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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Life‐cycle greenhouse gas emissions associated with nuclear power generation in the United States

Under the 2022 Inflation Reduction Act, tax credits of up to $3/kgH 2 are available to hydrogen producers if they generate emissions at levels below 0.45 kgCO 2 e/kgH 2 , spurring producers to explore how hydrogen production via electrolysis using electricity generated by nuclear power may qualify for such tax credits. With uranium as a primary fuel for nuclear power plants (NPPs) and no on-site emissions, the upstream emissions associated with nuclear fuel supply chains largely determine the carbon intensity of nuclear energy. Using the GREET (Greenhouse gases, Regulated Emissions, and Energy use in Technologies) model, we evaluated the life-cycle greenhouse gas (GHG) emissions of uranium production and the use of uranium to generate electricity in light water reactor (LWR) NPPs. We evaluated the process chemicals and energy inputs throughout the nuclear fuel supply chain to identify the major contributors to nuclear fuel cycle GHG emissions. Such emissions are estimated at 3.0 gCO 2 e/kWh at NPPs in the United States. The greatest share of nuclear fuel cycle GHG emissions—comprising 53% of total emissions—are associated with electricity consumption throughout the fuel supply chain. We extended the analysis to include an evaluation of the carbon intensity of H 2 production via electrolysis using nuclear power from LWRs. Finally, we examined the impact of future (2035 and 2050) electricity supply chain scenarios on nuclear fuel cycle GHG emissions. Our analysis revealed a decrease of 33% (2035) and 46% (2050) in the carbon intensity of nuclear electricity relative to current nuclear fuel cycle GHG emissions.

greenhouse gas emissions↗
DOE OSTI2024

Towards Net-Zero: Nuclear-Assisted Waste Biomass to Liquid Fuel in Eastern Idaho

A nuclear assisted carbon negative hybrid energy process that enables production of synthetic bio-crude oil and biochar from Eastern Idaho waste biomass is proposed. The process integrates nuclear powered electricity with high temperature steam electrolysis and biomass hydropyrolysis. The bio-crude oil is of sufficient composition and blended with traditional crude oil at a refinery. Hydrogen from the electrolyzer is pressurized and inserted into the pyrolyzer. Non condensable gases generated in the hydropyrolysis process are burned with oxygen from the electrolyzer to produce heat for the electrolyzer, biomass dryer, and pyrolyzer. The biochar is returned to the soil via fertilizer application and remains there for thousands of years. Since the total process uses nuclear generated electricity, the carbon in the biochar is ultimately sequestered from the atmosphere, thus making the process carbon negative. Using Eastern Idaho wheat or barley straw, this hybrid energy process has the potential to provide an alternative petroleum source. Two options exist for the system design: 1) send electricity from the nuclear plant and straw to a chemical processing plant to produce the bio-crude and biochar, 2) construct the biomass processing facility near the nuclear plant to allow use of nuclear-generated process heat to drive the chemical. Process model description and results are discussed. The process is sized to produce gasoline and diesel at the rate that the INL uses every day for fleet usage.

09 BIOMASS FUELS↗
DOE OSTI2021

Impact of Precipitation Parameters on the Specific Surface Area of PuO 2

Controlling the properties of PuO 2 through processing is of vital importance to environmental transport and fate, production of nuclear fuels, nuclear forensic analyses, stockpile stewardship, and storage of nuclear wastes applications. A number of processing conditions have been identified to control final product properties, including specific surface area (SSA), residual carbon content, adsorption of volatile species, morphology, and particle size. In this paper, a novel approach is developed for the prediction of PuO 2 SSA via the synthetic route of Pu(IV) oxalate precipitation followed by calcination. The proposed model utilizes multivariate regression methodology and leave one out formalism to link Savannah River Site (SRS) precipitation and calcination production data to the SSA of the final product. A comparison among the models provides insight into the accuracy and ability to identify variations amongst the processing data. Additionally, the models may also be used to fit new data outside of the parameters explored in a production facility. Finally, the trained model was compared to a similarly trained conventional model form to illustrate the influence of precipitation parameters on the prediction of the final SSA. The models presented here attempt to provide new methods for more accurate prediction of the PuO 2 product properties in a production scale environment for key environmental and nuclear applications.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

The Modeling of the Synfuel Production Process: Process models of Fischer-Tropsch production with electricity and hydrogen provided by various scales of nuclear plants

Synthetic fuels (synfuels), also known as electro-fuels (E-fuels), are hydrocarbon fuels produced from waste CO2 streams and water electrolysis, with electricity as the primary source of energy. To achieve substantial reductions in greenhouse gas (GHG) emissions, electricity sources must release zero carbon or near-zero carbon, as is the case with solar, wind, hydro, and nuclear power. Nuclear power is one of the largest and steadiest domestic sources of clean energy in the United States. Moreover, nuclear power has the potential to produce hydrogen economically for less than $2/kg, reaching the DOE near-term target price. Thus, using nuclear power to produce synfuels has the unique potential to significantly reduce the GHG emissions of hydrocarbon fuels production and end-use applications. Fisher-Tropsch or FT fuel (a mixture of naphtha, jet fuel, and diesel) is of great interest because it is a drop-in fuel that can be blended with conventional petroleum counterparts and is compatible with existing infrastructure. By using the ASPEN Plus model, this report develops FT fuel production models on three scales, corresponding to nuclear plants with capacities of 1000 MWe, 437 Mwe, and 100 MWe, respectively. The FT model case with energy from a 437-MWe nuclear plant is used as a baseline case. This report summarizes the baseline ASPEN Plus model results with a detailed mass and energy analysis. Our modeled facility produces 507 MT/day (185,000 gal/day) of FT fuel by converting 255 MT/day of hydrogen and 1,580 MT/day of CO2. The FT fuel production energy efficiency from hydrogen and electricity energy inputs is 70% (lowerheating-value or LHV-based). Including the high-temperature electrolyzer in the system boundary, the FT fuel production LHV efficiency from electricity and thermal energy inputs is 51%, considering 39.8 kWh/kg of electricity and 6.86 kWh/kg of thermal energy use from a nuclear plant for hydrogen production. The FT production efficiency can potentially be increased by further integrating the heat exchange between nuclear plant and FT process, and this study is underway. The carbon conversion ratio in the baseline case is 99%, with process CO2 capture and recirculation and oxy-combustion using the oxygen by-product from water electrolysis. The hydrogen consumption is 1.38 kg/gal-FT fuel and the CO2 consumption is 8.56 kg/gal-FT fuel in the baseline case. With different FT production scales determined by the nuclear plant capacity, the FT model was scaled using the same operating parameters, which led to the same conversion efficiency regardless of scale. However, the different FT plant scales will impact the economics of FT fuel production; this impact will be examined in the next phase of this study.

Zang, Guiyan↗
DOE OSTI2025

Vibroacoustic Process Monitoring of 5-Stage Centrifugal Contactors

Vibroacoustic monitoring was investigated to aid in informing process operators with parameters regarding aqueous separation techniques. These techniques have a wide range of applications including nuclear fuel reprocessing. A small-scale solvent extraction system was set up with five magnetic drive contactors and two piston pumps to collect realistic operation. Vibration sensors were placed on three of the contactors and both pumps, and a microphone was placed near the system. Contactor rotational speed and pump flow rate were varied with known values to compare against measured vibroacoustic signatures. Spectral analysis allowed for the determination of contactor rotational speed and the ratio of organic to aqueous flow rate, an important parameter for ensuring optimal extraction. Vibration metrics such as kurtosis and crest factor provided additional details regarding individual rotational speeds and have potential in examination of faults for predictive maintenance. Overall, results indicate the capability of vibroacoustic monitoring to improve operator awareness and process outcomes.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Microstructural heterogeneity of the buffer layer of TRISO nuclear fuel particles

Tristructural isotropic (TRISO) nuclear fuel particles contain a layered spherical shell designed to retain fission products; however, failure occurs in rare cases—commonly initiated in the porous pyrocarbon buffer layer. Achieving a comprehensive understanding of the buffer-initiated failure mechanisms requires detailed characterization of the buffer porosity and its heterogeneous distribution across multiple length scales. Here we performed FIB-SEM tomography across the buffer layer thickness (~100 µm) to produce 3D reconstructions of the buffer microstructure with 50 nm spatial resolution. We found an average overall porosity of ~14%, which does not solely account for the low density of the buffer (50% of the theoretical density). Additionally, the local porosity and its fluctuation increase from the kernel interface towards the inner pyrocarbon (IPyC) layer, which we attribute to the chemical vapor deposition process conditions during the TRISO particle fabrication. Detailed characterization of the porous microstructure—including analysis of the pore size, distribution, shape, and orientation—provides insight into the process-structure-property-performance relations of TRISO nuclear fuel particles and will inform multiscale models designed to predict the failure of TRISO particles under irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Radiolytic evaluation of select sulfur chlorides (S 2 Cl 2 and SOCl 2 ) for advanced low temperature chlorination of zirconium-based used nuclear fuel cladding

New sulfur chloride compounds—sulfur monochloride (S 2 Cl 2 ) and thionyl chloride (SOCl 2 )—have been proposed as alternative chlorinating agents for the chemical decladding of zirconium-based used nuclear fuel cladding materials. However, the radiation stability of these compounds has not been thoroughly evaluated, despite envisioned process conditions involving intense, multi-component radiation fields. Knowledge of the radiation chemistry of these sulfur chlorides is essential for the development and optimization of alternative chlorination technologies. To this end, we present preliminary findings on the gamma radiolysis of neat S 2 Cl 2 and SOCl 2 at ambient temperature up to ~ 34 MGy. In this study, Raman spectroscopy and density functional theory calculations were used to identify the Raman-active degradation products from S 2 Cl 2 and SOCl 2 radiolysis. Both sulfur chloride compounds exhibited significant radiation resistance, with respect to changes in the Raman signatures of the parent compounds and the ingrowth of Raman-active degradation products. For S 2 Cl 2 , molecular chlorine (Cl 2 ) was directly identified as the predominant degradation product, which was completely consumed at higher absorbed doses (> 9 MGy). Similarly, the main degradation product from SOCl 2 radiolysis was also Cl 2 , although in this system it continued to grow in with dose over the entire dose range. Further, an “S=O” containing degradation product(s) was also identified as a compliment to Cl 2 . Overall, the perceived radiation resistance of these sulfur chloride compounds makes them suitable for inclusion in a used nuclear fuel chemical decladding process, especially as radiolytically formed Cl 2 can be used to reform the parent compounds, thereby increasing the longevity of the solvents used.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Systems Engineering and Analysis in Support of a US Federal Staging Facility for UNF

The US Department of Energy Office of Nuclear Energy (DOE-NE) Office of Spent Fuel and High-Level Waste Disposition is examining a set of system options and conducting supporting analyses to inform the development of an integrated waste management system, which may include one or more federal staging facilities (FSFs) for used nuclear fuel (UNF ) sited using a collaborative siting process. This paper focuses on the ongoing activities in two systems engineering and analysis work areas: (1) data and tools development, validation, and maintenance and (2) systems engineering execution. Within the first work area, the STANDARDS 5.0 UNF data and analysis tool, formerly known as UNF-ST&DARDS, is being developed as a foundational resource to assist in the management of UNF data. It has the key capability to model UNF throughout the entire back end of the fuel cycle. STANDARDS also includes several compatible analysis tools for the time-dependent characterization of UNF and related systems by interfacing with the SCALE code system for nuclear analysis and COBRA-SFS for thermal analysis. Also, within the data and tools area is the Next Generation System Analysis Model (NGSAM), which is an agent-based simulation software tool expressly designed to be capable of modeling the waste management system, including the transportation of UNF to and from a FSF. NGSAM has been developed to enable informed decision-making by providing the capability to analyze various potential system options for the management of UNF and high-level radioactive waste. Finally, in the systems engineering execution area, the team has begun to apply a disciplined systems engineering approach at the system level along with supporting analysis to guide the development of the FSF project requirements (including associated transportation infrastructure). Systems engineering principles and practices and their adaptation/application to design and development activities will ensure that the waste management system is effectively implemented as work proceeds. Other activities include investigating the implications of changes in various assumptions and parameters related to waste management systems, such as UNF acceptance rates, receipt logic, facility capacities and capabilities, use of standardized canisters, and different assumed facility operation start dates. Keywords: federal staging facility (FSF), used nuclear fuel (UNF), integrated waste management (IWM) system, Next Generation System Analysis Model (NGSAM), STANDARDS, systems engineering

Joseph, Robert↗
DOE OSTI2024

Studies on the Vitrification Potential of High Risk Spent Nuclear Fuels

This research explored vitrification of U-Zr nuclear fuel that exhibits an explosion hazard when processed by common nitric acid dissolution. Glass frits were chosen based on properties that had been previously measured as well as being demonstrated successfully in production at the Defense Waste Process Facility (DWPF). Zirconium and Aluminum metal powders were chosen to represent the spent fuels after literature demonstrated that the potential phase that causes the issue for explosive reaction in the effluent stream is tied to the Zirconium-III phase field. DSC measurements were carried out on various simulants and waste forms to better understand thermal behavior of these materials. Zirconium (Zr) was tested with both nitric acid and the new effluent flowsheet that is under development by the SRNL Chemical Process Control (CPC) group. The nitric acid test did show the expected exothermic behavior as well as how the behavior is affected by particle size of the powders used. In contrast, the effluent samples did not show an exothermic peak. However, the concentration of metal in the samples was orders of magnitude lower than in the nitric acid testing and further testing of the effect of concentration on the exothermicity that could be measured confirmed a dampened signal. Tests were run that included frit, both DWPF Frit 510 and Iron Phosphate frit, exhibiting complex curves that included multiple endothermic and exothermic peaks during heating and cooling. The final crucible samples appear similarly to the bulk glasses pointing to these peaks being a part of the dissolution of the simulant into the base glasses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

EBR-II Driver Fuel Cladding Hull Analytical Data

Sharing EBR-II driver fuel cladding hull data with General Atomics under NDA No. 20NDA155. Data share is needed to advance discussions regarding process technologies to decontaminate separated cladding hulls

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

TRISO PARTICLE FUEL POLISHING PROCESS COMPARISON

The investigation of the TRISO coated particles is important in terms of verification the quality of the nuclear fuel samples before placing them in the reactor core. To determine the degree of damage to the coated particles, the samples must be properly polished in order to uncover the inner layers. The poster presents a comparison of two polishing methods. Mechanical polishing method, which is used frequently for sample preparations generates additional mechanical damages in the studied fuel particle, thus directly affecting the experimental results. Due to that, a novel way to polish the samples is presented - an ionic polishing method. Based on the obtained results, it was concluded that the ion polishing method is better because the level of interference with the structures of the individual layers of the tested samples is much lower than with the mechanical method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Production of anhydrous ƒ-element fluorides through the ionothermal treatment of ƒ-element oxalates

The pivotal role of uranium and plutonium fluorides in the nuclear fuel cycle, particularly in the pyrochemical reduction process, is well recognized. Traditionally, the fluorination of uranium and plutonium materials relies on the use of highly toxic and corrosive gases (e.g., HF (g) , F 2(g) ). Herein, we present an alternative approach using the ionic liquid 1‑butyl‑3-methylimidazolium hexafluorophosphate ([Bmim][PF 6 ] (l) ) and/or hexafluorophosphoric acid (HPF 6(aq) ) as fluorinating agents for the ƒ-element oxalates M$^{III}_{2}$(C 2 O 4 ) 3 ∙ 9H 2 O (s) M$^{III}_{2}$(C 2 O 4 ) 3 ∙ 9H 2 O (s) (M III = Ce, Pu) and M IV (C 2 O 4 ) 2 ∙ 6H 2 O (s) (M IV = Th, U). Our findings demonstrate that [Bmim][PF 6 ] (l) and HPF 6(aq) enable the ionothermal fluorination of ƒ-element oxalates, resulting in the formation of anhydrous CeF 3(s) , ThF 4(s) , and UF 4(s) within 2 hours at 200 °C. This method also facilitates the partial fluorination of plutonium(III) oxalate, yielding a mixture of anhydrous PuF 3(s) and an unidentified phase. Overall, the ionothermal treatment approach offers a safer and more efficient means of producing anhydrous ƒ-element fluorides than conventional methods involving hazardous gases. In addition, we describe the morphology of UF 4(s) materials as a function of production route and demonstrate the presence of morphological signatures that could be used during a nuclear forensic investigation.

Cerium↗
DOE OSTI2024

C Modules Enrichment

Module C1 discusses the step in the nuclear fuel cycle where the UF6 solid in cylinders from the conversion plant is processed to enrich the percentage of U-235 from 0.711% to the 3–5% typical of the enrichment used for light-water reactor nuclear fuel fabrication. It involves receipt of UF6 feed stock in 12.5 ton cylinders, enrichment operations, formation of enriched UF6 solid, and shipment of 2.3 ton cylinders to fuel fabricators. In this module, “SWU” is taken as shorthand for kg-SWU, the formal units for enrichment work, assuming that heavy metal mass flows will be gauged in kg.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Traveling Molten Zone Refining Process Development for Innovative Fuel Cycle Solutions

This ARPA-E ONWARDS project successfully developed and demonstrated an innovative immiscibility zone refining process for separating actinides from active fission products in metallic nuclear fuel applications. The technology addresses a critical challenge in advanced nuclear fuel cycles by providing a high-throughput, low-waste alternative to conventional electrorefining.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Optical vibrational spectroscopic signatures of ammonium diuranate process parameters

Ammonium diuranate (ADU) is commonly encountered in the nuclear fuel cycle; however, previous investigations have shown that ADU is a complex mixture of distinct compounds. Moreover, production parameters are known to heavily influence the composition of the resulting ADU. Here, we examine four samples of ADU prepared at Oak Ridge National Laboratory (ORNL), and one sample of ADU made at Pacific Northwest National Laboratory (PNNL), with the goal of further characterizing and elucidating the effect of processing parameters such as stir rate, strike direction, and temperature on material composition. Process parameters during ADU precipitation at ORNL and PNNL were well documented, and we relate process variables to optical vibrational spectroscopic signatures observed using Raman and infrared (IR) spectroscopy. In addition, powder X-ray diffraction (PXRD) reveals differences in the solid-phase composition of ADU precipitates, but we find that the primary phase is similar to the uranyl oxyhydroxyhydrate mineral metaschoepite. Despite the significant phase contributions of a metaschoepite-like phase, spectroscopic evidence of both nitrate and ammonium are observed for all samples. To gain a more holistic understanding of spectroscopic features of process parameters in ADU, principal component analysis (PCA) is employed and results in observable signatures that relate to the stir rate used during synthesis. These results provide further information about the process-dependence of ADU precipitate composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fuel Qualification for Molten Salt Reactors

This report supports the development and review of an efficient methodology or process for liquid salt fuel system qualification. The report describes the technical issues encountered when developing an adequate understanding of the fuel salt’s chemical and physical behavior under both normal and accident conditions. The concepts presented in this report align with the fundamental safety functions (FSFs) approach of the more general non-light water reactor (nonLWR) fuel qualification guidance provided in NUREG-2246. However, because liquid salt fuel has substantial technical differences from solid fuels, this report describes tailored approaches for achieving the common safety intents when using liquid salt. The purpose of this report is to illustrate how variations in fuel salt properties impact achievement of FSFs at liquid-salt fueled molten salt reactors (MSRs). The focus on FSFs supports methodology compatibility with both current and developing licensing processes. Customized fuel qualification guidance for liquid salt–fueled Nuclear Power Plants (NPPs) are also provided in this report to supplement the generic non-LWR guidance recently developed by the NRC. As with the broader non-LWR fuel qualification, this report focuses on identification and understanding of fuel life-limiting failure and fuel salt property degradation mechanisms that occur as a result of irradiation during reactor operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Quantitative Morphological Characterization of Carbide Inclusions in Uranium Metal

Uranium carbides (UCs) are prevalent inclusions in U metal that form during melting operations from interactions with crucible walls and the casting chamber atmosphere. Although UCs have been studied extensively since the beginning of U metal foundry operations, there are still unknowns regarding the effects of thermomechanical processing on their sizes and morphology. Here, we present the results of a series of controlled cooling experiments with molten uranium to elucidate the effect of cooling rate on inclusion morphology in a-U. Samples were melted using a vacuum induction melter and manually cooled at rates of 2.5, 1.7, 1.1, 0.8, 0.3 (± 1%) K/s from ~1600 K to <700 K in under 1 hour. Subsequent scanning electron microscopy (SEM) was performed on cross-sections of the samples, revealing a complex mixture of UC morphologies that are indicative of diffusion and growth influenced by the thermal processing of the U matrix. Image analysis using the Morphological Analysis of Materials (MAMA) software showed that UC sizes generally grew larger with slower cooling rates, and the two slowest cooling rates noticeably impacted the inclusion circularity and ellipse aspect ratio. These results indicate that UC morphology is sensitive to short cooling rates (<1 hour) and could therefore be controlled in the production of metallic nuclear fuels. Additionally, inclusion speciation and morphologies could potentially provide forensic clues about processing history of unknown metal samples. Understanding the driving forces involved in UC morphology evolution is beneficial for evaluating metal fuels for next generation nuclear reactors and for identifying signatures for nuclear forensics.

Athon, Matthew T.↗