The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YHx, x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YHx specimens were, thus, subjected to neutron irradiations in the range of 0.1-2 displacements per yttrium atom (dpa-Y) in the temperature range of 536-878 degrees C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YHx specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YHx exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. To accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300∘C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8×1021n/cm2 fast fluence (En>1.0MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

With the continued advancement of additive manufacturing (AM) techniques, interest has grown in the development and qualification of steels produced via these methods for use in the structural components of advanced nuclear reactors. Therefore, it is crucial that the properties of these materials such as the tensile strength and fracture toughness be investigated following neutron irradiation to support their use in industry. To that end, a test plan to irradiate several tensile and bend bar specimens in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory was proposed, along with the development of a new “MINBEN” capsule design that features MBS-1 bend bar specimens that have a larger cross-sectional area than those of previous designs, which allows for higher temperature out-of-pile testing. The specimens include AM 316H stainless steel with wrought 316H and A709 stainless steels for reference, and they will be irradiated at 2 dpa and 10 dpa—corresponding to roughly 1 and 5 cycles at the HFIR midplane—at temperatures of 400°C and 600°C. The MINBEN design was found to be capable of providing specimen test plane average temperatures in the range of 220–660°C for six specimens, with a min-max spread in this temperature of ~40°C and a 95% confidence interval of ~24°C. This analysis shows that the capsule provides an effective vehicle for gathering needed high-temperature fracture toughness data.

Tristructural isotropic (TRISO) fuel is being proposed for use in several high-temperature advanced reactor concepts because of its structural integrity under high operating temperatures and burnup. One of these advanced reactor concepts is the Kairos Power fluoride salt-cooled high-temperature reactor (KP- FHR) under development by Kairos Power, LLC. Previous TRISO irradiation experiments were focused on qualification for high-temperature gas reactors (HTGRs), which have higher operating temperatures but lower particle powers than the KP-FHR design. To study the performance of TRISO fuel designed for HTGRs under prototypical FHR conditions, a set of experiments was designed using the MiniFuel irradiation vehicle at the Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR). The experiments will irradiate 30 TRISO-containing carbon matrix compacts at inner small vertical experiment facilities in HFIR. Each compact will contain 20 TRISO particles (600 particles total) developed for HTGRs, consisting of either 14% enriched uranium dioxide uranium carbide, naturally enriched uranium dioxide, uranium carbide, or 9.6% enriched uranium dioxide fuel kernels with time- and volume-averaged silicon carbide layer temperatures between 500 and 900°C. This report summarizes the vehicle designs that have been developed, as well as the neutronic and thermal analyses completed for these irradiation experiments. These analyses show that MiniFuel compact irradiation is a versatile experiment that can be used to study a range of TRISO particle powers and fuel types while providing reasonable separation of burnup and temperature effects.

This report provides a summary of irradiation capsule design and analysis for the three main categories of materials to be used in the Transformational Challenge Reactor (TCR) core. These materials are yttrium hydride as the moderator, silicon carbide, and Grade 316L austenitic stainless steel. No neutron irradiation data exists on the former and the latter two are derived from advanced manufacturing techniques and therefor need to undergo rigorous testing to assess their behavior as function of dose and temperature. The detailed design and analysis summarized in this report is the critical step to facilitate safe and relevant irradiation of these materials to be conducted at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). These irradiations will ultimately yield critical materials properties data for reactor design and licensing purposes.

The safety of accident-tolerant fuel (ATF) candidates must be tested under simulated loss-of-coolant accident (LOCA) conditions to evaluate their improved safety margins. This evaluation eventually must include testing of irradiated materials. This report describes the fiscal year (FY) 2019 activities to (1) improve the information yield of integral design–basis LOCA tests by using digital image correlation (DIC) calculations on the engraved specimens’ surfaces and (2) to investigate the effects of reducing the length of nuclear-grade FeCrAl specimens tested in the Severe Accident Test Station (SATS) to improve the cost-effectiveness of irradiation campaigns for future work. Analysis showed the DIC technique to be effective at calculating strains, and that specimens as short as 4 inches could be used for LOCA experiments while maintaining similar balloon and burst behavior as 12-inch specimens. With these results, an irradiation vehicle design effort was initiated to encapsulate multiple 4-inch thin-walled tube cladding specimens and test them under irradiation in the flux trap of the High Flux Isotope Reactor. By utilizing thermal modeling and optimizing the design parameters, calculated specimen temperatures were predicted to vary less than 2% during irradiation. Future work is planned to finalize the design and safety basis approvals and initiate a HFIR irradiation in support of in-cell integral LOCA tests at the Irradiated Fuels Examination Laboratory (IFEL). the reactor’s known power profile. Simulation results with the same FeCrAl material from LOCA burst tests determined that all 4-inch specimen assemblies would experience excellent temperature uniformity and accuracy during the HFIR irradiation run. These results indicate the full-length HFIR target accommodating 4-inch long cladding specimens is a viable design for accelerated irradiation testing of thin-walled cladding tube segments.