACE file. For incoherent elastic scattering, however, a novel sampling method is derived from the double-differential cross section and can be used to preserve the true continuous angular distribution. In this case, the characteristic bound cross section and Debye–Waller integral are taken directly from an ENDF file. Testing on the H in ZrH evaluation from ENDF/B-VIII.0 has demonstrated how the new sampling method matches the true distribution whereas the traditional treatment based on discrete cosines can leaves significant gaps in the angular distribution.