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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 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Characterization Results for the August 2025 Tank Farm 2H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents the average characterization results for the August 2025 2H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC. These analyses were performed in duplicate, and all three measured radionuclide concentrations were within ETP WAC limits. A summary of the analytical results for this 2H evaporator overhead sample includes the following: The measured cesium-137 activity in the 2H evaporator overhead sample averaged 3.02E+02 dpm/mL, which is below the ETP WAC limit of 1.30E+03 dpm/mL. The strontium-90 activity in the 2H evaporator overhead sample averaged 3.33E+00 dpm/mL, which is below the ETP WAC limit of 1.76E+02 dpm/mL. The iodine-129 activity in the 2H evaporator overhead sample averaged 1.20E-01 dpm/mL, which is below the ETP WAC limit of 1.00E+00 dpm/mL. Although the measurements for Cs-137 and Sr-90 are lower compared to previous years, the results remain within a similar range to past measurements and are well below the WAC limits. Iodine-129 indicates detectable results compared to previous years, which can be attributed to recent salt batch processing trending close to the WAC limit for saltstone production. However, these results are still below the ETP WAC limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Effect of chlorine and chromium on sulfur solubility in low-activity waste glass

According to an empirical sulfur solubility model developed from over 200 simulated low-activity waste (LAW) glasses, chlorine and chromium show strong effects on lowering sulfur solubility in glass. This work was aimed at understanding the mechanism behind the negative effect of chlorine and chromium on sulfur solubility. A simplified LAW glass was prepared and saturated by sodium sulfate, sodium chloride, and sodium chromate salts as single components and as mixtures with different ratios at 25 mol%, 50 mol%, and 75 mol% of sulfate mixed with chloride or chromate. A crucible-scale salt-saturation was performed by mixing the crushed glass powder with an excess amount of salts, melting at 1150 °C for one hour, and quenching. Three mixing-melting-quenching cycles were applied to determine saturation concentrations of sulfur, chromium, and chlorine incorporated into the simplified LAW glass. The glass compositions were analyzed by inductively coupled plasma–atomic emission spectroscopy, ion chromatography, and X-ray emission spectroscopy. It is proposed that the negative effect of chlorine and chromium on sulfur solubility can be explained based on the simple hypothesis that sulfur, chlorine, and chromium are present in the glass as anions competing for the voids or interstitial sites of the glass matrix, and that their solubility strongly depends on the effective size of corresponding anions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

The Argonne SuperGel for CBRN Decontamination - 20160

The Argonne SuperGel was developed between 2003 and 2015 to fill a gap in our nation's capability to quickly decontaminate important structures following a radiological or nuclear release event. Specifically, the decontamination technology was developed to minimize damage to monuments, high valued structures, and critical infrastructure while reducing environmental and health impacts. An important criterion during its development, common reagents were employed that could be easily acquired in order to minimize the timeline for its deployment. Its current formulation uses off-the-shelf super-absorbing hydrogels common to the food and agricultural industry and common salts. Over the years, two formulations of the Argonne SuperGel have been developed to specifically target radioactive cesium contaminations and then, more generically, actinide and fission product contaminations. A biodegradable derivative of phosphoric acid is used in small amounts to promote the removal of insoluble actinide species. The original order of unit operations for the SuperGel technology was 1) applying the wash solution to promote mobility of the contaminants on the surface, 2) applying the hydrogel to absorb the wash solution and contaminants, and 3) removing the hydrogel for disposal. During the course of our studies, we tested a new formulation of the gel technology that eliminated the need for a separate step to apply the wash solution without a compromise in decontamination factors. Greater than 70% and >95% of Cs-137 were removed from concrete and tile coupons, respectively, after two decontaminations, with the best results using wash solutions of 1.0 M KCl and 1.0 M NH{sub 4}Cl formulations. We found no statistical difference between results at 30 deg. C and 40 deg. C or at 70% and 90% relative humidity. We tested the gel under strong UV irradiation to simulate extreme environments. While a tropical noontime UV flux did not adversely affect the gel system decontamination of Cs-137 from concrete or tile coupons, the noontime flux resulted in significant dehydration of the gel after two hours of constant exposure. The rehydration of the gel by the humid air (at 90% humidity and 40 deg. C) was slow and may not be sufficient to offset dehydration by the sun. Americium decontamination from concrete was 70% after optimizing the phosphoric acid derivative and carbonate salt concentration in the wash solution. Since its development, we have had the opportunity to test the SuperGel in the removal of legacy contaminations in hot cell facilities and former glovebox facilities at Argonne. This has provided a unique opportunity to evaluate the SuperGel on a range of contaminants outside the original specifications for its use. We will report on the origin of the gel formulation, some highlighted experimental data including independent testing by the US EPA which has never been reported, and its efficacy for removing legacy alpha contaminations and its potential use for removal of chemical and biological hazardous agents. (authors)

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DOE OSTI2021

Halogenation of used aluminum matrix test reactor fuel – a bench-scale demonstration with surrogate materials

In this work, experiments with surrogate materials were performed at bench scale to demonstrate a halogenation technique applicable to treatment of used aluminum matrix test reactor fuel. The technique involves dissolution and separation of aluminum from used aluminum matrix test reactor fuel in molten-halide salt systems prior to treatment and disposition of the fuel’s uranium and fission products. Demonstration of the halogenation technique was performed with neodymium metal as a non-radiological surrogate for uranium metal. Experiments involved blending forms of aluminum and neodymium metal with ammonium and lithium chloride or ammonium and lithium bromide, which upon heating decomposed into ammonia gas and the respective hydrogen chloride or bromide gas. The latter reacted with the metals to form the respective aluminum and neodymium halides. At elevated temperatures, aluminum halides gasified away from the respective neodymium halides, which fused with their respective lithium halides. Samples of fused and distillate salts were collected and analyzed, yielding extents of aluminum removal that ranged from 94.5–98.2% for chlorination runs and 91.4–97.8% for bromination runs. No neodymium was detected in the distillate fractions. Some experiments were repeated with excess reactants, and a portion of aluminum chloride distillate was processed into a consolidated waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report provides a description of the testing (in progress at the Savannah River National Laboratory) to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The details of the establishment of this plate test, the interim corrosion results (from visual examination), and the general plan for examination of the plate in CY21 to evaluate crack growth behavior (if any) are provided. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage. A “large plate” (approximately 51 x 46 cm with thickness 16 mm and weight about 30 kg) and containing a circumferential weldment from the mockup is the test specimen. The knowledge of the salt, temperature, humidity conditions for chloride-induced stress corrosion cracking (CISCC) including the threshold stress intensity factor, K iscc , were used to setup the large plate test in which a set of part-through-wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A marine salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a sustained 71-73% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Design of the Universal Neutron Irradiator

SHINE medical isotopes is developing a novel method to produce the medical isotope Mo-99, which is the parent of Tc-99m, the most widely used diagnostic compound in nuclear medicine. This method utilizes U-235 fission in a subcritical liquid target solution, driven by an accelerator-based neutron source that is based on a deuterium-tritium reaction in a gaseous target. In this approach, the neutron-generating target is surrounded by uranyl sulfate (low-enriched uranium); fission in the uranium salt solution efficiently produces the desired Mo-99, as well as other fission products. After irradiation, Mo-99 is separated and the uranyl sulfate solution can be irradiated again to produce Mo-99. This allows the solution to be recycled for a very long time and significantly reduces the amount of associated nuclear waste. SHINE has received U.S. Nuclear Regulatory Commission (NRC) approval to construct an Mo-99 production facility based on this method. In order to produce Mo-99 efficiently, the solution should be as close as possible to criticality. Because the solution is over moderated, any precipitation could push the solution to become critical and thereby hard to control. One of the possible precipitation scenarios is the formation of uranyl peroxide. Under irradiation, water decomposes into the products of radiolysis, some of which will form peroxide. To prevent precipitation, one has to develop a way to decompose the peroxide before it can react with uranium and form hard-to-dissolve precipitates. To study peroxide formation and develop methods to prevent precipitation, we are developing an experiment capable of mimicking radiation conditions in the SHINE system. This report describes the irradiation setup we designed to irradiate a small volume of uranyl sulfate solution for the duration of one SHINE irradiation (~6 days) at the relevant fission power (~0.5 W/cc)

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
NASA NTRS1998

An Advanced Approach to Simultaneous Monitoring of Multiple Bacteria in Space

The utility of a novel microarray-based microbial analyzer was demonstrated by the rapid detection, imaging, and identification of a mixture of microorganisms found in a waste water sample from the Lunar-Mars Life Support Test Project through the synergistic combination of: (1) judicious RNA probe selection via algorithms developed by University of Houston scientists; (2) tuned surface chemistries developed by Baylor College of Medicine scientists to facilitate hybridization of rRNA targets to DNA probes under very low salt conditions, thereby minimizing secondary structure; and (3) integration of the microarray printing and detection/imaging instrumentation by Genometrix to complete the quantitative analysis of microorganism mixtures.

Eggers, M.↗
DOE OSTI2019

Ion–ion interactions enhance aluminum solubility in alkaline suspensions of nano-gibbsite (α-Al(OH) 3 ) with sodium nitrite/nitrate

Despite widespread industrial importance, predicting metal solubilities in highly concentrated, multicomponent aqueous solutions is difficult due to poorly understood ion–ion and ion–solvent interactions. Aluminum hydroxide solid phase solubility in concentrated sodium hydroxide (NaOH) solutions is one such case, with major implications for ore refining, as well as processing of radioactive waste stored at U.S. Department of Energy legacy sites, such as the Hanford Site, Washington State. The solubility of gibbsite (α-Al(OH) 3 ) is often not well predicted because other ions affect the activity of hydroxide (OH – ) and aluminate (Al(OH) 4 – ) anions. In the study reported herein, we systematically examined the influence of key anions, nitrite (NO 2 – ) and nitrate (NO 3 – ), as sodium salts on the solubility of α-Al(OH) 3 in NaOH solutions taking care to establish equilibrium from both under- and oversaturation. Rapid equilibration was enabled by use of a highly pure and crystalline synthetic nano-gibbsite of well-defined particle size and shape. Measured dissolved aluminum concentrations were compared with those predicted by an α-Al(OH) 3 solubility model derived for simple Al(OH) 4 – /OH – systems. Specific anion effects were expressed as an enhancement factor (Al enhc ) conveying the excess of dissolved aluminum. At 45 °C, NaNO 2 and NaNO 3 -containing systems exhibited Al enhc values of 2.70 and 1.88, respectively, indicating significant enhancement. The solutions were examined by Raman and high-field 27 Al NMR spectroscopy, indicating specific interactions including Al(OH) 4 – –Na + contact ion pairing and Al(OH) 4 – –NO 2 – /NO 3 – ion–ion interactions. Dynamic evolution of the α-Al(OH) 3 particles including growth and agglomeration was observed revealing the importance of dissolution/reprecipitation in establishing equilibrium. These studies indicate that incomplete ion hydration, as a result of the low water activity in these concentrated electrolytes, results in: (i) enhanced reactivity of the hydroxide ion with respect to α-Al(OH) 3 ; (ii) increased concentrations of Al(OH) 4 – in solution; and (iii) stronger ion–ion interactions that act to stabilize the supersaturated solutions. This information on the mechanisms by which α-Al(OH) 3 becomes supersaturated is essential for more energy-efficient aluminum processing technologies, including the treatment of millions of gallons of Al(OH) 4 – -rich high-level radioactive waste.

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

Annual Radioactive Waste Tank Inspection Program - 2018

The Department of Energy (DOE) requires that Savannah River Remediation LLC (SRR), the Liquid Waste contractor, maintain a robust and timely Structural Integrity Program for waste tanks that are leak-tight and for waste tanks that have leaked or are suspect. These requirements are promulgated in DOE Manual 435.1-1, Section II.Q.(2). A key aspect of SRR's implementation of these requirements is its annual tank inspection program. In 2018, the SRR Visual and In-Service Inspection Program used two visual imagery techniques to ascertain the integrity of the Savannah River Site (SRS) waste tanks: high resolution digital photography and closed-circuit television. The primary inspection method made use of direct photograph techniques, e.g., making a series of photographs providing detailed views of the tank and wide-angle photography for obtaining overviews of large areas. Closed-circuit television was then typically used to further investigate conditions found during the initial inspection. During 2018, a total of 4,853 photographs were taken and 47 video inspections were performed as part of this inspection program. Of the 51 waste tank systems at SRS that were designed, constructed, and placed into active service storing radioactive liquid waste, 43 remain in active service. Eight tanks have been cleaned, stabilized by filling with engineered grout, and removed from service (also known as operationally closed) -Tanks 5, 6, and 12 (Type I tanks); Tank 16 (Type II tank); and Tanks 17, 18, 19, and 20 (Type IV tanks). For the 43 waste tanks, inspections were successfully conducted in 2018 through all accessible annulus risers and inspection ports for nine Type I tanks, three Type II tanks, and 27 Type III/IIIA tanks; at least one inspection was made in the interior of four Type IV tanks (i.e., the tanks without annular spaces) above the liquid level in the tank. Rainwater continues periodically to be observed migrating into the annuli of some tanks. Evidence of this infiltration was primarily noted by observation of surface stains, and occasionally by calciferous deposits, changed configuration of salt deposits in the associated annulus, and mild surface corrosion. The conditions of the 43 tanks remained essentially unchanged from the conditions reported in 2017. The one key exception was the identification of eight new cracks in the Tank 15 primary tank wall. These newly identified cracks are located at 65, 67, 70, 76, 150, 152, 176 inches and 200 inches above the tank bottom. The identified cracks in Tank 15 do not constitute an impact on the structural integrity of the tank, as evaluated and confirmed in the referenced report, SRNS-STI-2008-00028, A Structural Impact Assessment of Flaws Detected During Ultrasonic Examination of Tank 15, and calculation, T-CLC-H-00890, Calculation of Critical Crack Lengths in HLW Tank 15. The 2018 inspection program confirmed the structural integrity and waste confinement capability of all 43 waste tanks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Clay, Water, and Salt: Controls on the Permeability of Fine-Grained Sedimentary Rocks

We report that the ability to predict the permeability of fine-grained soils, sediments, and sedimentary rocks is a fundamental challenge in the geosciences with potentially transformative implications in subsurface hydrology. In particular, fine-grained sedimentary rocks (shale, mudstone) constitute about two-thirds of the sedimentary rock mass and play important roles in three energy technologies: petroleum geology, geologic carbon sequestration, and radioactive waste management. The problem is a challenging one that requires understanding the properties of complex natural porous media on several length scales. One inherent length scale, referred to hereafter as the mesoscale, is associated with the assemblages of large grains of quartz, feldspar, and carbonates over distances of tens of micrometers. Its importance is highlighted by the existence of a threshold in the core scale mechanical properties and regional scale energy uses of shale formations at a clay content X clay ≈ 1/3, as predicted by an ideal packing model where a fine-grained clay matrix fills the gaps between the larger grains. A second important length scale, referred to hereafter as the nanoscale, is associated with the aggregation and swelling of clay particles (in particular, smectite clay minerals) over distances of tens of nanometers. Mesoscale phenomena that influence permeability are primarily mechanical and include, for example, the ability of contacts between large grains to prevent the compaction of the clay matrix. Nanoscale phenomena that influence permeability tend to be chemomechanical in nature, because they involve strong impacts of aqueous chemistry on clay swelling. The second length scale remains much less well characterized than the first, because of the inherent challenges associated with the study of strongly coupled nanoscale phenomena. Advanced models of the nanoscale properties of fine-grained media rely predominantly on the Derjaguin-Landau-Verwey-Overbeek (DLVO) theory, a mean field theory of colloidal interactions that accurately predicts clay swelling in a narrow range of conditions (low salinity, low compaction, Na + counterion). An important feature of clay swelling that is not predicted by these models is the coexistence, in most conditions of aqueous chemistry and dry bulk density, of two types of pores between parallel smectite particles: mesopores with a pore width of >3 nm that are controlled by long-range interactions (the osmotic swelling regime) and nanopores with a pore width <1 nm that are controlled by short-range interactions (the crystalline swelling regime). Finally, nanogeochemical characterization and simulation techniques, including coarse-grained and all-atom molecular dynamics simulations, hold significant promise for the development of advanced constitutive relations that predict this coexistence and its dependence on aqueous chemistry.

58 GEOSCIENCES↗
NASA NTRS1972

Design and development of a prototype wet oxidation system for the reclamation of water and the disposition of waste residues onboard space vehicles

Laboratory investigations to define optimum process conditions for oxidation of fecal/urine slurries were conducted in a one-liter batch reactor. The results of these tests formed the basis for the design, fabrication, and testing of an initial prototype system, including a 100-hour design verification test. Areas of further development were identified during this test. Development of a high pressure slurry pump, materials corrosion studies, oxygen supply trade studies, comparison of salt removal water recovery devices, ammonia removal investigation, development of a solids grinder, reactor design studies and bearing life tests, and development of shutoff valves and a back pressure regulator were undertaken. The development work has progressed to the point where a prototype system suitable for manned chamber testing can be fabricated and tested with a high degree of confidence of success.

Jagow, R. B.↗
DOE OSTI2023

Informing and Engaging the Future Workforce on Generation IV International Forum Reactor Systems

Generation IV reactors are six nuclear reactor designs, Sodium Fast reactor (SFR), Very High Temperature Reactor (VHTR), Gas cooled Fast Reactor (GFR), Molten Salt Reactor (MSR), Lead Fast Reactor (LFR), and Supercritical Water Reactor (SCWR), that are considered to be the most promising in light of various criteria based on the following objectives: • continuation of the progress made by Generation III water reactors in terms of competitiveness and safety; • more effective use of uranium resources; • less radioactive waste, especially high-level, long-lived waste; • greater protection against malicious acts and the diversion or theft of nuclear materials. The GIF Education and Training Working Group (ETWG) launched two major initiatives in 2016 and in 2021 to not only inform and educate but also to engage the future workforce in support of these reactor systems. This paper will present an overview of the GIF EWTG webinar series as well as the 2021 Pitch your Gen IV Research Competition which is followed by the 2023 Pitch your Gen IV Research. Competition

Paviet, Patricia D.↗
DOE OSTI2020

Nanoscale TiO 2 coating improves water stability of Cs 2 SnCl 6

To improve the stability of Cs 2 SnCl 6 under aqueous/moisture environments, we applied a concept of artificial passivation by depositing a protective TiO 2 coating of 10 nm on the surface of Cs 2 SnCl 6 . Static leaching experiments results indicate that the initial release rates of Cs + and Cl - are decreased by 20–30 times with TiO 2 coating, suggesting its possibility to improve the short-term water/environmental stability of Cs 2 SnCl 6 . An amorphous-to-crystalline phase transition in TiO 2 film was observed, possibly resulting in degradation of Cs 2 SnCl 6 . However, the crystalline TiO 2 film still remains after 21 days water exposure and can still act as an effective passivation layer to reduce the release rates of Cs + and Cl - by as much as about 17 and 7 times, respectively, relative to static leaching without artificial coatings. Therefore, the water/environmental stability of metal halide perovskite Cs 2 SnCl 6 , which is a highly soluble molecular salt, can be enhanced by the nanoscale TiO 2 coating as an artificial passivation film.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

FY21 Status Report: Probabilistic SCC Model for SNF Dry Storage Canisters

Stress corrosion cracking (SCC) is an important failure degradation mechanism for storage of spent nuclear fuel. Since 2014, Sandia National Laboratories has been developing a probabilistic methodology for predicting SCC. The model is intended to provide qualitative assessment of data needs, model sensitivities, and future model development. In fiscal year 2021, improvement of the SCC model focused on the salt deposition, maximum pit size, and crack growth rate models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Apparatus and method for stripping tritium from molten salt

A method of stripping tritium from flowing stream of molten salt includes providing a tritium-separating membrane structure having a porous support, a nanoporous structural metal-ion diffusion barrier layer, and a gas-tight, nonporous palladium-bearing separative layer, directing the flowing stream of molten salt into contact with the palladium-bearing layer so that tritium contained within the molten salt is transported through the tritium-separating membrane structure, and contacting a sweep gas with the porous support for collecting the tritium.

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