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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Sulfur-functionalized solid-phase materials for the selective separation of arsenic and selenium

Radioactive arsenic (As) isotopes are of growing interest for applications in nuclear medicine, national security, and environmental research. Recent efforts at the Facility for Rare Isotope Beams (FRIB) have focused on aqueous harvesting of selenium-72,73 ( 72,73 Se) and their daughter isotopes, arsenic-72,73 ( 72,73 As), which are particularly valuable for medical applications and nuclear data studies, respectively. Both conventional isotope production and harvesting methods require chemical separations to purify radioactive As from parent and co-produced Se radioisotopes. While several solid-phase separation methods for As and Se exist, many depend on complex oxidation state control or highly acidic conditions. This study presents results for sulfur-based solid-phase materials selected to enable uptake at lower acidity and eliminate the need for intricate redox chemistry. Specifically, the performance of three covalently bound sulfur-based ligands were evaluated: (1) thiophenol-polystyrene, (2) propanethiol-silica, and (3) thiourea-silica. Uptake characteristics—including distribution coefficients (Dw), kinetics, and column separation behavior—were assessed using 75 Se and 73 As in hydrochloric (HCl) acid and nitric (HNO 3 ) solutions. The resins demonstrated high-yield (>95%) and high-purity As recovery across a range of HCl concentrations. Comparable results in HNO 3 were achieved when combined with anion exchange chromatography. Furthermore, the potential application of these materials for medical isotope generators was also investigated through ligand stability and repeated elution studies. Overall, sulfur leaching from the resins was negligible at the concentrations relevant for these separations but increased with higher acid concentrations.

Arsenic↗
DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Demonstration of Defense Waste Processing Facility (DWPF) Higher Fissile Content Glass

The Accelerated Basin De-inventory (ABD) Program has been proposed as an alternative for future spent nuclear fuel (SNF) and nuclear material processing at the Savannah River Site (SRS). This approach would change the baseline H-Canyon (HCAN), Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The ABD Program would require that all domestic and foreign research reactor SNF currently at SRS be dissolved, stored, and then transferred to CSTF without the recovery of uranium. Preliminary assessments in the ABD Program plan have shown that ~5000 extra SRS high-level waste (HLW) canisters would be produced if the fissile mass loading remains at the current 897 g/m 3 limit; however, increasing the limit to 2500 g/m 3 would result in ~520 extra canisters. Thus, the ABD Program plan requires an increase of the DWPF fissile mass loading limit to 2500 g/m 3 to minimize canister production. DWPF considers the following isotopes in the calculation of fissile mass loading: U-233, U-235, Pu-239 and Pu-241.

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

Global Security & Strategic Partnerships Programs

Serving the Nation: Our Missions include Providing the Nuclear Deterrent for our Nation and Allies; Leading Nuclear Nonproliferation and Counterterrorism / Counterproliferation efforts to make the world a safer place; Supplying Enriched Uranium Material for Reactors – Naval Propulsion, Army, NASA, Universities, Isotope Production Reactors; Providing High Explosives for the Stockpile and Strategic Partners; Delivering Solutions – Solve emerging national security challenges

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Radioisotope production at the Spallation Neutron Source: Design concept of experimental target station

Completion of the Proton Power Upgrade Project for the Spallation Neutron Source (SNS) accelerator at Oak Ridge National Laboratory opens an opportunity to utilize reserve beam power of more than 100 kW for applications beyond neutron production. One of these applications is the production of critical radionuclides. To demonstrate the feasibility of using the reserve beam power to produce radioisotope at SNS, a design concept of a small-scale experimental target station in the Linac Dump area has been developed. This experimental facility will provide isotope yield benchmarking data using protons in the GeV range. It will also enable additional research and development in isotope handling and radiochemical separation. The target station consists of a target module enclosed in a vessel and concrete shielding. Particle transport calculations and thermo-mechanical simulations are used to determine beam parameters, decay time, isotope yield, shielding dimensions, and target design parameters. Calculations verified that the irradiated capsule can be handled manually using hands-off tools and transported to a hot cell in a shielded container for post-irradiation characterizations.

Lee, Yong Joong [ORNL] (ORCID:0000000298381723)↗
DOE OSTI2025

Producing 236 U reference standards for Accelerator Mass Spectrometry at the University of Notre Dame

36 U is a rare isotope of uranium, naturally occurring in ores with an abundance of 236 U/ 238 U$<$ 1 x 10 -9 . The ability to detect it and make isotopic ratio measurements has applications ranging from nuclear forensics and nonproliferation to energy production and environmental protection. Currently, Accelerator Mass Spectrometry (AMS) is the only technique sensitive enough to accurately measure 236 U/ 238 U isotopic ratios as they exist in naturally occurring ores in the range of 236 U/ 238 U = 10 -12 $-$ 10 -9 . Some AMS facilities have demonstrated their capabilities to make these measurements. Historically, the lack of commercially available reference standards covering the range of naturally occurring 236 U/ 238 U abundances has necessitated the use of absolute measurements, notoriously difficult to do using AMS, resulting in increased uncertainties in measurements and a reliance on knowledge of systematic effects. To mitigate these issues, various AMS facilities have sought to develop their own reference standards. Using a reference standard prepared for other forms of mass spectrometry, a series of AMS suitable standards was created through dilution with low-background natural uranium. The techniques used to produce and characterize these materials as well as analysis of them using AMS will be discussed.

236U↗
DOE OSTI2025

Risk Analysis of Radiological Release from Pu-238 Targets During Manual Handling

Pu-238 Isotope Production Targets are routinely installed in the Advanced Test Reactor (ATR) core, transferred, and stored in the spent fuel canal. These evolutions involve manual handling and manipulation of the targets underwater using long handled tools. The ATR Safety Analysis Report (SAR) postulates a design basis accident which results in damage from manual manipulation of targets, and radiological consequences must be determined for receptors inside the reactor facility, as well as public receptors. This presentation presents the analysis used to determine the radiological consequences due to potential target damage in the ATR canal. The analysis considered radionuclide release fractions, damage ratios for handling evolutions, and entrainment of radionuclides in the canal water.

Advanced Test Reactor↗
DOE OSTI2024

The LLNL rabbit: determination of trace element levels in polyethylene

Rabbit facilities are systems that pneumatically transport a sample carrier (known as a rabbit) to and from a radiation source. However, trace impurities in the polyethylene (PE) rabbit may interfere with measurements of short-lived isotopes. Activated impurities with short half-lives as 28 Al interfere with the fission product γ-ray, therefore the rabbit cannot be measured immediately. Neutron activation analysis was performed on various PE samples obtained from different manufacturers and the highest quality PE was used to fabricate a high-purity PE rabbit. Here, the impurities identified in the PE study were 27 Al, 48 Ca, 37 Cl, 50 Cr, 65 Cu, 127 I, 26 Mg, 55 Mn, and 30 Si.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Polyethylene Upcycling to Liquid Alkanes in Molten Salts under Neat and External Hydrogen Source-Free Conditions

Development of facile approaches to convert plastic waste into liquid fuels under neat conditions is highly desired but challenging, particularly without noble metal catalysts and an external hydrogen source. Herein, highly efficient and selective polyethylene-to-gasoline oil (branched C 6 –C 12 alkanes) conversion was achieved under mild conditions (<170 °C) using commercially available AlCl 3 -containing molten salts as reaction media and to provide catalytic sites (no extra solvents, additives, or hydrogen feeding). The high catalytic efficiency and selectivity was ensured by the abundant active Al sites with strong Lewis acidity (comparable to the Al type in acidic zeolite) and highly ionic nature of the molten salts to stabilize the carbenium intermediates. Dynamic genesis of the Al sites was elucidated via time-resolved Al K-edge soft X-ray and 27 Al NMR, confirming the tricoordinated Al 3+ as active sites and its coordination with the as-generated alkene/aromatic intermediates. Further, the carbenium formation and polyethylene chain variation was illustrated by inelastic neutron scattering (INS) and an isotope-labeling experiment. Theoretical simulations further demonstrated the successive hydride abstraction, β-scission, isomerization, and internal hydrogen transfer reaction pathway with AlCl 3 as active sites. This facile catalytic system can further achieve the conversion of robust, densely assembled, and high molecular weight plastic model compounds to liquid alkane products in the diesel range.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

A Model for Radiolysis in a Flowing-Water Target during High-Intensity Proton Irradiation

At the Facility for Rare Isotope Beams (FRIB), interactions between heavy-ion beams and beam-dump water will create a wide variety of radionuclides which can be accessed by a technique known as “isotope harvesting”. However, irradiation of water is always accompanied by the creation of numerous radical, ionic, and molecular radiolysis products. Some of the radiolysis products have sufficiently long lifetimes to accumulate in the irradiated water and affect the harvesting chemistry. Here we investigate the formation of hydrogen peroxide, molecular hydrogen, and molecular oxygen during a high-intensity proton irradiation of a flowing-water isotope-harvesting target and compare the experimental results to simulations. The simulations kinetically model the chemical reactions occurring in the homogeneous phase of radiolysis in flowing water and establish an “effective yield”. In both the experiment and simulations, the bulk quantities of H 2 , H 2 O 2 , and O 2 are considerably lower than predicted by primary radiolysis yields (escape yields), meaning that in the high beam intensity regime the homogeneous phase reactions have a considerable impact on the overall chemical composition of the water. Further, it could be shown that for radiation which is characterized by a limited linear energy transfer, such as the here applied protons, the bulk outcome of the microscopic kinetic modeling could be estimated by a simplified steady-state model.

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

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

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

Large-scale production of 88 Y and 88 Zr/ 88 Y generators: A proof of concept study for a 70 MeV H – cyclotron

Here, the large-scale production of 88 Y with proton-induced reactions has been investigated from the perspective of new generation 70 MeV H – cyclotrons. Tandem target configurations are presented for both the direct production of 88 Y as well as for producing 88 Zr/ 88 Y generators. Based on the relevant excitation functions, physical yields have been derived for 88 Y production with Y 2 O 3 /SrCO 3 tandem targets and 88 Zr production with Zr/Y 2 O 3 tandem targets. Yields are presented for optimized targets (i.e. optimum yield) as well as for balanced thermal loads on the individual targets. Liquid 88 Zr/ 88 Y generators have been produced using both natural Zr and Nb target materials, the former for dedicated productions and the latter as a byproduct by processing spent irradiated Nb capsules which normally would constitute radioactive waste. These stock solutions, which contain both the target material and 88 Zr precursor, are retained virtually unchanged after processing except for the removal of 88 Y on AG MP-50 macroporous cation-exchange resin. Methods are presented for the preparation of Nb stock solutions in hydrofluoric acid and Zr stock solutions in sulfuric acid. It is shown that multi-Ci productions of 88 Y are feasible at a 70 MeV cyclotron facility, suitable for the needs of fracking applications. In addition, 88 Zr/ 88 Y generators can provide 88 Y with very high specific activity, suitable for labelling of biomolecules.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2026

Assessment of rare earth element fractionation in NIF implosions with radiochemically doped capsules

An ongoing experimental campaign at the National Ignition Facility (NIF) aims to measure neutron induced nuclear reaction cross–sections using radiochemically doped target capsules. Critical to this campaign is the ability to collect a representative sample of the reaction products using Solid Radiochemical Collectors (SRCs) fielded around the NIF chamber. The shot presented in this paper used a doped target capsule with a neopentane gas fill was to investigate the ratio of isotopes collected at three chamber angles. It was found that SRC samples of rare earth elements collected from NIF are representative of the ingoing dopant mix, thereby concluding that fractionation does not occur during a NIF implosion. This validates the doped capsule method for use in measuring neutron induced reaction cross-sections.

Capsule Doping↗
DOE OSTI2023

Production of Precursor Materials for the Development of Iron-Based Superconductor (IBS) Wire

The objective of the activities under this Annex is to address the extrinsic Jc limitations uncovered in pioneering work. In order to engineer the defects pinning landscape charged particle irradiations experiments on bulk samples, different doping and chemical additions will be performed at different National facilities (Test Beam Facility at Fermilab; Linac Isotope Producer at Brookhaven National Laboratory; Integrated Biology Laboratory at University of Wisconsin-Madison; Facility for Rare Isotope Beams at Michigan State University). The samples will be produced at ENEA by different techniques. Magnetization and advanced microscopy measures will be performed at Fermilab. After irradiation, depending on the level of radioactivity, the samples will be tested at Fermilab or where it is possible to measure radioactive materials as University of Wisconsin-Madison.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Oxygen stable isotopes in the nuclear fuel cycle: Assessment of the potential for determining the fabrication and provenance history of anhydrous and hydrous uranium oxides

Determining the origin and history of interdicted nuclear materials is a central challenge in nuclear forensics. The oxygen stable isotope composition of uranium oxide compounds has emerged as a promising forensic signature, attracting increasing attention since the early 2000s. This review examines analytical techniques for measuring oxygen isotope compositions in uranium oxides and evaluates how the nuclear fuel production cycle introduces or modifies these isotopic signatures. The potential for forensic geolocation is explored through workflows that calibrate the relationship between environmental water oxygen isotopes and those found in uranium oxides. Key strengths and limitations of this approach are assessed, including gaps in knowledge related to isotope fractionation during specific stages of the fuel cycle, and processing facility water inputs. The importance of proper sample handling and storage under inert atmospheres, as well as a deeper understanding of both intra-sample oxygen isotope heterogeneity, and hydrous uranium oxide phase formation, is highlighted for improving the reliability of forensic interpretations. In conclusion, the development of uranium oxide standards with well-characterized δ 18 O values and international collaboration toward consensus on their use are identified as essential steps for advancing the field.

Attribution↗
DOE OSTI2023

Cubical moderators for the Neutron Target Demonstator (NTD)

Neutron capture cross sections of unstable isotopes are important for neutron induced nucleosynthesis as well as for technological applications. The traditional time-of-flight method reaches its limits once the necessary detection of the reaction products is hampered by the size (mass) of the sample. Several factors may limit the sample mass: The decay properties of radioactive isotopes interfere with the signals from the neutron capture or neutron-induced fission reactions; The limited range of charged reaction products requires a thin sample. In both cases, an increased neutron fluence at the sample position with ever improved neutron sources overcomes the lack of reaction rate. Reference proposed a combination of a radioactive beam facility, an ion storage ring and a spallation neutron source surrounded by neutron moderator. This idea is part of the LANSCE strategy to stay a worldwide leader for neutron induced research. Figure 1 shows a sketch of the proposed future setup.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Production of Precursor Materials for the Development of Iron-Based Superconductor (IBS) Wire

The objective of the activities under this Annex is to address the extrinsic Jc limitations uncovered in pioneering work. In order to engineer the defects pinning landscape charged particle irradiations experiments on bulk samples with different doping and chemical additions will be performed at different National facilities (Test Beam Facility at Fermilab; Linac Isotope Producer at Brookhaven National Laboratory; Integrated Biology Laboratory at University of Wisconsin-Madison; Facility for Rare Isotope Beams at Michigan State University). The samples will be produced at TUAT by different synthesis techniques. Magnetization and advanced microscopy measures will be performed at Fermilab. After irradiation, depending on the level of radioactivity, the samples will be tested at Fermilab or where it is possible to measure radioactive materials as University of Wisconsin-Madison. In parallel, the precursor materials produced by TUAT will be used for IBS-wire fabrication at Fermi

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Dose reconstruction for plutonium-239 intakes at the Rocky Flats Plant

Purpose: The Rocky Flats (RF) Plant was a weapons manufacturing facility that operated from the early 1950s to 1989. Its primary missions were the production of plutonium (Pu) pits for thermonuclear weapons and the processing of retired weapons for Pu recovery. The purpose of this study was to estimate radiation doses to a cohort of 4499 RF workers from an intake of 239 Pu, the primary plutonium isotope handled at the site. Materials and methods: The latest biokinetic models of the International Commission on Radiological Protection, or site-specific variations of those models, were used to estimate 239 Pu intakes for each worker based on model fits to bioassay data often coupled with lung measurements. Results: Urinary excretion and lung retention data for most 239 Pu intakes could be fit reasonably well by a mixture of Pu dioxide and moderately soluble material. For some workers, better fits were obtained by application of other absorption types including Type S, 239 Pu nitrate, or pure 239 Pu dioxide, or by assuming intake via a wound. The lungs typically received the highest tissue doses, with fifty-year committed equivalent doses in the range of 0.5–1 Sv for 275 workers, 1–5 Sv for 115 workers, and greater than 5 Sv for 12 workers. Conclusions: RF was a unique site regarding a large number of lung measurements available for determining the appropriate absorption types for inhaled material. Further, this provided higher confidence in reconstructed 239 Pu doses than is generally gained from urinary data alone.

61 RADIATION PROTECTION AND DOSIMETRY↗