Tristructural Isotropic Particles Research Articles

Impact of temperature variations on BISON predictions of Ag release in AGR-1 and AGR-2 experiments

Understanding and quantifying the release of fission products like silver (Ag) from TRistructural ISOtropic (TRISO) fuel particles is important to assess the safe operation of advanced high temperature reactors. Although the silicon carbide (SiC) layer of TRISO particles is effective as the main fission product barrier, Ag can be released from intact TRISO particles. A mechanistic model for the effective Ag diffusivity, Deff, was previously developed as a function of temperature and microstructure variables informed by atomistic modeling of Ag diffusivity. In this study, we use this model to explore how experimental temperature uncertainties impact the overall predicted Ag release. This analysis indicates that calculated temperature uncertainties have a significant impact on the overall Ag release predictions. Furthermore, we show that the time average volume average (TAVA) temperature is not an appropriate proxy for the detailed temperature histories to predict fission product release due to the Arrhenius dependence of Ag diffusivity with respect to temperature. The detailed temperature histories, therefore, provide the most accurate results and are of most importance for modeling efforts. Overall, this work shows the importance of considering the experimental uncertainty of the temperature on computational predictions of fission product transport and release and the need for more accurate temperature histories from future experiments.

Fracture mechanics approach to TRISO fuel particle failure analysis

Weibull stress-based methods for failure probability assessment have been developed and analyzed to assess the integrity of tristructural isotropic (TRISO) fuel particles during fuel life cycles and accident operating conditions. While simple, these methods entail a number of drawbacks when stress concentrates near crack tips, including finite element mesh size dependency when the Weibull stress is averaged over the finite element domain. Fracture mechanics approaches involving the use of interaction integrals eliminate this lack of mesh convergence and produce consistent fracture predictions. In this work, we use an interaction integral approach to computing stress intensity factors for a crack in the inner pyrolytic carbon layer perpendicular to the silicon carbide layer, which is simplified representation of a failure mode in TRISO particles. The interface between these two TRISO layers has been shown to become porous, which we simulate by considering a transition of mechanical properties over such porous length, i.e. the layers are modeled as a functionally graded material. Aspects such as porosity and thermal and irradiation eigenstrains are considered in computing the stress intensity factor from a fracture mechanics approach and compared with the known Weibull stress failure approach. The methodology introduced in this paper enables a more general fracture probability assessment in TRISO particles and eliminates the need to identify best suited parameters when using local or averaged stress-based failure criterion. Finally, the numerical sensitivity studies show how parameters such as the porous transition zone length, the material stiffness, and creep affect the probability of TRISO fuel particle failure.

Numerical study of TRISO particles with random size and location distribution in cuboid and cylindrical matrix: A validation for two-regime heat conduction model

During reactor operation, the precise size values or specific size distribution of tri-structural isotropic (TRISO) fuel particles randomly dispersed within the fully ceramic microencapsulated (FCM) fuel elements are often unknown. To investigate the impact of the random distribution of positions and sizes of TRISO particles within a cuboid or cylindrical matrix on the temperature distribution of fuel elements and to establish a practical heat conduction model, TRISO fuel particles with randomly distributed positions, diameters following Gaussian or uniform distributions are generated using relevant algorithms. A cuboid and a cylindrical TRISO fuel element are modeled in ANSYS, and a three-dimensional temperature field distribution of the fuel element is obtained through numerical simulation. The probability density of the maximum temperature of the fuel element is calculated and fitted. Finally, the applicability of the two-regime heat conduction model over a broader range is validated based on the simulated data. The conclusions of this study provide reference information for predicting the temperature distribution of FCM fuel elements containing TRISO particles with random size and location distribution.

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

Thermal conductivity of SiC and PyC coatings in spherical nuclear fuel particles measured by nanosecond time domain thermoreflectance

TRi-structural ISOtropic (TRISO) nuclear particle fuels are intended for application in several designs of advanced high temperature nuclear reactors. Nanosecond time domain transient thermoreflectance has been applied to the silicon carbide (SiC) and the inner and outer pyrolytic carbon (PyC) coatings of surrogate TRISO particles, measuring mean thermal conductivities of 166 ± 18 W∙m−1K−1, 11.1 ± 1.9 W∙m−1K−1, and 8.6 ± 2.1 W∙m−1K−1, respectively, for each layer. The applicability of the analytical heat diffusion model used to fit the measured traces was examined and it was found that the boundary effect at the SiC/PyC interface becomes perceptible at a lateral distance approximately equal to the thermal penetration depth of SiC (7.1 µm). In addition, steady-state thermal modelling showed that the radial thermal resistance of the particle is dominated by the porous buffer layer. The thermal resistance will likely increase significantly upon circumferential detachment of the porous buffer carbon coating with irradiation. Implications for fuel design and reactor performance are discussed.

Accelerated thermal property mapping of TRISO advanced nuclear fuel

TRistructural ISOtropic (TRISO) fuel is a leading-edge nuclear fuel form representing a departure from the more traditional nuclear fuel forms utilized in the reactor fleet of today. Rather than a monolithic fuel pellet of uranium dioxide, integral fuel forms containing TRISO fuel are composed of thousands of microencapsulated uranium-bearing fuel kernels and individually coated with multiple layers of pyrolytic carbon and silicon carbide. These multilayered ceramic coatings serve as an environmental barrier to ensure radioactive and chemically reactive fission products are contained within the reactor fuel elements, but also participate in the transfer of heat generated in the nuclear fuel to the coolant – the primary purpose of a nuclear reactor. Since traditional thermal property measurement techniques, such as laser flash analysis, would be unable to resolve the thermal properties of the individual TRISO coating layers, a simplified frequency-domain thermoreflectance technique has been developed to rapidly map the thermal properties of TRISO particles. Using this technique, the thermal properties of TRISO particles have been mapped from room temperature up to 1000 °C to examine the spatial variation and temperature-dependency of the thermal properties within each layer. Additionally, spatial-domain thermoreflectance was used to examine the anisotropy of the thermal properties for each layer at different locations within a single TRISO particle, and across multiple TRISO particles to assess the intra- and inter-particle uniformity of thermal properties, respectively. To elucidate the underlying causes for the measured variations in thermal properties, scanning electron microscopy and Raman spectroscopy were used to examine variations in microstructure and chemical bonding within the different coating layers. Results from this work are then compared with previous examinations of TRISO fuel particles and microstructurally driven mechanisms for the variations in the measured thermal properties of the different carbonaceous layers are discussed.

Evaluation of the spatial self-shielding impact for TRISO-based nuclear fuel depletion

Reactor physics analyses of nuclear cores with nuclear fuel concepts containing tristructural isotropic (TRISO) particles, such as pebbles or compact fuel elements, rely on various degrees of simplification to keep these highly heterogeneous problems computationally tractable. One such limitation regards the level of spatial discretization employed during burnup calculations, where traditionally only a limited number of spatial zones are modeled at the full core level and assume that the spectrum is constant within the fuel elements and TRISO particles in this depletion zone. This type of assumption neglects the impact of spatial self-shielding effect within the kernels (microscale level) as well as within the compact or pebbles (mesoscale level). The Monte Carlo code Serpent 2 contains many relevant features for efficiently modeling this type of geometry, including a collision-based domain decomposition intended for very large burnup calculations, which we leveraged for this work to quantify the impact of capturing neutron flux variations occurring at the micro- and mesoscale level on a series of high-temperature gas-cooled reactor fuel element depletion problems. While spatial self-shielding is observed at both scales, with differences from a volume-averaged burnup of ±7% within the kernels and ±2% between TRISO particles within the fuel element, the conjugated effect on nuclide inventories and multiplication factor are negligible, hence confirming that assuming a single average spectrum value may be sufficient for most applications.

Ultra-high temperature microstructural changes of SiC layers in TRISO particles

The SiC layer is the critical layer in the tristructural-isotropic (TRISO) fuel particle with the ability to sustain internal pressure and retain fission products. Further research into the failure mechanisms and tolerance limits of SiC layers at ultra-high temperatures is required. The microstructural evolution of the SiC layer in TRISO particles heat-treated at1800∼2500 °C in argon atmosphere is the focus of this paper. The results show that SiC decomposition and phase transition occur at temperatures above 2200 °C, causing TRISO particles to fail. At 2200 °C, significant decomposition of the SiC layer occurred, resulting in an increase in carbon content on the surface and the appearance of pores. The decomposition reaction of SiC became more complicated at higher temperatures, and the carbon with the graphene structure remained after decomposition. At temperatures above 2300 °C, the SiC phase changed from 3C to 6H, with an abnormal increase in grain size. This study establishes a database for the performance of SiC layers under operation or accident conditions and serves as a reference for future applications of TRISO particles in novel ultra-high temperature reactors.

Simulation of random medium of fuel pebble with liquid salt coolant using PebMC code

Pebble-bed High Temperature reactors (HTR) are proposed to generate heat for power production and process heat application such as Hydrogen (H2) production. A HTR fuel pebble consists of random distribution of several thousands of fuel particles, called Tri-Structural Isotropic (TRISO), in fuel zone followed by a graphite layer. This configuration along with coolant layer results in double heterogeneity effect. A multi-group Monte Carlo code PebMC has been developed to accurately analyse molten salt cooled HTR lattice cell. Neutron transport in such a complex fuel distribution requires resonance self-shielded cross-section which is generated using rational approximation and Dancoff correction. Dancoff factor accounts for those neutrons leaving a fuel particle and directly colliding with another fuel particle, without any collision with moderator nuclei.In this study, an analytical expression based on a two-region model developed for gas cooled HTRs is adopted to calculate Dancoff factor for pebble with salt coolant. The implementation of these analytical expressions with 172groups WIMS library in PebMC code have been discussed. Subsequently, paper describes neutron tracking in random medium of HTR lattice using Monte Carlo method. The influence of fuel salt on Dancoff factor for various packing of TRISO particles and subsequent effect on infinite multiplication factor (K-inf) calculation have been presented. In addition, effect of temperature of molten salt on Dancoff factor have been studied.

Fabrication of beryllium oxide based fully ceramic microencapsulated nuclear fuels with dispersed TRISO particles by pressureless sintering method

In this research paper, two types of tristructural isotropic (TRISO) particles were dispersed within beryllium oxide (BeO) matrix using the pressureless sintering method to create fully ceramic microencapsulated (FCM) nuclear fuel. The findings indicated that TRISO particles with an outermost layer comprising silicon carbide (SiC) can establish interfacial bonding with BeO matrix, while those with an outermost layer comprising pyrocarbon (OPyC) and BeO matrix exhibit a noticeable gap after the sintering process. Results from transmission electron microscopy (TEM) revealed the presence of an amorphous phase transition zone between SiC and BeO. This zone, according to molecular dynamics (MD) simulation results, is an amorphous phase composed of Be, Si, and O atoms. Overall, TRISO particles with an outermost layer of SiC are more suitable for sintering with BeO matrix than those with an outermost layer of OPyC.

Bayesian uncertainty quantification of tristructural isotropic particle fuel silver release: Decomposing model inadequacy plus experimental noise and parametric uncertainties

Tristructural isotropic (TRISO) particle fuel is one of the most promising fuel concepts enabling high temperature and high burnup reactor operation. One dominant source of radioactivity released from the TRISO particles is silver (Ag), which is subject to a high release fraction and long decay life compared to other fission products. Previous modeling efforts using the fuel performance code BISON indicated nonnegligible uncertainties in modeling the diffusion process of fission products in TRISO compared to the Advanced Gas Reactor experiments. The overall uncertainties observed when modeling the fission product diffusion can result from uncertainties in model parameters, noisy experimental measurements, and deficiencies in the developed models. The three types of underlying uncertainties have not yet been properly quantified in open literature. This paper presents the Bayesian uncertainty quantification (UQ) using massively parallelizable Markov chain Monte Carlo samplers. The uncertainties due to model parameters, model inadequacy, and experimental measurement noise are quantified, with the σ term used to represent the sum of the model inadequacy and measurement noise uncertainties. It is worth noting that this is the first time the σ term is inferred for nuclear fuel experiments, as compared to using prescribed values for uncertainty quantification in previous work. The parallelizable Markov chain Monte Carlo samplers efficiently infer the model parameters and the σ term, giving insight into physical parameters like diffusion coefficients and the combined model discrepancy and measurement noise. A subsequent forward uncertainty quantification (UQ) is also performed based on the calibration results to generate more accurate predictions of the Ag release. The model inadequacy plus experimental noise is the most dominant source of uncertainty compared to the parametric uncertainty. All the UQ analyses presented in this work are based on the second series of the irradiation experiments in the Advanced Gas Reactor program.

Improved measurement of local thermal conductivities of coatings by the steady-state Raman spectroscopy method: A case study in TRISO particles

Tristructural-isotropic particle (TRISO) is a fully ceramic microencapsulated fuel of high-temperature gas-cooled reactors with complex microstructural morphology and heat transfer process, hindering the accurate measurement of its thermal conductivity. Here, a dual test wavelength steady-state Raman spectroscopy method (SSRS) for determining the thermal conductivities is proposed, resolving the issue of the conventional SSRS failing to measure the thermal conductivity of 3C-SiC. By demonstrating prerequisites for using the SSRS method and modeling the conductive-radiative coupled heat transfer, the thermal conductivities of all layers in TRISO have been accurately measured. It is also found that the thermal conductivity of an underlying material can be estimated by measuring the material on its top if the outer material has low absorption under the applied laser wavelength, offering the possibility of measuring the buried layers in localized structures. The work extends the usable range of the SSRS method for obtaining thermal conductivities in wide applications.

Failure analysis of nuclear transient-tested UN tristructural isotropic fuel particles in a 3D printed SiC matrix

Fully ceramic microencapsulated fuel elements containing UN tristructural isotropic (TRISO) fuel particles within a 3D printed SiC matrix were subjected to transient testing with varying energy depositions. Detailed post-irradiation examinations were performed, including leaching in hot HNO3 and post-leaching X-ray computed tomography, to quantify the percentage of failed TRISO particles and crack propagation within the particles and surrounding fuel matrix. In parallel, detailed finite element analyses were performed for comparison with experimental findings and to better evaluate transient failure modes. The lowest transient energy deposition—which still exceeded bounding values for high-temperature gas-cooled reactor applications—resulted in no detectable TRISO particle failures or matrix cracking, which was consistent with the simulations. Simulations of the higher-energy transients for which significant TRISO particle failure was expected were generally able to reproduce the transient temperatures and matrix cracking. Thus, the TRISO particle failures were explained based on the effects of local SiC matrix thickness and porosity. Results generally confirmed the high strength of the additively manufactured SiC matrix but also affirmed the need for a modified UN TRISO architecture to prevent SiC matrix cracks from propagating through TRISO layers. This unique failure mode has not historically been considered for TRISO fuels contained in weaker graphite matrices.

TRISO fuel performance analysis: Uncertainty quantification toward optimization

Tri-structural isotropic (TRISO) fuel particles are a fuel form being considered for potential use in next-generation nuclear reactors (i.e., high-temperature gas-cooled reactors). Though the TRISO fuel manufacturing process has continually advanced in recent years, particle comparisons still reveal statistical variations and uncertainties in terms of geometric configurations and material properties. Given that the physical processes ongoing in TRISO fuel particles during reactor operation are highly correlated with each other, a small degree of uncertainty in one model may lead to significant uncertainty in another. This makes appropriate uncertainty quantification of TRISO fuel particles essential. However, one may wonder about the extent to which the current version of TRISO particles has been optimized, and whether any room remains for further improvements.This paper quantifies TRISO fuel performance model uncertainties that stem from geometric and material data. For this analysis, the BISON code was used, and the Advanced Gas Reactor (AGR)-2 experiment served as a reference case. A total of 105 calculations was performed for the uncertainty and optimization analysis, altering the geometric and material data within their uncertainty range. Lastly, the optimization potential of TRISO particles is evaluated from a fuel performance perspective.

High temperature oxidation and microstructural evolution of SiC layer of TRISO particles in the presence of NaCl

With the different types of advanced nuclear fission reactors, the reliability of the SiC layer in tristructural-isotropic (TRISO) particles exposed to different corrosive environments is of vital important. In the present work, the oxidation behaviors of the SiC layer of TRISO particles exposed in NaCl salts with the concentrations of 0.2–2 wt% at 800–1400 °C were investigated. The effects of NaCl concentration and oxidation temperature on phase composition, microstructure, and chemical composition of the TRISO particle samples were systematically investigated. The results show that, with the increase of oxidation temperature and NaCl concentration, the oxidation rate of SiC layer in TRISO particles was accelerated. However, the effect of NaCl concentration alone on the oxidation rate of SiC layer was not significant. This could be attributed to the volatility of NaCl at high temperature. In addition, just like other halogen salts, NaCl salts also reacted with the oxide layer to form silicate compounds, such as Na2SiO3, which provided a channel for both oxygen and fluoride ions to diffuse into the inner SiC layer. The results obtained in the present work can provide a guide for the application of TRISO particles in molten salt stacks of chlorine salt.

Sensitivity analysis of cesium and strontium release from TRISO particle under irradiation and high temperature conditions

This paper presents a systematic sensitivity analysis of the diffusivity parameters of cesium and strontium inside Tristructural isotropic (TRISO) fuel particle and their dependence on irradiation and high temperature annealing conditions. The BISON fuel performance code was used to model the base irradiation and the high temperature tests performed within the AGR-1 experiment. The sensitivity analysis results quantified the contributions of the various diffusion parameters to BISON predictions of Cs and Sr release from compacts exposed to a number of different irradiation and thermal conditions. Informed by the Sobol indices, a calibration study provided an optimized set of diffusion parameters that improved the accuracy of BISON predictions. Finally, we describe an analysis of the influence of irradiation and thermal conditions on the Cs and Sr release and characterize their dependence on the testing conditions. The systematic approach in this study allows us to better understand the transport mechanism of Cs and Sr inside TRISO particle, and to identify a number of future needs to fully resolve the discrepancy between BISON predictions of Cs and Sr release with the AGR-1 experimental observations.

Neutronic and Thermal-Fluidic Analyses for an Additive Manufactured Reactor With SiC Matrix and TRISO Particle Fuel

Abstract Additive manufacturing (AM) is a transformational digital manufacturing technology featured with rapidity, customizability, precision, and economy, which is fundamentally altering the way components are designed and manufactured. AM enables the freedom of design, and makes full use of complexity of geometry which “comes for free”. Applying AM technology to nuclear industry can yield advanced reactor designs with function and structure matched for the best thermal, fluidic and mechanical performance. In this work, an AM-informed reactor core design with silicon carbide (SiC) matrix and tri-structural isotropic (TRISO) particle fuel is proposed and analyzed. The core is an integrated 3D-printed SiC bulk with helical cruciform coolant channels, and the UO2-bearing TRISO fuel particles are dispersed in the bulk. A multiphysics analysis framework for irregular geometry is developed to analyze and further optimize the reactor design. The TRISO particle positions are generated with discrete element method (DEM). The Reactor Monte Carlo code (RMC) and the commercial computational fluid dynamics (CFD) software star-ccm+ are used for the neutronic and thermal-fluidic analyses, respectively. RMC simulates the neutron transport to predict the effective multiplication factor and power distribution. star-ccm+ calculates the flow and heat transfer in coolant channels and heat conduction in solid matrix with the power distribution as the heat source. Preliminary results show that the power peaking factor FQ decreases below 1.65, the heat transfer area increases by 30.3% and the fuel peaking temperature decreases by 25 K. The optimized AM-informed design enjoys better neutronic and thermal-fluidic performance than those with regular geometry.

Seeing the whole picture: Methods for getting the most from micro X-ray computed tomography of TRISO nuclear fuel particles

Tristructural isotropic (TRISO) coated fuel particles are a nuclear fuel form under extensive study for use in advanced nuclear reactor concepts. TRISO fuels are subjected to high temperature neutron irradiations and then examined to assess their performance by determining fission product retention and studying morphological changes. Micro X-ray computed tomography is one method of nondestructively studying the effects of TRISO performance. This work addresses the need for image processing to remove X-ray tomographic reconstruction artifacts that prevent the study of TRISO features, as the TRISO particles’ high Z kernel can introduce metal artifacts that degrade the image quality in the surrounding low Z coating layers. These metal artifacts were reduced by imaging the TRISO particles with both high- and low-energy X-rays and applying a mask to the radiographs obtained with low-energy X-rays to digitally remove the dense fuel kernel region. These masked radiographs were then used to produce a tomographic reconstruction which was combined with the tomographic reconstruction of the high-energy data. This enabled the relatively-low-density TRISO buffer layer to be examined in more detail, providing information on irradiation induced dimensional changes of the coatings. This methodology, which helps see the full picture of a TRISO particle, is not limited to nuclear fuels but can be applied to systems that contain highly attenuating material surrounded by less dense materials.

Electrochemical behavior of matrix graphite in nitric acid by cyclic voltammetry

Post-irradiation examination of irradiated spherical fuel elements in high-temperature gas-cooled reactors (HTGRs) requires the separation of integrated tri-structural isotropic (TRISO) particles from matrix graphite (MG). The electrochemical method is a promising approach for separating intact TRISO particles and exploring the fission product source terms in MG. In this work, the electrochemical behavior of MG and its three compositions, i.e., artificial graphite (AG), natural graphite (NG), and binder carbon (BC) were investigated experimentally by cyclic voltammetry using nitric acid as an electrolyte, and these three compositions showed very different electrochemical properties. Compared with BC, AG and NG exhibited relatively higher electrochemical reactivity controlled by diffusion, while BC performed almost identical adsorbing-controlled electrochemical behavior with MG composites, which probably resulted from encased AG and NG flakes by bulk BC. According to cyclic voltammetric and characterizing results of functional groups on the surface of the specimens, the deconsolidation reacting procedures of MG could be further interpreted into a two-step process. The first step was the electrochemical oxidation of BC by unsaturated , defective, and sp2 carbon atoms. In the second step, the NG and AG flakes were intercalated, oxidized and exfoliated by combined reactions of hydrolysis, intercalation, and oxygen evolution. The proposed electrochemical procedures obtained in this study provide the theoretical foundation to PIE investigations of HTGR spent fuels.

Fission gas retention of densely packed uranium carbonitride tristructural-isotropic fuel particles in a 3D printed SiC matrix

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. This work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with ∼55% TRISO particles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low (<250 °C) surface temperatures. Initial testing at higher (∼700–750 °C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (∼300–750 °C) of both fueled compacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause complete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, resulting in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.