Results for “Graphite Irradiation” — Page 12

DOE OSTI2021

Advanced Microscopy Report on UCO Fuel Kernels from Selected AGR-1 and AGR-2 Experiments

A variety of neutron irradiation experiments were completed at Idaho National Laboratory (INL) on tristructural isotropic (TRISO) coated particles contained in a graphitic matrix as part of the Advanced Gas Reactor (AGR) fuel development and qualification program. Although the initial advanced microscopy and microanalysis studies on AGR-1 particles focused predominantly on the SiC layer?s role as the main fission-product containment, studies were further expanded to evaluate UCO kernels of selected AGR-1 and AGR-2 coated particles. Fuel kernels tested in the AGR-1 and AGR-2 experiments is composed of a mixture of UO2 and UC, with a nominal as-fabricated stoichiometry of UO1.37 C0.35 and UO1.43 C0.39, respectively. These as-fabricated kernel-stoichiometry values are determined at the completion of kernel fabrication and do not account for any changes that will occur during the subsequent coating and compacting processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

ART Graphite R&D Program Status - 2020

Structure of the program Graphite material issues of interest How do things fit together Status of the program Baseline AGC Experiment Update AGC-4 and HDG-1 Irradiation Oxidation Modeling ASME Code development Discussion of how the data can be used Modeling behavior of core components Licensing/code development Interaction with vendors Collaborations (domestic and international)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1984

Electron radiation effects on the thermal expansion of graphite/resin composites

The effects of 1 MeV electron radiation on the thermal expansion characteristics of two graphite reinforced resin matrix composite systems were studied. Specimens of both graphite/epoxy (T300/5208) and graphite/polyimide (C6000/PMR15) were irradiated to a total dose of 6 x 10 to the 9th rads at two different rates. Dynamic mechanical analyses (DMA) were performed to study changes in resin chemistry. Thermal expansion results indicate that radiation did produce permanent residual strains of up to -70 x 10 to the -6th for the graphite/epoxy when exposed to temperatures up to +280 F. However, no permaanent changes in the coefficient of thermal expansion (CTE) were observed. No permanent residual strains or changes in the CTE attributable to radiation were observed for the graphite/polyimide specimens. DMA results indicate that electron radiation caused chemical changes in the epoxy matrix. These changes resulted in a lower glass transition temperature and broader 'rubbery region' which extended into the temperature range of the thermal expansion tests.

Bowles, D. E.↗

NASA NTRS2021

Understanding the Space Weathering of Mercury Through Laboratory Experiments

Introduction: Airless surfaces across the solar system are continually modified by energetic particles from solar wind and micrometeoroid bombardment [1,2]. This process is known as space weathering, and it alters the chemical, microstructural, and optical properties of surface regoliths on airless bodies, including Mercury. On the Moon and S-type asteroids, the reflectance spectral signatures of space weathering include reddening (increasing reflectance with increasing wavelength), darkening (lowering of reflectance), and the attenuation of characteristic absorption bands [2]. Such spectral changes are driven by the production of Fe-bearing nanoparticles(npFe) through both solar wind irradiation and micrometeoroid bombardment. While our understanding of space weathering for the Moon and near-Earth S-types asteroids is advanced, insight into how these processes operate on other planetary bodies is limited. In particular, Mercury experiences a uniquely intense space weathering environment than planetary counterparts at 1 AU, including a moreintense solar wind flux and higher velocity micrometeoroid impacts [4]. Additionally, Mercury has a surface composition unique in the inner solar system, including regions of the surface with very low albedo known as the low reflectance material (LRM), which is enriched in carbon, likely graphite, up to 4wt.% [5].In addition, the concentration of Fe across Mercury’s surface islow (<2 wt.%) compared to the Moonor S-type asteroids asteroids[6]. Our understanding of the effects of space weathering on C-rich and Fe-poor phases is limited. Since Fe plays a critical role inthe development of space weathering characteristicson other airless surfaces(e.g., npFe), its limited availability may significantly affect the development of space weathering features in Mercury surface materials. We can simulate space weathering processes in the laboratory to explore their effects on the microstructural, chemical, and spectral characteristics of Mercury surface materials[7]. Here we used pulsed laser irradiation to simulate the short duration, high-temperature events associated with micrometeoroid impacts. We performed coordinated analyses including reflectance spectroscopy and electron microscopy to investigate the spectral, chemical, and microstructural changes in these mercurian analog samples. Methods: For these experiments, we usedforsteritic olivine with varying FeOcontents, a mineral phase proposed to be abundant on the surface of Mercury. We mixed each sample with graphite to simulate LRM regions of the surface. Wesynthesized the olivinesamplesat 1-bar at NASA’s Johnson Space Centerand prepared pressed powder pellets for laser irradiation [8].We prepared three samples, each with a base layer of olivineto maintain structural integrity and topped witha surface layer containing the graphite-olivine mixture: 1) Sample SC-001 San Carlos olivine(Fo90.91),2)Sample F-S-002 with0.05 wt.% FeO olivine, and 3) F-T-004 with 0.53 wt.% FeO olivine. Each sample was mixed with 5 wt.% powdered graphite and had grain sizes ranging from 45to 125μm. We irradiated each sample usinga pulsed Nd-YAG laser, (l=1064 nm, ~6 ns pulse duration, energy of 48 mJ/pulse) while undervacuumat Northern Arizona University. The laser was rastered1x and then 5x over the surfaceof each sample to simulate progressive space weathering. We collected in situreflectance spectra from the samples after each laser pulse witha Nicolet IS50 Fourier-Transform Infrared spectrometer (lfrom 0.65-2.5 μm). We used an FEI Nova NanoSEM200scanning electron microscope (SEM) and a Hitachi TM4000 Plus benchtop SEM at Purdue University to image the surface morphology and topography of the samples. We extracted thin sections for analysis in the transmission electron microscope(TEM)using the FEI Helios NanoLab 660 focused ion beam (FIB) SEM at the University of Arizona.We performed analysis of the microstructural and chemical characteristics of the samples using the 200 keV JEOL 2500 scanning TEM at Johnson Space Center. Reflectance Spectroscopy Results:Reflectance spectra for each sample are shown in Fig. 1.SC-001:The spectrum of the unirradiated sample exhibits a weak 1.0 μm absorption feature, associated with Fe2+in the olivine,and low overall reflectance (Fig. 1a). Thereflectance and the depth of the absorption band increases after 1x laser raster but are at their lowest after 5x laser rasters.F-T-004:The unirradiated sample has a blue-sloped spectrum with low reflectance without identifiable absorption features (Fig. 1b). With progressive laser irradiation, the sample reflectance increases and becomes strongly red-sloped.F-S-002:The unirradiated sample exhibits a dark, blue-slopedspectrum. The brightness of the sample increases significantly from <0.2average reflectance over >0.8 reflectance in the most irradiated sampleand thespectral slope also becomesslightly reddened (Fig. 1c). Microstructural and Chemical Analysis: Two primary alteration textures were observed in the samples exposed to simulated space weathering: 1) fluffy C-rich,and 2) vesiculated melt. The fluffy C-rich texture is composed oflow-densitydeposits distributed across the surface of the sample(Fig. 2A). Analysis of a FIB section extracted from a low-density C-rich region in sample SC-001 reveals multiple globule-type deposits, discrete from stacked graphite, likely produced via melting from the laser irradiation [9].The vesiculated melt textureis smooth and uniformly distributed across isolated regions of the sample surface. The vesicles measure up to 100s of nmin diameter. Analysis of a FIB section from this texture was extracted from sample F-T-004 reveals a layer of amorphous melt material, close to 100 nm thick and uniform across the FIB section (Fig. 2B). Isolated regions of this melt layer contain small nanoparticles, <5 nm in diameter. Chemical analysis through energy dispersive X-ray spectroscopy reveals the composition of this layer is enriched in Si and depleted in Mg and O compared to the underlying sample. Implications for Space Weathering on Mercury: Previous experiments simulating space weathering of Mercury have showndarkening and reddening of spectra[7,10].However, our use of low-Fe materials and graphite to create a sample set more analogous to the mercurian surface. Our results indicate that sample composition plays a significant and important role in the space weathering of Mercury. In particular, our spectral data demonstrates a strong correlation between spectral slope, Fe content, and simulated space weathering. While the variation in FeO content between samples F-S-002 and F-T-004 is <0.6 wt.%, the spectra deviate from flat to strongly red-sloped(F-T-004). This reddeningmay be linked to the presence of very small nanoparticles observed in the melt textures extracted from sample F-T-004. For the SC-001 sample, the fluffy C-rich textures may be developed by the amalgamation of small graphite particles into these unique morphologies. Such observations indicate that space weathering on Mercury may result in both familiar and new microstructural and chemical characteristics. References: [1]Hapke B. (2001) J. Geophys. Res.-Planet.,106,10039–10073. [2]Pieters C.M. and Noble S.K. (2016) J. Geophys. Res-Planet., 121, 1865–1884. [3] Lucey P.G., and Riner, M.A. (2011) Icarus,212, 451-462.[4]CintalaM.J.(1992)J. Geophys. Res.-Planet.,97,947–973.[5]Klima R.L.et al.(2018)Geophys.Res.Letters, 45, 2945–2953. [6]Nittler L.R., et al. (2011) Science 333, 1847-1850.[7]Sasaki S. and Kurahashi E. (2004) Space weathering on Mercury, Adv.Space Res., 33, 2152-2155.[8] Vander KaadenK.E., et al. (2018) LPSCXLIX, Abstract 1230. [9] McGlaun M.L. et al. (2019) LPSCL, Abstract 2019. [10] TrangD.et al. (2018)LPSCXLIX,Abstract2083

M S Thompson↗

NASA NTRS1983

Effects of high energy radiation on the mechanical properties of epoxy graphite fiber reinforced composites

The effects of high energy radiation on mechanical properties and on the molecular and structural properties of graphite fiber reinforced composites are assessed so that durability in space applications can be predicted. A listing of composite systems irradiated along with the maximum radiation dose applied and type of mechanical tests performed is shown. These samples were exposed to 1/2 MeV electrons.

Gilbert, R. D.↗

NASA NTRS1987

Influence of electron radiation and temperature on the cyclic, matrix dominated response of graphite-epoxy

The effects of electron radiation and elevated temperature on the matrix-dominated cyclic response of standard T300/934 and a chemically modified T300/934 graphite-epoxy are characterized. Both materials were subjected to 1.0 x 10 to the 10th rads of 1.0 MeV electron irradiation, under vacuum, to simulate 30 years in geosynchronous orbit. Cyclic tests were performed at room temperature and elevated temperature (121 C) on 4-ply unidirectional laminates to characterize the effects associated with irradiation and elevated temperature. Both materials exhibited energy dissipation in their response at elevated temperature. The irradiated modified material also exhibited energy dissipation at room temperature. The combination of elevated temperature and irradiation resulted in the most severe effects in the form of lower proportional limits, and greater energy dissipation. Dynamic-mechanical analysis demonstrated that the glass transition temperature, T(g), of the standard material was lowered 39 C by irradiation, wereas the T(g) of the modified material was lowered 28 C by irradiation. Thermomechanical analysis showed the occurrence of volatile products generated upon heating of the irradiated materials.

Reed, Susan M.↗

DOE OSTI2020

Irradiation Creep in Materials

A knowledge of the dimensional stability of reactor structural components, under irradiation conditions, is of major importance in the design of thermal, fast, and fusion reactors. When subjected to simultaneous mechanical loading and irradiation, structural materials exhibit a visco-plastic deformation phenomenon, referred to as irradiation creep, which can be more rapid than the deformation occurring out of irradiation. In this article, the phenomenology of this peculiar behavior is described after a short history of its discovery. Then, the theoretical mechanisms proposed in the literature during these past 60 years are presented, with a special focus on mechanisms based on stress induced preferred absorption of point defects by dislocation loops, dislocations and grain boundaries and on mechanisms based on climb-enhanced glide of dislocations. These mechanisms are discussed in the light of experimental evidences. Finally, irradiation creep in various materials, such as zirconium alloys, austenitic stainless steels, nickel-based alloys, ferritic-martensitic steels and graphite, is described.

Onimus, Fabien↗

DOE OSTI2020

Impact of Grain Size on Performance Degradation of TREAT LEU

We argue that radiation damage induced degradation of thermal conductivity does not set a lower limit on fuel grain sizes for the low enriched uranium fuel design of the Transient Reactor Test Facility (TREAT). Earlier work reports that smaller grains cause a larger degradation of thermal conductivity than larger grains constraining the smallest feasible size of fuel grains. This work assesses TREAT’s transient performance in the presence of radiation damage. The difference between the two studies is in treating damaged and fresh graphite as serial (this work) or parallel (previous) heat resistors. We use a multiphysics model of TREAT fuel grains to compute the reduction in transient capability measured by the total deposited energy as a function of irradiation dose. We find that radiation damage has a negligible effect on energy deposition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Testing and Characterization to Develop a Mechanistic Explanation for Unsaturated Drift of Fiber Optic Sensors during High-Dose Irradiation

The primary limitation for any optical fiber-based sensor for nuclear reactor applications is radiation-induced attenuation (RIA) of the transmitted and/or reflected signals. Based on several recent studies, RIA is tolerable for some fused silica optical fibers with the proper choice of sensing wavelength and fiber dopants. For extreme temperature applications (> 1000°C), sapphire optical fibers have been proposed; however, recent optical transmission measurements performed on bulk sapphire samples showed prohibitively large RIA. For some sensors, radiation-induced dimensional changes in the fiber materials can also cause significant drift. Moreover, the drift that was observed in numerous experiments performed in the High Flux Isotope Reactor (HFIR), the Advanced Test Reactor, the Massachusetts Institute of Technology Reactor, and other international facilities far exceeded what would be expected based on compaction of fused silica glass. Clearly, additional work is needed to better understand the origins of both RIA and radiation-induced drift in both silica and sapphire optical fiber-based sensors before these sensors can be reliably deployed for nuclear applications. This work evaluated the underlying mechanisms that may be responsible for RIA and drift in silica and sapphire materials. First, detailed characterization was performed on bulk fused silica glass samples that were previously irradiated to different neutron fluences at different temperatures to better understand the structural changes that drive radiation-induced drift in the absence of coating effects that are discussed later. Results show that the non-monotonic compaction that occurs with increasing neutron fluence continues up to fast neutron fluences approaching 10 22 n/cm 2 , which has important implications for physics-based models that may be used to compensate for the sensor drift. Initial Raman spectroscopy and synchrotron x-ray diffraction provide insights into the nature of the structural changes. Next, detailed characterizations were performed on silica fibers with various coatings that were subjected to several different thermal treatments. The hypothesis is that the coatings convert to carbon-rich materials that compact under irradiation, putting a large compressive strain on the fiber. Out-of-pile testing confirms that both polyimide and acrylate fiber coatings convert to glassy carbon (GC) materials when heated under inert conditions, and the degree of order (i.e., graphitization) increases with increasing temperature. The results provide increasingly strong evidence that the combination of polymeric coatings and inert (or vacuum) conditions render fiber optic sensors susceptible to significant radiation-induced drift that would not otherwise exist in uncoated fibers. Finally, transmission electron microscopy was performed on bulk sapphire samples that were irradiated to two neutron fluences at different temperatures to gain insights into the potential mechanisms driving the prohibitive RIA at higher neutron fluences and temperatures. Contrary to previous hypotheses, results show that scattering from radiation-induced voids cannot explain the observed RIA. Similarly, models for scattering losses from dislocation loops also do not agree with the experimental results. Instead, fitting to the experimental data shows that increased absorption from aluminum vacancy centers is the most likely explanation for the the prohibitively large RIA that was observed at high irradiation temperature and dose. In addition, the voids that formed in these single-crystal samples were found to align along the basal plane (a-axis) as opposed to that seen in previous observations of c-axis alignment in polycrystalline samples, which could have important implications for anisotropic swelling and other phenomena that could affect sensor performance at high neutron fluence.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Testing and Characterization to Develop a Mechanistic Explanation for Unsaturated Drift of Fiber Optic Sensors during High-Dose Irradiation

The primary limitation for any optical fiber-based sensor for nuclear reactor applications is radiation-induced attenuation (RIA) of the transmitted and/or reflected signals. Based on several recent studies, RIA is tolerable for some fused silica optical fibers with the proper choice of sensing wavelength and fiber dopants. For extreme temperature applications (> 1000°C), sapphire optical fibers have been proposed; however, recent optical transmission measurements performed on bulk sapphire samples showed prohibitively large RIA. For some sensors, radiation-induced dimensional changes in the fiber materials can also cause significant drift. Moreover, the drift that was observed in numerous experiments performed in the High Flux Isotope Reactor (HFIR), the Advanced Test Reactor, the Massachusetts Institute of Technology Reactor, and other international facilities far exceeded what would be expected based on compaction of fused silica glass. Clearly, additional work is needed to better understand the origins of both RIA and radiation-induced drift in both silica and sapphire optical fiber-based sensors before these sensors can be reliably deployed for nuclear applications. This work evaluated the underlying mechanisms that may be responsible for RIA and drift in silica and sapphire materials. First, detailed characterization was performed on bulk fused silica glass samples that were previously irradiated to different neutron fluences at different temperatures to better understand the structural changes that drive radiation-induced drift in the absence of coating effects that are discussed later. Results show that the non-monotonic compaction that occurs with increasing neutron fluence continues up to fast neutron fluences approaching 10 22 n/cm 2 , which has important implications for physics-based models that may be used to compensate for the sensor drift. Initial Raman spectroscopy and synchrotron x-ray diffraction provide insights into the nature of the structural changes. Next, detailed characterizations were performed on silica fibers with various coatings that were subjected to several different thermal treatments. The hypothesis is that the coatings convert to carbon-rich materials that compact under irradiation, putting a large compressive strain on the fiber. Out-of-pile testing confirms that both polyimide and acrylate fiber coatings convert to glassy carbon (GC) materials when heated under inert conditions, and the degree of order (i.e., graphitization) increases with increasing temperature. The results provide increasingly strong evidence that the combination of polymeric coatings and inert (or vacuum) conditions render fiber optic sensors susceptible to significant radiation-induced drift that would not otherwise exist in uncoated fibers. Finally, transmission electron microscopy was performed on bulk sapphire samples that were irradiated to two neutron fluences at different temperatures to gain insights into the potential mechanisms driving the prohibitive RIA at higher neutron fluences and temperatures. Contrary to previous hypotheses, results show that scattering from radiation-induced voids cannot explain the observed RIA. Similarly, models for scattering losses from dislocation loops also do not agree with the experimental results. Instead, fitting to the experimental data shows that increased absorption from aluminum vacancy centers is the most likely explanation for the the prohibitively large RIA that was observed at high irradiation temperature and dose. In addition, the voids that formed in these single-crystal samples were found to align along the basal plane (a-axis) as opposed to that seen in previous observations of c-axis alignment in polycrystalline samples, which could have important implications for anisotropic swelling and other phenomena that could affect sensor performance at high neutron fluence.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

AGC-4 As-Run Uncertainty Analysis

The concept of measuring neutron flux (or fluence rate) with activation foils has existed since the late 1940s and is a well-used methodology. Activation foils are selected for their unique isotopic cross sections that create reaction products with desirable properties: relatively simple radioactive decay scheme, measurable quantities of gamma-rays or betas, and half-lives appropriate for the irradiation and decay durations of the experiment. The activation materials, deemed flux wires, fluence wires, dosimeters, or flux foils, are fabricated to have a mass appropriate for the irradiation as well so the measurement using a radiation detector following the irradiation is sufficient but not overwhelming in terms of total activity. Activation measurement results are expressed in terms of reaction rate, which can then be used to “adjust” the neutron spectrum. The Advanced Graphite Creep (AGC)-4 experiment contained 26 flux wire packages, each containing an iron and a niobium wire inside a vanadium capsule. For all stackup positions, the value of fast to thermal neutrons is higher than expected, but the total neutron fluence is near the predicted value of ~2 x 1022 n/cm2. As compared to the predicted neutron spectrum, there are four takeaways in the results as compared to the predicted ATR irradiation: 1) Additional fast (fission) neutrons present, 2) Spectrum hardening due to temperature, 3) Spectrum hardening due to container attenuation, 4) Total neutron fluence is close to predicted values. Since the neutron energy spectrum is a key parameter when attributing neutron damage in materials, the adjusted energy-dependent spectrum and the total neutron fluence results from the dosimeters can be incorporated for future calculations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Ultrafast formation of a transient two-dimensional diamondlike structure in twisted bilayer graphene

Due to the absence of matching carbon atoms at honeycomb centers with carbon atoms in adjacent graphene sheets, theorists predicted that a sliding process is needed to form AA, AB', or ABC stacking when directly converting graphite into sp 3 bonded diamond. Here, using twisted bilayer graphene, which naturally provides AA and AB' stacking configurations, we report the ultrafast formation of a transient two-dimensional diamondlike structure (which is not observed in aligned graphene) under femtosecond laser irradiation. This photoinduced phase transition is evidenced by the appearance of bond lengths of 1.94 and 3.14 Å in the time-dependent differential pair distribution function using MeV ultrafast electron diffraction. Molecular dynamics and first-principles calculation indicate that sp 3 bonds nucleate at AA and AB' stacked areas in a moiré pattern. This work sheds light on the direct graphite-to-diamond transformation mechanism, which has not been fully understood for more than 60 years.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2020

Failure investigation of nuclear grade POCO graphite target in high energy neutrino physics through numerical simulation

In many high energy physics neutrino beamlines, targets are made up of isotropic POCO graphite grade for production of neutrinos for high energy particle physics research and are bombarded with highly energetic pulsed proton beam. The pulsed proton beam with a small beam size creates thermal stress waves as well as radiation damage such as displacement damage, void formation, swelling, and gas formation amongst others. As a result, after a few years of operation one of the longest operated targets appears to have undergone bulk swelling and fractured along the beam centerline. A complex interaction of dynamic loading due to beam, material degradation due to irradiation are thought to have caused such a failure. Here, we present a numerical simulation to capture the combined effects of swelling and the dynamic effect of thermal stress waves due to beam heating, to explain the failure of target. An empirical formula has been developed to take into account swelling as a function of temperature and proton fluence and has been implemented in a commercial finite element code. Extensive X-ray diffraction (XRD) were performed on graphite samples from the failed target to understand the lattice parameter changes due to irradiation and build a valid empirical model. Simulation results show that swelling plays a major role in elevating the stress state in the material exceeding the failure strength while the pulsed beam heating also introduces fatigue loading that would explain brittle fracture emanating from inside of the material and propagating outward. Such empirical formulations will help in identifying the threshold values of critical parameters to extend the service life of future multi-megawatt neutrino targets and targets in other accelerator environments.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

AGR-3/4 Post-Irradiation Examination

Presentation for the TCT Review meeting

36 MATERIALS SCIENCE↗

DOE OSTI2023

AGR-5/6/7 Capsule 1 Temperature Re-analysis_Grant Hawkes_Final

Slides for Gas-Cooled Reactor AGR-5/6/7 Thermal Model with Non-Uniform Gas Gaps. Includes investigation of offset through modeling and experimental results, ABAQUS finite element model created for the entire test train, Capsule 1 information, Three irradiation dates to compare the thermal model with the working thermocouples (TCs), Offset distances and direction of offset for the top and bottom of the graphite holder with computer runs, determination of offset by the root mean square error (RMSE) between the TCs and model predtictions, and the results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Development of Hydrogen Permeation Barrier and Matrix Interaction Protection Coatings for TZM Enclosure of Hydride Moderators (Microreactor Program M3 Report)

It is important to develop a high-performance enclosure solution for hydride-based microreactor moderation components to effectively suppress hydride thermal decomposition and prevent hydrogen loss in harsh in-reactor environments. Currently, the DOE-NE Microreactor Program (MR program) is investigating the titanium-zirconium-molybdenum (TZM) canned yttrium hydride (YH 2-x ) solution, which provides acceptable hydrogen retainment performance up to ~700 °C. However, improvements in the enclosure are necessary for increasing long-term operation temperature beyond 700 °C and enhancing short-term survivability at even higher temperatures during power transients. One potential solution is the addition of a dedicated hydrogen permeation barrier, which can potentially improve the enclosure's performance. Furthermore, the hydride moderator components will be inserted into the microreactor matrix, which may consist of materials that are not chemically inert against TZM. Therefore, the TZM might need to be protected from the external environment through the application of another surface protection coating. To address these critical issues, two types of barrier coatings optimized for TZM-based moderator enclosure has been developed. A multilayer coating approach has been used to identify TZM-compatible barrier materials. Behavior of these coating designs against H 2 permeation and long-term high-temperature graphite matrix interactions has been examined in detail. Additionally, the developed coatings' behavior when exposed to both high-temperature thermal cycling and heavy-ion irradiation has also been investigated. Overall, the development of these barrier coatings is essential for achieving maturity for the metal hydride moderator technology. The results in this report will provide crucial insights into the behavior of the developed coatings and their potential for improving the long-term performance of TZM-based moderator enclosures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

AGR-3/4 TRISO Fuel Compact Ceramography

The combined third and fourth irradiation in the Advanced Gas Reactor (AGR) program (AGR-3/4) contained tristructural isotropic (TRISO)-coated particle fuel and designed-to-fail (DTF) fuel particles. The DTF particles were only coated with highly anisotropic pyrocarbon (PyC) so they would purposely fail during the AGR-3/4 irradiation and provide a source of fission products for measurement. To observe the post-irradiation morphology of these DTF particles and the TRISO-coated particles, three AGR-3/4 compacts were mounted in epoxy, sectioned above their centerlines, and ground/polished. Three rounds of grinding/polishing and optical microscopy were performed so that the particles could be observed at multiple planes. Each compact contained approximately 1,872 TRISO particles and exactly 20 DTF particles. The three AGR-3/4 compacts examined covered a wide temperature range and featured both the hottest average irradiation temperature (1375°C) and the coldest/lowest burnup (872°C and 5.5% fissions per initial metal atom) of any compacts to undergo post-irradiation examination (PIE) in the AGR program to date. A total of 29 DTF particles were located and observed via microscopy across the three compacts. All the DTF particles observed via microscopy had completely failed PyC coatings. Some different DTF kernel and PyC morphologies were observed that have been attributed to the differences in irradiation temperature and/or burnup. At low and medium irradiation temperatures (roughly 850 to 1050°C), it appears irradiation-induced dimensional changes in the DTF PyC caused it to fracture and fold in on itself, and the kernel deformed to accommodate this PyC deformation. The extent of DTF kernel deformation to accommodate DTF PyC buckling tended to be greater for the medium-temperature fuel compared to the cold fuel. In the high-irradiation-temperature compact (1375°C), the DTF PyC layer appears to have completely reacted chemically with the kernel material. The only material surrounding the DTF particles in this hot compact resembles the compact matrix material. Observations of the TRISO-coated driver fuel particles and fuel compact graphitic matrix were also made. The TRISO particle kernels in the medium-temperature compact (1047°C), and many in the high-temperature compact (1375°C), showed morphologies consistent with what was seen in AGR-1 and AGR-2. In other TRISO particles from the high-temperature compact, a spatial gradient in the kernel appearance was evident. Here the cool side of the kernel (away from the center of the compact) had a much darker appearance (like that of the buffer and pyrocarbon). This is the first observation of this kind of kernel spatial gradient in AGR fuel. It is believed that the high irradiation temperature of that compact (an average of 1375°C) activates or accelerates an unidentified chemical reaction, and the temperature gradient within the fuel causes a sharp spatial gradient. The low-temperature/low-burnup compact also showed unique kernel morphologies where the TRISO-coated particles had clearly distinguishable oxide and carbide phases at the center of the kernel, an oxide rind surrounding this, and remnants of a carbide skin surrounding the oxide rind. These are features observed in the as-fabricated fuel that are no longer present in higher burnup fuel. Occasional gaps between the outer pyrolytic carbon (OPyC) and graphitic matrix material were observed in all compacts. Gaps were most often found in the small, matrix-filled spaces between adjacent particles. Finally, buffer morphologies in AGR-3/4 TRISO particles were like those observed in AGR-1 and AGR-2, and the buffer fracture frequencies in AGR-3/4 and AGR-2 TRISO particles were plotted versus fast neutron fluence and irradiation temperature. In AGR-3/4 the buffer fracture rate was highest (23%) for an irradiation temperature of 1047°C and a fast fluence of 5.18E25 n/m 2 . Increasing the irradiation temperature to 1375°C reduced the buffer fracture rate to 14%. Reducing the irradiation temperature to 872°C and the fast fluence by a factor of 3 reduced the buffer fracture rate to 7%. The temperature and fluence dependencies observed for AGR-3/4 buffer fracture were consistent with those observed in AGR-2. Lower fluences and/or higher temperatures significantly reduced the buffer fracture frequencies. Lower buffer fracture frequencies were also observed at lower temperatures as long as the fluence was also significantly reduced. This is believed to be due to lower fluence resulting in less irradiation-induced dimensional change (shrinkage) of the buffer, and a higher temperature promoting enhanced creep relaxation of stresses within the buffer, leading to less buffer fracture.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Radiation Response of Low Dimensional Carbon Systems

The project is aimed at understanding the fundamentals of radiation response of low dimensional carbon systems and irradiation-induced mechanical property changes, with a focus on the unique phenomena caused by their geometry, boundary, and quantum size effects. In comparison with their bulk counterparts (graphite), radiation responses of low-dimensional carbon systems are substantially different. The differences exist at almost every stage of defect development, namely, displacement creation, damage cascade, thermal spike formation defect recombination, defect clustering, and structural reconstruction. Many traditional concepts in ion-solid interaction theory do not apply at the nanoscale or require substantial modification. Through this project, we systematically studied various aspects of radiation damage development in graphene, carbon nanotubes (CNTs), and graphite.

36 MATERIALS SCIENCE↗