Advanced methodologies for modeling and assessment of SFR safety functions

The present doctoral work was performed to contribute to the conceptual design development and safety assessment of a Generation IV Sodium Fast Reactor (SFR) in the frame of the European Sodium Fast Reactor Safety Measures Assessment and Research Tools (ESFR-SMART) project. The contribution to the safety assessment is divided into two main areas. 1) To the development of new safety assessment methodologies using mainly Phenix and Superphenix reactor data, and 2) to the new safety measures assessment of the European Sodium Fast Reactor (ESFR). The safety assessment of ESFR was structured according to the safety functions the specific analyzed safety measure relates to, belonging mainly to the heat removal and reactivity control safety functions.In the first part of the work, the conceptual design of the ESFR was synthesized by using the operational experience of former reactor designs and design options from reactor concepts currently under development. Based on these concepts, the safety measures were optimized and updated for use within the ESFR conceptual design. Within this part, a 3-Dimensional (3D) Computer-Aided Design model has been prepared, which served as the base for input preparation for much of the safety assessment within the research.In the safety assessment, various calculation tools were involved, either already available for SFR analysis or newly developed methods as part of the doctoral work. The first of these newly developed tools is the currently available coupled TRACE-PARCS neutronic/thermal-hydraulic analysis toolset being further advanced by introducing a unique XS preparation technique at PSI. Through this development, increased accuracy 3D core power evaluation can be performed for transient simulations. The second tool developed in the research is the core mechanics analysis methodology, allowing general core distortion evaluation and the calculation of the resulting reactivity effect of the deformation. As part of the development, benchmark analysis or verification and validation work have been performed, testing the accuracy of the abovementioned techniques.The ESFR safety measure analysis started with the various decay heat removal system (DHRS) assessment, which is one of the key new safety measures of ESFR, belonging to the reactivity control safety function. The decay heat removal capability of the new DHRSs were assessed, using the Protected Station Blackout (PSBO) accidental scenario, together with the general reactor behavior moving from forced to natural convection. In the study, each of the DHRS performance was evaluated as well as the system temperature evolution during the simulated accidental condition. Thus, it was assessed if the specific DHRS can keep the reactor from reaching the defined temperature limits or if certain modifications are required to achieve this, for which recommendations were made. Finally, an unprotected loss of flow (ULOF) simulation has been performed, assessing the core behavior under such conditions. In the study, the low void effect core safety measure of ESFR was specifically targeted, corresponding to the reactivity control safety function. The sodium boiling progression has been assessed in the core for the reference SA design, as well as a modified design, affecting the vapor propagation in the SA.