Speaker
Description
The accurate and efficient mapping of the radiation environment in a nuclear fusion reactor requires the most advanced radiation transport tools. The Monte Carlo method has long been deployed to deal with the complexity of fusion relevant geometries, with MCNP the adopted industry standard code among the European and wider international community in this field. However, reliance on a single code, particularly relating to the real and perceived limitations of MCNP and strict licensing conditions have driven explorations into alternatives. It is imperative that the transport codes meet: i) stringent modelling and analysis requirements for fusion, and ii) may be used within an integrated engineering design workflow that can support ITER, DEMO, STEP analysis as well as experimental devices such as JET and MAST. The radiation transport codes, Serpent, OpenMC as well as the framework for allowing CAD based transport, DAGMC, are being actively researched and have shown promise to improve the efficiency of at least one aspect of the current best practices in fusion neutronics workflows. To date the codes have largely been explored by researchers in isolation through simplified (and in limited cases, more complex) fusion benchmarks against MCNP and to a lesser extent to experiment. The paper will present a detailed assessment of the explored transport codes adopting both CSG and CAD based workflows as well as rigorously validating the ability to interface with nuclear data, including reading in and correctly interpreting the data captured in ACE file format. Specifically, FENDL and JEFF neutron transport libraries will be compared in determining several fusion relevant responses such as DPA, nuclear heating and particle flux for several fusion-relevant benchmarks. The study will be extended to the evaluation of dose rates through interfacing with the MCR2S method calling TENDL activation files. A suitable set of metrics for assessing the relative capabilities of the transport codes has already been established, as well as identifying and preparing a workflow for converting the computational and experimental models, consistent run parameters and tallies into the relevant code input formats. For the latter, several experiments from the SINBAD database will be explored, with contributions from this work supporting the activities of the Working Party on International Nuclear Data Evaluation Co-operation Subgroup 47 (WPEC SG47). The capabilities and outstanding needs for the application of each respective code to full-scale fusion neutronics will be concluded.