To understand the impact of stoichiometry on the thermomechanical performance of zirconium carbide (ZrC), 3 mm diameter, 0.25 mm thick, truncated disc specimens with varying C/Zr ratios (0.84, 1.05, and 1.2) were subjected to three-point flexure at temperatures between 300 and 1673 K and strain rates between 0.0001 and 0.01/s in a high-vacuum environment. Although the sub-stoichiometric (C/Zr-0.84) composition exhibited higher flexural strengths because of fine grains and increased microstructural and compositional homogeneity, a significant decrease in strength with increasing temperature was observed as a result of the onset of plasticity. Further evidence of plastic behavior beyond 1270 K in this composition was indicated by drastically decreasing strain rate sensitivity. Although samples with higher carbon contents (C/Zr ratios of approximately 1.05 and 1.2) exhibited overall lower strengths compared with the sub-stoichiometric composition, the dependence of the strength and the strain rate sensitivity on the test temperature was less pronounced. The overall reduction in strength and the preservation of brittle behavior even at high temperatures in ZrC1.05 and ZrC1.2 could be attributed to the carbon-rich phases in these samples. Microstructural observations of the fractured specimens indicated strong stoichiometry dependence of the cleavage facet morphology. As the carbon content increased, cleavage steps became finer as the cleavage facets appeared more fragmented. This led to an overall increase in the fracture surface roughness, indicating the preservation of brittle behavior.

Coated fuel particle architectures with ZrC coatings are candidate fuels for advanced power reactors and space nuclear propulsion (SNP) concepts. Owing to its relevance to SNP, the composition, microstructure, and mechanical properties of eight ZrC coatings prepared by fluidized bed chemical vapor deposition were evaluated. Evaluation by SEM and EBSD showed that all grains were columnar. Across the various examined samples, minor axis diameters varied between 0.3 and 1.1 μm, and major axis diameters varied between 0.4 and 2.3 μm. Major and minor diameters increased with thickness particularly at higher deposition temperatures in which the major grain axis (from an ellipse fit to the grain shape) increased by 2.5 μm over the entire coating. Coatings with higher reactive gas flows and Zr/C concentrations closer to 1 were observed to contain nanocrystalline graphite deposits. Reactive gas flow doubling led to increases in coating thickness from around 10–15 μm to around 22–27 μm.

Specimens composed of a novel nuclear fuel architecture underwent testing in the Transient Reactor Test Facility (TREAT) to investigate their behavior in thermomechanical states only possible with this type of testing. Uranium mononitride fuel kernels, with coatings typical of Tristructural Isotropic (TRISO) systems, were loaded into additively manufacturing silicon carbide (SiC) canisters and treated by chemical vapor infiltration so that the TRISO particles were held in a fully ceramic SiC matrix. A metallic heat sink capsule was used in TREAT to test fresh fuel specimens under progressively higher energy depositions. Thermomechanical modeling and post transient examinations, including a unique application of x-ray computed tomography, showed that more energetic transients led to stronger temperature gradients, higher stresses, and increased level of fracturing in the specimens. While some specimens were subjected to challenging conditions, the level of fracturing observed did not show a measurable dimensional change or loss of specimen geometry. These results indicate that this fuel architecture has resilience to catastrophic failure under power pulse transients. Future experiments are recommended for specimens which have been previously irradiated in order to investigate the effects of fuel burnup and neutron fluence on transient fuel performance.

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. This work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with-55% TRISO parti-cles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low ( < 250 degrees C) surface temperatures. Initial testing at higher (-700-750 degrees C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (-300-750 degrees C) of both fueled com-pacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause com-plete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, result-ing in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.(c) 2023 Elsevier B.V. All rights reserved.

Uranium carbides are receiving renewed interest as a preferred nuclear fuel composition for advanced reactors due to their numerous favorable properties. Like many other refractory and transition metal carbides these compounds exist in both hypo- and hyper-stoichiometric compositions which results in significant variations in physical, thermal, and mechanical properties. This manuscript surveys both historic and recent literature to compile important properties data mainly as a function of temperature and carbon content to support fuel performance modeling. Expressions and fits for such properties are also suggested when allowed by complete data sets. The manuscript also attempts to highlight gaps and discrepancies in the reported data to assist the nuclear fuels community identify key areas that may require further attention in terms of future research and development.(C) 2021 Elsevier B.V. All rights reserved.

the most influential factors on the integrity of TCR core, testing and evaluation have focused on producing mechanical and thermophysical properties data for the binderjet/CVI SiC. Baseline mechanical and thermophysical properties were measured from the disk specimens printed for different and sizes orientations, which included equibiaxial flexural failure strength, elastic constants, thermal diffusivity and conductivity, density, and the coefficient of thermal expansion. Flexural failure strength datasets showed similar Weibull distributions regardless of sample variants including different orientations. The mean failure strengths of the 3D-printed SiC variants were in the range of 280–310 MPa, which are slightly lower than that of the chemical vapor deposition (CVD) SiC. Thermophysical test results showed that specific heat and thermal expansion are not sensitive to the build direction of SiC samples, while thermal conductivity is highly dependent on the build direction and can be correlated to the anisotropic character of the 3D-printed SiC. Neutron irradiation tests were carried out on the 3D-printed 6-mm diameter SiC disk specimens. Irradiation was carried to 2.3 dpa over a temperature range of 360–880°C. No significant degradation in strength was observed in SiC after irradiations in various conditions and with different orientations. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

The fuel form developed for the Transformational Challenge Reactor demonstration program leverages recent advances in manufacturing, materials, and computational sciences, delivering a new architecture for production of high-performance microencapsulated nuclear fuels. The fuel consists of conventionally manufactured uranium nitride tristructural isotropic fuel particles embedded inside a 3D-printed silicon carbide matrix. This paper describes the overall architecture and manufacturing process for this fuel form, its properties and behavior, and the ongoing development activities.

This report presents the mechanical and thermophysical properties of 3D-printed SiC before and after neutron irradiation that have been evaluated to assess the fuel matrix material for the Transformational Challenge Reactor (TCR). The TCR fuel form consists of an additively-manufactured silicon carbide (SiC) matrix and uranium nitride tristructural isotropic (UN TRISO) fuel particles, which is manufactured through a newly developed processing route combining binderjet 3D printing, TRISO fuel particle loading, and chemical vapor infiltration (CVI). Because the fuel matrix is a primary component of the TCR core and its response to mechanical and thermal loads during operation is one of the most influential factors on the integrity of TCR core, testing and evaluation have focused on producing mechanical and thermophysical properties data for the binderjet/CVI SiC. Baseline mechanical and thermophysical properties were measured from the disk specimens printed for different and sizes orientations, which included equibiaxial flexural failure strength, elastic constants, thermal diffusivity and conductivity, density, and the coefficient of thermal expansion. Flexural failure strength datasets showed similar Weibull distributions regardless of sample variants including different orientations. The mean failure strengths of the 3D-printed SiC variants were in the range of 280–310 MPa, which are slightly lower than that of the chemical vapor deposition (CVD) SiC. Thermophysical test results showed that specific heat and thermal expansion are not sensitive to the build direction of SiC samples, while thermal conductivity is highly dependent on the build direction and can be correlated to the anisotropic character of the 3D-printed SiC. Neutron irradiation tests were carried out on the 3D-printed 6-mm diameter SiC disk specimens. Irradiation was carried to 2.3 dpa over a temperature range of 360–880°C. No significant degradation in strength was observed in SiC after irradiations in various conditions and with different orientations. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

The Transformational Challenge Reactor (TCR) is a reactor technology development project led by Oak Ridge National Laboratory. The project is focused on adoption of technology advancements, particularly advanced manufacturing, into nuclear application. The seminal goal of this project is a brief operation of a micro-scale reactor made with novel core materials. A rendering of the TCR core, vessel, and shielding structures can be seen in Figure 1 [1]. The TCR core is cooled by inert gas and comprised of various novel materials such as yttrium hydride moderator and additively manufactured structural components. The focus of this paper, however, is upon TCR’s unique fuel form.

INTRODUCTION The Transformational Challenge Reactor (TCR) is a unique advanced gas cooled reactor concept with an advanced manufactured reactor core [1,2]. By exploiting advanced manufacturing, the TCR demonstration targets implementation of advanced insitu monitoring methods that are expected to immensely reduce the financial and temporal burden of the current nuclear qualification and licensing process while enhancing its rigor.

INTRODUCTION The Transformational Challenge Reactor (TCR) is a helium-cooled, yttrium-hydride moderated reactor being designed for the US Department of Energy Office of Nuclear Energy to demonstrate and deploy advanced manufacturing technologies in nuclear design and manufacturing processes [1]. To inform the TCR safety analysis, sample fuel elements will be placed in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) and exposed to pulses representative of TCR reactivity-insertion accident (RIA) conditions. Using TRACE and RELAP5-3D, we calculated fuel temperatures and average energy deposition in TCR fuel at both hot full power (HFP) and hot zero power (HZP) for both sub-prompt and prompt reactivity insertions. Though RELAP has been used for gas-cooled reactor safety analysis in the past [2], the prompt generation time in TCR is an order of magnitude shorter than for other gas-cooled reactors due to the use of yttrium-hydride instead of graphite as the moderator [3]. The TREAT facility at INL is an air-cooled, graphitemoderated reactor fueled with uranium oxide dispersed into the graphite blocks. It has been used for fuel testing in both fast and thermal reactors, and following its nearly 25-year shutdown, the facility has resumed operation and demonstrated an ability to match historical tests. This provides confidence in the TREAT team’s ability to design tests for a wide range of transients including pulses representative of reactivity insertions terminated by either temperature feedback or negative reactivity insertion and pulses providing a power profile representative of a loss of coolant accident in a light-water reactor [4]. The demonstrated capabilities of TREAT for both fast and thermal reactors give us confidence that TREAT can create pulses that mimic anticipated TCR transients. To gain a better understanding of the impact of our transients on the heterogeneous fuel form, the RELAP boundary conditions were also used to inform a thermomechanical analysis using the fuel performance code

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We present results from atomistic simulations of sputtering and near-surface implantation of concurrent He, D, and T bombardment of cubic silicon carbide (3C-SiC). This is achieved by first establishing a many-body interatomic potential parameter set to treat interactions of He and hydrogenic species in 3C-SiC informed by abinitio calculations. To obtain sputtering yields we perform both classical molecular dynamics and binary collision approximation simulations for normal incident particles having energies ranging from 25 to 800 eV. We find that due to differences in species surface binding energy of various crystallographic surfaces in 3C-SiC, the sputtering yield of Si is significantly less than that of C, but sputtering yields show limited sensitivity to crystallographic surface orientation. An exception to this occurs when the terminating crystallographic surface plane is more rich in Si rather than C, resulting in comparable sputtering yields of Si and C. The influence of temperature on sputtering is explored and shows limited effect. Finally, the nature of implanted He, D, and T within 3C-SiC surfaces is investigated to understand implantation profiles and stability of defects.

To obtain valuable data that will help in the design and commissioning of the Transformational Challenge Reactor, a series of graded, increasingly complex irradiation tests are planned that will demonstrate the feasibility of advanced manufactured reactor core components. In the first irradiation test, conventionally- fabricated uranium nitride tristructural isotropic (UN TRISO) fuel kernels will be embedded into compacts that are fabricated using advanced manufacturing methods. These compacts are to be irradiated at Massachusetts Institute of Technology’s Nuclear Reactor Laboratory and Idaho National Laboratory’s Transient Reactor Test Facility. This report presents the characteristics of the proposed samples and experiments.

Increasing demand for sustainable battery technologies has necessitated substantive global research efforts towards delivering improvements in terms of both, the underlying battery materials and processes. Due to favorable attributes such as high gravimetric energy density, batteries based broadly on inorganic-ion, and more specifically on Lithium-ion (Li-ion), electrochemistry have emerged as the technologies of choice for current and next-generation consumer electronic devices and electric automobiles. Improvements in critical system metrics such as energy/power density, cycle-life, cost, safety, and technological sustainability rely heavily on multiparameter, multi-scale, and multi-physics optimization of the materials, interfaces, and processes that constitute a battery system. Over the past decade, Scanning Probe Microscopy (SPM) techniques have delivered important solutions for addressing these technological challenges through new characterization tools at the atomic-tonanoscale. This article presents a comprehensive review of these SPM techniques of relevance for the battery systems. Specifically, each SPM operation mode of relevance, which include atomic force microscopy, electrochemical atomic force microscopy, Kelvin probe force microscopy, electrochemical strain microscopy, scanning ion conductance microscopy, and scanning electrochemical microscopy, have been discussed in terms of their state-of-the-art, applications, advantages, limitations, and future directions. Furthermore, a detailed analysis of the following information and insights revealed by these methodologies has been provided: ionic transport and diffusion dynamics, electrochemical activity, surface and interfacial studies, morphological and dimensional changes, and evolution of mechanical properties/their degradation.

This paper reports on plastic recovery and self-healing behavior in longitudinally-twinned and [112] orientated SiGe nanowire (NW) beams when they are subjected to large bending strains. The NW alloys are comprised of lamellar nanotwin platelet(s) sandwiched between two semi-cylindrical twins. The loading curves, which are acquired from atomic force microscope (AFM) based three-point bending tests, reveal the onset of plastic deformation at a characteristic stress threshold, followed by further straining of the NWs. This ductility is attributed to dislocation activity within the semi-cylindrical crystal portions of the NW, which are hypothesized to undergo a combination of elastic and plastic straining. On the other hand, the lamellar nanoplatelets undergo purely elastic stretching. During the unloading process, the release of internal elastic stresses enables dislocation backflow and escape at the surface. As a result, the dislocations are predominantly annihilated and the NW samples evidenced self-healing via plastic recovery even at ultra-large strains, which are estimated using finite-element models at 16.3% in one of the tested devices. Finite element analysis also establishes the independence of the observed nanomechanical behavior on the relative orientation of the load with respect to the nanoplatelet. This first observation of reversible plasticity in the SiGe material system, which is aided by a concurrent evolution of segmented elastic and plastic deformation within its grains during the loading process, presents an important new pathway for mechanical stabilization of technologically important group-IV semiconductor nanomaterials.

Atomic force microscopy (AFM) based nanomechanics experiments involving polytypic todorokite-like manganese dioxide nanobelts reveal varied nanomechanical performance regimes such as brittle fracture, near-brittle fracture, and plastic recovery within the same material system. These nanobelts are synthesized through a layer-to-tunnel material transformation pathway and contain one-dimensional tunnels, which run along their longitudinal axis and are enveloped by m × 3 MnO6 octahedral units along their walls. Depending on the extent of material transformation towards a tunneled microstructure, the nanobelts exhibit stacking disorders or polytypism where the value for m ranges from 3 to up to ∼20 within different cross-sectional regions of the same nanobelt. The observation of multiple nanomechanical performance regimes within a single material system is attributed to a combination of two factors: (a) the extent of stacking disorder or polytypism within the nanobelts, and (b) the loading (or strain) rate of the AFM nanomechanics experiment. Controllable engineering of recoverable plasticity is a particularly beneficial attribute for advancing the mechanical stability of these ceramic materials, which hold promise for insertion in multiple next-generation technological applications that range from electrical energy storage solutions to catalysis.