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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Burnable absorbers in nuclear reactors – A review

Burnable absorbers can benefit nuclear reactors of virtually any design by providing reactivity control for extended fuel cycles, tritium production, burning of long-lived radionuclides, and reactor safety. When selecting the ideal burnable absorber type and its design, one must consider the resulting impact on the reactor’s fuel cycle design and cost, reactivity, thermal hydraulics, manufacturing, and radiation response. These selection criteria, as well as neutronic and thermophysical material property requirements, may be vastly different depending on whether the burnable absorber is intended for use in a commercial water-cooled reactor, a research reactor, or a next-generation advanced reactor system. A recent integration, synthesis, description of past and present technologies, and identification of existing gaps and areas of future research is lacking on these important topics. Here, this paper includes a fundamental description of the use of burnable absorbers and their impacts on reactivity, absorber depletion, self-shielding, basic thermophysical properties, and the use of burnable absorbers in next-generation nuclear applications..

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Tristructural isotropic (TRISO)-coated fuel particles are designed for use in high-temperature gas-cooled nuclear reactors, featuring a structural SiC layer that may be exposed to oxygen-rich environments over 1000 °C. Surrogate TRISO particles were tested in 0.2–20 kPa O 2 atmospheres to observe the differences in oxidation behavior. Oxide growth mechanisms remained consistent from 1200–1600 °C for each P O$_2$ , with activation energies of 228 ± 7 kJ/mol for 20 kPa O 2 and 188 ± 8 kJ/mol for 0.2 kPa O 2 . At 1600 °C, kinetic analysis revealed a change in oxide growth mechanisms between 0.2 and 6 kPa O2. In 0.2 kPa O 2 , oxidation produced raised oxide nodules on pockets with nanocrystalline SiC. Oxidation mechanisms were determined using Atom probe tomography. Active SiC oxidation occurred in C-rich grain boundaries with low P O$_2$ , leading to SiO 2 buildup in porous nodules. Here, this phenomenon was not observed at any temperature in 20 kPa O 2 environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Understanding Platinum Ionomer Interface Properties of Polymer Electrolyte Fuel Cells

A well-designed cathode catalyst layer with optimal ionomer distribution is critical to minimizing amount of Platinum (Pt) content in polymer electrolyte fuel cells (PEFCs). The impact of Pt loading, ionomer content and carbon support types on the catalyst/ionomer interface were also investigated at dry and wet conditions. Higher Pt loadings resulted in higher double layer capacity (C dl ) and similar electrochemical surface area (ECSA) due to well dispersed ionic phase material. Higher ionomer content resulted in higher ionic conductivity but also showed similar SO 3 − group coverage. High surface area (HSA) carbon support had larger ECSA and C dl at both dry and wet conditions, as less agglomerated Pt was well dispersed in the meso-pores of the support. Lower SO 3 − group coverages were observed for HSA carbon than for Vulcan carbon due to Pt particles being buried within the porous HSA carbon support. The effect of cell conditioning and voltage recovery on the PEFC cathode catalyst layer was shown to have minimal impact on SO 3 − group coverage despite a decrease in C dl and ECSA due to the size increase of Pt particles. At dry condition, a significant increases in SO 3 − group coverage were observed for all MEAs due to higher adsorptivity of ionomer in dry conditions.

25 ENERGY STORAGE↗
DOE OSTI2020

Advanced fuels from ethanol – a superstructure optimization approach

We develop a superstructure framework for the design of biorefineries for ethanol upgrading into advanced biofuels to replace gasoline, jet fuel or diesel. The framework integrates catalysis, process synthesis, and fuel property modelling towards the design of biorefineries producing fuels with specified properties. The proposed framework is applied to identify strategies for the upgrading of ethanol into one or more fuels with specific properties. We discuss the trade-off between profit and biorefinery complexity, as well as the relation among fuel property constraints, the optimal upgrading strategy selected, and process economics. Lastly, we show how to find the optimal biorefinery associated with a particular chemistry or catalyst. The results presented constitute the first systematic study of ethanol upgrading considering, simultaneously, fuel and process design.

09 BIOMASS FUELS↗
DOE OSTI2024

Credible Criticality Safety Margin in the 30B Package with LEU+ UF 6 and Hypothetical Water Ingress

The commercial nuclear industry is pursuing advancements in fuel and reactor design that increase the uranium enrichment above 5 wt. % 235 U. These advancements will necessitate the ability to transport bulk quantities of UF 6 at increased enrichments. Currently, the 30B cylinder is the primary container used by the industry for UF 6 storage and transportation and has a long history of successful shipments. This container can support up to 2,277 kg of UF 6 at a maximum enrichment of 5 wt. % 235 U. Previous evaluations have assessed the potential impact of criticality safety for 30B transport at higher enrichments but assumed moderator intrusion would not require evaluation. Although current regulations allow for the exception of moderator intrusion for UF 6 packages through the design and quality control of the package content, this exception is limited to enrichments up to 5 wt. % 235 U. Thus, an investigation of moderator intrusion into a 30B cylinder should be performed. Moderator intrusion into a 30B cylinder is a unique condition for criticality safety evaluation in transportation because of the violent chemical reactions that occur between UF 6 and H 2 O. The resulting intrusion is strongly dependent on the breach size, breach location, breach interface solid/ullage of the UF 6 content, the UF 6 distribution (which is temperature dependent), and temperature/pressure conditions which are dynamic in accident conditions. Additionally, the complexity of the HF-UO 2 F 2 -H 2 O interface that occurs during the event can influence the potential solubility of uranium in the system, as well as influence the amount of UF 6 reaction with H 2 O. With the operating experience from Orano Federal Services LLC, the UF 6 chemical expertise from Oak Ridge National Laboratory (ORNL), and the criticality safety expertise from ORNL, this paper evaluates the neutronic conditions (i.e., k eff ) that apply the understood chemistry and experimental conditions that occur during moderator intrusion of a breached 30B cylinder under postulated accidents. This report examines the historical evaluations of UF 6 transport and expands these evaluations for the enrichments expected for nuclear industry advancement. These simulations primarily feature a homogeneous mixture of UF 6 and H 2 O as an infinite media system and as a sphere with water reflection given an impurity limit of 0.5 wt. % UF 6 . This report demonstrates that this H/U limit is valid up to 8 wt. % enrichment for both H 2 O and HF as the moderating mixture. The water-reflected homogeneous mixed spheres evaluated demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. Additionally, this report evaluates moderator ingress scenarios of a 30B cylinder. The 30B cylinder simulations vary the cylinder orientation, the mass of UF 6 in the system, the mass of H 2 O in the system, and how much H 2 O has reacted with UF 6 . The moderator ingress of a 30B cylinder was evaluated for the water-reflected homogeneous mixed spheres to demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. These simulations are standard practice and are independent of the UF 6 -H 2 O reaction. This work also explored more complex simulations that layer the H 2 O over the UF 6 , which incorporates the amount of reacted UF 6 as a separate layer and considers solubility limits of UO 2 F 2 in H 2 O as it is produced. This approach is intended to simulate the postulated event of a large hairline crack occurring on a submerged cylinder. Finally, this work considers the potential of H 2 O mixing into the UF 6 solid heterogeneously using a sponge-like model. The water ingress into the UF 6 is treated as random size spheres, and the reaction products form into a layer over the spheres.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mitigating Electronic Conduction in Ceria‐Based Electrolytes via External Structure Design

Doped ceria electrolytes are the state of the art low‐temperature solid oxide electrolytes because of their high ionic conductivity and good material compatibility. However, cerium tends to reduce once exposed to reducing environments, leading to an increase in electronic conduction and a decrease in efficiency. Here, the leakage current is mitigated in ceria‐based electrolytes by controlling the defect chemistry through an engineered cathode side microstructure. This functional layer effectively addresses the problematic electronic conduction issue in ceria‐based electrolytes without adding significant ohmic resistance and increases the ionic transference number to over 0.93 in a thin 20 µm ceria‐based electrolyte at 500 °C, compared to a of 0.8 for an unmodified one. Based on this design, solid oxide fuel cells (SOFCs) are further demonstrated with the remarkable peak power density of 550 mW at 500 °C and excellent stability for over 2000 h. This approach enables a potential breakthrough in the development of ceria‐based low‐temperature solid oxide electrolytes.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Enhancing Direct Electrochemical CO 2 Electrolysis by Introducing A-Site Deficiency for the Dual-Phase Pr(Ca)Fe(Ni)O 3-δ Cathode

High-temperature CO 2 electrolysis via solid oxide electrolysis cells (CO 2 –SOECs) has drawn special attention due to the high energy convention efficiency, fast electrode kinetics, and great potential in carbon cycling. However, the development of cathode materials with high catalytic activity and chemical stability for pure CO 2 electrolysis is still a great challenge. In this work, A-site cation deficient dual-phase material, namely (Pr 0.4 Ca 0.6 ) x Fe 0.8 Ni 0.2 O 3-δ (PCFN, x = 1, 0.95, and 0.9), has been designed as the fuel electrode for a pure CO 2 –SOEC, which presents superior electrochemical performance. Among all these compositions, (Pr 0.4 Ca 0.6 ) 0.95 Fe 0.8 Ni 0.2 O 3-δ (PCFN95) exhibited the lowest polarization resistance of 0.458 Ω cm 2 at open-circuit voltage and 800 °C. The application of PCFN95 as the cathode in a single cell yields an impressive electrolysis current density of 1.76 A cm -2 at 1.5 V and 800 °C, which is 76% higher than that of single cells with stoichiometric Pr 0.4 Ca 0.6 Fe 0.8 Ni 0.2 O 3-δ (PCFN100) cathode. The effects of A-site deficiency on materials' phase structure and physicochemical properties are also systematically investigated. Such an enhancement in electrochemical performance is attributed to the promotion of effective CO 2 adsorption, as well as the improved electrode kinetics resulting from the A-site deficiency.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2023

Long-term sinking of nuclear waste canisters in salt formations by low-stress creep at high temperature

Abstract Rock salt has a self-sealing capacity, low permeability, and high thermal conductivity, making it a potential host for heat-generating nuclear waste. The feasibility of nuclear waste disposal within salt formations has been investigated mostly for small-sized canisters. Geologic disposal of larger-sized canisters originally designed for spent fuel storage and transportation has lately been examined as a cost-effective alternative. This raises questions about their long-term vertical movement due to their weight and high decay heat. Low-stress creep governs this movement; however, most salt constitutive models do not incorporate it. In this paper, the Norton and the WIPP creep models are compared with the Lux/Wolters/Lerche (LWL) model and a simpler model that combines linear and Norton creep laws (named combined creep model). The LWL and combined creep models consider pressure solution creep, though all incorporate dislocation creep. The models are first applied to creep tests under various stress levels. The LWL and the combined creep models results fit the experimental data well in both high and low stress ranges, whereas the Norton and WIPP models results only fit in higher-stress ranges. The different models are further applied for analyzing long-term canister movement. A sinking rate of $$-4.4\times 10^{-7}$$ - 4.4 × 10 - 7 mm/year was predicted using the Norton and WIPP models versus $$-2.1\times 10^{-2}$$ - 2.1 × 10 - 2 mm/year and $$-3.1\times 10^{-2}$$ - 3.1 × 10 - 2 mm/year using the LWL and the combined creep models, respectively. This comparative study confirms that creep models calibrated exclusively against high-deviatoric stress data might result in an inaccurate estimation of waste packages sinking rate in salt formations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Nondestructive measurements of residual 235 U mass of Israeli Research Reactor-1 fuel using the Advanced Experimental Fuel Counter

In 2018, a measurement campaign took place with participants from Los Alamos National Laboratory (LANL), the Nuclear Research Centre-Negev (NRCN) and Soreq Nuclear Research Center (SNRC) at the Israeli Research Reactor-1 (IRR-1) in which 14 of the reactor’s used fuel assemblies (FAs) with varied amount of depletion were measured with the nondestructive assay instrument Advanced Experimental Fuel Counter (AEFC). Designed for safeguards purposes, the AEFC measures both neutrons emitted from the FA (passive neutrons) and fission neutrons induced by an external neutron source (in this experiment, 252 Cf). Signals recorded with the AEFC include total neutron count rates (Singles), time-correlated neutron count rates (Doubles), and total gamma-ray count rates. The 235 U content of the FAs was previously assessed by two independent methods: (1) measurement of the transparency of the FA to low-energy gamma rays from an activated rhenium source (rhenium gamma transmission, or the RGT method) and (2) calculation of the 30-year burnup history of the core using detailed three-dimensional Monte-Carlo core depletion calculations. The results from the FAs that had been measured via the RGT method were used to construct the calibration curves, which translate the AEFC count rates to 235 U mass. Then, the calibration was evaluated using AEFC measurements of six additional FAs that were not measured via the RGT method. From the results, it was determined the Doubles calibration curve was more reliable than that of the Singles and follows a simple second-order polynomial fit for the whole range of residual 235 U mass content, albeit with larger statistical uncertainty. Detailed uncertainties quantification was conducted for both the AEFC Singles and Doubles. This includes the analysis of statistical uncertainties, calibration uncertainty, and random uncertainties due to the sensitivity of the AEFC to several sources of uncertainty, namely the FA position, FA orientation, interrogation source position, and ambient pool temperature. Finally, an overall total uncertainty of 6 g of 235 U is estimated for the Singles and Doubles, which is mainly due to calibration uncertainty (for the Singles) and statistical uncertainty (for the Doubles), and which constitutes 3%–6% of the 235 U total mass in the FAs, depending on their level of depletion.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Thermomechanical analysis and modeling of involute-shaped fuel plates using the Cheverton–Kelley experiments for the High Flux Isotope Reactor

Three research reactors with involute-shaped fuel plates are pursuing conversion from highly enriched uranium to low-enriched uranium fuel. Various core design and safety evaluation studies are essential to assess the feasibility of the conversion. The use of 3D computational multiphysics codes is being explored in these analyses and therefore they must undergo a thorough evaluation and quality assurance process due to their potential impact on nuclear safety. Here, the Cheverton and Kelley physical tests performed in the late 1960s to investigate the deflections of HFIR’s outer plate under uniform pressure and temperature fields are simulated by employing commercially available computational codes, with the goals to (1) verify and validate the models and numerical solvers implemented in the codes for thermomechanical analysis of involute reactor plates and (2) to develop a benchmark computational test to evaluate future versions of existing software or newly developed computational codes. The results of the simulations showed good agreement with each other as well as against the Cheverton–Kelley experimental data. Some minor deviations were observed for a few multiphysics cases and their potential origins and impact on the analysis results is investigated in the paper. The validated models increase the confidence in using multiphysics codes to evaluate existing or new LEU designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

A scalable membrane electrode assembly architecture for efficient electrochemical conversion of CO2 to formic acid

Abstract The electrochemical reduction of carbon dioxide to formic acid is a promising pathway to improve CO 2 utilization and has potential applications as a hydrogen storage medium. In this work, a zero-gap membrane electrode assembly architecture is developed for the direct electrochemical synthesis of formic acid from carbon dioxide. The key technological advancement is a perforated cation exchange membrane, which, when utilized in a forward bias bipolar membrane configuration, allows formic acid generated at the membrane interface to exit through the anode flow field at concentrations up to 0.25 M. Having no additional interlayer components between the anode and cathode this concept is positioned to leverage currently available materials and stack designs ubiquitous in fuel cell and H 2 electrolysis, enabling a more rapid transition to scale and commercialization. The perforated cation exchange membrane configuration can achieve >75% Faradaic efficiency to formic acid at <2 V and 300 mA/cm 2 in a 25 cm 2 cell. More critically, a 55-hour stability test at 200 mA/cm 2 shows stable Faradaic efficiency and cell voltage. Technoeconomic analysis is utilized to illustrate a path towards achieving cost parity with current formic acid production methods.

08 HYDROGEN↗
DOE OSTI2020

Inflationary stimulated Raman scattering in shock-ignition plasmas

In the shock-ignition inertial confinement fusion scheme, high-intensity lasers propagate through an inhomogeneous coronal plasma, driving a shock designed to cause fuel ignition. During the high-intensity ignitor laser pulse, in the long scale length coronal plasma, back-scattered stimulated Raman scattering (SRS) is likely to be in the kinetic regime. In this work, we use one-dimensional particle-in-cell simulations to show that there is a non-linear frequency shift caused by kinetic effects, resulting in the growth of SRS in an inhomogeneous plasma far exceeding the predictions of fluid theory, the so-called inflationary SRS or iSRS. We find that iSRS occurs over a wide range of density scale lengths relevant to shock-ignition and other directly-driven inertial confinement fusion schemes. The presence of iSRS in shock-ignition plasmas has implications for the theoretical gains from shock-ignition inertial confinement fusion. Here, we quantify the intensity threshold for the onset of iSRS for shock-ignition relevant parameters.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Methane storage in flexible and dynamical metal–organic frameworks

Recently, there has been significant interest in methane as an abundant and cleaner burning alternative to fossil fuels. Consequently, the design of media capable of the storage of methane under practical conditions has become an area of significant interest to the scientific community. While metal−organic frameworks have seen pronounced examination for this application, flexible metal−organic framework variants have been little examined despite having tremendous promise for methane storage applications. This work provides an overview of the current state of the art regarding the investigation of these systems for the purpose of providing a baseline for future research.

Forrest, Katherine A.↗
DOE OSTI2022

A mechanism for reduced compression in indirectly driven layered capsule implosions

High-yield implosions on the National Ignition Facility rely on maintaining low entropy in the deuterium–tritium fuel, quantified by its adiabat, in order to efficiently couple energy to the hot spot through high compression of the fuel layer. We present very-high-resolution xRAGE simulation results that study the impacts of interfacial mixing and the jetting of materials due to surface defects, defects on internal interfaces, voids, and engineering features on fuel layer compression. Defects and voids are typically neglected in implosion simulations due to their small size and three-dimensional geometry. Our results showed that supersonic jets of material arise through weak spots in the shell at peak implosion velocity that prevent uniform compression of the fuel layer even when they do not introduce contaminant into the hot spot. This occurs despite maintaining low fuel entropy, since the formation of the weak spots involves nonradial displacement of fuel mass. In contrast, simulations show that fuel–ablator mixing due to interfacial instabilities has a much smaller impact on compression. We show that defects on interior interfaces of plastic capsules decrease compression by 15% to 25% and interfacial mixing between the ablator and fuel decreases compression by less than 1% for implosions with plastic or high-density carbon (HDC) ablators. For low adiabat implosions, the impact of jetting seeded by the support tent can also decrease the compression by 25%. We demonstrate that the inclusion of interior defects in simulations can explain the inferred compression in two fielded plastic capsule implosions and that the inclusion of voids, for which available characterization has large uncertainties, in simulations of HDC capsule implosions has a qualitatively consistent impact. This mechanism offers a potential explanation for persistently overestimated fuel compression in design simulations of layered implosions on the National Ignition Facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Forward Modeling of Gamma Reaction History Signatures From Anticipated Deuterium-Tritium Filled MagLIF Implosions on Sandia’s Z-Machine

Nuclear reaction history measurements provide a bang time and burn width of Inertial Confinement Fusion (ICF) implosions and are essential for understanding implosion performance to constrain ICF capsule design. When fusion fuel contains Deuterium (D) and Tritium (T) gas, reaction history is informed by measuring the 16.75 MeV gamma rays generated from the D(T,γ) 5 He reaction. Such DT based reaction history measurements have not been made on the Magnetized Laser Inertial Fusion (MagLIF) platform on Sandia’s Z-Machine due to the lack of Tritium being used. The recent development of ICF implosions with tritiated fuel will open the possibility of measuring the gamma reaction history on the Z-Machine. A forward model of the Gamma Reaction History diagnostic on Z (GRH-Z) has been developed using the MCNP6.3 (Monte-Carlo N-Particle) radiation transport code. The model included the Z-Machine geometry of interest to characterize the impact of neutron induced gamma rays on the DT reaction history measurements. In addition, the impulse response functions of the GRH-Z diagnostic to understand the temporal response of the detector, and the minimum yields required to make a reaction history measurement were calculated. This approach also predicted that with T 2 gas doping of MagLIF implosions a reaction history may be made for high performance shots >8e12-2.4e13 depending on the chosen threshold for the detector, with a maximum signal to background ratio of 25%. It was found that for long duration ICF implosions that additional collimation will be needed to prevent the neutron induced gamma rays from modifying the shape of the measured DT reaction history curve.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Improvement and Verification of Online Cross Section Generation Capability of Griffin for TRISO-fueled Reactors

Griffin, a MOOSE-based reactor multiphysics code jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE Office of Nuclear Energy’s NEAMS program, has pursued the development of an online multigroup cross section generation capability for a few years to enable high-fidelity, problem-dependent neutronics analyses of advanced thermal reactors. Recent advancements in Griffin’s online multigroup cross section generation capability have significantly improved the accuracy, robustness, and efficiency of self-shielding calculations for both prismatic and pebble-bed TRISO-fueled reactor applications. Key developments include a unified fuel self-shielding method applicable to both TRISO and annular compact/spherical shell fuel zone geometries; an advanced Dancoff Category-based Equivalence Theory using a bell function for non-fuel resonance treatment, achieving more than an order-of-magnitude speedup compared to the Tone method; an on-the-fly multigroup equivalence approach to mitigate group condensation errors; and a streaming correction method for pebble-bed homogenization. A proof-of-concept demonstration of on-the-fly group condensation with consistent P0 transport correction was also achieved. The method reproduced direct fine-group solutions with excellent accuracy (eigenvalue errors within 10 pcm and pin-power differences within 0.5%), but due to performance limitations of the current fixed-source solver, improvements to solver efficiency will be addressed in future work. Verification tests were performed on graphite-moderated TRISO-fueled two-dimensional core benchmark problems representing gas-cooled microreactors, heat pipe-cooled microreactors, gas-cooled pebble-bed reactors, and fluoride salt-cooled high-temperature reactors. Across all cases, Griffin showed excellent agreement with Serpent2 continuous energy Monte Carlo solutions: eigenvalue errors within 200 pcm, pin-power root-mean-square errors within 2%, and control rod and drum worth errors less than 2%. It should be noted that, for the benchmark problem, cross section generation contributed less than 3% of the total simulation times. These results demonstrate that Griffin’s online cross section generation capability delivers accurate and efficient reactor physics solutions across a wide spectrum of TRISO-fueled advanced reactor designs. With further improvements to the fine-group fixed-source solver and planned extensions to depletion, transients, and coupled neutron–gamma transport, Griffin will be well-positioned to become a powerful and comprehensive tool for advanced reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Irradiation Performance: High-Temperature Gas Reactor Fuels

The response of TRISO fuel under irradiation is complex and dictated by the integrated nature of all components which define the TRISO fuel architecture. Potential failure mechanisms have been well understood and iterative improvements to the TRISO fuel architecture have been made to ensure optimized performance for different reactor designs. Ultimately, modern TRISO fuel for HTGR applications has exceeded performance targets.

Gerczak, Tyler↗