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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 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Advanced Computational Modeling of High-Level Waste Vitrification at the Hanford Site

The U.S. Department of Energy (DOE) has selected vitrification for stabilizing legacy tank waste at the Hanford site, where radioactive waste from plutonium production was historically stored in underground tanks. This waste will be separated into low-activity waste (LAW) and high-level waste (HLW) fractions and processed at the Waste Treatment and Immobilization Plant (WTP). At WTP, glass melters are used for the vitrification of radioactive tank waste, transforming it into a stable borosilicate glass form for safe long-term storage. The melter vessel is constructed from highly durable and heat-resistant materials, where the vitrification process occurs. The main regions that are modeled are the melt pool, plenum, cold cap, riser/discharge chamber, and surrounding structure with insulation layers. Forced convection induced by air bubblers at the base of the melter ensure uniform temperature distribution and provide heat to the cold cap layer. The cold cap is a region of reacting batch feed that floats on top of the molten glass and is where the batch-to-glass reactions occur. Joule heating provided by electrodes mounted along the vertical walls of the melter and immersed directly in the glass, generates the necessary heat for the net endothermic conversion processes that occur in the cold cap. The high temperatures, radioactivity, and opaque nature of the glass prevent direct observation inside the melters. Therefore, computational models are essential for providing insight into factors that affect melter throughput. Thermocouples in the plenum provide operators with plenum temperature measurements. Operational adjustments include bubbling rate, voltage supplied to the electrodes, feed adjustments, and glass removal rate. Different computational fluid dynamics (CFD) models have been developed, each serving a specific purpose. There are CFD models of different scale melters, as well as models that capture the two-phase flow interfaces of rising bubbles in the molten glass or models with a simplified molten glass region so that the surrounding structure and plenum can be feasibly incorporated. Pilot-scale melter models have been developed to serve as validation of the methods employed in the simulation of the full-scale WTP melters. Models incorporating resolved bubbling are used to develop momentum source terms to implement into a single phase, multi-region, steady-state flow model that is being validated by measured process parameters such as glass production rate, voltage, input power, plenum temperatures, etc. The resolved bubbling model uses the multiphase volume of fluid approach to model the system with a high-resolution interface capturing scheme to maintain sharp interfaces between the molten glass and the air phase. The suite of CFD models is continually being improved to incorporate more realistic physics and achieve faster turnaround time. For example, an incremental controller is implemented to automatically adjust electrode voltage within the simulation to a molten glass set point temperature of 1150°C. Newer models feature improved meshes to ensure conformal meshes between regions and eliminate unnecessary mesh refinement in areas that are not of interest (such as boundary layers in offgas ports). Instead of explicitly modeling the structural, refractory, and insulation layers of the melter, a thermal resistance approach is used with published correlations used for boundary conditions. The development of robust and efficient CFD models will be instrumental in enabling the WTP to successfully fulfill its mission of safely stabilizing legacy nuclear waste.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2026

Utilization of Unmanned Aircraft Systems for Environmental Purposes at the Savannah River Site – 26578

Born in the 1950s, the Savannah River Plant was constructed as a plutonium and tritium production plant. As the Cold War era came and went, a change of direction was signaled when the name of the facility changed from Savannah River Plant to Savannah River Site (1989) – the main mission at Savannah River shifted from “production” to “cleanup.” The Department of Energy is responsible for managing the 310 square miles of land that is the Savannah River Site and the cleanup/remediation activities that occur. The Savannah River National Laboratory utilizes unmanned aircraft systems to assist with some of those environmental monitoring and remediation activities. One function of unmanned aircraft systems at the Savannah River National Laboratory is conducting aerial photography and videography. Various infrastructure evaluations have been performed with unmanned aircraft – facility rooftop inspections, water tower lock verifications, earthen dam surveys, etc. An unmanned aircraft system has been used for progress footage of remediation projects – Dunbarton Bay remediation, Saltstone Disposal Unit construction, etc. Thermal cameras on an unmanned aircraft system are used to help identify cracks/leaks in structures from vantage points not readily accessible by personnel. Virtual tours of different waste units are conducted with unmanned aircraft systems for Site groups and federal/state regulators to save costs and travel time. Additionally, advanced remote sensing technologies are used on unmanned aircraft systems at the Savannah River Site. Light Detection and Ranging uses laser pulses to measure distances to the Earth's surface or other targets to create highly detailed topographic maps that accurately portray terrain (e.g., elevation changes, slopes, and contours). Data collection with a Light Detection and Ranging unit mounted on an unmanned aircraft system is quick and efficient – large-area surveys are conducted in shorter time frames. Because Light Detection and Ranging can penetrate through foliage and vegetation to ground level, it is being used in conjunction with a watershed model to study the General Separations Area of the Site to determine stormwater flow volume and direction, monitor stream stability, and forecast river flow behavior. A hyperspectral camera captures a wide spectrum of electromagnetic wavelengths across numerous narrow bands, beyond what the human eye can see. It provides detailed spectral information about the objects or surfaces in an image, which can be used to identify and analyze materials based on their spectral signatures. A hyperspectral camera mounted on an unmanned aircraft system has been used at the Savannah River Site for non-destructive evaluation of concrete and concrete structures. Unmanned aircraft systems at the Savannah River Site are also used for the application of herbicide on identified rooftops. P-Reactor and R-Reactor are entombed nuclear reactors at the Savannah River Site. Unwanted vegetation is often present on these rooftops – over time roots can penetrate seams, cracks, and joints of the general roof and concrete caps, leading to water ingress and deterioration of the structural components. For time and cost efficiencies as well as in the interest of personnel safety, an unmanned aircraft is used to dispense herbicide on rooftop areas where vegetation is noticed to help alleviate the issues/hazards.

Lorier, Troy [Savannah River National Laboratory (↗
DOE OSTI2021

Rapid dissolution of PuO 2 analytical samples using mediated electrochemical oxidation

Current methods for quantitative dissolution of solid PuO 2 involve lengthy dissolution times. The adaptation of mediated electrochemical oxidation, which uses an electrochemically produced oxidizing agent to catalyze dissolution, has been explored for radiochemical analysis scale dissolutions of PuO 2 . The effects of temperature, electrode material, and electrocatalyst on the dissolution kinetics were explored. Under conditions examined in this study, complete dissolution of tens of milligrams of solid PuO 2 can be achieved in less than 1 h at ambient temperature. Overall, this methodology could be adapted to dissolve other metal oxides.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Progress on Pu-238 production at Idaho National Laboratory from February 2022 to July 2023

Idaho National Laboratory (INL) has continued to qualify irradiation positions in the Advanced Test Reactor (ATR) for Pu-238 production to support NASA deep space missions. Over the past year, INL qualified Np-237 targets for ATR’s North East Flux Trap (NEFT), Inner-A and H positions. Work has begun to requalify the South Flux Trap (SFT) and to qualify the East Flux Trap (EFT) for the ATR GEN I target and is midway through the qualification process. This paper gives an overview of operational and technical activities from February 2022 to July 2023.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Biogeochemistry of Pond B (Savannah River Site, South Carolina, USA): Sediment Core, Total extraction data, Pond B Savannah River Site July 2019. Subsurface Biogeochemistry of Actinides SFA

Pond B at Savannah River Site (SRS, South Carolina) is a monomictic reservoir that received SRS R reactor cooling water from 1961–1964. Previous studies conducted between the 1980s–1990s on the water column and sediments of Pond B measured trace amounts of Pu (33 MBq 238Pu and 430 MBq 239,240Pu), 241Am, and 137Cs. Since then, the pond has been relatively isolated and the radionuclide concentrations have not been monitored over time. Herein, about 30 years after the last publication on Pond B, we are re-evaluating the geochemistry and radionuclide distribution within Pond B at four locations along a horizontal transect from the inlet to outlet.This study investigated the distribution of anthropogenic radionuclides Pu-239 and Cs-137 along with total organic carbon, iron, and trace element in contaminated sediments of Pond B at the Savannah River Site (SRS). Pond B received reactor cooling water from 1961 to 1964, and trace amounts of Pu-239 and Cs-137 during operations. Our study collected sediment cores to determine concentrations of Pu-239, Cs-137, and major and minor elements in solid phase, pore water and an electrochemical method was used on wet cores to determine dissolved elemental concentrations.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Experimental Investigations into the Corrosion of Alloy 625 Using NaCl-PuCl3 Molten Salt in a Natural Circulation Microloop

Molten salt reactors (MSRs) can potentially revolutionize the nuclear industry by providing a path to a near-zero nuclear waste fuel cycle, contributing to more sustainable energy sources. As a plethora of MSR developers in the United States work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have limited operational testing with nuclear material. Licensing and operating these reactors require an understanding of corrosion effects on reactor materials of construction under operational conditions. The TerraPower Molten Chloride Fast Reactor (MCFR) is a liquid-fueled chloride-salt fast reactor which has received notable interest from the utility sector based on its desirable economic characteristics. The reactor operates at low pressure but does not require the use of highly reactive chemicals, leading to a reduced use of concrete and steel during construction. Additionally, liquid fuel allows for inherently stable behavior and natural circulation during a loss-of-site-power scenario. MCFR can be refueled while operating which makes it compatible with variable generation sources such as wind and solar. MCFR is a breed-and-burn in-situ reactor that does not implement any chemical processing or separations in the fuel cycle. Only mechanical filtration of noble metals and off-gassing of noble gases are utilized while the actinides stay mixed with the fuel at all times. The MCFR will require technology development to reach commercialization. With a breed-and-burn in-situ reactor like MCFR, the transmutation of fertile U-238 to fissile Pu-239 allows for much greater fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The electronic Raman scattering spectrum of PuO 2

Here the Raman spectrum of PuO 2 was measured up to 13,000 cm –1 with three different laser excitation wavelengths spanning the resonance (405 nm), near-resonance (457 nm), and preresonance (514 nm) energy range. Approximately 26 never-before-seen bands were observed between 3500 and 13,000 cm –1 . Given the very high energy of the Raman shifts of these bands and the relative insensitivity of their spectral position to the interrogating laser wavelength, they are believed to arise from an electronic origin. These bands are present in both freshly calcined and radiolytically aged PuO 2 , although a broad luminescence is observed in the aged material, which obscures many of the high frequency features. In situ laser annealing of the material attenuated this luminescence and allowed for clear observation of these never-before-seen spectral features. Discovery of these high-energy bands presents a new way of identifying PuO 2 for nuclear nonproliferation and forensics purposes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Characterization of PuO 2 With Visible and UV Raman Spectroscopy: Discrimination Between the Bulk, Surface, and an Intermediate Disordered Layer

Raman spectroscopy is an ideal tool in the characterization of materials including PuO 2 . The wavelength-dependent absorptivity of the material defines the light penetration depth and the relative Raman scattering contribution from the bulk and the surface. Here, the surface contribution to the total Raman scattering was investigated for PuO 2 calcined at various temperatures and recorded with laser wavelengths of 355, 325, and 244 nm. These experiments provided the first glimpse of the wavelength-dependent disappearance and emergence of new phonons and electronic bands from the PuO 2 surface layers. The first indication of the wavelength transition in the Raman spectra was the loss of the 2LO2 (overtone, ~1155 cm -1 ) band and the weakening intensity of the Г 1 → Γ 5 electronic band (~2135 cm -1 ) with the 355-nm excitation laser. The Γ 5 electronic band was barely visible with the 244-nm excitation. The electronic band located at ~1050 cm -1 , corresponding to the Г 1 → Γ 4 electronic transition was observed to dramatically increase in intensity while the Г 1 → Γ 3 electronic band (2640 cm -1 ) sharpened as the UV wavelength was increased in energy from the near- to deep-UV (355–325–244 nm). The FWHM of the T 2g band was found to vary with calcination temperature (450°C and 900°C) with the 325-nm laser and the 244-nm laser. The T 2g band attributes, the strong emergence of the Г 1 → Γ 4 electronic band, and the disappearance of the 2LO2 overtone acquired with the 244-nm excitation for the different calcination temperatures suggest a shallow penetration depth.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Neptunium redox speciation and determination of its total concentration in dissolved fuel simulant solutions by spectrophotometry

Here, two new approaches to measure Np concentration in dissolved used nuclear fuel simulant (aqueous feed for PUREX process) by spectrophotometry are developed. The first approach is based on chemical reduction of Np in the feed to its tetravalent state using ascorbic acid with simultaneous conversion of Pu(IV) to Pu(III). Interfering effects from light absorbing fission and corrosion products are accounted for by measuring optical absorbance spectrum of aqueous raffinate after extraction of U, Np, and Pu by tributyl phosphate in dodecane. The second approach uses no chemical treatment at all and relies on spontaneous valency adjustment of Np to Np(V) by dilution of the feed with water to reduce its acidity to low decimolar range of nitric acid concentration. Results of Np determination in the feed by spectrophotometry are in good agreement with its concentration measured by ICP-MS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Machine learning for fundamental spectroscopic and thermodynamic data of actinides and lanthanides

Accurately modeling optical spectra with absolute radiometric intensities is vital for nuclear forensics applications that depend on characterizing optical emissions from energetic nuclear phenomena. This requires precise knowledge of the individual atomic transition probabilities, known as Einstein A-coefficients, for each emission line. Obtaining these values theoretically or experimentally is often impractical due to the complex electronic structures and the number of transitions involved in atoms relevant to nuclear applications. In this study, we explore the use of machine learning to predict the Einstein A coefficients for atomic transitions. Seven models were evaluated that ranged from deep learning to decision tree algorithms, and found that gradient boosting performed best, specifically the Extreme Gradient Boosting (XGB) architecture, achieving a precision of 86% across transitions of 36 elements. Furthermore, the model was cross-validated using published transition probabilities reported in the literature and applied to estimate Pu plasma temperatures from a previous experiment conducted at Savannah River National Laboratory.

Atomic spectroscopy↗
DOE OSTI2020

Use of a Pressurized Water Treatment to Prevent Cracking of Internal Gelation Sol-Gel Microspheres

Pressurized water treatments were investigated as a means to prevent cracking of internal gelation sol-gel microspheres upon drying and heating. A batch of gelled cerium oxide microspheres was prepared, split into two samples that were processed with and without a pressurized water treatment, and dried in air at room temperature. Thermal gravimetric analysis was used to compare the mass loss profiles between the two samples and to understand volatile emissions from cerium oxide microspheres during heat treatments up to 600 °C. Microspheres that were processed without a pressurized water treatment lost approximately 4.6 times more mass upon heating than microspheres that were subjected to a pressurized water treatment. Comparison of the cracking behavior of microspheres prepared with and without a pressurized water treatment indicated that microspheres subjected to a pressurized water treatment were virtually crack-free after heat treatments. Conversely, microspheres processed without a pressurized water treatment, which is typical of traditional washing processes, exhibited extensive cracking. Analysis of residues dried from pressurized water treatment effluents indicated that hexamethylenetetramine and urea, which are excess reactants from the internal gelation process, were removed from gelled microspheres during the pressurized water treatment step. Traditional processing methods for preparing internal gelation microspheres use specific feed solution and gelation parameters to reduce subsequent cracking upon drying and heating. Results from this study indicate that more aggressive gel-forming conditions may be used to produce crack-free microspheres, if a pressurized water treatment step is included after washing and before air-drying.

cerium oxide, internal gelation, microsphere, TRIS↗
DOE OSTI2021

Effects of a Pressurized Water Treatment on Internal Gelation Sol-Gel Microspheres

Based on previous observations that a pressurized water treatment (PWT) prevented cracking of sol-gel microspheres, we investigated the effects of a PWT on microsphere crystallinity, density, and specific surface area (SSA). Results were used to determine how a PWT alters the properties of microspheres upon drying and heating. Microspheres with diameters near 100-200 µm were prepared with and without a PWT and measured using x-ray diffraction (XRD), transmission electron microscopy (TEM), nitrogen adsorption (BET), and pycnometry. Properties of air-dried microspheres processed with and without a PWT are compared. Additionally, the properties of microspheres processed using a PWT are reported after heating to 150 °C, 450 °C, 750 °C, 1050 °C, and 1350 °C. X-ray diffraction measurements indicate that the PWT step improves the crystallinity of air-dried microspheres. XRD data was also used to calculate crystallite size, which increases with higher heat treatment temperatures. TEM images support crystallite size calculations from XRD data and provide an indication of the range of crystallite sizes, particularly for samples processed at higher temperatures where crystallite sizes are too large for estimation using the Scherrer formula. Density and SSA measurements performed as a function of heat treatment temperature indicate that a PWT increases the density of air-dried microspheres, creates a pore network, and that significant densification occurs between 450 - 750 °C. These results may be used to inform decisions on internal gelation flowsheet parameters to optimize the microsphere formation and gelation step, prevent microsphere cracking, and produce microspheres suitable for subsequent coating operations or pressing into pellets.

cerium oxide, TRISO, internal gelation, nuclear fu↗
DOE OSTI2021

Nuclear Criticality Safety [Book Chapter]

Nuclear Criticality Safety is a field of nuclear engineering that involves worker and public safety during the handling, processing, transportation, and storage activities of fissile isotopes ( 233 U, 235 U, and 239 Pu) outside of nuclear reactors. Criticality accidents release energy as a result of accidentally producing a self-sustaining fission chain reaction. If these occur during hands-on operations the results can be lethal to worker within about 4.6 m (15 ft.) of the accident. There have been 22 process criticality accidents in the world between 1953 and 1999 resulting in physical injury, on and off-site dose, or death. Nuclear Criticality Safety involves work by qualified persons to preclude criticality accidents and their consequences by ensuring activities with fissionable materials remain subcritical, i.e., safe, during all normal and credible abnormal conditions. This article presents information about those nuclear fuel cycle steps presenting Nuclear Criticality Safety concerns that must be considered to ensure worker and public safety is assured.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Properties and reaction mechanism of phosphoric acid activated metakaolin geopolymer at varied curing temperatures

Reaction mechanism of phosphoric activated metakaolin is studied by techniques of calorimetry, ICP-OES, FTIR and NMR taking curing temperature and [H{sub 3}PO{sub 4}] as variables. A two-stage curing method including pre-curing at 40 °C for 24 h and second curing at 60/80 °C for another 24 h effectively mitigates thermal-cracking of the obtained silico-aluminate phosphate (SAP) samples, whose maximum compressive strength reaches 120 MPa. Dealumination of metakaolin in H{sub 3}PO{sub 4} solution produces low aluminate-containing silicate units and subsequently condensation of silicate tetrahedrons occurs. Higher temperature and H{sub 3}PO{sub 4} concentration accelerate the two processes. After complete reaction, A{sub VI} structure from metakaolin disappears and the dissolved Al preferably forms Al{sub VI}-O-P units. During reaction, metastable P-O-P intermediate appears in SAPs with higher P/Al ratio and finally transforms into Al-O-P structure upon PO{sub 4}{sup 3−} consumption. Phosphate units as P(OAl){sub x}(H{sub 2}O){sub 4-x} are connected with Si or Al structure units. Major structural units of the SAPs include Al-O-P, Si-O-P, Si-O-Si, and Si-O-Al.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A CALPHAD-informed approach to modeling constituent redistribution in Zr-based metallic fuels using BISON

Here, a CALPHAD-informed (Computer Coupling of Phase Diagrams and Thermochemistry) constituent redistribution model was developed for Zr-based metallic fuels and incorporated into the BISON fuel performance code. Three uncertain model parameters associated with β and γ phase kinetics were calibrated using integral test data from U-Zr fuel elements irradiated in Experimental Breeder Reactor II. The calibrated constituent redistribution model was shown to predict the behavior of U-Zr fuels with excellent accuracy. Model predictions for U-Pu-Zr fuels were physically reasonable but less accurate. Reduction of uncertainties in the ternary phase transition temperatures and collection of kinetic data for the ζ phase are expected to improve the model’s ternary predictions. Finally, the new model was coupled to existing thermomechanics models in BISON to simulate irradiation of an entire U-Zr fuel element, demonstrating its ability to accurately predict the behavior of U-Zr fuels with realistic geometries and mesh resolutions at the engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗