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DOE OSTI2022

Behavior of helium cavities in ion-irradiated W-Ni-Fe ductile-phase toughened tungsten

This study reports on the distribution of helium (He) cavities in a hot-rolled W-Ni-Fe ductile-phase toughened tungsten (DPT W) composite irradiated to a dose and a helium concentration that are comparable to those in the material after 5-year irradiation in a conceptual fusion power plant. Here, the DPT W sample consists of W particles embedded in a ductile-phase NiFeW matrix with a nominal composition of 90W-7Ni-3Fe by weight. It was hot-rolled to a thickness reduction by 87% (87R DPT W). Sequential irradiation of the material with 1.2 MeV Ni + ions to a fluence of 2.15×10 16 Ni + /cm 2 and 90 keV He + ions to 6.5×10 15 He + /cm 2 was performed at 973 K. Larger He cavities with a lower number density are observed in NiFeW than W. Helium cavities are aggregated preferentially along the NiFeW/W interphase boundary. This behavior is not observed along the W/W grain boundary under the same irradiation conditions. A similar distribution of He cavities with comparable sizes is observed in 87R DPT W irradiated with He + ions only, suggesting that the He cavity distribution is not significantly affected by pre-irradiation of Ni + ions at 973 K. The data from this study does not provide any evidence for formation of visible voids or Ni precipitates in W phase irradiated with Ni + ions only. Diffusion and clustering of He atoms and mobile vacancies during ion irradiation at 973 K are responsible for the formation and growth of the He cavities.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Orientation-selected micro-pillar compression of additively manufactured 316L stainless steels: Comparison of as-manufactured, annealed, and proton-irradiated variants

In this work, irradiation response and deformation mechanisms of additively manufactured (AM) 316L stainless steel were studied by atomic scale characterization and micro-pillar compression. The AM 316L stainless steels were fabricated by direct energy deposition, a laser-based additive manufacturing process. Irradiation with 2 MeV protons at 360°C was performed to create ~1.8 displacements-per-atom (dpa) damage in AM 316L. Deformation behaviors of the as-manufactured, annealed, and proton-irradiated variants were studied, focusing on the effects of manufacturing-induced pores, residual stress, and irradiation-introduced defects (dislocation loops and voids). Micro-pillars were prepared from grains of pre-selected orientation, avoiding contributions of grain boundaries and allowing determination of resolved shear stress on {111} glide planes. Transmission electron microscopy was used to characterize the pre- and post-deformation microstructure. It was found that in the as-manufactured alloy variant, moving dislocations were the major deformation carrier, with noticeable blocking by fabrication-induced pores, In the annealed variant, hardness was reduced, and deformation was also accomplished by dislocation gliding. In the proton-irradiated variant, significant twinning was observed. Comparing measured resolved shear stress and predicted critical stress for dislocation dissociation, we conclude that irradiation hardening became high enough to activate twinning. Therefore, the deformation mechanism changes from dislocation gliding to twinning. The study is important for both processing optimization and performance evaluation of AM alloys for reactor applications.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Dynamics of deuterium retention and desorption from plasma-facing materials in fusion reactor-relevant conditions

Hydrogen isotopes retention and desorption during and after discharges in fusion devices are still not well understood due to the complex device conditions and limitations of in-situ diagnostics and measurements. Here, we simulated well-diagnosed recent experiments at the DIII-D facility to benchmark our ITMC-DYN integrated package of modeling deuterium diffusion, retention, and desorption during and after D discharge irradiation. Modeling results were compared with detail experimental data of D desorption fluxes for various irradiation conditions. We predicted the temporal evolution of free and trapped D distribution in tungsten (W) plasma-facing material (PFM). Effects of key parameters namely diffusion coefficient, recombination rate, trapping energies against different defect types, were examined in these simulations. Existing experimental data of these parameters in literature varies significantly which makes it harder to identify key mechanisms and physics responsible for hydrogen isotope retention and desorption. The purpose of this work is to accurately simulate recent well-diagnosed reactor experiments given the uncertainties in such parameters and identify mechanisms responsible for the retention and desorption. We implemented the best identified diffusion, recombination, and trapping parameters in ITMC-DYN package that integrate both various collisional and thermal processes. We predicted, for example, that sample cooling between discharges in DIII-D operations can significantly affect the spatial distribution of trapped D in W under reactor irradiation conditions. Correct prediction of desorption spectra from samples irradiated during 10 DIII-D discharges showed that up to 35% of D can be retained in high binding energy defects such as vacancy clusters or voids.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Assessment of effective elastic constants of U-10Mo fuel: A multiscale modeling and homogenization study

The significant microstructural changes that U-Mo fuel undergoes during operation degrades its mechanical properties and structural integrity. Microstructural evolution entails the formation, evolution, and redistribution of porosity in conjunction with grain refinement. In the present paper, we employ numerical approaches to assess the impact of the various microstructural features—grains, nanoscale intragranular fission gas bubbles, and mesoscale intergranular voids—on the degradation of elastic constants. Phase-field microstructure models are combined with the asymptotic expansion homogenization technique in order to derive the effective elastic constants as a function of porosity and fission density. Here the results are verified and compared against theoretical bounds. Using this approach, elastic degradation in operating nuclear fuels can be quantified when the distributions of microstructural features are known.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

TEM characterization of two variants of fuel cladding chemical interaction in a HT-9 Clad U-10Zr Fuel. Variant 1: FCCI with a Zr Rind

Here, this study investigated the fuel cladding chemical interaction (FCCI), a key factor that limits operational temperature and burnup, in an HT-9 clad U-10Zr nuclear fuel sample irradiated to a high burnup of 13.1 at.% at a time-averaged peak inner cladding temperature (PICT) of 530 °C. Previous results showed this fuel sample exhibited two distinct levels of FCCI at d. This paper analyzed the FCCI at an azimuthal position showing an interdiffusion layer of <10 µm using transmission electron microscopy to examine chemical and crystallographic nature of phases at the fuel-cladding interface at the nanoscale level. A ZrC layer and a Zr 3 Si phase were identified at the interface; these, along with the relatively low local temperature, potentially contributed to limit interdiffusion, behaving as inhibitors for deleterious interactions. Lanthanides (Ln) partially consumed the ZrC layer and interacted with Fe, forming a Zr-Ln compound and a (Zr,Ce)Fe 2+x phase while also infiltrating up to 4 µm into the cladding. Neither U nor Zr were observed in the cladding, whereas Fe diffused up to 3–5 µm in the fuel. Fe infiltration formed a ternary U-Zr-Fe ε-phase and likely promoted the precipitation of a Cr-rich α’ phase on the cladding interface. Additionally, a Cr-rich χ-phase, likely formed by the dissociation of pre-existing M 23 C 6 carbide precipitates, was identified about 2–5 µm from the fuel-cladding interface. Irradiation-induced nano-voids were also observed in the HT-9 bulk. These findings provide critical insights into FCCI mechanisms at representative irradiation conditions, essential for developing models simulating in-pile metallic fuel behaviors for next-generation reactors.

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Micro-tensile characteristics of As-fabricated and irradiated AGR-2 TRISO fuel particle buffer, IPyC, and buffer-IPyC interlayer regions

A recently developed micro-tensile sample preparation technique was implemented to evaluate the tensile strengths of the buffer, IPyC, and buffer-IPyC interlayer regions of the unirradiated and irradiated AGR-2 TRISO fuel particles. Understanding the mechanical properties of the buffer-IPyC interlayer is essential for developing thermomechanical models of buffer-IPyC separation, yet there is a lack of experimental data on its micro-tensile properties. TEM analysis was conducted on these regions to determine the microstructural changes relevant to the samples' tensile properties. In the unirradiated TRISO particle samples, the buffer layer demonstrated the weakest tensile strength, while the IPyC layer exhibited the highest. Conversely, in the irradiated TRISO particle samples, the buffer-IPyC interlayer region showed the lowest tensile strength, with the IPyC layer being the strongest. Fractures in the samples from the buffer-IPyC region predominantly occurred either in the buffer layer or at the buffer-IPyC interface. However, some buffer-IPyC interlayer samples displayed stress-strain and fracture behaviors more akin to the IPyC layer than the buffer layer. Analysis of diffraction patterns suggests that irradiation may have increased anisotropy in the three regions tested. Despite this suggested increase in anisotropy, there was no evidence that it affected the measured strengths. The irradiated TRISO particles demonstrated a considerable increase in void space and a decrease in ultimate tensile strength within the buffer-IPyC interlayer region due to the densification and contraction of the buffer layer. Minor variations in diffraction ring patterns were also observed. These changes, coupled with a significant reduction in the Weibull modulus/shape parameter, imply that irradiation-induced densification leads to tearing between the buffer and IPyC layers at locations of elevated porosity in the buffer-IPyC interlayer region.

Tristructural isotropic (TRISO)↗

DOE OSTI2025

Interpreting experimental measurements of helium bubbles using stochastic cluster dynamics models of heterogeneous nucleation and growth in irradiated ferritic alloys

Among a number of other advantageous features, ferritic/martensitic steels are being considered as fusion reactor structural materials due to their low intrinsic swelling under irradiation. However, under high-energy neutron irradiation, He produced through (n, α) reactions stabilizes vacancy clusters, which then act as seeds for bubble formation and growth, negating the intrinsic swelling resistance of these alloys. Standard models of irradiation damage accumulation and microstructural evolution consider homogeneous nucleation as the basis for bubble formation and growth. However, this generally does not explain the large bubble densities and sizes observed experimentally under a number of different conditions. Here, we propose a more realistic physical model of bubble nucleation, formation, and growth designed to capture recent experimental measurements of He-bubble formation and evolution during co-implantation of He and Fe ions in model ferritic alloys. We find that experimental results are explained only when the following three features are considered simultaneously: (i) heterogeneous nucleation of He-vacancy bubbles at defect sinks (e.g., dislocations, grain boundaries, and second-phase precipitates), (ii) nucleation and growth of bubbles via the ‘trap mutation’ mechanism (i.e., spontaneous production of Frenkel pairs due to absorption of He atoms), and (iii) transition from a growth-limited, He-stabilized bubble regime to a ‘runaway’ void-type growth scenario. The model is implemented into a stochastic cluster dynamics framework that takes advantage of cluster size grouping methods to accelerate the simulations, allowing us to reach 10 dpa of simulated irradiated dose, and to capture cluster sizes in excess of 20 nm. Further, a careful extrapolation exercise conducted assuming classical nucleation theory leads to excellent agreement with the experimental measurements at 50 dpa of irradiation.

36 MATERIALS SCIENCE↗

DOE OSTI2026

A comparison of wrought and powder metallurgical FeCrAl claddings under simulated LWR accident transients

Iron-chromium-aluminum (FeCrAl) alloys are potential accident tolerant fuel (ATF) cladding candidates for light-water reactors but are difficult to fabricate as thin-walled tubes via conventional cast-and-wrought routes. Powder metallurgy (PM) offers a manufacturing alternative with improved compositional control, but its transient accident performance has not been directly benchmarked against wrought variants. This study evaluates the burst behavior of commercially developed PM-processed FeCrAl alloys, PM-C26M (Fe-12Cr-6Al-2Mo) and the high precipitate density FA-SMT (Fe-22Cr-5Al-3Mo), under simulated light-water reactor accident transient conditions. Burst testing was conducted using the Severe Accident Test Station with heating rates of 5 °C/s and 50 °C/s and internal pressures ranging from 25 MPa to 100 MPa. PM-C26M reproduced wrought C26M burst behavior within 7–37 °C across the stress range, indicating that PM processing does not compromise transient strength. FA-SMT exhibited markedly higher burst temperatures and reduced heating-rate sensitivity, consistent with its engineered precipitate strengthening. FA-SMT rupture exhibited axial "unzipping" rather than the lateral tearing characteristic of PM- and wrought C26M. Post-test EBSD and fractography indicate that this behavior is strongly correlated with strain-gated intergranular void nucleation associated with the dense precipitate architecture of FA-SMT, a response absent in the comparatively clean PM-C26M matrix and consistent with rupture morphologies reported for oxide-dispersion strengthened (ODS) FeCrAl of similar base-matrix chemistry to PM-C26M. These findings highlight the potential of powder metallurgy as a viable fabrication route for ATF claddings from an accident performance standpoint.

Bell, Sam [ORNL] (ORCID:0000000251905657)↗

DOE OSTI2022

In situ feature analysis for large-scale multiphase flow simulations

The study of multiphase flow is essential for designing chemical reactors such as fluidized bed reactors (FBR), as a detailed understanding of hydrodynamics is critical for optimizing reactor performance and stability. An FBR allows scientists to conduct different types of chemical reactions involving multiphase materials, especially interaction between gas and solids. During such complex chemical processes, the formation of void regions in the reactor, generally termed as bubbles, is an important phenomenon. The study of these bubbles has a deep implication in predicting the reactor’s overall efficiency. But physical experiments needed to understand bubble dynamics are costly and non-trivial due to the technical difficulties involved and harsh working conditions of the reactors. Therefore, to study such chemical processes and bubble dynamics, a state-of-the-art computational simulation MFIX-Exa is being developed. Despite the proven accuracy of MFIX-Exa in modeling bubbling phenomena, the large-scale output data prohibits the use of traditional post hoc analysis capabilities in both storage and I/O time. Herein, to address these issues and allow the application scientists to explore the bubble dynamics in an efficient and timely manner, we have developed an end-to-end analytics pipeline that enables in situ detection of bubbles, followed by a flexible post hoc visual exploration methodology of bubble dynamics. The proposed method enables interactive analysis of bubbles, along with quantification of several bubble characteristics, enabling experts to understand the bubble interactions in detail. Positive feedback from the experts has indicated the efficacy of the proposed approach for exploring bubble dynamics in very-large-scale multiphase flow simulations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2023

Structural and electrochemical evolution of alloy interfacial layers in anode-free solid-state batteries

“Anode-free” solid-state batteries feature high energy density since there is no anode active material. Although the beneficial effects of interfacial layers at the anode-solid electrolyte interface have been demonstrated, the mechanisms through which they influence lithium plating/stripping are unclear. Here, in this study, we reveal the evolution of 100-nm silver and gold interfacial layers during lithium plating/stripping using electrochemical methods, electron microscopy, X-ray microcomputed tomography, and modeling. The alloy layers improve Coulombic efficiency and resistance to short circuiting, even though the alloys form solute regions or particulates that detach from the current collector during plating. In situ electrochemical impedance spectroscopy shows that the alloys return to the interface and mitigate contact loss at the end of stripping, avoiding a critical vulnerability of anode-free cells. Contact retention is driven by even Li thickness that promotes spatially uniform stripping, as well as local alloy delithiation in response to current concentrations that homogenizes current and diminishes voiding.

25 ENERGY STORAGE↗

DOE OSTI2020

Long term durability test and post mortem for metal-supported solid oxide electrolysis cells

Hydrogen is a renewable energy carrier, and electrolysis to split water is the most environmentally friendly method to produce hydrogen. This work reports long-term durability and degradation mode analysis for metal-supported solid oxide electrolysis cells (MS-SOECs). Catalyst screening showed that MS-SOECs with composite electrode catalysts (samarium-doped ceria-nickel [SDC-Ni] serving as a fuel electrode catalyst, and praseodymium oxide [PrO x ]-SDC or La 0.6 Sr 0.4 Co 0.2 Fe 0.8 O 3 [LSCF]-SDC serving as an air electrode catalyst) exhibit the highest electrochemical performance at 700 °C. The degradation rate of cells with LSCF-SDC as the air electrode catalyst was as low as 1.3%/100 h in long term durability tests at a current density of 0.33 A cm -2 , in contrast to rapid degradation observed for a cell with a PrO x -SDC air electrode. Furthermore, post-mortem analysis reveals the degradation is dependent on the primary modes of fuel electrode catalyst coarsening and Cr poisoning on the air electrode catalyst, as well as secondary modes of oxidation of the metal support and local elemental accumulation of Ni. Other degradation modes reported in conventional anode-supported SOECs, such as Ni migration, foreign element contamination, delamination of the cell, and nano-voids on the electrolyte, are not observed in the present MS-SOECs.

25 ENERGY STORAGE↗

DOE OSTI2021

The effect of aspect ratio on the mechanical behavior of Li metal in solid-state cells

The Li metal anode is necessary for solid-state batteries to dramatically exceed the energy density of state-of-the-art conventional Li-ion batteries. To enable the Li metal anode, the mechanical properties of Li metal under compression in a commercially relevant cell configuration must be determined. This work examines the compression creep behavior of Li metal coupled with a relevant model solid-state electrolyte (Li 6.5 La 3 Ta 0.5 Zr 1.5 O 12 ) and current collector (Ni) as a function of applied stress (1–10 MPa) and Li foil thickness (100–700 μm). At 1 MPa applied stress, the strain rate of all Li foil thicknesses under compression was almost four orders of magnitude lower than what is predicted from tensile creep tests. Steady-state creep behavior was observed for Li foil ≥350 μm at 5 MPa and ≥250 μm at 10 MPa. Conversely, sustained creep deformation was not observed for 100 and 150 μm Li foils at any applied stress. It is suggested that friction at the Li/LLZO and Li/Ni interfaces leads to the formation of hydrostatic stress that controls Li deformation at low aspect ratios. Finally, these results have implications for battery pack stack pressure requirements, particularly relating to void formation during stripping and Li metal filament penetration during plating.

25 ENERGY STORAGE↗

DOE OSTI2026

The structure, composition, and performance impact of a YSZ-GDC interdiffusion layer in solid oxide electrolysis cells

This study provides a combined experimental and computational investigation into the structure and impact of the cation interdiffusion layer that appears at the gadolinium doped ceria (GDC)/yttria stabilized zirconia (YSZ) interface in solid oxide electrolysis cells (SOECs). Scanning transmission electron microscopy (STEM) illustrates that a ∼0.4 μm interdiffusion layer (IDL) with an intermixed cation distribution and fine grain size forms upon sintering. STEM identifies that the interdiffusion layer exists in the cubic fluorite structure despite changes in cation composition. The interdiffusion layer microstructure formed during sintering does not change during SOEC testing at either 1.3V or heightened voltage pulse testing. Modeling predicts that ionic conductivity may decrease in the interdiffusion layer due to Coulombic trapping between mobile oxygen vacancies and excess Gd 3+ acceptor dopants. Yet, the density and continuous nature of the layer should benefit cell stability by substantially reducing the formation of SrZrO 3 , which is corroborated by STEM and Synchrotron X-ray diffraction (XRD). We conclude that the interdiffusion layer acts as a beneficial barrier to Sr diffusion, when operating in a regime where electrolyte void formation is not observed.

organic↗

DOE OSTI2021

Factors controlling pore network development of thermally mature Early Palaeozoic mudstones from the Baltic Basin (N Poland)

Understanding the formation of pore space, especially in low porosity shales (as source rocks and as unconventional resources), is critical to the oil and gas industry, since pores control the space available for hydrocarbon and participate in hydrocarbon transport. We examined 87 Ordovician and Silurian mudstone samples collected from four wells located in the Pomeranian part of the Baltic Basin (northern Poland), one of the primary Polish targets for hydrocarbon exploration. These samples represent the Pelplin, the Paslek, the Jantar, the Prabuty, and the Sasino Formations, which still requires more detailed porosity studies. Our study aimed to identify factors controlling porosity development, by applying bulk techniques (organic petrology and TOC analyses, quantitative mineralogy, and porosimetry) as well as nano- to microscale techniques (thin section petrography, electron microscopy). The studied samples are mainly argillaceous mudstones. The results of porosimetry measurements, combined with image analysis, indicate that the pores of all studied rocks are dominated by micropores (pores < 2 nm in diameter), mesopores (2-50 nm in diameter) and small macropores. The SEM images showed three main pore types: a) voids related to clay mineral aggregates, b) pores inside organic matter particles, and c) pores between other mineral grains. In the Jantar and Sasino mudstones, the organic matter content and its thermal maturity control porosity. The occurrence of solid bitumen in the rocks from these formations reduces samples’ mesoporosity because of the pore-clogging effect. In contrast, in the Paslek and Prabuty Formations, there is low organic matter content and specific surface area and the volume of mesopores increase with clay minerals content. In the Pelplin mudstones, there are no prevailing factors controlling porosity. As a result, we suggest that a combination of SEM image analysis and dual liquid porosity (DLP) measurements is a powerful method to assess porosity available for petroleum flow in mudstones.

03 NATURAL GAS↗

DOE OSTI2021

Neutron Irradiation Induced Changes in Isotopic Abundance of 6Li and 3D Nanoscale Distribution of Tritium in LiAlO2 Pellets Analyzed by Atom Probe Tomography

Tritium, a radioactive isotope of hydrogen is produced by neutron irradiation of 6Li enriched LiAlO2 pellets in Tritium producing burnable absorber rods (TPBARs). Three-dimensional nanoscale mapping of Tritium in irradiated pellets is critical to understand its spatial variation and its association with various microstructural features in the pellets. Spatially resolved analysis of 6Li isotopic enrichment in such ceramic pellets before and after irradiation can provide insights to heterogeneities in Li isotopic distribution. However, Li being a light isotope, its spatially resolved compositional or isotopic analysis is a challenging task to most analytical methods. Here we used atom probe tomography, to evaluate the Li isotopic enrichment in both the LiAlO¬2 matrix and the Li deficient secondary phase LiAl5O8 before irradiation. These results were then compared with 6Li isotopic ratio measured by APT analysis of neutron irradiated pellets to clearly demonstrate the unique capability of APT to quantitatively analyze the isotopic enrichment of 6Li as well as for other light elements present in these materials. Evidence for heterogenous nanoscale distribution of 3H and O3H within irradiated pellet microstructure is also provided, which is attributed to 3H trapping in irradiation induced vacancy clusters or voids within the LiAlO2 matrix. The characterization results given here, now prove the feasibility for using APT for analyzing nanoscale spatially resolved isotopic enrichment of Li as well as other light elements such as Tritium in TPBARs. This work also paves way to adopt APT for analyzing light elements such as Li and tritium in materials used for nuclear fusion reactor applications, geochemistry, astrophysics as well as for energy storage applications.

Devaraj, Arun↗

DOE OSTI2021

Microstructure and mechanical properties of friction stir welded Haynes 282

Advanced ultrasupercritical (A-USC) steam plants are designed to operate at high temperatures and pressures due to the necessity for higher operational efficiency. The extreme operating conditions of A-USC requires the deployment of precipitation strengthened Ni-base alloys that exhibit elevated temperature strength and good fabricability. Fusion welding of precipitation strengthened nickel alloys lead to solidification cracking in fusion zone and/or liquation cracking in heat affected zone. Therefore, an alternative non-melting welding technique is a necessity to efficiently join nickel alloys. In the current study, friction stir welding, a solid state joining technique was implemented on a precipitation strengthened nickel-based superalloy, Haynes 282. Detailed microstructural and mechanical properties characterization was carried out. The processed region exhibited wrought, fine-grained microstructure, absence of weld defects such as voids and cracks and absence of elemental segregation. Both hardness and cross-weld tensile tests demonstrated that the weld region was stronger than the base material. And the cross-weld tensile samples failed in the base material. Based on joint efficiency analysis, friction stir welded Haynes 282 outperformed fusion welds.

36 MATERIALS SCIENCE↗

DOE OSTI2021

A study on the creep behavior of alloy 709 using in-situ scanning electron microscopy

In this research, an experimental evaluation of creep properties of Alloy 709 in the temperature range of 750–850 °C was undertaken. Alloy 709 is a novel austenitic stainless steel with 20% Cr and 25% Ni by wt% that was developed for application in structural components of nuclear power plants. Creep rupture tests were conducted in an in-situ heating-loading and Scanning Electron Microscope (SEM) unit equipped with Electron Backscatter Diffraction (EBSD) detector and Energy Dispersive Spectroscopy (EDS). “Real-time” creep damage mechanisms of Alloy 709 at various stresses and temperatures using a flat, un-notched sample with continuously reducing cross-section is studied so that the failure and maximum creep damage occurred at the center of the sample where the in-situ SEM imaging could be focused. Accelerated creep tests at temperatures and stresses above service conditions were performed by employing multiple blocks of constant loads where the loads were increased once the sample attained constant creep rate, indicating a secondary creep regime. This technique ensures multiple data points can be obtained from the same test, saves the time required for an otherwise long-term creep test and usage of SEM. Further, Coincident Site Lattice (CSL) boundary maps were collected as control maps before testing, and the grain boundaries were observed during the creep test to understand the effect of grain boundary character on the creep damage mechanism. Void growth, grain boundary separation, and sliding were found to be the main creep mechanisms whose rate is dependent on stress and temperature. Failure mechanisms studied on the fracture surface using SEM fractography were correlated to the sample surface observations to create complementary information to better understand the underline creep mechanism of Alloy 709.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Multiscale characterization and comparison of historical and modern nuclear graphite grades

Beginning with Chicago Pile I, graphite has been used as a moderator material in nuclear power stations and is considered a potential material for use in future Generation IV advanced reactors. The microstructure of graphite is responsible for much of its mechanical and thermo-physical properties, and how it responds to irradiation. To understand graphite microstructure, it is necessary to understand its porosity at the macro- and micro-scales; and to understand its porosity, it is necessary to characterize the morphological connectivity of the void content and the two main phases of graphite: filler and binder. Here, using several microscopy and analytical techniques, a detailed examination of the heterogeneity, microstructure and pore structure of different graphite grades and their binder and filler phases is presented. Significant differences were found between coarser and finer nuclear grades. Coarse grades have a more diverse range of filler particles, pores and thermal cracks. Finer grades have a more well-defined pore size distribution, fewer variations of filler particles sizes and do not contain as many large thermal cracks. Fine grades tend to have a well-connected network of pores whereas coarser grades contain a larger content of closed porosity. The framework developed within this work can be applied and used to assess the various graphite grades that would down-select materials for specific use in graphite moderated reactor designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗