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Fission Gas Monitoring System Efficiency Calibration Summary Report

The Fission Gas Monitoring System was designed to work in conjunction with the Fuel Accident Condition Simulator Furnace and is used to measure fission gas releases during post-irradiation fuel heating tests. A high-purity helium sweep gas is swept past the heated fuel sample in the furnace and routed to a fission gas monitoring system, which cryrogenically traps the krypton (Kr) and xenon (Xe) fission gases. As the radioactive species of the trapped gas decays, high-purity germanium detectors monitor the amount of fission gas in each trap. The primary fission product present in this test is 85Kr, which has a 10.76-year half-life. The presence of 85Kr indicates particle defects and iodine release. There is also the potential for re-irradiating the fuel compacts in the Neutron Radiography Reactor in the Hot Fuel Examination Facility basement. With re-irradiation, it is possible to have xenon fission gas present, specifically 133Xe, which has a 5.243-day half-life. Re-irradiation also produces 131I, which is important for design-basis accidents and regulatory licensing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Demonstration of Integrated PostIrradiation Process Steps forMultikilogram Amounts of Low- and High-Burnup Used Nuclear Fuels

sponsored research and development (R&D) on the advanced, closed nuclear fuel cycle, including post-irradiation processing of used fuels, recovery of separated components, and preparation of recycle nuclear fuels. During the period 2007–2010, kilogram-scale demonstrations of the integrated processing steps were performed for actual light water reactor (LWR) used nuclear fuels (UNF). These integrated processing demonstrations were called “the Coupled-End-to-End” (CETE) campaigns and were associated with the Global Nuclear Energy Partnership program, which ran concurrently. The demonstrations were carried out in the shielded hot cells and associated gloveboxes located in the Irradiated Fuels Examination Laboratory (IFEL) and the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). Limited circulation program reports and a few technical papers were issued. The subject document provides a comprehensive summary of results, technical comparisons, and conclusions obtained from the demonstration campaigns and subsequent associated R&D.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advanced Multiscale Microscopy Characterization of High Burnup LWR UO 2 Before and After LOCA Testing

Previous microscopy work within the campaign has observed increased porosity coupled with subgrain formation in the mid-radial region of the fuel. However, examinations of high burnup fuel taken from multiple reactors with different operating conditions have shown that this structure can move radially inward depending on the fuel operating conditions. It is currently theorized that inter- and intragranular fission gas bubble nucleation is precursory to the grain subdivision observed in the mid-radial and central regions of the fuel. Characterization of samples after loss-of-coolant accident (LOCA) testing has revealed that these restructured regions with a high density of bubbles and subgrains pulverize during the transient test. It appears that the increase in fission gas bubbles coupled with grain subdivision renders the fuel mechanically weaker during a LOCA transient and thus susceptible to fuel fragmentation relocation and dispersal (FFRD). Work this past fiscal year has prioritized understanding this behavior by using advanced microscopy to investigate fission gas behavior in the as-irradiated and post-LOCA state. Additional work has been performed to verify the theory described above by analyzing multiple post-LOCA optical micrographs and comparing that to as-irradiated microstructural data. This document reports progress in the post-irradiation characterization of high burnup nuclear fuel with emphasis on the restructured fuel regions, particularly how microstructural features influence FFRD and fission gas release behavior in LOCA conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preparation and Analysis of Austenitic Steel Samples Irradiated at Very High Damage Doses

This report describes the preliminary results of the visual inspection, cutting, preparation for IASCC (irradiation-assisted stress corrosion cracking) testing, and property evaluation of austenitic stainless steel specimens irradiated at very-high-damage doses (up to 125 dpa) in the BOR-60 fast reactor. Work was performed at Oak Ridge National Laboratory using the Irradiated Materials Examination and Testing (hot cell) Facility and the Low Activation Materials Development and Analysis (LAMDA) laboratory. Report sections describe specific aspects of the research activity. Section 1 reviews the general background, material, and specimen origin, and provides detailed information on material composition and damage dose level. Section 2 describes operations performed with irradiated specimens at the hot cell facility and LAMDA laboratory. At the moment, tensile specimen preparation is complete, and the tensile bars are ready for shipment to the University of Michigan for stress corrosion cracking testing. Section 3 discusses the microstructure of archive alloys and post-irradiation material properties; because of the very limited timeframe, only selected results are given on the magnetic properties of the irradiated specimens. Section 4 summarizes the work performed, presents conclusions, and discusses future activities within the line of research.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Delayed expression of hpS2 and prolonged expression of CIP1/WAF1/SDI1 in human tumour cells irradiated with X-rays, fission neutrons or 1 GeV/nucleon Fe ions

PURPOSE: Differences in gene expression underlie the phenotypic differences between irradiated and unirradiated cells. The goal was to identify late-transcribed genes following irradiations differing in quality, and to determine the RBE of 1 GeV/n Fe ions. MATERIALS AND METHODS: Clonogenic assay was used to determine the RBE of Fe ions. Differential hybridization to cDNA target clones was used to detect differences in expression of corresponding genes in mRNA samples isolated from MCF7 cells irradiated with iso-survival doses of Fe ions (0 or 2.5 Gy) or fission neutrons (0 or 1.2 Gy) 7 days earlier. Northern analysis was used to confirm differential expression of cDNA-specific mRNA and to examine expression kinetics up to 2 weeks after irradiation. RESULTS: Fe ion RBE values were between 2.2 and 2.6 in the lines examined. Two of 17 differentially expressed cDNA clones were characterized. hpS2 mRNA was elevated from 1 to 14 days after irradiation, whereas CIP1/WAF1/SDI1 remained elevated from 3 h to 14 days after irradiation. Induction of hpS2 mRNA by irradiation was independent of p53, whereas induction of CIP1/WAF1/SDI1 was observed only in wild-type p53 lines. CONCLUSIONS: A set of coordinately regulated genes, some of which are independent of p53, is associated with change in gene expression during the first 2 weeks post-irradiation.

NASA Discipline Radiation Health↗

Ionizing Radiation Affects Gene Expression in Mouse Skin and Bone

Future long-duration space exploration beyond low earth orbit will increase human exposure to space radiation and microgravity conditions as well as associated risks to skeletal health. In animal studies, radiation exposure (greater than 1 Gy) is associated with pathological changes in bone structure, enhanced bone resorption, reduced bone formation and decreased bone mineral density, which can lead to skeletal fragility. Definitive measurements and detection of bone loss typically require large and specialized equipment which can make their application to long duration space missions logistically challenging. Towards the goal of developing non-invasive and less complicated monitoring methods to predict astronauts' health during spaceflight, we examined whether radiation induced gene expression changes in skin may be predictive of the responses of skeletal tissue to radiation exposure. We examined oxidative stress and growth arrest pathways in mouse skin and long bones by measuring gene expression levels via quantitative polymerase chain reaction (qPCR) after exposure to total body irradiation (IR). To investigate the effects of irradiation on gene expression, we used skin and femora (cortical shaft) from the following treatment groups: control (normally loaded, sham-irradiated), and IR (0.5 Gy 56Fe 600 MeV/n and 0.5 Gy 1H 150 MeV/n), euthanized at one and 11 days post-irradiation (IR). To determine the extent of bone loss, tibiae were harvested and cancellous microarchitecture in the proximal tibia quantified ex vivo using microcomputed tomography (microCT). Statistical analysis was performed using Student's t-test. At one day post-IR, expression of FGF18 in skin was significantly greater (3.8X) than sham-irradiated controls, but did not differ at 11 days post IR. Expression levels of other genes associated with antioxidant response (Nfe2l2, FoxO3 and Sod1) and the cell cycle (Trp53, Cdkn1a, Gadd45g) did not significantly differ between the control and IR groups at either time point. Radiation exposure resulted in a 27.0% increase in FGF18-positive hair follicles at one day post-IR and returned to basal levels at 11 days post-IR. A similar trend was observed from FGF18 gene expression analysis of skin. In bone (femora), there was an increase in the expression of the pro-osteoclastogenic cytokine, MCP-1, one day after IR compared to non-irradiated controls. FGF18 expression in skin and MCP- 1 expression in bone were found to be positively correlated (P less than 0.002, r=0.8779). Further, microcomputed tomography analysis of tibia from these animals showed reduced cancellous bone volume (-9.9%) at 11 days post- IR. These results suggest that measurements of early radiation induced changes in FGF18 gene expression in skin may have value for predicting subsequent loss of cancellous bone mass. Further research may lead to the development of a relatively simple diagnostic tool for bone loss, with the advantage that hair follicles and skin are relatively easy to acquire from human subjects.

space radiation↗

Cardiovascular Responses to Simulated Spaceflight: Molecular Signatures and Surrogate Outputs to Measure CVD Risk

During extended space missions beyond low Earth orbit, astronauts will encounter prolonged periods of weightlessness and low dose space radiation. Previous studies have shown that exposure to small doses of high LET radiation (< 50 cGy) can lead to both short-term and long-term alterations in heart function, structure and underlying molecular mechanisms. In this study, we aim to identify the molecular signature associated with the cardiovascular response to simulated galactic cosmic radiation (5-ion GCR) alone or in combination with simulated weightlessness at time intervals relevant to mission length and recovery. Additionally, we aim to determine whether sex impacts cardiovascular responses to these spaceflight factors. Our overarching goal is to enhance our understanding of the cardiovascular risks associated with extended space missions and the clinical endpoints they suggest. We hypothesize that exposure to simulated space radiation leads to enduring alterations in the transcriptome, redox signaling and cytokine environment of cardiovascular tissue, some which have known links with reduced cardiovascular performance, aging, and increased risk of cardiovascular disease (CVD). Furthermore, we posit that simulated space radiation exposure in combination with simulated microgravity exacerbates cardiovascular deficits compared to single factor exposure. Female and male C57BL/6J mice, aged 23-24 weeks, were exposed to a single dose of 5, 15, or 50 cGy of 5-ion GCR, or sham-treated (0 cGy). Euthanasia was performed at 14 days and ~4 months post-irradiation. Hearts, aorta and blood plasma were collected shortly thereafter. RNA-sequencing of left ventricles at ~4 months post-GCR exposure revealed sex differences in the heart transcriptome with a few genes showing radiation-dependent changes in expression levels. Notably, some of the differentially expressed genes in 15 and 50 cGy GCR groups are known to play roles in the development of CVD. Analysis of protein levels of a subset of inflammatory cytokines in the heart indicated sex differences but no differences between sham and 50 cGy groups. Results also showed correlations among differentially expressed genes and a subset of inflammatory cytokines, with some correlations altered by GCR exposure. These findings suggest that GCR exposure can modify protein and gene networks linked to inflammation and CVD progression. In the aorta, telomere lengths were comparable across treatment groups sexes. Mitochondrial copy number is a biomarker for mitochondrial function with decreased copy numbers associated with cardiometabolic disease traits. Mitochondrial copy numbers of aorta also showed no sex nor dose differences. In a second study, mice underwent one week of simulated microgravity by hindlimb unloading (HU) and then exposed to a single dose of 15 cGy of 5-ion GCR. HU was conducted for an additional two weeks following GCR exposure. Single factor exposure groups (HU or GCR only) also were included in the study. Euthanasia was then performed and the same tissues were collected. Protein levels of select inflammatory cytokines in the heart showed sex-dependent differences in expression. In the aorta, telomere lengths and mitochondrial copy number also showed sex differences. In summary, our results indicate differences between sexes in biomarkers related to cardiovascular health. Exposure to 5-ion GCR or HU, alone or in combination, did not result in changes in most of the cardiovascular biomarkers that were examined. However, in the heart, simulated space radiation at doses of 15 and 50 cGy led to long-term alterations in the expression levels of a small group of genes known to be associated with the progression of CVD. The long-term transcriptomic changes resulting from exposure to simulated space radiation should be carefully investigated to mitigate adverse cardiovascular events during and after deep space missions. Our results also highlight the importance of sex-specific strategies in monitoring and maintaining cardiovascular health during and after deep space missions.

cardiovascular↗

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2017, the activities for the INL ART TDO fuel development are focused on preparations for the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea. With the successful transfer of fuel fabrication technology to BWXT, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing technical support, verification of BWXT laboratory results, and fabricating surrogate fuel materials for university collaborations. ORNL will also provide analytical support to BWXT and INL for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

AGR-5/6/7 Fuel Specification

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) program is developing and qualifying fuel for a first of a kind high temperature gas-cooled reactor (HTGR) and associated systems to generate electrical power and produce hydrogen or process heat. One of the critical technology development areas of the program is nuclear fuel development and qualification. Fuel qualification is needed to support licensing, operation, and near term deployment of an HTGR reactor.5 Toward this end, the Advanced Gas Reactor (AGR) program includes manufacturing fuels for a series of irradiation experiments in the Advanced Test Reactor (ATR) at INL.5 The fifth, sixth, and seventh irradiation experiments, called “AGR 5/6/7,” will use fuel manufactured entirely in engineering scale equipment to demonstrate the performance of fuels fabricated in a production environment. This fuel will be composited from multiple low enriched uranium carbide/oxide (LEUCO) kernel forming runs, multiple tristructural isotropic (TRISO) coating runs, and multiple overcoating runs. The fuel will be composited into lots as kernels, TRISO particles, and compacts. Previous AGR fuel specifications1,2,8,9 provide the basis for most of the following fuel specifications, which are specific to the AGR 5/6/7 experiment, and may not be immediately applicable to fuel produced for an HTGR reactor. Fuel specifications will be influenced by the final reactor design.

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2017, the activities for the INL ART TDO fuel development are focused on preparations for the AGR-5/6/7 experiments and technical support to the Generation IV (GenIV) pre-irradiation leach-burn-leach round robin effort with China and South Korea. With the successful transfer of fuel fabrication technology to BWXT, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing technical support, verification of BWXT laboratory results, and fabricating surrogate fuel materials for university collaborations. ORNL will also provide analytical support to BWXT and INL for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

Support for Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) and the transition to the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Technology Development Office (TDO) since the inception of the program. In this effort, ORNL has fabricated tristructural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2, AGC-4, and AGR 3/4 experiments and for university studies, and provided characterization, coating, and compaction support for INL and BWX Technologies Nuclear Operations Group (BWXT). In fiscal year (FY) 2018, the activities for the INL ART TDO fuel development are focused on final characterization to support the AGR-5/6/7 experiments and technical support to the Generation-IV International Forum (GIF) pre-irradiation leach-burn-leach (LBL) round robin effort with China and South Korea.

AGR-5/6/7 Fuel Fabrication Feedstock Chemical Purity Specifications

This specification has been prepared to provide fuel fabricators with minimum chemical purity specifications intended to ensure that fuel components have consistent properties and that impurity levels in the as-fabricated fuel is sufficiently low to minimize in-pile TRISO particle failures and to enhance fission product retention. Process chemical purity specifications are given for the following processes: • Kernel forming and sintering • TRISO particle coating • Matrix preparation • TRISO overcoating • Compact carbonization and heat treatment

Support for INL ART TDO Advanced Gas Reactor Fuel Fabrication and Characterization

Oak Ridge National Laboratory (ORNL) has participated in the Advanced Gas Reactor (AGR) program for the Next Generation Nuclear Plant (NGNP) since the inception of the program. In this effort, ORNL has fabricated tri-structural isotropic (TRISO) fuel and fuel compacts for AGR-1, AGR-2, and AGR-3/4 experiments, provided matrix only components for AGC-2 and AGR-3/4 experiments and for university studies, and provided characterization, coating, and compaction support for Babcock and Wilcox (B&W). In FY 2017, the activities for the Very-High Temperature Reactor (VHTR) Technology Development Office (TDO) fuel development are focused on preparations for the AGR-5/6/7 experiments. With the successful transfer of fuel fabrication technology to B&W, ORNL’s involvement has shifted toward transferring fuel characterization methods, providing support for fabricating surrogate fuel materials international collaborations. ORNL will also provide analytical support to B&W for uranium dispersion analysis and confirmatory fuel characterization analyses as needed.

Demonstrate Thermal Property Measurement on Irradiated Fuel in IMCL

In order to assist in the post-irradiated measurement of thermal conductivity, the Advanced Fuels Campaign (AFC) of the DOE’s Nuclear Technology Research and Development (NTRD) program has invested in the design and build of a thermal conductivity microscope (TCM) system [3] based upon thermoreflectance methods. The TCM was installed into the Thermal Properties Cell (TPC) in the Irradiated Materials Characterization Laboratory at the Idaho National Laboratory’s (INL) Materials and Fuels Complex (MFC) in December 2018. In addition to the TCM, a Netzsch Laser Flash analyzer (LFA) instrument (model LFA427HC) was designed and built via vendor contract and installed into the TPC in April 2019. These instruments were specifically designed for the measurement of thermal conductivity and diffusivity on irradiated fuels and materials. The TPC began radiological operations on September 12, 2019. In order to demonstrate the successful installation of this new measurement capability the U-Zr fuel system, specifically U-10 wt.% Zr, was chosen as the first material to be evaluated due to its importance to the NTRD program. The results of the first four measurements conducted in the TPC on the LFA and the TCM are given in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NDMAS System and Process Description

The U. S. Department of Energy (DOE) has made a significant investment in research to develop the next generation of reactor technologies as well as to improve the performance and lengthen the life cycle of existing nuclear reactors. Data collected to demonstrate new concepts may also be used in the future to support licensing of these technologies. Provenance of these data must be preserved. The Nuclear Data Management and Analysis System (NDMAS) was established to manage and preserve data collected by fuels and materials research conducted by the high-temperature, gas-cooled reactor program. The scope of NDMAS is expanding to include other nuclear research programs that have the shared need to preserve the provenance of research data. Nuclear research funded by DOE is conducted by Idaho National Laboratory (INL), universities, other national laboratories, foreign research partners, and private companies. This research will generate a large amount of data from a variety of sources over a period of many years. Managing the data generated by the research and development projects presents a significant challenge for retaining data integrity and availability.

Statistical Sampling Plan for AGR-5/6/7 Fuel Materials

Kernels, coated particles, and compacts will be manufactured to predefined specifications for irradiation in the AGR 5/6/7 experiments. Many tests will be performed for quality assurance and most of these tests correspond to criteria that have to be met with 95% confidence, based on random samples. This report describes the statistical acceptance testing methods and gives minimum sample sizes for kernels to achieve 95% confidence. Coated particles and compacts will be addressed in revisions to this report. The methods used for sample size determination are described in detail in INL/EXT 05 00349, Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials (Einerson 2005). The statistical approach is to collect a representative sample, apply an acceptance test to the sample, and accept or reject the product based on this test for each criterion. The criteria or “product specifications” are contained in SPC-1352, “AGR 5/6/7 Fuel Specification,” Revision 1 and are duplicated in this plan.

The Impact of Furnace-Testing on the Microstructure of an Irradiated U-Mo Fuel Plate

The Material Management and Minimization Program (formerly known as the Reduced Enrichment for Research and Test Reactor Program) is responsible for converting research reactors that use highly enriched uranium (HEU) fuels to ones that use low-enriched uranium (LEU) fuels. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. In addition to the reactor tests, there is interest in generating information about the transient behavior of irradiated fuel plates. For example, it is of interest to determine the temperature at which a blister will form on the surface of a fuel plate, due to fission gas release from the fuel meat, for a particular fuel plate design. Since in-pile testing to determine the 'blister threshold' (the temperature where a blister forms on the surface of a fuel plate) in plate-type fuel is very difficult and expensive and has historically been deemed impractical for fuel qualification testing, an out-of-pile 'blister test' has been developed and employed. This post-irradiation blister anneal test was developed as a conservative simulation of in-pile behavior and has been routinely used for many research reactor fuel designs currently in use. In this test, the irradiated fuel plate is inserted into a furnace at a given temperature, soaked for a period of time, and removed for inspection. If no visible blisters appear, the furnace temperature is increased, and the plate is soaked and inspected again. This process continues until blisters are observed. Recently, a section from an irradiated U-7Mo dispersion fuel plate with Al-4043 matrix, which has a nominal composition of 4.81Si-0.20Fe-0.14Ti-0.16Cu-0.01Cr 0.01Mn-bal Al, was tested in a furnace located in the Hot Fuel Examination Facility, located at the Idaho National Laboratory, up to 500 deg. C, at which point blisters were observed on the surface of the fuel plate section. A sample near the blistered region of the fuel plate section was then characterized using scanning electron (SEM) and transmission electron microscopy (TEM). This paper discusses the results of this characterization and how they compare to those generated for a similar sample characterized after fabrication but before irradiation and for a sample characterized after irradiation but before furnace testing. One goal of this work is to identify the changes in the microstructure during blister testing that contribute to the eventual release of fission gas that results in the development of blisters. SEM analysis of the furnace-tested sample showed that in some locations of the U-7Mo particles, Al-4043 matrix constituents (primarily Si) had penetrated the microstructure along grain boundaries. This provided potential pathways for increased fission gas migration in the fuel meat microstructure during the furnace heating process. Some attack of these boundaries was also observed in the as-fabricated U-7Mo/Al-4043 sample, but there appeared to be additional attack during the furnace annealing. Besides the changes observed in grain boundary features, the TEM analysis showed that the fission gas bubbles that are typically present in the intra-granular regions of the as-irradiated U-7Mo microstructure as a fission gas bubble superlattice had changed in size, morphology, and distribution. The result of this may be an increase in the mobility of the fission gas in intra-granular regions of the fuel. The complete implications of the changes in fission gas morphology and distribution in the fuel meat microstructure and their impact on blistering continues to be investigated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fission Product and Actinide Distribution in the SiC Layer of Fissioned Triso Fuel Particles

Tri-structural Isotropic (TRISO) encapsulated fuel particles are used in next generation high-temperature gas-cooled nuclear reactors. The integrity of the fuel over a wide range of temperatures is integral to the safety of an operating nuclear reactor. It has been observed that fission products are released at high temperatures that simulate hypothesized accident scenarios. The particles under study were from the AGR-1 fuel development and qualification program. They consist of an enriched UCO fuel kernel surrounded by a dense carbon buffer layer (100 {mu}m), an inner pyrolytic carbon (IPyC) layer (40 {mu}m), a silicon carbide (SiC) layer (35 {mu}m), and an outer pyrolytic carbon (OPyC) layer (40 {mu}m). The SiC layer acts as the main diffusion barrier to fission products, but has been demonstrated to fail due to release of Ag through intact particles and local corrosion from metallic fission products such as Pd. In this work, we use transmission electron microscopy and diffraction techniques to study the structure and dispersion of release products within the SiC of TRISO particles as a function of annealing temperature used to simulate margin temperatures for in-reactor accident scenarios. In order to improve our understanding of fission product retention as a function of temperature, we have the unique opportunity to examine irradiated, fissioned TRISO particles to study post-irradiation effects on these materials. This data will contribute to current models used to predict the behavior of fission products in the SiC layer of TRISO fuels.

07 ISOTOPE AND RADIATION SOURCES↗