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International audience; The Euratom Horizon-2020 project European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) aims to enhance the safety and performance of the European sodium-cooled fast reactor (ESFR) considering safety objectives envisaged for generation-IV reactors and the update of European and international safety frameworks, taking into account the Fukushima accident. Further, the project aims to support the development and validation of the computational tools for the situations to be considered at each defense-in-depth level in order to support the safety assessments using data produced in the project as well as selected legacy data. Within this activity, the focus is on the assessment and when needed further development of computer codes for the analysis of the sodium thermal-hydraulics phenomena in sodium cooled fast reactor (SFR) subassemblies under operational and accidental conditions including sodium boiling and transitional convection from forced to natural/mixed convection. In support of this activity, a review is performed for the sodium boiling models used in the codes participating in benchmark activities within the project. The objective of this paper is to summarize the result of the review, which encompasses both the phenomenological and mathematical models implemented in the codes. In particular, the review addresses the physical bases of sodium boiling models, the analytical models in connection with the numerical implementations in the codes, and the geometry representations ranging from a one-dimensional single channel and finite volume to full three-dimensional computational fluid dynamics (CFD) and porous media representations. Finally, the needs for further developments are discussed.

International audience; Abstract The computational codes used in the evaluation of the European sodium fast reactor—safety measures assessment and research tools(ESFR-SMART) reactor performance and specifically their sodium boiling models are assessed using two KNS-37 loss of flow (LOF) experiments, i.e., L22 and L29 tests, where boiling onset and two-phase flow regime up to dry-out occurred. The well-equipped KNS-37 experimental facility provided very valuable information for understanding the physical phenomena occurring in a 37-pin subassembly under LOF conditions, as well as experimental data to be used for computational tools validation. NATOF-2D, SAS-SFR, TRACE, ASTEC-Na, CATHARE-2, CATHARE-3, and NEPTUNE_CFD codes have been used in this exercise in order to compare the various boiling models and conclude on the advantages and limitations of them based on the comparison against the experimental data. Beyond boiling onset, the various sodium two-phase flow approaches determine the ability of the code to correctly represent the rewetting and voiding phases as well as cladding dry-out onset. A simulation performed by a computational fluid dynamics (CFD) approach (NEPTUNE_CFD code) taking into account liquid–vapor interfaces by an interface-tracking method is also shown and compared with the others approaches. Conclusions on each code performance are presented where the improvements needed to solve the issues encountered are included. This paper provides a first step in the process to investigate the required evaluation of the sodium two-phase flow models able to assess the safety of new SFR core designs (e.g., low void cores) under accidental conditions such as unprotected loss of flow (ULOF) transients.