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DOE OSTI2021

Surface chemistry models for GaAs epitaxial growth and hydride cracking using reacting flow simulations

Hydride vapor phase epitaxy (HVPE) is a promising technology that can aid in the cost reduction of III-V materials and devices manufacturing, particularly high-efficiency solar cells for space and terrestrial applications. However, recent demonstrations of ultrafast growth rates (~500 µm/h) via uncracked hydrides are not well described by present models for the growth. Therefore, it is necessary to understand the kinetics of the growth process and its coupling with transport phenomena, so as to enable fast and uniform epitaxial growth. In this work, we derive a kinetic model using experimental data and integrate it into a computational fluid dynamics simulation of an HVPE growth reactor. We also modify an existing hydride cracking model that we validate against numerical simulations and experimental data. Here, we show that the developed growth model and the improved cracking model are able to reproduce experimental growth measurements of GaAs in an existing HVPE system.

CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSICS↗

DOE OSTI2022

Tritium Transport Modeling at the Pore Scale in Ceramic Breeder Materials Using TMAP8

Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders (CCBs) are dense materials that can maintain a high tritium breeding ratio while promoting tritium release because of highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport [i.e., the Tritium Migration Analysis Program (TMAP8)], which is itself built on top of the finite-element multiphysics framework multiphysics object-oriented simulation environment (MOOSE). This study demonstrates some of TMAP8’s capabilities and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Tribocorrosion of stainless steel sliding against graphite in FLiNaK molten salt

The molten salt reactor (MSR) concept recently gained renewed interest in developing Generation IV nuclear reactors. One MSR design uses fluoride molten salts to cool tri-structural isotropic particle fuel encapsulated by a carbon matrix into spherical pebbles, which would inevitably contact the reactor's stainless steel container wall during salt circulation. Characterizing this interaction is crucial for reactor safety. Here, in this work, we report the tribocorrosion behavior of graphite sliding against Type 316H stainless steel lubricated by a FLiNaK (LiF:NaF:KF; 46.5:11.5:42 mol %) molten salt in an argon environment. Accelerated wear loss was observed at a higher temperature because of a lower molten salt viscosity and a higher corrosion rate. The graphite had a more rapid material loss at a higher sliding speed than stainless steel because of its higher vulnerability to vibration-induced microfracture. The salt-starved condition caused more material loss than either the no-salt or the salt-flooded condition because neither a graphite transfer film nor stable boundary lubrication could be established at salt starvation. An interesting dual-layer surface film was discovered on the stainless steel worn surface: a chromium-rich top film and a nickel-accumulated but chromium-depleted interlayer. The film composition and structure provide fundamental insights to the mechanochemical interactions between stainless steel and graphite in a molten salt environment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Modelling the ejection of primary aerosols during the fast pyrolysis of biomass anisotropic particles

A model for the fast pyrolysis of anisotropic biomass particles is presented which considers bubbling dynamics within the liquid intermediate phase (metaplast) and aerosol ejection from this phase. The model employs the population balance equation and the method of moments to estimate the production rate and resultant size distribution of aerosol ejections, incorporating a detailed CRECK reaction mechanism, and considers the effect of anisotropic biomass microstructure on the intraparticle transport of mass and energy. Here, this study investigates the impact of particle size, heating rate (heat transfer coefficient), and lignocellulosic composition on aerosol ejection. The model predicts that, at high heating rates (convective heat transfer coefficient of 359 W/m 2 .K), aerosols can contribute over 20% to the heavy fraction yield in bio-oil for small particles (1 mm diameter, 4 mm length). The model can predict aerosol size distribution and surface area, indicating an average size of 20 μm for bubbles and 5 μm for aerosols during increased bubble production and aerosol ejection rates. These findings are consistent with prior experimental results and provide essential information for future modeling of extra-particle reactions of the aerosols as they progress through the reactor.

09 BIOMASS FUELS↗

DOE OSTI2023

Insights into Active Sites and Mechanisms of Benzyl Alcohol Oxidation on Nickel–Iron Oxyhydroxide Electrodes

The electrochemical oxidation of bio-derived molecules has recently garnered interest for its potential in opening electrified synthetic pathways toward value-added products. Herein, we investigate the electrochemical conversion of benzyl alcohol (BA) to benzaldehyde and benzoate on nickel–iron (Fe ~ 7–18%) electrodes as a model system to understand reaction mechanisms and environmental conditions that can transform these molecules. Our results indicate a strong correlation between benzyl alcohol oxidation (BAO) onset potentials and Ni(II/III) redox peak positions, highlighting the potential role that lower oxidation states of nickel, i.e., Ni 3+ , can play in BAO catalysis. Our work on the Ni 2+/3+ system complements mechanisms that involve higher oxidation states of Ni as reported by others. We note that the Ni redox position and thus BAO onset is impacted by Fe incorporation during electrochemistry from unpurified electrolytes, which can resemble standard reactor operating conditions. We perform a systematic computational investigation into BAO and provide density functional theory (DFT) insights into how the redox mechanism has been such a prominent focus of alcohol oxidations. This includes the mode of BA adsorption and the nature of the adsorption site; upon conversion of the Ni 2+ surface to active Ni 3+ via hydroxyl deprotonation, BAO is thermodynamically downhill. Our DFT study also introduces the possibility of a vacancy-driven mechanism, though expected to be less prevalent during catalysis than the redox mechanism for a Ni 3+ surface. Through the systematic investigation of experimental reaction conditions and computational free energy thermodynamics, we have gained valuable insights into BAO reaction mechanisms that inform catalytic activity. Our study opens avenues for further design and development of catalyst active sites for the oxidation of related organic molecules.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Hydrogen isotope analysis in W-tiles using fs-LIBS

Abstract Laser-Induced Breakdown Spectroscopy (LIBS) is a promising technology for in-situ analysis of Plasma-Facing Components in magnetic confinement fusion facilities. It is of major interest to monitor the hydrogen isotope retention i.e. tritium and deuterium over many operation hours to guarantee safety and availability of the future reactor. In our studies we use ultraviolet femtosecond laser pulses to analyze tungsten (W) tiles that were exposed to a deuterium plasma in the linear plasma device PSI-2, which mimics conditions at the first wall. A high-resolution spectrometer is used to detect the Balmer- $$\alpha$$ α transition of the surface from implanted hydrogen isotopes (H and D). We use Calibration Free CF-LIBS to quantify the amount of deuterium stored in W. This proof-of-principle study shows the applicability of femtosecond lasers for the detection of low deuterium concentration as present in first wall material of prevailing fusion experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

CO 2 methanation reaction pathways over unpromoted and NaNO 3 -promoted Ru/Al 2 O 3 catalysts

Catalytic CO 2 sorbents, materials that adsorb and pre-concentrate CO 2 on the catalyst surface prior to subsequent conversion, are becoming important materials in CO 2 capture and utilization. In this work, a prototypical CO 2 methanation catalyst – Ru/Al 2 O 3 – and a related catalytic sorbent – NaNO 3 /Ru/Al 2 O 3 – are used for CO 2 methanation in flowing hydrogen in a fixed bed reactor at temperatures ranging from 220 to 280 °C. Activation energies for the NaNO 3 /Ru/Al 2 O 3 material are slightly higher than unpromoted Ru/Al 2 O 3 catalysts, and the reaction orders vary more significantly. In situ IR spectroscopy and steady-state isotopic kinetic analysis (SSITKA) using in situ IR/MS spectroscopy show that bicarbonate and linear carbonyl species are the likely reaction intermediates over unpromoted Ru/Al 2 O 3 , while bidentate carbonate, formate and linear carbonyl species are among likely reaction intermediates over NaNO 3 /Ru/Al 2 O 3 . Rate laws consistent with the obtained experimental data are proposed after kinetic modeling of multiple plausible reaction pathways. In conclusion, results suggest that the pathway over the NaNO 3 /Ru/Al 2 O 3 catalyst likely has an additional kinetically relevant irreversible step in the CO 2 methanation reaction pathway.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Sister Rod Destructive Examinations (FY23): Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water

The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed. Consistent with existing literature, the leaching followed a trend in which isotopes of certain elements such as Cs and Mo were among the first species in the matrix to dissolve, and with the highest release rates. This was followed by a gradual matrix dissolution consisting of uranium and other actinides and a slower-than-matrix release from some isotopes, including Ru and Rh. It was also observed that the circumferential samples having less exposed fuel surface area (to begin with) leached more than the axial samples for a majority of the isotopes during the timespan of the study. Previous experiments were conducted by decladding or exposing the pellet-clad surface completely, but the samples used in this study retained the cladding. One possible explanation for the higher leaching rate of the circumferential samples is that the pellet-clad interface, which has a greater density of grain boundaries and defects, may be the most vulnerable area to leaching of fuel in the presence of water.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Panel Session 4 and 15: Japan Fukushima Daiichi D and D Update and Technological Challenges at Japan Fukushima Daiichi D and D and Update on Nuclear Overview and Development in Japan - Nuclear Fuel Cycle

Sessions 4 and 15 represent a two-part panel series that discusses progress and challenges associated with cleanup at Fukushima. Severe limitations on availability of original panelists from Japan due to strict restrictions put in place to alleviate the spread of coronavirus necessitated changes to both panels. The result was a significantly modified panel for Session 04 (shown above) and the elimination of all panelists for Session 15. The 4 and 15 Panel Sessions provide an overview of activities related to both the progress and challenges of cleanup and decommissioning of the Fukushima Daiichi Nuclear Power Station (NPS) in Japan. Five panelists discussed perspectives of the cleanup following a Tokyo Electric Power Company (TEPCO) video showing the progress on site since the devastating Great East Earthquake and tsunami that caused the explosions at three of the six reactors on the site. Three of the five panelists discussed on-going work being performed for the effort, while the other two provided expert perspectives of on strategic efforts at the site. The panel was attended by over 80 technologists and policy makers spanning the globe and was opened by Dr. Monica Regalbuto of Idaho National Laboratory and a short video provided by TEPCO. The video described changes at the site that spanned the cleanup efforts from stabilizing water intrusion into the contaminated reactor buildings to construction of new administrative facilities. The video explained the processes underway to retrieve spent fuel rods and challenges in retrieval of the compromised fuel debris. The video highlighted working condition improvements that included establishment of rest housing and a small convenience store on the site, and the rollback of protective equipment around the site due to decreases contamination. Panelists with presentations: Revision of 'the Mid-and-Long-Term Road-map towards the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station' (Paul Dickman); Sharing UK experience at Fukushima Daiichi (Adrian Simper); SRNL Japan (Andrew Fellinger); ABLE's Initiative to Dismantle the Exhaust Stack (Daniel Walter); JAEA R and D in Fukushima (Tokio Fukahori); TEPCO - Overview and Update of the Fukushima Decommissioning Process (Monica Regalbuto); Toshiba's Involvement in the Decommissioning of the Fukushima Daiichi Nuclear Power Plant (Yasuhiro Yuguchi); Remote Dismantling of the Exhaust Stack At Fukushima Dai-ichi NPS (Takashi Okutsu)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Evaluation of Changes to the MTR Fuel Dissolution Flowsheet to Increase the Number of L-Bundles Charged to a Dissolver

Material Test Reactor (MTR) fuels are dissolved in the H-Canyon facility using a Hg-catalyzed, HNO₃ dissolution flowsheet. The flowsheet is based on experimental work in which laboratory-scale experiments were used to provide bounding values for both offgas and H₂ generation rates. The generation of H₂ gas, primarily from the dissolution of Al, is controlled by limiting the surface area exposed to the solution. The number of L-Bundles charged and their submergence level defines the exposed surface area which is directly proportional to the H₂ generation rate. For the MTR fuel dissolution flowsheet, the maximum number of L-Bundles which can be charged to a dissolver was specified as a function of the Al concentration in the solution and the submergence level of the bundle. The MTR fuels are currently dissolved in the 6.1D dissolver. To reach the desired terminal Al concentration, a total of 12 L-Bundles of fuel are dissolved per batch. Depending on the submergence level being used in the dissolver, the 12 bundles are dissolved in three charges of four bundles per charge or in a sequence of three, four, and five bundles. However, if six L-Bundles could be dissolved in the initial charge, a dissolver batch could be completed in two dissolutions instead of three. To assess the feasibility of dissolving six bundles in the first charge, the Savannah River National Laboratory evaluated changes to the MTR flowsheet which increase the number of bundles which can be charged to a dissolver. Since the development of the existing MTR fuel dissolution flowsheet, improved offgas measurement methods have been developed to more accurately calculate the offgas composition and generation rate during laboratory-scale dissolutions. In the first part of this study, improvements in the calibration model for the Raman spectrometer (used for offgas characterization) and a more effective data smoothing routine were applied to existing data. The uncertainty in the offgas composition data used to define the MTR fuel dissolution flowsheet were subsequently re-evaluated which resulted in an approximate 50% reduction in the uncertainty of the H₂ gas concentration measurements. The reduced uncertainty was used in new calculations to define the maximum number of L-Bundles which can be charged to either the 6.1D or 6.4D dissolver as a function of the Al concentration. With the reduced uncertainty in the H₂ concentration measurements, a flowsheet was defined which supported initially charging 6 bundles to a dissolver. This flowsheet is based on the use of 0.002 M Hg, an L-Bundle immersion depth of 54 in., an air sparge/purge flowrate of 40 scfm, an iodine reactor operating temperature of 200 °C, and not exceeding 60% of the H₂ LFL in the dissolver offgas. In the second phase of this study, calculations were performed to assess the impact of changing various operating parameters to increase the number of L-Bundles charged to an H-Canyon dissolver. Applying the improved offgas measurement techniques including the reduced uncertainty in the H₂ concentration measurements to the existing data, the number of bundles of MTR fuel which can be charged to either the 6.1D or 6.4D dissolver were calculated based on varying the bundle immersion depth (44-60 in.), air sparge/purge flowrate (40-60 scfm), iodine reactor operating temperature (150-200 °C), and the percentage of the H₂ LFL which must not be exceeded (60-76%). Qualitatively, the percentage of the LFL had the largest influence on the maximum number of bundles which can be charged to a dissolver followed by the air sparge/purge flowrate, L-Bundle immersion depth, and the iodine reactor operating temperature. During the evaluation of each variable, the other variables were held at the baseline flowsheet conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

The chlorination and separation of aluminum using low-temperature sulfur chloride reagents

There is currently no strategy for the permanent waste disposal or recycling for used nuclear fuelsfrom research reactors. For this reason, low-temperature reactions have been developed for thechlorination of the Al alloys and subsequent separation from used nuclear fuels to reducethe volume of high-level waste in storage. Three sulfur chloride reagents – S 2 Cl 2 , SOCl 2 , and SO 2 Cl 2 —were tested, and two were found to quantitatively chlorinate Al metal and Al alloys undermild conditions. Further, these low-temperature reactions proceed between 298 and 411 K, and up to 5 gof metal is chlorinated in 1–3 h. Preliminary results indicate that the reactivity and exothermicity ofthe reaction between the Al and sulfur chloride reagents is highly dependent on the surface area-to-volume ratio of the metal and the volume of solvent. Elemental S is produced as a by-productduring the chlorination with S 2 Cl 2 but can be quantitatively rechlorinated under mild conditions toregenerate the initial chlorination reagent. Therefore, in this case, chlorine is the only elementconsumed in the reaction, thus minimizing the waste generated during the chlorination process.The AlCl 3 may then be separated from other materials present in Al 6061 or Al 8001 because of itshigh solubility in the sulfur chloride reagents. This process may also be extended to chlorinate Alfrom research reactor fuels.

aluminum↗

DOE OSTI2025

Large radiation back-flux from Monte Carlo simulations of fusion neutron–material interactions

Abstract Fusion power reactors will generate intense neutron fluxes into plasma-facing and structural materials (SMs). Radiation back-fluxes, generated from neutron–material interactions under these fluxes, can dramatically impact the plasma dynamics, e.g. by seeding runaway electrons during disruptions via Compton scattering of background electrons by wall-emitted gamma radiation. Here, we quantify these back-fluxes, including neutrons, gamma rays, and electrons, using Monte Carlo calculations for a range of SM candidates and first wall (FW) thicknesses. The radiation back-flux magnitudes are remarkably large, with neutron and gamma radiation back-fluxes on the same order of magnitude as the incident fusion neutron flux. Electron back-fluxes are two orders of magnitudes lower, but are emitted at sufficiently high energies to impact the sheath and boundary plasma dynamics. Material configuration plays a key role in determining back-flux magnitudes. The SM chiefly determines the neutron back-flux magnitude, while the FW thickness principally attenuates the gamma ray and electron back-fluxes. In addition to prompt back-fluxes, which are emitted immediately after fusion neutrons impact the surface, significant delayed gamma ray and electron back-fluxes arise from nuclear decay processes in the activated materials. These delayed back-flux magnitudes range from 2% to 7% of the prompt back-fluxes, and remain present during transients when fusion no longer occurs. During disruptions, build-up of delayed gamma radiation back-flux represents potential runaway electron seeding mechanisms, posing additional challenges for disruption mitigation in a power reactor compared with non-nuclear plasma operations. This work highlights the impact of these radiation back-fluxes plasma performance and demonstrates the importance of considering back-flux generation in materials selection for fusion power reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Metalorganic chemical vapor deposition of ZnGeN 2 films on GaN: effects of cation stoichiometry on surface morphology and crystallinity

Novel optoelectronic device designs based on the heterostructures of III-N and II-IV-N 2 are promising to advance device performance significantly. For example, by utilizing InGaN-ZnGeN 2 quantum well (QW) structures instead of pure InGaN, the band structure engineering in the QW active region can lead to improved electron-hole wavefunction overlap, and thus enhance the radiative efficiency for photon generation. These novel heterostructures have great potential to address the current challenge of low quantum efficiency in InGaN QW based light emitting diodes emitting in green and beyond. The materials development of ZnGeN 2 is still at an early stage as compared to the much matured GaN material system. In an ideal octet rule preserving ordered structure of ZnGeN 2 , every N atom is coordinated by exactly two Zn and two Ge atoms. However, local violation of octet rule can be caused by non-ideal coordination of N by the cations. The disordered structure is thermodynamically less favorable but can still be achieved, for example, in kinetics-limited growth regime. The ordered ZnGeN 2 has a bandgap very close to that of GaN (~3.4 eV) and a lattice mismatch of <0.1% with GaN. Interestingly, the valence band of ZnGeN 2 has been predicted to be ~1 eV above that of GaN, which has inspired novel designs for high efficiency light emitters. In this work, we investigated the metalorganic chemical vapor deposition (MOCVD) of ZnGeN 2 films on GaN/c-sapphire templates. Diethylzinc (DEZn), germane (GeH4) and ammonia were used as the precursors for Zn, Ge and N, respectively. A systematic study was conducted to investigate the cation stoichiometry as a function of growth temperature (TG), total reactor pressure (P) and DEZn/GeH 4 molar flow rate ratio (RII/IV). Under the investigated growth window, the Zn/(Zn+Ge) composition in the films, determined from energy dispersive X-ray spectroscopy, decreased monotonically with increase in TG but increased with increase in P and RII/IV. Atom probe tomography data did not indicate the presence of any secondary phases such as Zn 3 N 2 or Ge 3 N 4 . The surface morphology and crystallinity of the grown films had strong correlation with the Zn/(Zn+Ge) composition. The scanning electron microscopy images showed that the near-stoichiometric films have planar surfaces whereas Zn-rich films had crystallites on their surface and the Zn-poor films had faceted surface. Scanning transmission electron microscopy (STEM) imaging revealed that the Zn-rich and Zn-poor films have columnar and filament-like morphology, respectively, whereas the near-stoichiometric films have continuous film-like cross-sectional morphology. TEM nano-diffraction patterns as well as X-ray diffraction 2θ-ω scan profiles indicate that the near stoichiometric films are single crystalline. Nano-diffraction pattern of the stoichiometric films resembled that of a disordered ZnGeN 2 structure. Room temperature Raman spectra of near-stoichiometric films showed only the phonon density of states like features of a cation disordered ZnGeN 2 . Cathodoluminescence and photoluminescence spectra measured at different temperatures had similar features with peak emission wavelength at ~ 2 eV. In conclusion, the stoichiometry of ZnGeN 2 films can be widely tuned by tuning the MOCVD growth parameters. The surface morphology and the crystallinity of the films were found to have strong correlation with the Zn/(Zn+Ge) composition. The stoichiometric ZnGeN 2 films grown on GaN were demonstrated with uniform surface morphology and high crystalline quality. The results from this work will provide pathway to implement ZnGeN 2 in device structures.

Karim, Md Rezaul↗

DOE OSTI2025

Back to functional hydrides: Effects of neutron-irradiated microstructure on hydrogen retention in yttrium hydride

Functional hydrides are promising candidates for advanced nuclear reactors, particularly in portable or transportable applications, due to their high hydrogen-retention capabilities, enabling efficient neutron moderation, and compact reactor design. However, hydrogen mobility in hydrides at elevated irradiation temperatures poses significant technological challenges, necessitating a comprehensive understanding of their irradiation behavior. Furthermore, this study investigated the microstructural and chemical stability of neutron-irradiated yttrium hydrides to assess their hydrogen-retention capacity. A targeted literature review was also conducted to contextualize neutron-irradiation effects on functional hydrides with regards to structural stability and hydrogen retention. Experimental characterizations revealed that, at high temperatures, irradiated hydrides retained their phase stability, which was likely enhanced by irradiation-induced microstructure evolution. Notably, an amorphous yttrium and oxygen -rich surface layer was present at the free surface of the hydride. Its thickness decreased while a continuous crystalline Y-O-rich layer was formed with increasing neutron damage. Additionally, the number density of dislocation loops and cavities generally increased as a function of neutron dose. First-principles calculations of hydrogen behavior within yttrium vacancy clusters in yttrium hydrides demonstrated vacancy-size-dependent hydrogen stability and configuration, highlighting the role of vacancy geometry in regulating hydrogen retention. Thus, the presence of irradiation-induced dislocation loops and cavities were hypothesized to improve hydrogen retention. Collectively, these findings advance the understanding of hydride behavior under neutron irradiation as well as their technological readiness for portable or transportable nuclear reactors.

Defect clusters↗

DOE OSTI2022

Implement and Test 3D Mortar Contact in BISON

We leverage the extension of the generation of mortar segment meshes to three dimensions in MOOSE’s framework to extend thermomechanical modeling capabilities to problems with three dimensions. A modular approach to gap heat transfer physics using the mortar finite element method was created and documented, mechanical contact was extended to three dimensions—including frictional behavior, performance and ease of use were improved, and steps towards scalability of solid mechanics problems involving contact were taken. Many of these new developments are demonstrated in the simulation of 3D light-water reactor (LWR) problems, where the thermomechanical interface problem is solved using the mortar finite element method. Usage of the mortar framework has improved convergence in 2D problems and has enabled employing friction in 3D problems, of which we show results of a short, local stack of 3D pellets. Consequently, the benefits of mortar in terms of solution convergence and quality are extended to three dimensions. Section 2 discusses fundamental developments that enabled the simulation of practical mortar problems in three dimensions and other general improvements, including the reduction of the derivative container size, the modification of dual basis computations when edge dropping (lack of secondary element projection) takes place, the improvement of conditioning when employing the VCP in-edge dropping conditions, and code usability and quality improvements. These latter code enhancements include the migration of tests using “old” mortar contact constraints to using dual mortar with a semi-smooth Newton solution strategy and the reuse of lower dimensional domains for straightforwardly setting up a mortar thermomechanical LWR problem, i.e. the MOOSE action is employed for mechanical contact and the thermal LWR action is employed to capture the gas conductance, contact, and radiation components of gap heat transfer physics. Independently of the mortar LWR thermal action, we developed a modular approach to gap heat transfer that resides in MOOSE and can be leveraged, e.g., in metallic fuel problems. This approach, whose code design based on MOOSE’s user objects to model specific physics was proposed by the maintenance activity, is detailed in Section 3. Based on the dual mortar finite element method, the frictional contact constraints were extended to three dimensions. A block sheared in two directions in and out of contact with a rigid plane is employed in Section 4 to show the way the approach handles changes in frictional states (e.g. stick to slip) within a competitive number of Newton iterations. Equations and numerical results on the use of the VCP with Cartesian Lagrange multipliers, whose combination enables their direct condensation, are described in Section 5.3. Two-dimensional and three-dimensional BISON LWR simulations are discussed in Section 6. Particularly, a stack of five eccentric pellets with a surface defect is simulated and the effect of pellet-cladding friction is assessed. Finally, conclusions are outlined in Section 7.

42 ENGINEERING↗

DOE OSTI2023

A novel method for quantifying irradiation damage in nuclear graphite using Raman spectroscopy

Raman spectroscopy has long been used in studying irradiation damage in carbon/graphite materials. It is however unclear if the measurements from different types of materials are directly comparable. Further, decoupling the contribution from irradiation temperature and fluence to the total damage in nuclear graphite possessing irradiation damage gradient is currently not readily feasible as it requires a complete set of test reactor irradiation experiments over the relevant temperature and fluence range. A novel methodology has therefore been proposed and developed to quantify total irradiation damage evolution at crystal level based on graphite Raman G-band position shift. Specifically, G-band positions derived from a large number of spectra collected from the fractured surface of microfine-grained POCO ZXF-5Q graphite possessing proton irradiation damage gradient were plotted with open literature Raman data on HOPG, BEPO (AXGP graphite), PCEA and IG-110 graphite as a function of dpa. The total damage level within 2σ beam radius in this POCO graphite was estimated to be equivalent to ~2–5 dpa at ~350–370 °C. Derived G-band positions were then mapped to the three-stage amorphization trajectory model indicating beam centre area has entered the second stage, i.e., transitioning from nanocrystalline graphite into amorphous carbon. G-band position relative shift (ΔG) curve with a ‘turn-around’ peak as a function of total damage can be used for in-service lifetime prediction. Importantly, the developed methodology has the potential to ‘unify’ total damage levels across different grades of nuclear graphite caused by ion, neutron and proton irradiation at different temperatures.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Perspectives and challenges of ultra-high temperature ceramics for fusion plasma-facing applications

Ultra-high temperature ceramics (UHTCs) offer several potential advantages as plasma-facing components (PFCs) in fusion reactors due to their extreme melting points, tailorable thermal conductivity, and attractive unirradiated mechanical properties including fracture toughness comparable or superior to tungsten. Here, recent developments and material properties of UHTCs are briefly reviewed, along with an overview of limited studies on their responses to neutron irradiation and an evaluation of plasma-surface interactions. Five key research pathways, primarily focused on irradiation effects, for advancing UHTCs in PFC applications are discussed: (1) assessing irradiation effects on the coupled thermal–mechanical performance (2) addressing the lack of studies on irradiation, plasma-surface interactions, and their synergistic effects; (3) investigating high-temperature (>1000 °C) neutron irradiation effects critical for PFC performance; (4) optimizing multi-component UHTC compositions or composites to improve thermal or mechanical properties; (5) enhancing radiation resistance to mitigate microcracking and void swelling through strategies such as increasing sink strength by reducing grain size, introducing fine particles, and leveraging complex concentrated alloy concepts.

36 MATERIALS SCIENCE↗