Work

Extending OFFBEAT capabilities to modelling uranium hydride fuel for the AGR-like FHR reactor

Year: 2026

Type: article

Abstract: 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.

Source: Nuclear Engineering and Design

DOI: 10.1016/j.nucengdes.2026.114891

Authors: A. de Lara, A. Scolaro, E. Shwageraus

Institutions: University of Cambridge, École Polytechnique Fédérale de Lausanne

Language: English

Topic: Nuclear Materials and Properties

Subfield: Materials Chemistry

Field: Materials Science

Domain: Physical Sciences

SDG: Affordable and clean energy

Open Access status: hybrid

Estimated APC paid: $2,710