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No abstract, annual program review. Name of presentation is "Correlative Atom Probe Tomography of the Buffer-IPyC Interlayer Region of TRISO-coated Particles." While APT and TEM data were collected, they were collected separately and not correlatively as originally planned. APT and TEM data are paired with micro-tensile data from the same regions.
A novel micro tensile sample fabrication technique for determining the tensile strength of the buffer, IPyC, and buffer -IPyC interlayer regions of surrogate (ZrO2) TRISO fuel particle layers was refined and implemented. Copper micro tensile samples served as baseline materials to verify the methods used. Tensile tests performed in this study, while limited in number, were analyzed using standard and Weibull statistics. As expected, the buffer layer was weakest, with an average ultimate tensile strength of 138.70 MPa, and the IPyC layer samples, were strongest, with an average ultimate tensile strength of 189.74 MPa. In the buffer -IPyC interface samples, all breaks occurred in the buffer region, though the average ultimate tensile strength of the samples, 159.80 MPa, was between the pure buffer and IPyC samples. These results suggest the interlayer region has unique properties, perhaps associated with pyrocarbon infiltration into the buffer layer during particle coating. All interlayer samples fractured within the buffer side; however, the stress strain behavior of some of these samples resembled the behavior of the IPyC layer samples. Here, the buffer and IPyC layer strengths had a normal distribution under Weibull analysis, while the interlayer region had a Rayleigh distribution. Further testing is needed to clarify both the standard and Weibull statistical results
Tristructural isotropic (TRISO) coated nuclear fuel particles are proving to be a versatile fuel form for new reactor designs. Understanding the bounding strength and failure mode of each coating interface is important to both fuel quality evaluation and failure prediction. A mechanism of key significance is failure of the silicon carbide (SiC) layer to retain fission products due to incomplete tearing of the buffer layer. This is a two-step mechanism involving both mechanical failure in the buffer and inner pyrolytic carbon (IPyC) layers and degradation of the SiC layer through palladium silicides at the IPyC-SiC interface. However, the mechanical properties of TRISO particle coating layers have yet to be fully characterized due to the small dimension of TRISO fuel particles and high radioactivity. To investigate this mechanism, in situ micro-tensile properties of the buffer, IPyC, SiC, buffer-IPyC, and IPyC-SiC interlayer regions of fueled TRISO particles have been tested at both as-fabricated and irradiated conditions. Determination of the mechanical properties of these TRISO particle regions will lead to a better understanding of the SiC layer failure mechanism and enable progress towards TRISO fuel qualification.
Investigating the microstructural, mechanical, and chemical behaviors of Tristructural Isotropic (TRISO) fuel particles is vital for its qualification and use in advanced reactors. Central to the study of TRISO particles is understanding the silicon carbide (SiC) layer's ability to confine fission products, with failure mechanisms linked to chemical degradation following mechanical degradation of the buffer and IPyC layers. Research has been done to quantify the micro-tensile properties of the buffer, inner pyrolytic carbon (IPyC), and buffer-IPyC interlayer regions and their interactions within both irradiated and un-irradiated TRISO particles. Techniques such as atom probe tomography (APT) and transmission electron microscopy (TEM) have also been deployed to examine microstructural defects and fission product distribution in detail. The goal is to understand layer delamination, establish connections between microstructure and mechanical attributes, and inform computational predictions of fuel performance. This work may help refine predictive models of TRISO fuel behavior and facilitating its certification for use in advanced reactors.
An international round-robin test to examine the consistency in leach-burn-leach (LBL) analysis of tristructural-isotropic- (TRISO-) coated particle fuel was conducted by three research organizations from the Generation IV International Forum member countries of the People’s Republic of China, the Republic of Korea, and the United States of America. Two sets of round-robin test samples were exchanged for analysis. One set of samples consisted of a series of nonuranium-bearing, TRISO-coated zirconium dioxide particles seeded with up to four depleted uranium-bearing, TRISO-coated uranium dioxide (UO 2 ) particles, which had intentionally damaged coating layers to simulate either particles with either exposed-kernel defects (i.e., particles with a cracked TRISO coating that should be detected during preburn leaching) or particles with silicon carbide (SiC) defects (i.e., particles with an intact pyrocarbon coating and a hole in the SiC layer that should be detected during postburn leaching). These simulated samples also contained added powder with known quantities of impurities from a coal standard. The other sample set consisted of representative sublots of UO 2 -TRISO particles fabricated in a production-scale coater, except they all contained depleted uranium instead of enriched uranium. In this report, the methodology used at Oak Ridge National Laboratory to conduct LBL analysis of the round-robin samples is presented, and the general results are summarized.
The technical bases for the TRISO-coated particle fuel specifications used by the AGR Program during its fabrication process development and fuel qualification efforts are provided based on historical HTGR fabrication experience in the U.S. and Germany and improvements and insights gained during execution of the AGR Program. The technical bases are provided for specified properties of the kernel, the TRISO coating layers, the fuel particle, and fuel compact in terms of fabricability considerations, in-pile fuel performance, and fission product release under normal and off-normal conditions.
The main objective of this technical report is to ensure that research and development (R&D) needs are identified to address the significant (highly ranked) technical challenges related to TRISO-coated particle fuel, high-temperature gas-reactor (HTGR) technologies, and other TRISO-related advanced designs by recognizing the existing foundation of identified R&D needs and augmenting that foundation with insights drawn from current TRISO-related research results and R&D needs related to novel applications of TRISO-related technologies that extend beyond the traditional HTGR designs.
AGR-5/6/7 was the last in a series of irradiation experiments sponsored by the U.S. Department of Energy in support of the development and qualification of TRISO coated particle fuel for use in a high-temperature gas-cooled reactor. This experiment was conducted to verify the performance of the reference-design TRISO-coated low-enriched UCO fuel for modular high-temperature gas-cooled reactor normal operating conditions and to explore fuel performance at temperatures substantially beyond those typical of normal operation. A total of 194 UCO fuel compacts in five capsules were irradiated in the Advanced Test Reactor for 360.9 effective full-power days, achieving final burnup ranging from 5.66% to 15.26% fissions per initial heavy metal atom and fast neutron fluence ranging from 1.62 × 10 25 to 5.55 × 10 25 n/m 2 (E > 0.18 MeV). Calculated time-averaged fuel temperatures ranged from 467 °C to 1432 °C, with a peak fuel temperature of 1536 °C. During the first five irradiation cycles (∼180 effective full-power days), 85m Kr fission gas release-rate-to-birth-rate ratios were 10 −7 –10 −6 , indicating no particle failures. Near the end of the sixth cycle, a substantial number of in-pile particle failures occurred in Capsule 1. The fission gas release from this capsule impacted the readings from the other four capsules and led to unreliable fission gas release measurements for all capsules during the last four cycles. A preliminary post-irradiation examination of Capsule 1 fuel and internal components revealed the in-pile particle failures resulted from operational issues with the capsule, not subpar performance of the fuel particles.
Safety functions are the actions, passive or active, that structures, systems, and components of nuclear facility that contribute to the safety of the workers, the public, or the environment. Well-defined safety functions are the foundation of a solid safety case for a reactor. For reactors that use TRISO-coated particles, the TRISO fuel plays an important part of the safety case because of its ability to contain radionuclides in the fuel itself. This ability enables the use of a functional containment strategy for the reactor where radionuclide retention is the primary safety function supported by the safety functions of controlling reactivity control and controlling heat rejection. This paper establishes at a deeper level the role that TRISO fuel plays in each of these safety functions and associated quality assurance and testing requirements for TRISO particle manufacturing to ensure these safety functions. Safety limits necessary to protect these safety functions include: operational limits, time at temperature limits, and fission gas release activity limits. In conclusion, this approach demonstrates the role that specific aspects of TRISO fuel play in protecting the safety of workers, the public, and the environment.
Modeling results used to assess the fuel performance of the TRISO-coated fuel particles as a function of kernel/buffer volume fraction include SiC tangential stress, formation of the buffer/IPyC gap, particle temperature profile, internal particle pressure, fission gas released from the kernel, probability of fuel particle failure, and fission product diffusion. These results were evaluated at two burnup levels and irradiation temperatures to bound expected steady-state irradiation conditions. In general, increasing the kernel/buffer volume fraction increases the SiC stress and subsequently the failure probability of a fuel particle when compared to the AGR designed particle. There was little impact on the fission product diffusion through the particle as the kernel/buffer volume fraction increased.