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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Social support and cognitive function in Chinese older adults who experienced depressive symptoms: is there an age difference?

Objective This study examined the moderating effect of overall social support and the different types of social support on cognitive functioning in depressed older adults. We also investigated whether the moderating effect varied according to age. Methods A total of 2,500 older adults (≥60 years old) from Shanghai, China were enrolled using a multistage cluster sampling method. Weighted linear regression and multiple linear regression was utilized to analyze the moderating effect of social support on the relationship between depressive symptoms and cognitive function and to explore its differences in those aged 60–69, 70–79, and 80 years and above. Results After adjusting for covariates, the results indicated that overall social support (β = 0.091, p = 0.043) and support utilization (β = 0.213, p < 0.001) moderated the relationship between depressive symptoms and cognitive function. Support utilization reduced the possibility of the cognitive decline in depressed older adults aged 60–69 years (β = 0.310, p < 0.001) and 80 years and above (β = 0.199, p < 0.001), while objective support increased the possibility of cognitive decline in depressed older people aged 70–79 years (β = −0.189, p < 0.001). Conclusion Our findings highlight the buffering effects of support utilization on cognitive decline in depressed older adults. We suggest that age-specific measures should be taken when providing social support to depressed older adults in order to reduce the deterioration of cognitive function.

Jing, Yurong↗
DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types, such as light- and heavy-water-moderated reactors, graphite-moderated reactors, and breeder reactors, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. Because of the wide variety and condition of SNF, a robust canister termed the DOE Standardized SNF Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard SNF Canister had four variations consisting of a 10- or 15-foot length and an 18- or 24-inch diameter. For ultimate disposal in the Yucca Mountain Repository these canisters were to be grouped with 24-inch diameter HLW canisters in a codisposal waste package. The 18-inch DOE Standardized SNF Canister could be placed in the middle of five HLW canisters. The 24-inch DOE Standardized SNF Canister would take the place of one the five HLW canister on the outer ring in the codisposal waste package. No DOE Standardized SNF Canisters have been loaded, and many fuel types can be loaded in multiple canisters. Depending on the selected diameter of the DOE Standardized SNF Canister for a particular fuel, the number of potential codisposal waste packages could vary. In addition, different types of DOE SNF may be able to be combined in a single canister. This could significantly change the number of DOE SNF canisters that are loaded and result in a decrease of the number of total potential codisposal waste packages. Ultimately, if too many HLW canisters need disposition, then a reduction in the number of DOE Standardized SNF Canisters will change the final total of codisposal waste packages very little. Likewise, if the ratio of DOE Standardized SNF Canisters to HLW canister produced is too high, then DOE Standardized SNF Canisters will not have enough HLW canisters to surround them in the codisposal waste package. This paper examines the results of differing loading strategies for a few DOE SNFs. It also compares the ranges of DOE Standardized SNF Canisters that may be produced to the ranges of HLW canisters that may be produced in order to determine potential scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Yttrium Hydride Post-Irradiation Examination Plan

Micro nuclear reactors are expected to use metal hydride moderators to ensure thier compact design for the flexible operational purposes. Department of Energy’s Microreactor program therefore aims to provide infrasture and qualification bed for the development of micro nuclear reactors. Because micro nuclear reactors must be compact to offer a flexible operational purposes, metal hydrides are expected to be used as neutron moderator materials due to their high hydrogen number density. Los Alamos National Laboratory (LANL) directed by the DOE to provide a qualified fabrication techniques of metals hydride, especially the yttrium hydride (YHx) which provides advantages over other metal hydride candiates. Qualification of YHx involves fabrication, property, and performance pillars, which must be met for the deployment of microreactors. In the support of the qualification process metal hydride, LANL has developed fabrication techniques based in direct hydriding and powder compacting. For both manufacturing techniques, fresh material properties of metal hydrides have been reported in the literature. To establish the performance limits and the qualification parameters, an irradiation campaign has been initated at Idaho National Laboratory (INL). The irradation campaign aims to investigate the critical irradiated YHx properties as a function of neutron fluence and temperaute. The first phase of the irradiation campaign includes the effect of the operating temperature under neutron irradiation which provides initial data on the metal hydrides and informs the manufacturing processes. So that, drop-in irradations in Advanced Test Reactor (ATR) has been designed. This document depicts the post-irradation examination (PIE) plan of the ATR irradiated YHx. Objective of the PIE plan is to serve as a qualification basis for YHx as a neutron moderator material by understanding the impacts irradiation has on the physical and thermophysical properties; to establish a link between fabrication and the irradiated YHx properties; to institute a PIE strategy for other solid-state neutron moderator candidates, and; to inform micro nuclear reactor fuel performance and safety codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Electrochemically modulated molten salt reactor

An electrochemically modulated molten salt reactor (EMMSR) that contains a vessel and a power source. The vessel houses a fuel salt, at least a portion of a neutron moderator, and at least a portion of an insulator. The fuel salt includes enough dissolved fissile isotopes to cause continued self-sustaining fission reactions during the operation of the EMMSR. The neutron moderator is configured to slow down fast neutrons produced by the dissolved fissile isotopes. The insulator is configured to electrically isolate the neutron moderator from the vessel. The power source has a positive potential and a negative potential. The positive potential is received by the neutron moderator and the negative potential is received by the vessel.

Jiang, Junhua↗
DOE OSTI2023

Microstructural characterization of the oxidation of nuclear graphite under chronic and accident conditions via XCT and synchrotron XCT

Graphite is a candidate material to moderate fast neutrons and for structural components in the US next-generation graphite-moderated reactors. A graphite core is conceived as a large formation of interconnected bricks that primarily serves as the moderator of fast neutrons, holds instrumentation, fuel elements, control rods, and is a receptacle for molten salts for Molten Salt Reactors (MSRs) designs. During the operation of a graphite-moderated power plant, graphite components might be subjected to chronic oxidation during normal operating conditions or aggressive oxidation as a result of accidental ingress of air that reacts vigorously with the graphite core. Using synchrotron, x-ray computed tomography (XCT), this research systematically characterized microstructural changes that accompanied these two oxidation scenarios. Chronic oxidation was studied by characterizing IG-110, PCEA, and NBG-18 specimens that were gradually oxidized in air at a low temperature (520°C). The accidental ingress of air into the graphite reactor was simulated by subjecting the grades mentioned above to high-temperature oxidation at approximately 750°C. This research is the first, in situ, systematic characterization of nuclear graphite microstructural evolution that can be associated with the two possible oxidation scenarios and provide insight into related repercussions. The results show that the microstructure and pore connectivity influence the rate of oxidation and evolution of the microstructure under the two oxidation regimes. These results are crucial to understanding which grades are more resilient to each type of oxidation and documenting the damage created in the graphite.

Arregui Mena, Jose'↗
DOE OSTI2024

Sum-of-Fractions Method

Sum-of-fractions is a method intended to make sure a subcritical margin for aqueous solutions and slurries of fissionable isotopes exists. The method indicates that a system is subcritical if the sum of the ratios of the mass of each isotope (in a mixture) to its individual minimum subcritical mass limit is less than or equal to one. Historically, the basis of the sum-of-fractions has been derived from allowances given in the American National Standards Institute (ANSI)/ American Nuclear Society (ANS)-8.15-1981. However, the allowance was removed in ANSI/ANS-8.15-2014 due to a lack of technical basis. A methodology was developed to assess the validity of using the sum-of-fractions for water- or polyethylene-moderated systems for the following nuclides: 232U, 233U, 234U, 235U, 237Np, 236Pu, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 242mAm, 243Am, 242Cm, 243Cm, 244Cm, 245Cm, 246Cm, 247Cm, 249Cf, and 251Cf. The methodology uses available benchmark data for mixtures of 233U, 235U, and 239Pu to establish the calculational margin, and a mass limit reduction to establish the margin of subcriticality. Water- or polyethylene-moderated and -reflected mixtures containing the nuclides are evaluated with the code system, SCALE 6.2.4. Including the calculational margin, subcritical mass limits for each nuclide were computed for optimally water- or polyethylene-moderated and fully reflected systems. These masses were used to create nuclide mixtures in which the sum of the mass to subcritical mass limit ratios is one. The various nuclide mixtures were modeled over a range of moderation and demonstrate the keff does not exceed the calculational margin. For additional assurance of subcriticality, a significant mass reduction is applied to each computed minimum critical mass of the nuclides without adequate benchmark data consistent with the method in ANSI/ANS-8.15-2014.

criticality safety, Actinide↗
DOE OSTI2020

Elastic moduli of high-density, sintered monoliths of yttrium dihydride

Yttrium dihydride has been considered as potential nuclear reactor moderator for micro-reactors operating in the thermal energy spectrum due to their ability to retain hydrogen to high temperatures while having a relatively low impact on neutron economy. Despite these advantages, it is difficult to prepare yttrium dihydride in high-density, near net-shape monoliths for moderator applications due to volume expansion upon hydriding and the pyrophoric nature of the material. In this work, high-density monoliths of yttrium dihydride were prepared by direct hydriding and using powder metallurgical methods. Resonant ultrasound spectroscopy was used to measure the elastic moduli of the sintered monoliths, while nanoindentation was used to determine the Young’s modulus and hardness of the directly-hydrided yttrium and the sintered monoliths. Single crystal elastic constants and related properties of yttrium dihydride were also determined using density functional theory (DFT) for comparison. We report mechanical properties of materials produced by both methods were observed to be consistent with the values from literature for yttrium dihydride. Due to the novelty of producing yttrium dihydride by powder metallurgy, this result indicated that the monoliths produced by powder metallurgy were of high quality. To that end, powder metallurgy is believed to be a viable method for large-scale production of yttrium dihydride monoliths for nuclear reactor moderator applications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Thermophysical properties of high-density, sintered monoliths of yttrium dihydride in the range 373–773 K

Yttrium dihydride is a promising nuclear reactor moderator for high-temperature, thermal micro-reactors due to its ability to retain hydrogen to high temperatures while having a relatively low impact on neutron economy. However, it is difficult to fabricate yttrium dihydride in high-density, near net-shape monoliths for moderator applications due to challenges associated with the hydrogen absorption process. Here, high-density monoliths of yttrium dihydride were prepared by powder metallurgical methods. The lattice strain and thermal expansion coefficient were measured using neutron diffraction of powders of directly-hydrided yttrium, while the molar heat capacity and thermal diffusivity of sintered yttrium dihydride were measured using differential scanning calorimetry (DSC) and laser flash analysis (LFA), respectively. The molar heat capacity of yttrium dihydride was also calculated using density functional theory (DFT) for comparison. These parameters were used to then calculate the thermal conductivity and resistivity of yttrium dihydride as a function of temperature. The thermophysical properties of materials produced by both methods were observed to be consistent with the values from literature for yttrium dihydride. Due to the novelty of producing yttrium dihydride by powder metallurgical methods, this result indicated that the sintered monoliths were of high quality and that powder metallurgy is a viable method for large-scale production of yttrium dihydride monoliths for nuclear reactor moderator applications.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Characterization of the microstructure of yttrium hydride under proton irradiation

High moderation per unit volume solid moderator materials like yttrium hydride (YH x ) are necessary for compact nuclear microreactors. However, the phase stability and hydrogen transport processes of YH x under high-temperature irradiation are largely unknown. Proton irradiation was conducted on YH x at 300 °C and 580 °C to 0.2 dpa using 1 MeV or 2 MeV protons in a high-vacuum environment. The hydrogen concentration was determined before and after irradiation using elastic recoil detection analysis, and microstructural evolution was examined via post-irradiation scanning transmission electron microscopy and Raman spectroscopy. Dislocation loops and cavities were observed in all conditions; their distribution was correlated with the bombarding proton energy and ion irradiation temperature. This work revealed that hydrogen retention is proportional to the formation of traps for hydrogen gas atoms and identified pathways for hydrogen release. The relative contributions of bulk or fast diffusion paths, such as grain boundaries, delamination boundaries, and stacking faults are discussed; the primary mechanisms of hydrogen loss are likely based on diffusion, ruling out artefacts of the experimental design. In conclusion, the study suggests proton irradiation may be a strong surrogate to study hydrogen transport in hydride moderator materials under irradiation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

High temperature elastic properties of sub-stoichiometric yttrium dihydrides

Yttrium hydrides are considered as candidate materials for neutron moderation applied in microreactors (akin transportable tiny nuclear reactors) owing to their superior thermal stability and hydrogen retention. The evolution of elastic properties of these materials at elevated temperatures, needed for predicting the thermomechanical response and performance of the moderator during in-service reactor conditions, however, is lacking. Here, we report the Young’s and shear elastic moduli of three stoichiometries of bulk yttrium hydride (YH x , x = 1.61, 1.82, and 1.84) from room temperature to 1000°C. In situ temperature-dependent measurements of the longitudinal and shear wave velocities were performed using a laser ultrasonic technique while heating the sample in a vacuum-pumped heating stage. The elastic moduli increased linearly with increasing hydrogen content and decreased by ~10% during heating from room temperature to 1000°C in the three YH x compositions. The linear relationship between the elastic moduli and the hydrogen content in yttrium hydride was verified by atomistic calculations based on density functional theory (DFT). The absence of abrupt changes in the temperature-dependent measurements of elastic modulus of the YH x samples suggested negligible loss of hydrogen at elevated temperatures. Excellent agreement was found between the measured and calculated dependence of the elastic moduli on the stoichiometry, thereby providing a new approach for investigating the effects of fabrication-induced parameters (such as porosity) on the elastic moduli. Furthermore, this study demonstrates the utility of the combined approach involving DFT-based atomistic calculations and measurements of the elastic moduli for the informative development of metal hydrides and can be used as a metric for novel moderator materials investigations for emerging microreactors and beyond.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Post Irradiation Examination Results of Irradiated Yttrium Hydride

Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.

08 HYDROGEN↗
DOE OSTI2024

Rejection of low-molecular weight neutral organics is highly sensitive to reverse osmosis system design and operation

A computational model was developed to investigate the significance of system design and operating conditions on the rejection of neutral, low-MW organics by reverse osmosis for potable reuse. Here, the model demonstrated that the decrease in local rejection as net driving pressure decreases is substantially greater for moderately rejected compounds than for highly rejected compounds. At recovery values less than 70%, the local permeate concentration can exceed the pressure vessel feed concentration for moderately rejected compounds. System-level rejection of moderately rejected compounds is likewise substantially more sensitive to operating conditions than highly rejected compounds. The findings highlight a drawback of relying on rejection results from bench-scale testing that operates at low recovery, which invariably has higher rejection than full-scale systems operating at similar pressure. The analysis demonstrates a trade-off in which the low-pressure, high-recovery operation desired for potable reuse systems can be detrimental to the removal of low-MW neutral organics. The removal of low-MW neutral organics can be improved if organics rejection is explicitly evaluated during the design process.

42 ENGINEERING↗
DOE OSTI2022

Observational study of the South American low-level jet during the SALLJEX

The South American Low-Level Jet (SALLJ) east of the Andes is a narrow maximum northerly wind present just above the boundary layer. It is an important component of the tropical-extratropical exchange of heat and moisture in South America and favors deep moist convection in Southeastern South America (SESA). In the present study, the SALLJ characteristics during the 2002-2003 summer season are investigated using upper-air measurements collected during the SALLJ Experiment (SALLJEX). At most stations, the highest wind speed was observed below 3000 m between 06 and 12 UTC, mainly in association with a maximization of the northerly component of the wind while the zonal component reached its minimum. Here, an increase in the maximum wind speed from 10°S to Santa Cruz de la Sierra’s latitude and from there a weakening southward was found throughout the SALLJEX network. Additionally, SALLJEX observations allowed to locate SALLJ core purely from observations in a region just east of the Andes in agreement with previous studies. A new localized kinematic classification was developed and applied in this study to analyze the relationship between the SALLJ intensity and the thermodynamic properties of the lower layers of the atmosphere. Weak, moderate, and strong SALLJ categories were defined with a local threshold of the northerly wind speed maximum at each station. The strongest SALLJs were frequently observed at nighttime, while weaker SALLJs were likely to occur at any time of the day. A stabilization of the nocturnal boundary layer was observed between weak and moderate cases whereas a destabilization between moderate and strong cases were found near the time of the wind speed maximum (06 UTC) at Santiago del Estero and Resistencia stations. A deepening of the convective boundary layer and an increase of its mean temperature between weak and strong category during the previous afternoon to nocturnal SALLJs, reveal a possible contribution to the mechanisms that generate the maximum wind speed.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Surrogate-driven design optimization with uncertainty constraints in Monte Carlo simulations

In multi-objective design tasks, the computational cost increases rapidly when high-fidelity simulations are used to evaluate objective functions. Surrogate models help mitigate this cost by approximating the simulation output, simplifying the design process. However, under high uncertainty, surrogate models trained on noisy data can produce inaccurate predictions, as their performance depends heavily on the quality of training data. This study investigates the impact of data uncertainty on two multi-objective design problems modelled using Monte Carlo transport simulations: a neutron moderator and an ion-to-neutron converter. For each, a grid search was performed using five different tally uncertainty levels to generate training data for neural network surrogate models. These models were then optimized using NSGA-III. The recovered Pareto-fronts were analyzed across uncertainty levels: in the moderator problem, normalized hypervolume dropped from 0.886 at 1.0% uncertainty to 0.748 at 10% uncertainty, while in the converter problem it remained near 0.50 for all cases. Average simulation times were also compared to evaluate the trade-off between accuracy and computational cost. Results show that the influence of simulation uncertainty is strongly problem-dependent. In the neutron moderator case, higher uncertainties led to exaggerated objective sensitivities and distorted Pareto-fronts, reducing normalized hypervolume. In contrast, the ion-to-neutron converter task was less affected—low-fidelity simulations produced results similar to those from high-fidelity data. These findings suggest that a fixed-fidelity approach is not optimal. Surrogate models can recover the Pareto-front under noisy conditions, and multi-fidelity studies help identify suitable uncertainty levels for each problem to balance efficiency and accuracy.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Characterizing the Experiment for Calibration with Uranium (Excalibur) neutron source for use in warhead verification

Neutron sources can play a variety of roles in warhead verification. For transmission radiography, a source of directed high energy neutrons is required, while for applications to detect fissile isotopes, sub-MeV neutrons are preferred. The Excalibur (Experiment for Calibration with Uranium) neutron source has been built and used in a variety of verification-related experiments. Excalibur is based on a commercial deuterium-tritium neutron generator specified and measured to be capable of producing 14 MeV neutrons at rates of up to 8.2 × 10 8 neutrons/s. Here, the generator is enclosed in a carbon-steel 32" diameter, 23.62" high carbon-steel cylinder that moderates the mean neutron energy to under 500 keV. This, in turn, is encased in 5%-borated polyethylene such that the entire assembly is a 48" x 48" box that is 30" tall. For radiographic applications, a narrow, tapered channel in the steel and polyethylene allows 14 MeV neutrons to stream directly from the generator to a test object. Its collimating capability is demonstrated by measuring the neutron flux profile. In the moderated mode of operation, the generator is fully enclosed in the steel, but a large section of the polyethylene is removed, providing a flux of sub-MeV neutrons from a wide range of angles. Neutron angular and spectral measurements using both a nested neutron spectrometer and a commercial liquid scintillator coupled with a 3 He detector show the expected softer neutron spectrum in moderated mode in good agreement with MCNP6 calculations. The gamma-ray spectrum from Excalibur is also in good agreement with MCNP modeling. Based on these findings, the future application of Excalibur in its two configurations is discussed.

Fissile material detection↗
DOE OSTI2020

Transformational Challenge Reactor preconceptual core design studies

In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. Furthermore, the designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability. Additionally, the TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Computational modeling of graphite degradation in molten salt reactors: Role of infiltration

Molten salt reactors (MSRs) often employ graphite as a moderator and reflector. An important challenge for deploying graphite in these reactors is that, due to limited experimental data, our understanding of graphite’s structural integrity in molten salt environments remains incomplete. Here, this study addresses heat generation from fuel-bearing salt that has infiltrated open pores in the graphite, driven primarily by pressure differentials. This is one of multiple identified physical and chemical mechanisms through which molten salt could potentially degrade graphite. Thermally driven stresses are quantified using the Molten-Salt Reactor Experiment (MSRE) graphite moderator elements as a case study. Finite element simulations predict stress distributions at varying infiltration levels, indicating that thermal stresses increase with higher infiltration. Rare-event simulations using the parallel subset simulation framework identify the combinations and corresponding ranges of input parameters that lead to stresses above a specified threshold. In particular, combinations involving high infiltration amounts, high power density, and low thermal conductivity tend to induce the highest stresses. Under the inputs and assumptions considered in this work, the magnitudes of the thermally driven stresses are quite low, with a very low likelihood of causing failure due to exceeding the graphite’s tensile strength. Additionally, rare-event simulations were performed for two more scenarios: a scaled-up moderator geometry and a localized hotspot in the original geometry. Both cases resulted in increased susceptibility to failure, though not to a detrimental extent. Furthermore, the combined effects of irradiation and infiltration-induced thermal stresses were evaluated. The results showed that thermal stresses from infiltration were negligible compared to those caused by irradiation. The findings of such a study are inherently component-specific, but the methodology presented here could be used for similar assessments of salt-infiltration effects in other graphite components.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Differential localization patterns of pyruvate kinase isoforms in murine naïve, formative, and primed pluripotent states

Highlights: • PKM1/2 protein abundance is greater in formative mEpiLCs compared to naïve mESCs or primed mEpiSCs. The ratio of PKM1/2 is maintained. • PKM1/2 are localized in both nuclear and cytoplasmic regions relative to GAPDH and OCT4 across the pluripotent continuum. • PKM1 localization is strongly correlated to OCT4 localization, and moderately to GAPDH in formative mEpiLCs. • PKM1/2 localization is moderately correlated to OCT4 and GAPDH localization in naïve mESCs. • PKM1/2 localization are moderately correlated to GAPDH localization in primed mEpiSCs. Mouse embryonic stem cells (mESCs) and mouse epiblast stem cells (mEpiSCs) represent opposite ends of the pluripotency continuum, referred to as naïve and primed pluripotent states, respectively. These divergent pluripotent states differ in several ways, including growth factor requirements, transcription factor expression, DNA methylation patterns, and metabolic profiles. Naïve cells employ both glycolysis and oxidative phosphorylation (OXPHOS), whereas primed cells preferentially utilize aerobic glycolysis, a trait shared with cancer cells referred to as the Warburg Effect. Until recently, metabolism has been regarded as a by-product of cell fate, however, evidence now supports metabolism as being a driver of stem cell state and fate decisions. Pyruvate kinase muscle isoforms (PKM1 and PKM2) are important for generating and maintaining pluripotent stem cells (PSCs) and mediating the Warburg Effect. Both isoforms catalyze the final, rate limiting step of glycolysis, generating adenosine triphosphate and pyruvate, however, the precise role(s) of PKM1/2 in naïve and primed pluripotency is not well understood. The primary objective of this study was to characterize the cellular expression and localization patterns of PKM1 and PKM2 in mESCs, chemically transitioned epiblast-like cells (mEpiLCs) representing formative pluripotency, and mEpiSCs using immunoblotting and confocal microscopy. The results indicate that PKM1 and PKM2 are not only localized to the cytoplasm, but also accumulate in differential subnuclear regions of mESC, mEpiLCs, and mEpiSCs as determined by a quantitative confocal microscopy employing orthogonal projections and airyscan processing. Importantly, we discovered that the subnuclear localization of PKM1/2 changes during the transition from mESCs, mEpiLCs, and mEpiSCs. Finally, we have comprehensively validated the appropriateness and power of the Pearson's correlation coefficient and Manders's overlap coefficient for assessing nuclear and cytoplasmic protein colocalization in PSCs by immunofluorescence confocal microscopy. We propose that nuclear PKM1/2 may assist with distinct pluripotency state maintenance and lineage priming by non-canonical mechanisms. These results advance our understanding of the overall mechanisms controlling naïve, formative, and primed pluripotency.

60 APPLIED LIFE SCIENCES↗