Verification and validation of Monte Carlo simulations using swiss PWR HZP data

The use of Monte-Carlo (MC) simulations for reactor cores is currently of high interest. Resulting MC models would be relevant, e.g., for high-resolution assessments of the local neutron flux and power gradients, generally beyond the modelling capabilities of deterministic codes for regular full core simulations. Such MC simulations would be of particular importance for the advancement of reactor operation and safety assessments or, for example, to support designing new materials testing experimental programs at operational reactors. Nevertheless, MC-based core-follow burnup calculations are still very challenging for routine applications. Therefore, the LRT/PSI has developed in recent years a "cycle-check-up" (CHUP) concept, which allows the transfer of the operating conditions (coolant and fuel temperatures, density of moderator, boron concentration, position of control rods) and burned fuel isotopic compositions from validated reference core-follow models, based on the state-of-the-art deterministic codes CASMO/SIMULATE/SNF to MC codes, such as Serpent or MCNP®. Thanks to the availability of the validated deterministic core-follow models and access to additional reference data for Swiss power plants, such as start-up tests at PWRs, there is a valuable opportunity to extend the validation database for criticality safety evaluation methodology using Swiss reactors data. This paper presents the ongoing work towards the verification and validation (V&V) of MC full core simulations for burned configurations against results of the validated deterministic models, as well as real measurements for hot zero power (HZP) reactor conditions. (author)