Evaluation of local safety parameters in nodal coupled neutronics/thermal hydraulics reactor analysis

The present paper gives a description of recent developments related to the evaluation of local safety parameters in nodal reactor calculations. These new methods are realized in the coupled neutron kinetics/thermal hydraulics program system PANBOX. It is designed for steady-state and transient analysis of pressurized water reactors with rectangular or hexagonal fuel assembly geometry. The underlying physical models and solution methods for neutronics and thermal hydraulics are presented. In particular, the intra-nodal flux reconstruction and the subsequent local thermal margin and DNB evaluation methodology are discussed in some detail. The capability to determine thermal safety margin based on local hot channel fuel pin values is especially valuable in safety analyses for transients resulting in highly non-uniform power density distributions and/or low mass flow through the core. It can also serve for high accuracy core design applications, achieving economically efficient reload strategies. (author)