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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Measurement of Fission Product Concentration Profiles in AGR-3/4 TRISO Fuel Graphitic Matrix and Nuclear Graphites

The third Advanced Gas Reactor (AGR) irradiation experiment, AGR-3/4, was designed to investigate the migration of fission products in fuel compact graphitic matrix and reactor graphite components. Using destructive methods, radial fission product concentration profiles were measured for gamma-emitting fission products (e.g., Ag-110m, Cs-134, and Eu-154) and beta-emitting Sr-90 in irradiated graphitic and graphite components from six different AGR-3/4 irradiation capsules. These new measured concentration profiles can now be compared to non-destructive measurements and fission product transport simulations and will be used to derive new diffusivities and sorptivities to support refinement of fission product transport models and high-temperature gas-cooled reactor (HTGR) source-term analyses. Each capsule in the AGR-3/4 experiment had four fuel compacts in the middle of two concentric rings of graphitic matrix material, PCEA graphite, or IG-110 graphite. In addition to the approximately 1898 tristructural isotropic (TRISO) coated particles in each compact, there were 20 designed to fail (DTF) particles coated only in pyrocarbon so that they released fission products into the surrounding cylindrical rings of carbonaceous materials. Destructive sampling of the rings involved machining/milling material from around the circumference of the rings, collecting that material, and performing radiochemical analyses on it. Milling operations were performed in multiple steps or segments, and each segment was generally 0.508 mm (0.020 in) thick. Knowing the radial position at which each segment was milled, the volume of the milled material at each segment, and the fission product content in each segment, the radial fission product concentrations were constructed for select isotopes in each ring. Ag-110m profiles had the most variation. Some profiles were peaked at an inner or outer surface. Some were peaked at the middle of the ring wall thickness. Some increased radially outward, and some decreased radially outward. These types of variations and the fact that the measured profiles do not generally compare favorably with the transport model employed for AGR-3/4 may adversely impact the ability to extract reasonable transport parameters for this isotope. In many cases, the Cs-134 profiles decreased somewhat linearly in the outward radial direction, and in cursory comparisons, the shapes of these profiles appeared similar to those from model predictions. The step changes in concentration across the inner-outer ring gap were generally consistent with the model predictions as well. In some cases, there were local maxima in concentration at the outer surface of the rings. This suggests that fission products could have transported in the small gaps between the inner ring and the outer ring and between the outer ring and the sink ring such that some portion of a given fission product can bypass diffusion through the ring itself. The analysis of the small nubs on the outer surfaces of some of the outer rings revealed fission product concentrations in the nubs that were often higher than in the outermost segments of the rings. This further supports the hypothesis that short-circuit, gap transport occurred, causing relatively high surface concentrations on the outer surfaces of the rings. Eu-154 and Sr-90 profiles tend to have very similar shapes, suggesting that they transport via the same mechanisms. The observed profiles were indicative of a transport process where the isotopes are sorbed on the inner surface of the ring, but diffusion into the ring from that surface is quite slow. Some elevated concentrations of Sr-90 (relative to Eu-154) on the outer surface of a ring suggested that rapid, gas-gap transport of gaseous precursor Kr-90 and volatile Rb-90 could have occurred prior to their decaying to Sr-90. Overall, the Eu 154 and Sr-90 profiles were still very similar, which indicates that the transport of short-lived Sr-90 precursors is not a major effect. In some capsules, the qualitative Sr-90 behavior across the ring gaps was consistent with the model (using the available legacy Sr-90 transport parameters), but in other capsules the model was inconsistent with the measurements and seems to underestimate the amount of Sr-90 in the outer rings. The total ring Sr-90 inventories were estimated for all the rings that were subject to physical sampling. These results will be used to adjust the predicted particle and/or compact releases used in the AGR-3/4 fission product transport model. Given the different irradiation temperatures among the capsules and the rings, it was not possible to discern fundamental differences in the transport of isotopes within the different carbon materials, i.e., graphitic matrix, IG-110, or PCEA. It may be possible to do this in the course of determining transport from the concentration profiles in future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reconstruction of Fission Product Distribution from Tomographic Scans in TRISO Fuel Graphitic Matrix and Nuclear Grade Graphites

An image reconstruction method was developed to rectify shortcomings of earlier methods that became apparent as destructively sampled data became available. This reconstruction method was applied to the tomographic gamma scans of nuclear graphite and graphitic matrix samples from AGR-3/4. There is generally agreement between profiles measured via destructive sampling of these rings and profiles from tomographic reconstruction.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reconstruction of Fission Product Distribution from Tomographic Scans in TRISO Fuel Graphitic Matrix and Nuclear Grade Graphites

An image reconstruction method was developed to rectify shortcomings of earlier methods that became apparent as destructively sampled data became available. This reconstruction method was applied to the tomographic gamma scans of nuclear graphite and graphitic matrix samples from AGR-3/4. There is generally agreement between profiles measured via destructive sampling of these rings and profiles from tomographic reconstruction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Determining the oxidation behavior of matrix graphite

This work presents the oxidation behavior of matrix graphite in air. Matrix graphite, graphite powder/flakes bonded by a small amount of non-graphitic carbon, surrounds coated fuel particles in order to form cylindrical fuel compacts (in prismatic core designs) or spheres (in pebble-bed reactor designs). This work focuses on oxidation tests conducted on two matrix graphite materials, one provided by Kairos Power and the other A3 matrix graphite. Some of the tests followed American Society for Testing and Materials (ASTM) oxidation testing standards using a vertical furnace system and others were performed in a thermogravimetric analyzer (TGA). It was determined that, at temperatures of 450 °C–700 °C, the oxidation rate of the Kairos matrix graphite follows the Arrhenius equation. In comparison with A3 matrix graphite, the Kairos matrix graphite shows better oxidation resistance at high temperatures (≥550 °C), but also a higher oxidation rate at low temperatures. Both the A3 matrix graphite and the Kairos matrix graphite materials may experience preferential oxidation of the partially graphitized binder. An oxygen penetration gradient was also observed when using the three characterization methods (i.e., optical microscope, x-ray tomography [XCT], and density profile by the lathe) enlisted in this research. In conclusion, the oxygen penetration depth increases with decreasing isothermal oxidation temperature, while the center of the oxidized samples (10% weight loss) remains almost untouched even at 500 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Determination of oxidation rates and volatile oxidation products for HTGR graphite matrix material exposed to steam atmospheres

High-temperature gas-cooled reactors (HTGRs) in operation use tristructural isotropic (TRISO) particles embedded in graphite and carbonized resin matrix to form the fuel element. This graphite matrix material serves as a supportive structural element, heat transfer medium, and neutron moderator. In HTGR designs, fuel compacts are exposed to helium coolant, which facilitates high outlet temperatures (750°C 2 O, 800°C 2 , and H 2 , are quantified in varied oxidant atmospheres using a coupled thermogravimetric analyzer and mass spectrometer. Furthermore, oxidation rates reported here for varied steam (H 2 O [g]) atmospheres are predominantly linear and comparable with literature values in the range of tested temperatures (800–1200°C). Changes in dominant matrix oxidation products from primarily CO to a mixture of CO, CO 2 , and H 2 were observed at higher temperatures (≥1000°C) and steam atmospheres (≥5 kPa pH 2 O). Kinetic data indicates that there was no shift in oxidation regime with chemical oxidation occurring at all temperatures and H2O (g) atmospheres tested. These data provide insight into the oxidation behavior of graphite matrix material and will inform future testing conditions, notably mixed atmospheric conditions, of HTGR fuel elements.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Hydrogen uptake in graphite matrix at high temperature

Tritium management is a critical challenge for the next generation of nuclear reactors, such as Fluoride Salt Cooled High Temperature Reactors (FHRs) and High Temperature Gas-cooled Reactors (HTGRs), due to the higher production rate (up to 10,000 times) than conventional Light Water Reactors (LWRs). Graphitic materials employed as moderator, reflector, and fuel pebbles offer a potential pathway for tritium recovery by serving as a sink for tritium. Prediction of uptake capacity under reactor relevant conditions remains a challenge due to a lack of low partial pressure data and significant inter-grade variability of graphite. This study addresses these gaps by providing a comprehensive characterization of hydrogen (as a tritium surrogate) uptake and release behavior in the A3-3 graphite matrix (GM) used in fuel pebbles. Uptake measurements are performed at reactor relevant temperatures of 600- 800 °C, 1-200 Torr hydrogen pressure, and 15-120 min equilibration time, followed by thermal desorption spectroscopy up to 1100 °C. Uptake experiments at different equilibration times demonstrate the role of kinetics in hydrogen uptake, which can be modeled as a diffusion-with-trapping process. In the thermodynamics limit, the Sips adsorption model is shown to capture the uptake in A3-3 GM well. Our campaign provides a set of new results for hydrogen uptake in A3-3, including limiting uptake capacity at 600 °C, apparent diffusion coefficient at 600 °C, and the first estimates of the FHR/HTGR relevant (600 °C, 20 Pa partial pressure) equilibrium uptake capacity and time to saturation. Desorption data highlights a new site for hydrogen uptake, not observed in nuclear graphite, which we attribute to the non-graphitized binder. Using the Kissinger method, we estimate activation energy for release from the desorption peaks, confirming the activation energy for release from the basal planes and providing the first estimate for the activation energy of release from the binder.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Matrix Graphite Material Models In Pebbles and Compacts For Bison

The cores and reflectors in high-temperature gas-cooled reactors (HTGRs) are made of graphite materials, with the graphite acting as a moderator, a fuel host matrix, or the foundation for various structural components. This study aims to survey the models in the literature for graphite materials being used as host matrices in pebble/fuel compacts and to implement those surveyed models into Bison to conduct an early assessment of graphite's thermo-mechanical response under various reactor conditions. In this study, thermal (e.g., thermal conductivity, and specific heat capacity) and mechanical (e.g., elastic properties, thermal expansion, irradiation-induced dimensional changes, and irradiation-induced creep) material models for various graphite grades (e.g., H-451, IG-110, G-348, 2020, A3-3, and A3-27) are incorporated into Bison. Two benchmark problems are then exercised utilizing these new graphite-related capabilities: (1) modeling an Advanced Gas Reactor (AGR)-2 fuel compact, and (2) modeling the debonding of a particle-matrix interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal energy storage composites with preformed expanded graphite matrix and paraffin wax for long-term cycling stability and tailored thermal properties

Harvesting solar energy, preventing hot spots in electronics, transport of temperature-sensitive materials, and capture and repurposing of thermal energy require a latent heat thermal energy storage (TES) system to store/discharge heat repeatedly. For the practical application of phase change material (PCM) composites within TES systems, reliable thermal performance throughout its operational lifetime is essential. Nevertheless, the reliability of thermal conductivity in multi-phase composites over relevant numbers (>10 3 ) of melt/freeze cycles has barely been studied, particularly for composites containing fillers for thermal conductivity enhancement. Here, we introduce a preform-type expanded graphite (EG)/paraffin wax composite possessing highly robust heat transfer and storage properties even after 10,000 melt/freeze cycles. To achieve such excellent reliability, comparative studies on the combined influence of fabrication process, particle size, EG vol%, binder amount, and compaction on both magnitude and robustness of thermal conductivity were undertaken. Our parametric study has yielded a trade-off between thermal conductivity and latent heat. Based on our modeling, 20 vol% EG approaches the case where all EG particles are well-connected thermally while 10 vol% EG is close to loosely connected fillers in the matrix. Thermal conductivity of our paraffin composites containing 20 vol% EG (25.1 W·m -1 ·K -1 ) is highest among other EG/paraffin composites without aligned EG in the literature. After 10,000 thermal cycling, negligible conductivity fading was observed for the 10 vol% EG composite, while reduction in latent heat remained within 10% for all 10, 14, 17 and 20 vol% EG samples. Here we anticipate this work provides insight on suitable recipe for desirable magnitude and robustness of thermal conductivity of EG/paraffin composites.

25 ENERGY STORAGE↗
DOE OSTI2021

Oxidation behaviors of matrix-grade graphite during water vapor ingress accidents for high temperature gas-cooled reactors

Water ingress into cores of high temperature gas-cooled reactors (HTGRs) can cause serious safety problems in graphitic components, especially for matrix graphite, which is more vulnerable than nuclear graphite. Here in this study, oxidation behaviors of a matrix-grade graphite, ARB-B1, were investigated under simulated water ingress accidental conditions. The oxidation tests were conducted in water vapor/helium mixed atmospheres at elevated temperatures up to 1200 °C. Oxidation rates and activation energy were evaluated by mass loss measurements. The activation energy was 121.5 kJ/mol in the considered temperature range, indicating only one oxidation regime-in-pore diffusion assisted by cracks. The cracks originated from the fabrication process, acting as penetration paths for the oxidant, played an important role in the oxidation behaviors. Ungraphitized binder oxidation occurred at all temperatures, while apparent oxidation of filler particles started from 1100 °C. Moreover, oxidation effects on compressive properties and hardness of the matrix graphite were discussed in terms of oxidized microstructures. This first detailed study on matrix graphite oxidation in water vapor provides important kinetics knowledge and microstructural and mechanical data for accidental scenarios of HTGRs as well as for development of new matrix graphite materials.

36 MATERIALS SCIENCE↗
NASA NTRS1991

Thermal conductivity and thermal expansion of graphite fiber/copper matrix composites

The high specific conductivity of graphite fiber/copper matrix (Gr/Cu) composites offers great potential for high heat flux structures operating at elevated temperatures. To determine the feasibility of applying Gr/Cu composites to high heat flux structures, composite plates were fabricated using unidirectional and cross-plied pitch-based P100 graphite fibers in a pure copper matrix. Thermal conductivity of the composites was measured from room temperature to 1073 K, and thermal expansion was measured from room temperature to 1050 K. The longitudinal thermal conductivity, parallel to the fiber direction, was comparable to pure copper. The transverse thermal conductivity, normal to the fiber direction, was less than that of pure copper and decreased with increasing fiber content. The longitudinal thermal expansion decreased with increasing fiber content. The transverse thermal expansion was greater than pure copper and nearly independent of fiber content.

Ellis, David L.↗
DOE OSTI2025

Neutron and x-ray computed tomography of a natural uranium tristructural isotropic (TRISO) fuel compact

A natural uranium-based, unirradiated tristructural isotropic (TRISO) fuel compact was nondestructively imaged using both X-ray (XCT) and neutron computed tomography (nCT). While XCT of compacts can provide information on fuel kernels, imaging artifacts preclude examination of the graphite matrix. In this work, nCT was used for the first time on a TRISO compact to examine the graphite matrix. A crack was clearly resolved within the graphite matrix, proving that nCT is a viable tool for nondestructive volumetric examination of the matrix material in TRISO fuel compacts. The XCT and nCT data were then fused together to create a more comprehensive dataset containing both matrix and fuel kernels.

36 - MATERIALS SCIENCE↗
NASA NTRS1980

Improved fiber retention by the use of fillers in graphite fiber/resin matrix composites

A potential problem in the use of graphite fiber reinforced resin matrix composites is the dispersal of graphite fiber during accidental fires. Airborne electrically conductive fibers originating from burning composites could enter and cause shorting in electrical equipment located in surrounding areas. A variety of matrix fillers have been tested for their ability to prevent loss of fiber from graphite fiber/PMR polyimide and graphite fiber/epoxy composites in a fire. The fillers tested included powders of boron, boron carbide (B4C), lime glass, lead glass, and aluminum. Of these fillers, boron was the most effective and prevented any loss of graphite fiber during burning. Mechanical properties of composites containing boron filler were measured and compared to those of composite containing no filler.

Gluyas, R. E.↗
NASA NTRS1996

Graphite Fiber Textile Preform/Copper Matrix Composites

Graphite fiber reinforced/copper matrix composites are candidate materials for critical heat transmitting and rejection components because of their high thermal conduction. The use of textile (braid) preforms allows near-net shapes which confers additional advantages, both for enhanced thermal conduction and increased robustness of the preform against infiltration and handling damage. Issues addressed in the past year center on the determination of the braid structure following infiltration, and the braidability vs. the conductivity of the fibers. Highly conductive fibers eventuate from increased graphitization, which increases the elastic modulus, but lowers the braidability; a balance between these factors must be achieved. Good quality braided preform bars have been fabricated and infiltrated, and their thermal expansion characterized; their analytic modeling is underway. The braided preform of an integral finned tube has been fabricated and is being prepared for infiltration.

Filatovs, G. J.↗
NASA NTRS1995

Graphite fiber textile preform/copper matrix composites

Graphite fiber reinforced/copper matrix composites have sufficiently high thermal conduction to make them candidate materials for critical heat transmitting and rejection components. The term textile composites arises because the preform is braided from fiber tows, conferring three-dimensional reinforcement and near net shape. The principal issues investigated in the past two years have centered on developing methods to characterize the preform and fabricated composite and on braidability. It is necessary to have an analytic structural description for both processing and final property modeling. The structure of the true 3-D braids used is complex and has required considerable effort to model. A structural mapping has been developed as a foundation for analytic models for thermal conduction and mechanical properties. The conductivity has contributions both from the copper and the reinforcement. The latter is accomplished by graphitization of the fibers, the higher the amount of graphitization the greater the conduction. This is accompanied by an increase in the fiber modulus, which is desirable from a stiffness point of view but decreases the braidability; the highest conductivity fibers are simply too brittle to be braided. Considerable effort has been expended on determining the optimal braidability--conductivity region. While a number of preforms have been fabricated, one other complication intervenes; graphite and copper are immiscible, resulting in a poor mechanical bond and difficulties in infiltration by molten copper. The approach taken is to utilize a proprietary fiber coating process developed by TRA, of Salt Lake City, Utah, which forms an itermediary bond. A number of preforms have been fabricated from a variety of fiber types and two sets of these have been infiltrated with OFHC copper, one with the TRA coating and one without. Mechanical tests have been performed using a small-scale specimen method and show the coated specimens to have superior mechanical properties. Final batches of preforms, including a finned, near net shape tube, are being fabricated and will be infiltrated before summer.

Gilatovs, G. J.↗
NASA NTRS1991

Graphite Fiber/Copper Matrix Composites for Space Power Heat Pipe Fin Applications

High specific thermal conductivity (thermal conductivity divided by density) is a major design criterion for minimizing system mass for space power systems. For nuclear source power systems, graphite fiber reinforced copper matrix (Gr/Cu) composites offer good potential as a radiator fin material operating at service temperatures above 500 K. Specific thermal conductivity in the longitudinal direction is better than beryllium and almost twice that of copper. The high specific thermal conductivity of Gr/Cu offers the potential of reducing radiator mass by as much as 30 percent. Gr/Cu composites also offer the designer a range of available properties for various missions and applications. The properties of Gr/Cu are highly anisotropic. Longitudinal elastic modulus is comparable to beryllium and about three times that of copper. Thermal expansion in the longitudinal direction is near zero, while it exceeds that of copper in the transverse direction.

McDanels, David L.↗
NASA NTRS1981

New kind of type 3 chondrite with a graphite-magnetite matrix

Four clasts in three ordinary-chondrite regolith breccias are discovered which are a new kind of type 3 chondrite. As with ordinary and carbonaceous type 3 chondrites, they have distinct chondrules, some of which contain glass, highly heterogeneous olivines and pyroxenes, and predominantly monoclinic low-Ca pyroxenes. Instead of the usual, fine-grained, Fe-rich silicate matrix, however, the clasts have a matrix composed largely of aggregates of micron- and submicron-sized graphite and magnetite. The bulk compositions of the clasts, as well as the types of chondrules (largely porphyritic), are characteristic of type 3 ordinary chondrites, although chondrules in the clasts are somewhat smaller (0.1-0.5 mm). A close relationship with ordinary chondrites is also suggested by the presence of similar graphite-magnetite aggregates in seven type 3 ordinary chondrites. It is thought that this new kind of chondrite is probably the source of the abundant graphite-magnetite inclusions in ordinary-chondrite regolith breccias and that it may be more common than indicated by the absence of whole meteorites made of chondrules and graphite-magnetite.

Scott, E. R. D.↗
NASA NTRS1980

Improved fiber retention by the use of fillers in graphite fiber/resin matrix composites

A variety of matrix fillers were tested for their ability to prevent loss of fiber from graphite fiber/PMR polyimide and graphite fiber/epoxy composites in a fire. The fillers tested included powders of boron, boron carbide lime glass, lead glass, and aluminum. Boron was the most effective and prevented any loss of graphite fiber during burning. Mechanical properties of composites containing boron filler were measured and compared to those of composites containing no filler.

Gluyas, R. E.↗