This study evaluates the feasibility and performance of uranium-zirconium hydride (U-ZrH x ) fuel in a graphite-free AGR-like Fluoride Salt-Cooled High-Temperature Reactor (AGR-like FHR). This reactor concept combines the pin-type geometry of Advanced Gas-cooled Reactors (AGR) with the low-pressure, high-efficiency characteristics of molten salt coolants and Hastelloy-N as cladding, aiming to improve thermal efficiency and passive safety. While conventional UO 2 fuel has been previously considered for this configuration, issues such as excessive fission gas release (FGR), swelling, and high fuel temperatures limit its viability. To address these challenges, a promising solution relies on using U-ZrH x fuel. The study explores its potential benefits, particularly its intrinsic neutron moderation, and includes an analysis of hydrogen transport accounting for diffusion, thermal gradients, and permeation across cladding. A new graphite-free 4-ring fuel pins cluster geometry was also introduced, increasing pin count while maintaining the core average power density; this increase from 36 to 60 pins reduces the pin-average power and linear heat rate proportionally (36/60 ≈ 0.60), thereby flattening the power distribution and reducing peak temperatures. The analyses were carried out using the OFFBEAT multidimensional fuel performance code and the Serpent Monte Carlo code for neutronics analyses. Simulation results show that U-ZrH 1.6 45 wt% U fuel used in the AGR-like FHR can potentially reduce peak fuel temperatures and azimuthal temperature variation, while maintaining a relatively constant temperature drop across the gap and internal pressure. Hydrogen migration to the gap and through the cladding can be mitigated through an inner liner coating. Verification against MARVEL fuel computational benchmark cases shows the accuracy of the hydrogen transport model. These findings are promising to support the viability of U-ZrH x fuel for the AGR-like FHR design, offering improved safety margins, mechanical stability, and simplified core geometry. • Extension of the OFFBEAT code to model U-ZrH x hydride fuel behaviour. • Development of a hydrogen transport model including diffusion and boundary loss. • Introduction of a graphite-free 4-ring AGR-like FHR core geometry. • U-ZrH x fuel reduces peak temperature by ∼ 250 ℃ and improves temperature uniformity. • Hydrogen migration identified as main pressure driver; liner coating mitigates leakage.

The Management and Uncertainties of Severe Accident (MUSA) project aimed to establish a harmonised approach for applying BEPU methods in severe accident analyses, with particular focus on the source term. The second work package “ Identification & Quantification of Uncertainty Sources” addressed one of the most crucial steps in BEPU analysis applying forward uncertainty propagation. Its main outcome is a knowledge-based matrix compiling over 600 uncertain input parameters related to source term assessment, including their associated models, phenomena, and (partial) quantifications. It was utilised and continuously updated in subsequent calculation work packages (Phebus FPT1, reactor and SFP case). First a consistent set of Figure of Merits, such as iodine and caesium release, was established. A systematic backward selection and characterisation approach of the associated uncertain input parameters was established, considering among others influencing phenomena, modelling choices, code capabilities, etc. While some parameters, such as shape factors, were common across codes - though different characterisations depending on code/user were provided - others were code-specific, appearing in only one or few codes. Establishing physically justifiable distributions proved challenging largely due to the scarcity of data and consequently heavy reliance on expert judgement and code-specific knowledge. Nevertheless, each parameter's characterisation had to be defendable on scientific grounds. In the absence of direct evidence, the characterisation was justified based on physical behaviour, while where direct association was not possible due to system complexity, on reasonable indirect evidence. This paper presents an overview of the work, main findings and conclusions, including the data base, and highlights key challenges and future research needs identified during the project. • Identification and quantification of uncertainty sources related to source term. • Systematic backward approach to select and quantify uncertain input parameters associated with the source term. • Comprehensive data base consisting of over 600 uncertain input parameters, including their quantification. • Scientifically defendable characterisation including justification. • Identification of future needs: further harmonisation, cross comparison as well as extension of the data base.

The CODEX-ATF integral bundle test was conducted within the framework of the IAEA Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS) project at the CODEX (COre Degradation Experiment) facility in Hungary. The electrically heated seven-rod bundle consisted of four Cr-coated and three uncoated Zr alloy cladding tubes, enabling a direct comparison of their behavior under high-temperature accident conditions. The experiment primarily aimed to investigate fuel failure and degradation mechanisms. During the test, several rods exhibited ballooning and burst phenomena. The maximum temperature exceeded 1600 °C. The transient was terminated by a bottom-up water quench. The total hydrogen generation was approximately 3 g, indicating substantial oxidation of the zirconium-based components. Intensive Zr-Cr eutectic interaction was observed in the hottest region of the bundle on the Cr-coated claddings. Post-test examinations revealed pronounced deformation and failure in both coated and uncoated claddings. • The CODEX-ATF test was performed with electrically heated bundle composed of three uncoated and four Cr-coated optZIRLO™ cladding tubes. • Under high-temperature (>1600 °C) accident conditions in steam atmosphere, both Cr-coated and uncoated fuel rods would lose their integrity. • In those parts of the core where the temperature remains below 1400 °C, oxidation would dominate the degradation process. • At temperatures above 1400 °C, the Cr layer may be lost and the formation of a Cr-Zr eutectic can cause localized melting.