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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 1,747 records · Page 97
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Modeling and experimental validation of a W f /W-fabrication by chemical vapor deposition and infiltration

Tungsten (W) has a unique combination of excellent thermal properties, low sputter yield, low hydrogen retention, and acceptable activation. Therefore, W is presently the main candidate for the first wall material in future fusion devices. However, its intrinsic brittleness and its further embrittlement during operation bears the risk of a sudden and catastrophic component failure. As a countermeasure, tungsten fiber-reinforced tungsten (W f /W) with extrinsic toughening is being developed. A possible synthesis route is chemical vapor deposition (CVD) using heated W fabrics as substrate. The challenge is that the growing CVD-W can isolate domains from precursor access leading to strength-reducing pores. To deepen the process understanding and to optimize the CVD parameters, models were developed with COMSOL Multiphysics and validated experimentally. W deposition rate equations as function of the temperature and the partial pressures of the precursors H 2 and WF 6 were experimentally validated in previous work. In the present article, these equations are applied to obtain partial pressures within the CVD reactor. The results are taken as input for transient simulations in the microscale, in which W coatings, growing onto multiple adjacent W fibers, were simulated via mesh deformation and remeshing. The surface-to-surface contact of the W coatings and the corresponding potential pore formation were simulated by implementing sophisticated deposition rate stop conditions. Within the measuring uncertainties of ≃ ±1%, the models are validated successfully by experimental comparison regarding the deposition rate, pore structure, and relative densities ranging from 0.6 to 0.9.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

NSUF CoMET Project, Implementation Plan FY2019

The Combined Materials and Experiment Toolkit (CoMET) is an NSUF information management resource designed to give users access to the information and tools necessary for designing meaningful experiments and writing clear, structured proposals. NSUF has a milestone to deploy an experiment design expert system to the NSUF website to support applications for the FY2020 CINR, due on August 30, 2019. To meet this milestone, the CoMET project work for FY2019 focuses on three main objectives: 1.) Integrate the three NSUF web applications by linking data fields in one application with the data found in the others. The NSUF web applications are: The Rapid Turnaround Experiment (RTE) Proposal Solicitation system, the Nuclear Fuels and Materials Library (NFML) and the Nuclear Energy Infrastructure Database (NEID) and the researcher and technical staff account and profile management systems. 2.) Build an Experiment Design Wizard (expert system) that will guide users through the process of writing a statement of work for a CINR or RTE experiment. 3.) Build an Irradiation Effects Calculator to help a PI choose the best neutron irradiation facility for their needs, estimate the time needed to achieve their goals, and estimate the resulting radioactivity of their samples, post-irradiation. These objectives are designed to support the mission of the NSUF and optimize the experiment design portion of the proposal processes. The new researcher database will make finding collaborators easier. The wizard will provide the PI with early NSUF expertise so that their proposed experiments, at the “Letter of Intent” and “Pre-Application” stages are better designed to work within the available NSUF facilities. Utilization of these tools should reduce the proportion of proposals that are judged to be infeasible or “not ready” for NSUF access. It should also improve the efficiency of the NSUF technical leads since the PI will be starting with a better statement of work for the project. The tools will also provide resources for NSUF staff to more efficiently screen and score incoming proposals. This implementation plan will layout the work breakdown structure, activities, resources required, and plan for completion of the activities for CoMET in FY2019.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Evaluation of the Reconstruction Accuracy of the Ultrasound Model–Based Image Reconstruction (U-MBIR) Method for Concrete to Optimize Damage Detection

Reinforced concrete is a critical structural material used to construct nuclear power plants (NPPs). As such, its safety and performance must be thoroughly examined throughout the life cycle of the NPP infrastructure system. Ultrasonic measurements have been an industry standard for both surface and subsurface inspections. To support the development of these techniques, Oak Ridge National Laboratory (ORNL) has researched and developed advanced image reconstruction algorithms to capture internal damage. The results and discussion presented herein summarize the current state of the ultrasonic model–based iterative reconstruction (U-MBIR) algorithm developed at ORNL. In the work documented in this report, the U-MBIR methodology was applied to four sets of ultrasonic data collected from concrete specimens. The results demonstrate that the U-MBIR algorithm can successfully detect defects within the four concrete samples. The reconstruction images help identify the specimen thickness, regions of delamination, and location of rebar embedded within the concrete. The reconstruction images allow engineers and technicians to characterize the internal defects within concrete specimens and structural members. Ultimately, this knowledge can guide engineers in making informed decisions regarding the performance, safety, and reliability of structural materials (i.e., reinforced concrete) throughout the life cycle of NPPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Neutron capture measurements at UMass lowell research reactor

A new facility was designed around the thermal column beam port of the 1 MW Research Reactor at the University of Massachusetts Lowell. Thermal neutrons are collimated to a 1-inch diameter beam and incident on samples to induce the radiative neutron capture. New measurements of capture gamma rays are planned for Mn, Cu, Ni, Cr, and Gd samples in the next few years. The experiments will be carried out in close collaboration with the ENSDF (Evaluated Nuclear Structure Data File) evaluation group at Brookhaven National Laboratory. The gamma rays are measured using an array of high-resolution HPGe detectors. Some of the HPGe detectors use active Compton shields that were designed from scintillation detectors and improve significantly the signal-to-background in the measured gamma-ray spectra with HPGe. The experimental spectra will be validated by the GEANT4 simulations of the array and theoretical models of the emission of gamma rays from the compound nucleus. In conclusion, the new experimental results on the capture gamma ray intensities will be incorporated in future ENSDF evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Elucidating the corrosion mechanism of commercial Ni-based superalloys in UCl3 containing-chloride molten salt systems

Elucidating the role of UCl3 in the corrosion mechanism of Ni-based superalloys exposed to chloride molten salts Trishelle Copeland-Johnson1, Michael Woods1, Ruchi Gakhar1, Daniel J. Murray1, Guoping Cao1, Lingfeng He1 1Idaho National Laboratory, Idaho Falls, ID, United States The United States Department of Energy (DOE) aims to diversify the domestic energy portfolio towards more sustainable options, including implementation of molten salt reactor (MSR) technology. Chloride molten salts have been investigated as an appropriate MSR coolant and fuel because their relatively inexpensive, abundant, and exhibit favorable thermophysical properties. However, the corrosivity of chloride molten salts have not been extensively studied, especially with the inclusion of actinide products, such as UCl3. Accordingly, the development of nuclear structural materials with excellent corrosion performance is critical to the successful implementation of MSRs, particularly from a mechanistic perspective. In this investigation, we attempt to elucidate the interfacial corrosion mechanism between Ni-based structural materials, such as Inconel 617, and UCl3-containing salt systems through a multi-modal advanced characterization approach, including electron microscopy techniques. The findings from this investigation will expand the knowledgebase of chloride molten salt corrosion of MSR structural materials for strategic property-to-performance design.

36 - MATERIALS SCIENCE↗
DOE OSTI2018

An Assessment of MELCOR 2.1: Containment Thermal-Hydraulic Tests in the Heissdampfreaktor (HDR) Facility

MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. It provides a capability for independently auditing analyses submitted by reactor manufacturers and utilities. In order to assess the adequacy of containment thermal - hydraulic modeling incorporated in the MELCOR code, a key containment test facility was analyzed. This report documents MELCOR code calculations for simulating steam - water blowdown tests performed in the Heissdampfreaktor ( HDR) de-commissioned containment facility located near Frankfurt , Germany . These tests are a series of blowdown experiments in a large scaled test facility ; including some tests with the addition of hydrogen release which are intended to simulate a variety of postulated break s inside large containment buildings. The key objectives of this MELCOR assessment are to study: (1) the expansion and transport of high energy steam - water releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment gas mixing and stratification. Moreover, MELCOR results are compared to the CONTAIN code for the same tests.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Scalable Synthesis of Ti 3 C 2 T x MXene

Scaling the production of synthetic 2D materials to industrial quantities has faced significant challenges due to synthesis bottlenecks whereby few have been produced in large volumes. These challenges typically stem from bottom‐up approaches limiting the production to the substrate size or precursor availability for chemical synthesis and/or exfoliation. In contrast, MXenes, a large class of 2D transition metal carbides and/or nitrides, are produced via a top‐down synthesis approach. The selective wet etching process does not have similar synthesis constraints as some other 2D materials. The reaction occurs in the whole volume; therefore, the process can be readily scaled with reactor volume. Herein, the synthesis of 2D titanium carbide MXene (Ti 3 C 2 T x ) is studied in two batch sizes, 1 and 50 g, to determine if large‐volume synthesis affects the resultant structure or composition of MXene flakes. Characterization of the morphology and properties of the produced MXene using scanning electron microscopy, X‐ray diffraction, dynamic light scattering, Raman spectroscopy, X‐ray photoelectron spectroscopy, UV–visible spectroscopy, and conductivity measurements show that the materials produced in both batch sizes are essentially identical. This illustrates that MXenes experience no change in structure or properties when scaling synthesis, making them viable for further scale‐up and commercialization.

Shuck, Christopher E.↗
DOE OSTI2018

Dimensional Stability of the Structural Parts of VVER-1000 FA Made of E-635 Zirconium Alloy

The dependence of dimensional changes occurring in the structural parts made of E-635 zirconium alloy on the metallurgical and operational factors as a result of thermal creep, radiation growth, and radiation creep was studied. Tests close to operating conditions showed high resistance to radiation growth and thermal and radiation creep of the structural parts of VVER-1000 fuel assemblies made of E-635 zirconium alloy. Parts with greater resistance to creep without irradiation also possess high resistance to creep in the presence of irradiation. However, in-reactor creep can be almost 10 times greater than creep with no irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

An experimental and kinetic modeling study of the oxidation of hexane isomers: Developing consistent reaction rate rules for alkanes

Alkanes are key components in gasoline, jet and diesel fuels and considerably influence the combustion behavior of these fuels because of their wide range of reactivity. An improved understanding of their combustion behavior and the development of chemical kinetic models that can accurately simulate their combustion behavior are important for the development of next-generation internal-combustion and gas-turbine engines. The current work provides here improved insight into oxidation mechanisms of a representative family of hydrocarbon fuels, specifically the hexane isomers: n-hexane, 2-methylpentane, 3-methylpentane, 2,2-dimethylbutane and 2,3-dimethylbutane. These isomers provide carbon “skeletons” ranging from straight-chained to highly-branched and provide a framework for the subsequent development of kinetic mechanisms for larger alkanes. New ignition delay times for the four branched hexane isomers were measured in a high-pressure shock tube and in a rapid compression machine, both at stoichiometric conditions (φ = 1), p = 15 bar and X O2 = 21% over temperatures from 600 to1300 K. These data were combined with previously published measurements under the same conditions for the remaining n-hexane isomer to provide a complete body of experimental data for kinetic modeling analysis. In addition, very recent experimental measurements of individual intermediate chemical species concentrations from all five hexane isomers in a jet-stirred reactor are also included and provide another family of data for further assessment of hexane isomer reactivity. Different reactivities were observed for each hexane isomer in each experimental facility, resulting from differences in their molecular structures. Consistent reaction rate rules have been applied to develop a combined detailed chemical kinetic model for all five hexane isomers. Kinetic model validation studies are reported to show that the current model reproduces well the ignition delay times of all five alkane isomers, as well as their variations in reactivity over a wide range of temperatures and other operating conditions. Equally important, these results show that it is not necessary to have a separate, different kinetic model for each isomer of a family of alkane fuels and that a single, coherent, integrated set of reaction rate classes and rules is sufficient to accurately describe combustion rates of combustion of straight-chain n-alkanes and branched-chain alkane fuels. This suggests strongly that a single set of reaction classes and rate rules should be sufficient to describe combustion kinetics of alkane fuels of any size and degree of branching.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Electric-field-assisted-sintering of rare-earth oxide dispersion strengthened Fe-Cr-Mo alloys

Oxide dispersion-strengthened (ODS) alloys are widely recognized for their exceptional high-temperature strength, creep resistance, and radiation tolerance, making them indispensable for advanced nuclear reactors, aerospace, and energy systems. Achieving a fine and stable dispersion of oxide nanoparticles is critical, as these particles act as strong barriers to dislocation motion and effective sinks for irradiation-induced defects, ensuring structural integrity under extreme conditions. Here, in this study, Fe–Cr–Mo-based ODS alloys were fabricated via mechanical alloying and consolidated using electric-field-assisted sintering (EFAS) with additions of Y 2 O 3 , La 2 O 3 , and CeO 2 . EFAS processing produced ultrafine-grained microstructures (average grain size <1 μm) with uniformly distributed oxide clusters (2–4 nm). Atom probe tomography revealed that La 2 O 3 -containing alloys exhibited the highest nanoparticle number density, resulting in superior tensile strength compared to yttria- and ceria-bearing counterparts. The combined effect of grain refinement and rare-earth oxide dispersion significantly enhanced mechanical performance, demonstrating the potential of EFAS for developing high-strength ferritic alloys for demanding environments such as nuclear systems.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

The Use of Microelectrodes in Molten Salt Electrochemistry

Molten salts have attracted considerable interest as essential media for advanced high-temperature technologies, including molten salt reactors, thermal energy storage, high-temperature electrolysis, and pyrochemical processing. Their ability to remain stable in liquid form at elevated temperatures, combined with favorable thermophysical properties and wide electrochemical windows, makes them highly suitable for applications involving heat transfer, energy storage, and hightemperature electrochemical processing. However, despite these advantages, molten salts present challenges due to their chemically reactive nature at high-temperatures, especially in the presence of oxidizing impurities. Salt chemistry can fluctuate through interactions with impurities over time, or fuel burnup in the case of molten salt reactors, often leading to the dissolution of metal species. This dynamic environment not only results in complex redox behavior but also promotes corrosion, which is rarely uniform and frequently manifests as localized degradation driven by structural materials’ compositional differences, electrochemical imbalances, and microstructural susceptibilities.

Kim, Changkyu [University of Wisconsin-Madison, WI↗
DOE OSTI2019

Direct Deposition of Crystalline Ta 3 N 5 Thin Films on FTO for PEC Water Splitting

Tantalum Nitride is a promising photoanode material for solar water splitting, but the further study and practical use is constrained by the harsh conditions of the synthesis from Ta metal. Here we report the direct deposition of crystalline Ta 3 N 5 on FTO substrate via a custom-built ALD system. A combination of TaCl 5 (Ta-precursor) and ammonia (N-source) were sequentially pulsed into the ALD reactor with the substrate heated to 550 °C to deposit compact and thin films of Ta 3 N 5 with controllable thicknesses on FTO substrates. Importantly, it is shown that the FTO is chemically and structurally stable under the reducing conditions of ammonia at 550 °C. These electrodes produced an exceptional photocurrent onset potential of ~ 0.3 V vs. RHE with a maximum photocurrent of ~ 2.4 mA cm -2 at 1.23 V vs RHE. Results of photoelectrochemical investigations as a function of film thickness and illumination direction reveal that the performance of Ta 3 N 5 is controlled by a hole diffusion length of ~50 nm. These results are crucial for the successful integration of Ta 3 N 5 in efficient unassisted water splitting applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Ce stabilized Ni–SrO as a catalytic phase transition sorbent for integrated CO 2 capture and CH 4 reforming

Integration of carbon dioxide capture from flue gas with dry reforming of CH 4 represents an attractive approach for CO 2 utilization. The selection of a suitable bifunctional material serving as a catalyst/sorbent is the key. This paper reports Ni decorated and CeO x -stabilized SrO (SrCe 0.5 Ni 0.5 ) as a multi-functional, phase transition catalytic sorbent material. The effect of CeO x on the morphology, structure, decarbonation reactivity, and cycling stability of the catalytic sorbent was determined with TEM-EDX, XRD, in situ XRD, CH 4 -TPR and TGA. Here, cyclic process tests were conducted in a packed bed reactor. The results indicate that large Ni clusters were present on the surface of the SrNi sorbent, and the addition of CeO 2 promoted even distribution of Ni on the surface. Moreover, the Ce–Sr interaction promoted a complex carbonation/decarbonation phase-transition, i.e. SrCO 3 + CeO 2 ↔ Sr 2 CeO 4 + CO 2 as opposed to the conventional, simple carbonation/decarbonation cycles (e.g. SrCO 3 ↔ SrO + CO 2 ). This double replacement crystalline phase transition mechanism not only adjusts the carbonation/calcination thermodynamics to facilitate SrCO 3 decomposition at relatively low temperatures but also inhibits sorbent sintering. As a result, excellent activity and stability were observed with up to 91% CH 4 conversion, >72% CO 2 capture efficiency and ~100% residual O 2 capture efficiency from flue gas by utilizing the CeO 2 ↔ Ce 2 O 3 redox transition. This renders an intensified process with zero coke deposition. Moreover, the SLDRM with SrCe 0.5 Ni 0.5 has the flexibility to produce concentrated CO via CO 2 -splitting while co-producing a syngas with tunable H 2 /CO ratios.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Application of Refractory High-Entropy Alloys for Higher-Reliability and Higher-Efficiency Brayton Cycles and Advanced Nuclear Reactors

An exceptional set of newly-discovered advanced superalloys known as refractory high-entropy alloys (RHEAs) can provide near-term solutions for wear, erosion, corrosion, high-temperature strength, creep, and radiation issues associated with supercritical carbon dioxide (sCO2) Brayton Cycles and advanced nuclear reactors. In particular, these superalloys can significantly extend their durability, reliability, and thermal efficiency, thereby making them more cost-competitive, safer, and reliable. For this project, it was endeavored to manufacture and test certain RHEAs, to solve technical issues impacting the Brayton Cycle and advanced nuclear reactors. This was achieved by leveraging Sandia’s patents, technical advances, and previous experience working with RHEAs. Herein, three RHEA manufacturing methods were applied: laser engineered net shaping, spark plasma sintering, and spray coating. Two promising RHEAs were selected, HfNbTaZr and MoNbTaVW. To demonstrate their performance, erosion, structural, radiation, and hightemperature experiments were conducted on the RHEAs, stainless steel (SS) 316 L, SS 1020, and Inconel 718 test coupons, as well as bench-top components. The experimental data is presented, analyzed, and confirms the superior performance of the HfNbTaZr and MoNbTaVW RHEAs vs. SS 316 L, SS 1020, and Inconel 718. In addition, to gain more insights for larger-scale RHEA applications, the erosion and structural capabilities for the two RHEAs were simulated and compared with the experimental data. The experimental data confirm the superior performance of the HfNbTaZr and MoNbTaVW RHEAs vs. SS and Inconel. Most importantly, the erosion and the coating material experimental data show that erosion in sCO2 Brayton Cycles can be eliminated completely if RHEAs are used. The experimental suite and validations confirm that HfNbTaZr is suitable for harsh environments that do not include nuclear radiation, while MoNbTaVW is suitable for harsh environments that include radiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels

Thermomechanical processing (TMP) of ferritic–martensitic (FM) steels, such as HT9 (Fe–12Cr–1MoWV) steels, involves normalizing, quenching, and tempering to create a microstructure of fine ferritic/martensitic laths with carbide precipitates. HT9 steels are used in fast reactor core components due to their high-temperature strength and resistance to irradiation damage. However, traditional TMP methods for these steels often result in performance limitations under irradiation, including embrittlement at low temperatures (<~430 °C), insufficient strength and toughness at higher temperatures (>500 °C), and void swelling after high-dose irradiation (>200 dpa). This research aimed to enhance both fracture toughness and strength at high temperatures by creating a quenched and tempered martensitic structure with ultrafine laths and precipitates through rapid quenching and unconventional tempering. Mechanical testing revealed significant variations in strength and fracture toughness depending on the processing route, particularly the tempering conditions. Tailored TMP approaches, combining rapid quenching with limited tempering, elevated strength to levels comparable to nano-oxide strengthened ferritic alloys while preserving fracture toughness. For optimal properties in high-Cr steels for future reactor applications, this study recommends a modified tempering treatment, i.e., post-quench annealing at 500 °C or 600 °C for 1 h, possibly followed by a brief tempering at a slightly higher temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Microstructural Characterization of Irradiated U{sub 0.17}ZrH{sub 1.6} Using Ultrasonic Techniques

In recent years, there has been an increased level of effort to understand the changes in microstructure that occur due to irradiation of nuclear fuel. The primary driver for this increased effort is the potential for designing new fuels that are safer and more reliable, in turn enabling new and improved reactor technologies. Much of the data on microstructural change in irradiated fuels is generated through a host of post irradiation examination techniques such as optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM) to determine grain structure, porosity, crack geometry, etc. in irradiated fuels. Such 'traditional' examination techniques were recently used to characterize a novel new fuel consisting of U{sub 0.17}ZrH{sub 1.6} pellets bonded to zircaloy-2 cladded with lead-bismuth eutectic before and after irradiation. In addition to the conventional suite of post-irradiation examination tools, several nondestructive interrogation techniques exist which bring the potential for in-situ assessment of microstructural changes. One such method, ultrasonic examination, was tested on each of the samples. This paper describes the results of this examination. Ultrasonic measurements have a long and successful history of use for nondestructive materials characterization, including nondestructive evaluation (NDE) of degradation and damage, microstructure characterization, and measurement of physical parameters for process control. Ultrasonic NDE is critical to the nuclear power industry's in-service inspection program for maintaining pressure boundary integrity, and is being actively investigated for PIE of fuels. Ultrasonic measurements, typically performed at 10 MHz or higher, performed post irradiation show that fuel microstructural parameters, such as porosity and grain size, can be correlated to ultrasonic velocity. The objective of this examination was to assess the feasibility of using ultrasonic measurements on irradiated fuel to determine sensitivity to microstructural changes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Preliminary Benchmark Uncertainties for Deimos, a HALEU-Fueled and Graphite-Moderated Advanced Reactor Testbed

Many advanced reactor concepts will make use of various uranium fuels with levels of enrichment higher than previously seen in current light water reactors. In particular, High-Assay Low Enriched Uranium (HALEU), that is uranium enriched to 235 U ≈ 20 w/o%, is planned to be used in over ten new reactor concepts. HALEU is attractive for advanced reactors as it enables longer intervals between refueling. Unfortunately, little to no experience with HALEU is available in experimental literature raising concerns for not only licensing advanced re actors but also fabrication and transportation of HALEU fuels. This is where Deimos, a Los Alamos National Laboratory internal project, comes in. Deimos is a new critical experiment scheduled for FY24 at the National Criticality Experiments Research Center (NCERC). Deimos is a graphite moderated, graphite and beryllium reflected critical experiment making use of HALEU TRi-structural ISOtropic (TRISO) fuel from the Compact Nuclear Power System (CNPS). This transaction entails a brief description of efforts to benchmark Deimos for inclusion into the International Criticality Safety

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗