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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

High-speed particle analysis using forward and backward two-dimensional angular optical scattering

Measurement of two-dimensional angle-resolved optical scattering (TAOS) patterns is an attractive technique for detecting and characterizing micron-sized high-speed flying particles. The elastic-scattering intensity pattern from a single particle as a function of spherical coordinate angles θ and ϕ provides detailed information on the particle’s morphology. By use of an ellipsoidal reflector and a high-speed CCD camera, scattering from a CW-laser was detected and resolved over a large angular range (θ from 12° to 168° and ϕ from 0° to 360°) from spherically shaped particles (e.g., amorphous silica and borosilicate glass) flowing through the reflector’s focal plane at speeds of 20 to 30 m/s. The scattering pattern from individual particles flying through the focal plane of the ellipsoid mirror provides insights about a particle’s size, clustering, roughness, and velocity.

Optics and optical instruments↗
DOE OSTI2025

Final Design for Additional Thermal/Epithermal eXperiments (TEX) with Sodium Chloride Absorbers to Provide Validation Benchmarks for TerraPower

The first set of Thermal/Epithermal eXperiments (TEX) with chlorine absorbers (TEX-Cl) were executed in Q4FY24 and are in the process of being benchmarked for the ICSBEP. TEX-Cl builds upon the TEX-HEU baseline cases that were published in the 2022 International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. TEX-HEU, like TEX-Pu, was designed to be modular to allow for the incorporation of various absorbers and reflectors to test nuclear data and application case needs. For example, TEX-HEU with hafnium (TEX-Hf) utilizes hafnium plates as both absorbers and reflectors, depending on the tested configuration. A second set of chlorine experiments, dubbed More TEX-Cl, are laid out in this report to meet the needs of TerraPower for chlorine validation for their Molten Chloride Fast Reactor (MCFR) systems. TerraPower’s Molten Chloride Reactor Experiment (MCRE) and MCFR are fast molten salt reactors that utilize sodium chloride (NaCl) salt eutectics as the fuel and coolant. The MCRE eutectic is a mixture of NaCl and uranium trichloride (UCl 3 ). An abundant need for chlorine absorption validation has been expressed by multiple members of the community, including Y-12 (whose needs were addressed with the first set of experiments), LANL (whose needs were addressed with the Chlorine Worth Study (CWS)), TerraPower, institute de radioprotection et de sûreté nucléaire (IRSN), Savannah River Nuclear Solutions (SNRS), and others. Of the members who have expressed interest in this validation, most are interested in the fast neutron energy region, where the 35 Cl(n,p) reaction is most prominent. New 35 Cl(n,p) differential cross section measurements performed by LANL at LANCSE show substantial changes to the cross sections (Figure 1) and may be validated through these experiments as some configurations are optimally sensitive to this cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Frequency and time multiplexing for multiphoton state generation

One of the primary challenges with photonic quantum information processing is that single-photon states are created and heralded probabilistically, rather than being created on demand. One solution to this problem is multiplexing , which is attempting probabilistic generation of a single photon at multiple locations, times, or frequencies, and switching one successfully generated photon into an output mode. Motivated by the development of frequency-based photonic quantum information processing, we consider multiplexing using frequency and time as our options to use to get many photon generation opportunities. We devise an approach for generating multiphoton states, with photons populating multiple frequency modes in the same spatiotemporal mode. This method uses a variable-length optical delays to manipulate the temporal mode of the photons, and spaced fiber Bragg grating (FBG) reflectors to jointly manipulate the frequency and temporal modes of the photons. Experimental progress toward implementing this multiplexing scheme is proceeding along 2 fronts, each with different ways to achieve the variable-length optical delay. First, we will implement this scheme using a free-space optical quantum memory with multiple discretely adjustable free-space delays. For this implementation method, I calculate multiphoton generation rates, accounting for loss, that are realistically achievable with commercially available hardware. This work will appear in a theory paper, currently in preparation. Second, we will implement this multiplexing scheme in an integrated manner: the variable-length optical delay will happen on-chip using a long-lifetime Q-switchable Fabry-Perot cavity. The joint manipulation of the frequency and temporal modes of the photons will still happen off-chip in fiber with FBG reflectors. I report on experimental progress in constructing integrated Q-switchable Fabry-Perot cavities.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Flattop-HEU Benchmark Reevaluation Summary

The Flattop critical assembly was first constructed in the 1950’s at Los Alamos National Laboratory as a follow-on to the Topsy experiment. Flattop is composed of a sphere of special nuclear material (SNM) surrounded by a thick spherical reflector made of natural uranium (NU). Two SNM cores currently exist: a highly-enriched-uranium (HEU) core and a plutonium core. The reflector is composed of three parts: a stationary hemisphere and two movable quarter spheres. For fine reactivity control, there are three control rods of NU that are inserted into the stationary hemisphere from underneath the assembly. The final components that allow for reactivity adjustment are the glory hole pieces, mass adjustment buttons, and hemispherical caps. These pieces can be loaded in various configurations to change the available reactivity loaded in the system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Nuclear Requalification of the ATR Core

The Advanced Test Reactor (ATR) creates a unique, high neutron flux materials testing environment that subsequently causes neutron embrittlement damage to its beryllium reflector. The reflector is eventually required to be replaced during a reactor Outage. Concurrently, a piece-for-piece core replacement is conducted to maintain reactor functionality called a core internals changeout (CIC). Once changeout is complete, ATR Reactor Engineering utilizes previous CIC data, notably 1977-1994 CIC data, to conduct predictive criticality analysis on the new, unirradiated core. However, CIC data has been inconsistently reported over the years making it difficult, and in some cases impossible, to utilize. These inconsistencies arise in various forms of not using correct units, inconsistent diagram usage, and not portraying the same kind of data similarly across the reports. To work around this inconsistency issue, Reactor Engineering is implementing the use of several new neutronic modeling codes that utilize modern computation speeds and methods to model the new core in different criticality environments. Further validation is then conducted by nuclear testing and bringing the reactor critical and conducting various irradiation tests. Upon completion and certification that the reactor passed all testing requirements, ATR can once again re-start normal operations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Assessment of Measurement Uncertainties in the Jupiter High-240 Experiment

The Jupiter High-240 experiment performed in May of 2019 was previously discussed as a variant of the original Jupiter experiment incorporating plutonium metal alloy fuel plates with higher 240Pu content and lead plates, using both a reference configuration and a second configuration where eight lead plates were replaced with aluminum to simulate voiding. Measurements were recorded for experiment period, the “pressure” of the Comet ram upon closure for each near-critical measurement, and temperature. The experiment reactor period is the time it would take to increase the neutron population by a factor of e. For this experiment, the copper reflectors and upper third of the fuel sits upon a support structure with the lower fuel arrays raised up into the center of the reflectors using a ram (see Fig. 1). The recorded logbook temperature for each measurement corresponds to a resistance temperature detector (RTD) located at the top center of the upper fuel array. This paper summarizes the evaluated uncertainties for the Jupiter High 240 experiment as contributed via the recorded measurements and nuclear data and their assessed impact upon the computation of system reactivity and eigenvalue.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Progress Towards the Validation of a new RELAP5-3D model of the High Temperature Test Facility

Validation is a key step in the development of any type of systems model. As the next generation of reactors approaches, the need for codes that have been validated for these new types of systems continues to grow. An example of a prominent option is the Reactor Excursion Leak Analysis Program (RELAP5-3D), developed by Idaho National Laboratory. This code was developed for the purpose of systems level thermal-hydraulic modeling of light water reactors (LWRS) and postulated transients that can occur in LWRS.RELAP5-3D has been substantially validated against LWR data. Due to its long history as a reactor safety analysis tool, there has been an effort to adapt RELAP5-3D for the purposes of advanced reactor concepts such as prismatic high-temperature gas-cooled reactors (HTGRs). However, RELAP5-3D has not nearly been validated and verified for HTGRs to the degree of LWRs, warranting verification and validation opportunities with computational benchmarks and existing experimental facilities. Examples of such facilities include the modular high-temperature gas-cooled reactor (MHTGR) 350 and the high temperature engineering test reactor (HTTR) from Japan. The MHTGR 350 is a benchmark design concept for code-to-code verification purposes; therefore, it does not provide any experimental data for validation opportunities The HTTR provides useful multiphysics validation data but does not have the in-core instruments to generate thermal-hydraulic experimental data to help with RELAP5-3D validation. Consequently, a facility that could provide key in-core temperatures for thermal-hydraulic validation was still needed. The High Temperature Test Facility (HTTF) is an integral effects facility for HTGR thermal hydraulics developed and operated by Oregon State University. HTTF represents ¼ length scale of the General Atomics MHTGR and is rated for a total power of 2.2 MW. Axially, the core consists of an upper and lower reflector and 10 blocks, numbered from bottom to top (Block 1 is right above lower reflector). The core is heated via graphite resistive heater rods, with respective channels distributed throughout the core. The primary coolant is helium and heat can radiate out of the core to the reactor cavity cooling system (RCCS), which is cooled by water. The primary purpose of the facility is to investigate pressurized conduction cooldown (PCC) and depressurized conduction cooldown (DCC) transients, which are also referred to as the pressurized and depressurized loss of forced cooling respectively. Two experiments were chosen to perform the validation study with a RELAP5-3D model of HTTF. These experiments are PG-27 (PCC) and PG-29 (DCC). These were chosen based off of the quality of available experimental data before and during the experiment which led to their inclusion in the HTGR Thermal Hydraulics Benchmark.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Linewidth narrowing and wideband frequency modulation of a DBR laser

We present a scheme to phase-lock a 240 mW, 852 nm distributed Bragg reflector (DBR) laser to a fixed-frequency narrow-linewidth laser with a rapidly tunable frequency offset near 9 GHz. The phase-lock is accomplished by electronic feedback on the beatnote between the two lasers. The frequency offset can be swept 200 MHz in 300 us, limited by the feedback loop bandwidth, allowing for its use in complex cooling and state preparation schemes needed in atomic physics experiments. Here, we find that the phase-lock reduces the linewidth of the DBR laser below its natural linewidth of ~400 kHz to ~100 kHz

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Solar Energy Technical Publications Catalog: Solar Thermal Technology

The research and development described in these documents was conducted within the U.S . Department of Energy's (DOE) Solar Thermal Technology Program. The goal of this program is to advance the engineering and scientific understanding of solar thermal technology and to establish the technology base from which private industry can develop solar thermal power production options for introduction into the competitive energy market. Solar thermal technology concentrates the solar flux using tracking mirrors or lenses onto a receiver where the solar energy is absorbed as heat and converted into electricity or incorporated into products as process heat. The two primary solar thermal technologies, central receivers and distributed receivers, employ various point and line-focus optics to concentrate sunlight. Current central receiver systems use fields of heliostats (two-axis tracking mirrors) to focus the sun's radiant energy onto a single, tower- mounted receiver. Point focus concentrators up to 17 meters in diameter track the sun in two axes and use parabolic dish mirrors or Fresnel lenses to focus radiant energy onto a receiver. Troughs and bowls are line-focus tracking reflectors that concentrate sunlight onto receiver tubes along their focal lines. Concentrating collector modules can be used alone or in a multimodule system. The concentrated radiant energy absorbed by the solar thermal receiver is transported to the conversion process by a circulating work fluid. Receiver temperatures range from l00 degrees C in low-temperature troughs to over 1500 degrees C in dish and central receiver systems. The Solar Thermal Technology Program is directing efforts to advance and improve each system concept through solar thermal materials, components, and subsystems research and development and by testing and evaluation. These efforts are carried out with the technical direction of DOE and its network of field laboratories that works with private industry. Together they have established a comprehensive, goal-directed program to improve performance and provide technically proven options for eventual incorporation into the nation's energy supply. To successfully contribute to an adequate energy supply at reasonable cost, solar thermal energy must be economically competitive with a variety of other energy sources. The Solar Thermal Technology Program has developed components and system-level performance targets as quantitative program goals. These targets are used in planning research and development activities, measuring progress, assessing alternative technology options, and developing optimal components. These targets are pursued vigorously to ensure a successful program. This catalog represents part of an effort to provide information on publications about solar thermal research and development activities conducted by DOE's laboratories. Publications listed include technical and research reports and special publications. The following national laboratories are represented in this edition: Sandia National Laboratories, Solar Energy Research Institute, Jet Propulsion Laboratory. This catalog is a product of the DOE Solar Technical Information Program, which is dedicated to providing information to scientific and industrial users in ways most convenient and useful to them. This catalog focuses on solar thermal technologies, and its purpose is to keep the scientific and industrial communities informed of the latest developments in federally sponsored research in this technology.

140000* -- Solar Energy↗
DOE OSTI2026

Modeling a generic TRISO-fueled heat pipe microreactor using SCALE: Depletion, transportation criticality, and shielding

This paper demonstrates the applicability of the SCALE code system to tristructural-isotropic (TRISO)-fueled heat pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study conducted supports US Nuclear Regulatory Commission code readiness efforts for advanced non–light-water reactor technologies and is intended as a code capability demonstration rather than as an optimization of a specific microreactor design. The modeled reactor employs high-assay low-enriched uranium (HALEU) uranium oxycarbide (UCO) TRISO fuel and beryllium oxide (BeO) reflectors and operates at 7.5 MWth with a nominal lifetime of about 3 effective full power years. Representative cases for fresh and irradiated cores were selected to exercise SCALE methods relevant to reactor operation and post-irradiation transport. The discharged-core decay heat is approximately 6% of operating power immediately after shutdown. Transportation criticality calculations show that internal water ingress is the dominant reactivity effect, with fully flooded fresh core and irradiated core configurations remain above the subcriticality criterion, even with the available control mechanisms. Shielding calculations for a simplified transportation package indicate that normal-condition dose rates are governed mainly by shielding thickness and cooling time, whereas the breached hypothetical accident case is governed primarily by cooling time. Overall, the study shows that SCALE supports depletion, transportation criticality, and shielding evaluations efficiently for TRISO-fueled heat pipe microreactors within a single code system.

Criticality↗
DOE OSTI2026

Hydrogen uptake in graphite matrix at high temperature

Tritium management is a critical challenge for the next generation of nuclear reactors, such as Fluoride Salt Cooled High Temperature Reactors (FHRs) and High Temperature Gas-cooled Reactors (HTGRs), due to the higher production rate (up to 10,000 times) than conventional Light Water Reactors (LWRs). Graphitic materials employed as moderator, reflector, and fuel pebbles offer a potential pathway for tritium recovery by serving as a sink for tritium. Prediction of uptake capacity under reactor relevant conditions remains a challenge due to a lack of low partial pressure data and significant inter-grade variability of graphite. This study addresses these gaps by providing a comprehensive characterization of hydrogen (as a tritium surrogate) uptake and release behavior in the A3-3 graphite matrix (GM) used in fuel pebbles. Uptake measurements are performed at reactor relevant temperatures of 600- 800 °C, 1-200 Torr hydrogen pressure, and 15-120 min equilibration time, followed by thermal desorption spectroscopy up to 1100 °C. Uptake experiments at different equilibration times demonstrate the role of kinetics in hydrogen uptake, which can be modeled as a diffusion-with-trapping process. In the thermodynamics limit, the Sips adsorption model is shown to capture the uptake in A3-3 GM well. Our campaign provides a set of new results for hydrogen uptake in A3-3, including limiting uptake capacity at 600 °C, apparent diffusion coefficient at 600 °C, and the first estimates of the FHR/HTGR relevant (600 °C, 20 Pa partial pressure) equilibrium uptake capacity and time to saturation. Desorption data highlights a new site for hydrogen uptake, not observed in nuclear graphite, which we attribute to the non-graphitized binder. Using the Kissinger method, we estimate activation energy for release from the desorption peaks, confirming the activation energy for release from the basal planes and providing the first estimate for the activation energy of release from the binder.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Improving the modeling of near-wall interphase heat transfer in porous media models of Pebble Bed Reactors

Here, this work aims to improve capabilities for modeling localized effects in porous media models of Pebble Bed Reactors. The wall-channeling effect is the primary local phenomenon of interest in a PBR, where the presence of the reflector wall disrupts the pebble packing, causing the pebbles near the wall to pack less efficiently and creating large void regions. Accurate modeling of the near-wall region is important as it will affect core bypass flow and temperature predictions. Porous media models are commonly used for design scoping and plant-level simulations of PBRs. Although these models have some capabilities to model the near-wall region, the correlations that are available in porous media codes are often inaccurate when a multi-region model is used to discretize the near-wall region. This work employs a high-to-low analysis to study the accuracy of available interphase heat transfer closures. NekRS, a spectral element computational fluid dynamics code, is used to perform Large Eddy Simulations. These LES simulation results are compared to porous media model results from the Pronghorn porous media code. The friction term of the KTA drag closure is first improved, reducing the error in the prediction of the near-wall velocity from over 50% to less than 5%. This is combined with improvements to the form term from previous works to produce a drag closure that is capable of accurately modeling the wall-channeling effect across a variety of flow conditions. The Nusselt number predictions of several heat transfer correlations are compared to the high-fidelity results where it is found that the KTA heat transfer correlation is capable of accurately predicting the local Nusselt numbers that were determined in the high-fidelity simulation. Comparison of the radial solid temperature profiles, however, reveal discrepancies between NekRS and Pronghorn. It is discovered that the implementation of the interphase heat transfer coefficient that exists in many current porous media codes is not valid when local porosities are modeled. Instead, it is suggested that the interphase heat transfer coefficient should be dependent on the local porosity, the Nusselt number, and the local solid surface-to-volume ratio. Implementation of this change produces improvement in the agreement between the results obtained by NekRS and Pronghorn while using the KTA heat transfer correlation.

interphase heat transfer↗
DOE OSTI2025

Sensitivity analysis of thermal contact conductance modeling to inform MiniFuel irradiation capsule designs

The MiniFuel irradiation platform has been developed by Oak Ridge National Laboratory as a flexible, high-throughput separate effects testing capability within the High Flux Isotope Reactor (HFIR). Finite element thermal models are relied upon to design MiniFuel experiments to achieve a specific time-averaged irradiation temperature for experimental objectives. A previous study identified that uncertainty in the component heat generation rates and thermal contact conductance (TCC) model are the most significant contributors to predicted fuel temperature variance. To address both sources of uncertainty, this work performs sensitivity analysis on the TCC model to identify high-impact, high-uncertainty parameters that contribute to fuel temperature variance. The TCC model is analyzed in increasing detail, first using a standalone Python code, then again after coupling Python to the BISON fuel performance code. Furthermore, the parameters with the largest contributions to fuel temperature variance which can be reduced through design changes are identified as the initial subcapsule gas pressure, contact pressure between the fuel and dish, and the effective surface roughness of the interface. A set of design recommendations for future capsule designs has been established and applied to reduce the previously quantified average fuel temperature uncertainty ranges of ± 40 °C in the HFIR vertical experiment facilities (VXF) and ± 80 °C in the removable beryllium (RB) reflector to approximately ± 32 °C and ± 53 °C, respectively. This equates to a 21 % and 33 % reduction in the uncertainty range of the average fuel temperature for VXF and RB, respectively.

BISON↗
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 OSTI2026

Transient uncertainty quantification and Global Sensitivity Analysis of the open-source Molten Chloride Reactor Experiment (MCRE) using GP-PCA surrogate models

Uncertainties in the thermophysical properties of molten salts impact both the steady-state and transient behavior of Molten Salt Reactors (MSRs). In this work, we aim to quantify the influence of such uncertainties on the transient operation of the Molten Chloride Reactor Experiment (MCRE), utilizing the open-source specifications provided for this reactor. Seven representative transient scenarios are considered. For each scenario, we evaluate the impact of thermophysical property uncertainties on four key multiphysics model output variables of interest (VoIs): maximum power density, maximum fuel temperature, maximum reflector temperature, and average fuel velocity magnitude. In addition, we perform a Global Sensitivity Analysis (GSA) by computing Sobol’ indices for the uncertain input parameters to determine their contribution to the variability of each VoI. Conducting GSA is computationally intensive due to the large number of required evaluations of the high-fidelity multiphysics model. To mitigate this cost, we develop a surrogate modeling framework that combines Gaussian Process (GP) regression with Principal Component Analysis (PCA), enabling efficient sample generation for the GSA. Our results show that for energy-related VoIs, thermal conductivity is the dominant contributor to uncertainty. In contrast, for flow-related VoIs, density and dynamic viscosity are the primary sources of uncertainty. The specific heat of the fuel salt was found to play a secondary role in the transient analyses.

42 - ENGINEERING↗
DOE OSTI2026

Polariton Control of Molecular Charge Transfer in Perylene Diimide Semiconductors

We report the modulation of molecular charge transfer in a bay-substituted perylene diimide derivative embedded in a planar distributed Bragg reflector microcavity. Angle-resolved reflectance spectra confirm the formation of upper and lower polaritons with clear Rabi splitting, indicating strong coupling between the cavity mode and molecular excitons. Using broadband transient absorption spectroscopy, we compare the ultrafast dynamics of cavity and non-cavity films. While excited-state absorption and stimulated emission pathways remain largely unchanged, kinetic modeling reveals a moderate increase in the charge transfer rate and yield under strong coupling. This enhancement is attributed to a reduction in the effective driving force via the formation of the lower polariton, placing the system deeper into the Marcus inverted regime. Our results demonstrate a promising non-chemical method leveraging cavity quantum electrodynamics to modulate charge separation processes in molecular semiconductors.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Shock compression of FeOOH and implications for iron-water interactions in super-earth magma oceans

Iron(Fe)-water reactions in a magma ocean can influence water storage and density of planets. These reactions can form Fe-O-H phases, whose density, melting, and electronic properties at planetary interior conditions are important for informing planetary models. Here, we study natural goethite (α-FeOOH) that is shock-compressed along its principal Hugoniot. Analysis of our velocity interferometer system for any reflector (VISAR) results extends the equation of state to over 800 GPa. X-ray diffraction and VISAR reflectivity results indicate the onset of melting occurs at ~95 GPa with complete melting by 166 GPa, which may be relevant to low seismic velocity anomalies observed above the core-mantle boundary. Analysis of X-ray emission spectroscopy results up to 285 GPa shows the spin crossover of Fe, with dominantly low spin Fe above ~265 GPa in the melt, supporting formation of dense basal magma oceans in terrestrial planets. Using our measured FeOOH densities, we model planetary interiors up to 10 Earth masses. Assuming FeOOH forms via iron-water reactions, the radius decreases by up to 28%, while the density increases by up to 165% compared to the unreacted cases, providing an avenue to investigate water storage and evolution in super-Earths and sub-Neptunes.

Zhang, Yanyao [Stanford Univ., CA (United States)]↗
DOE OSTI2025

Bayesian calibration of irradiated graphite property models under high temperatures

Graphite under high temperatures and irradiation is central to advanced reactors. We develop a Bayesian calibration framework for graphite property models that explicitly represents model-data mismatch via a Gaussian-process discrepancy. The approach propagates uncertainty from parameters, experimental noise, and model form, with a hierarchical variance structure to capture group and cross-group noise. Using two predictive models across five grades (IG-110, NBG-18, PCEA, NBG-17, 2114) and four properties-irradiation-induced dimension change, creep, Young’s modulus change ratio, and coefficient of thermal expansion change ratio-we obtain average predictive-error reductions of 54%, 65%, 17%, and 17% when discrepancy is included. We illustrate engineering impact with a multiphysics model of a very-high-temperature reactor prismatic reflector brick, analyzing stresses under high fluence and temperature. Accounting for model discrepancy markedly improves predictive accuracy and provides a robust basis for reliable graphite component design in advanced reactors.

36 - MATERIALS SCIENCE↗