The U.S. Department of Energy's Oak Ridge National Laboratory (ORNL) announced on September 4, 2026, that it had completed post-irradiation mechanical property measurements on TRISO fuel particles containing high-assay low-enriched uranium (HALEU). This data, which quantifies how TRISO fuel changes after irradiation at the level of the particle's individual layers, represents a category of measurement that can be fed directly into reactor design models—the latest addition to the multi-stage dataset built up under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.
This latest effort follows neutron scattering measurements conducted earlier this year (which analyzed the chemical composition of unirradiated particles), but differs in a key respect: it quantifies structural changes that occur after irradiation. Unless actual data confirms how far fuel particles deviate from their as-manufactured specifications, the basis for allowable burnup limits remains dependent on assumptions embedded in computational models rather than empirical evidence.
How Irradiation Affects the SiC and PyC Layers
A TRISO fuel particle's structure consists of five layers, from the inside out: a UCO or UO2 kernel, a buffer layer, an inner pyrolytic carbon (IPyC) layer, a silicon carbide (SiC) layer, and an outer pyrolytic carbon (OPyC) layer. This roughly 1-millimeter-diameter particle occupies a unique position among nuclear fuels because its coating layers are designed to retain fission products even under high-temperature gas reactor (HTGR) accident conditions.
However, this retention capability assumes that the as-manufactured specifications are maintained. During in-reactor irradiation, neutron flux creates lattice defects in the materials. In SiC, transmutation converts silicon into aluminum and phosphorus, affecting bulk properties, while pyrolytic carbon's shrinkage and swelling behavior under irradiation varies depending on the deposition conditions used during manufacturing (specifically, the degree of isotropy). If the elastic modulus and hardness of the SiC layer change, its function as a mechanical barrier contributing to fission product containment will operate under conditions different from those at manufacture. Measuring the magnitude of this pre- and post-irradiation difference is a prerequisite for validating the reliability of fuel performance models.
For this work, ORNL used a specialized instrumented indentation testing system built specifically for radioactive materials and nuclear fuel. Researchers in ORNL's Nanomechanics Laboratory measured the IPyC, OPyC, and SiC layers of the particles individually, quantifying how hardness and elastic modulus had changed. The results confirmed a decrease in both the elastic modulus and hardness of the SiC layer, along with changes in the properties of the PyC layers.
"These measurements let us compare as-fabricated behavior with behavior after irradiation at different temperatures and burnups," said Katherine Montoya, an ORNL research and development staff member in the Particle Fuel Forms Group. "This will help us improve nuclear fuel performance models and support the deployment of gas reactors."
Two Rounds of Measurements This Year, Linked to Safety Testing
Before completing this round of mechanical property measurements, ORNL had already finished a separate set of measurements earlier in 2026. Using a diffractometer at the lab's Spallation Neutron Source (SNS), researchers directed a pulsed neutron beam just 1 millimeter in diameter onto HALEU-containing TRISO particles. The goal was to characterize chemical and phase-level changes—specifically, how effectively the UCO kernel absorbs oxygen to suppress CO2 formation. Whether this mechanism, which prevents overpressurization and corrosion of the coating layers, functions properly has a direct bearing on fission product containment performance.
Whereas neutron scattering examined the chemical integrity of the fuel's interior, this round of instrumented indentation testing examined structural integrity. Combined, the two measurement approaches allow researchers to simultaneously assess what is happening chemically and how well the particle is holding up mechanically after irradiation.
Post-irradiation examination (PIE) across the broader AGR program is also progressing. The AGR-2 experiment completed PIE on engineering-scale UCO and UO2 TRISO fuel, confirming fission product retention performance exceeding HTGR design requirements. For AGR-5/6/7, following irradiation to 360.9 effective full-power days in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), PIE and high-temperature safety testing are still ongoing at ORNL. Safety testing at 1600°C detected no significant release of Kr-85m (krypton), confirming the integrity of the TRISO coating. Safety testing at 1800°C also found fission product retention behavior broadly consistent with previous AGR campaigns. On the other hand, depletion of Ag-110m (silver) and loss of Eu (europium) were observed at 1600°C, and refinement of performance models continues to focus on the known challenge of silver diffusion through SiC.
Once final AGR-5/6/7 PIE results are confirmed, they will be cross-referenced with mechanical property data of the kind reported here, making it possible to have concrete discussions about which temperature and burnup conditions push which component closest to its limits first.
Beyond Qualification: Two Remaining Hurdles—Supply and Regulation
While the data ORNL is accumulating forms the scientific basis for fuel qualification, commercialization still faces separate hurdles: establishing mass-production capacity and securing regulatory approval.
On the manufacturing side, 2026 marked the inaugural year of commercial supply infrastructure. In February 2026, X-energy's subsidiary TRISO-X received a 40-year special nuclear material (SNM) license from the U.S. Nuclear Regulatory Commission (NRC) for its TX-1 facility in Oak Ridge, Tennessee—the first NRC license issued for a Category II fuel fabrication facility using HALEU. As of August 2026, the facility had moved into the phase of interior construction and installation of process equipment. If TX-1 comes online as planned, it will be capable of producing approximately 700,000 TRISO pellets per year (equivalent to 5 metric tons of uranium), enough to supply fuel for up to 11 of X-energy's Xe-100 small modular reactors (SMRs).
Meanwhile, Framatome received NRC approval for a license amendment in July 2026 for its facility in Richland, Washington. The approved enrichment ceiling was raised from below 6.5% to below 10%, and TRISO fuel fabrication was also authorized (SNM-1227). Commercial production is scheduled to begin in 2027. Framatome has stated it is also considering a further amendment to enable 20% HALEU enrichment down the line, and that the necessary facility design work has already been completed.
Supply infrastructure development and data accumulation are proceeding in parallel, but their timelines are not aligned. With the timing of TX-1's full-scale operation still undetermined, even as ORNL continues to build out its measurement dataset, integrating that data into performance models tied to actual manufacturing specifications will require a full development and qualification cycle. The publication of final AGR-5/6/7 PIE results—and how those results connect to the process of finalizing commercial TRISO fuel specifications—will be the practical checkpoint that determines the pace of commercial deployment.