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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Gas generation during waste treatment of acidic solutions from the dissolution of irradiated LEU targets for 99 Mo production

The goal of the Reduced Enrichment for Research and Test Reactors Program is to limit the use of high-enriched uranium (HEU) in research and test reactors by substituting low-enriched uranium (LEU) wherever possible. The work reported here documents our work to develop the calcining technologies and processes that will be needed for 99 Mo production using LEU foil targets and the Modified Cintichem Process. The primary concern with the conversion to LEU from HEU targets is that it would result in a five- to six-fold increase in the total uranium. This increase results in more liquid waste from the process. We have been working to minimize the increase in liquid waste and to minimize the impact of any change in liquid waste. Direct calcination of uranium-rich nitric acid solutions generates NO 2 gas and UO 3 solid. We have proposed two processes for treating the liquid waste from a Modified Cintichem Process with a LEU foil. One is an optimized direct calcination process that is similar to the process currently in use. The other is a uranyl oxalate precipitation process. The specific goal of the work reported here was to characterize and compare the chemical reactions that occur during these two processes. In particular, the amounts and compositions of the gaseous and solid products were of interest. A series of experiments was carried out to show the effects of temperature and the redox potential of the reaction atmosphere. The primary products of the direct calcination process were mixtures of U 3 O 8 and UO 3 solids and NO 2 gas. The primary products of the oxalate precipitation process were mixtures of U 3 O 8 and UO 2 solid and CO 2 gas. Higher temperature and a reducing atmosphere tended to favor quadrivalent over hexavalent uranium in the solid product. These data will help producers to decide between the two processes. In addition, the data can be used to design appropriate off-gas systems for pilot and production facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of a Self-Consistent Model of Plutonium Sorption: Quantification of Sorption Enthalpy and Ligand-Promoted Dissolution

This university lead SBR project is a collaboration lead by Dr. Brian Powell (Clemson University) with co-principal investigators Dan Kaplan (Savannah River National Laboratory), Yuji Arai (presently at the University of Illinois), Udo Becker (U of Michigan) and Rod Ewing (presently at Stanford University). Hypothesis: The underlying hypothesis of this work is that strong interactions of plutonium with mineral surfaces are due to formation of inner sphere complexes with a limited number of high-energy surface sites, which results in sorption hysteresis where Pu(IV) is the predominant sorbed oxidation state. The energetic favorability of the Pu(IV) surface complex is strongly influenced by positive sorption entropies, which are mechanistically driven by displacement of solvating water molecules from the actinide and mineral surface during sorption. Objectives: The overarching objective of this work is to examine Pu(IV) and Pu(V) sorption to pure metal (oxyhydr)oxide minerals and sediments using variable temperature batch sorption, X-ray absorption spectroscopy, electron microscopy, and quantum-mechanical and empirical-potential calculations. The data will be compiled into a self-consistent surface complexation model. The novelty of this effort lies largely in the manner the information from these measurements and calculations will be combined into a model that will be used to evaluate the thermodynamics of plutonium sorption reactions as well as predict sorption of plutonium to sediments from DOE sites using a component additivity approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Optimized Dissolution Flowsheet for Uranium Metal

The U.S. Department of Energy’s National Nuclear Security Administration (NNSA) has entered into cooperative agreements with commercial businesses to establish a reliable domestic source of 99 Mo that will be produced without the use of highly enriched uranium (HEU).

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DOE OSTI2019

Dissolution of Light Curium Oxide with a Catalyzed Electrolytic Process

The ability to produce transcurium elements and isotopes, such as 249 Bk and 252 Cf, requires the appropriate feed material and a high thermal neutron flux. Currently, at Oak Ridge National Laboratory (ORNL), the High Flux Isotope Reactor (HFIR) provides one of the world’s highest steady state thermal neutron fluxes. The feed material consists of recycled curium from the decades of production of 249 Bk, 252 Cf, 253 Es, and 255/257 Fm at the Radiochemical Engineering Development Center (REDC). Over time, the feed material has been depleted and needs to be replenished. The current feed material is called “heavy recycled curium” and consists of over 50 wt% of 246/248 Cm isotopes. In order to replenish this stock material, additional curium is needed. The only curium material that is currently available consists predominantly (greater than 80 wt%) of 244 Cm. This material also contains 240 Pu from the radioactive decay of the 244 Cm. Because of the amount of 244 Cm is greater than 50% of the total Cm, this material is termed “light curium.” Today, 249 Bk and 252 Cf are produced in only two places in the world: the United States and Russia. In the United States, the US Department of Energy and its predecessors have made these isotopes and others available to research agencies and commercial users since the late 1960s as part of the Department of Energy heavy-element production program located at ORNL. Currently these materials are provided through the DOE Office of Science Nuclear Physics-Isotope Development and Production for Research and Applications and the Basic Energy Science-Heavy Element Chemistry. Feedstock for the production of 252 Cf and other transcurium element by-product isotopes was originally prepared by means of irradiation of multi-kilograms of 239 Pu in the large production reactors at the Savannah River Site. Products of the Savannah River Site irradiation were predominantly 242 Pu, 243 Am, and 244 Cm. These materials were shipped to ORNL for production of multi-milligram amounts per year of 252 Cf and other transcurium element isotopes by means of irradiation in the HFIR, followed by separation and purification processing in the Radiochemical Engineering Development Center (REDC). The transcurium element isotopes produced were distributed to other US and collaborating foreign laboratories for research studies to determine the nuclear and chemical properties of these rare materials. The Cm feed material plays a key role in the efficient production of 252 Cf. The early targets irradiated in HFIR contained only the 242 Pu, 243 Am, and 244 Cm produced at Savannah River Site, and the rate of production of 252 Cf was less than 1 mg per 8 g of target material as shown in Figure 1. Also, shown in Figure 1 is the significant difference in 252 Cf production between the light and heavy curium; it differs by approximately 2 orders of magnitude.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Characterization of Tank 9h Salt Dissolution Batch 2C In Support of Tank Closure Cesium Removal (TCCR) 1A Batch 2 Preparations

Savannah River Mission Completion (SRMC) is currently preparing the second batch of material to be processed through the Tank Closure Cesium Removal (TCCR) 1A system. The feed for TCCR 1A consists of dissolved saltcake from Tank 9H. The third batch of salt to make up processing Batch 2 (Batch 2C) has recently been dissolved in Tank 9H and transferred to Tank 10H where it was composited with the first part of the batch (Batches 2A and 2B) in preparation for processing through the TCCR 1A unit. Savannah River National Laboratory (SRNL) received samples from the recent batch (2C) of dissolved salt for characterization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Testing of the Defense Waste Processing Facility Cold Chemical Dissolution Method in Sludge Batch 10 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) tests the applicability of the digestion methods used by the DWPF Laboratory for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) Receipt samples and SRAT Product process control samples. DWPF SRAT samples are typically dissolved using a method referred to as the DWPF Cold Chemical or Cold Chem Method (CC), (see DWPF Procedure SW4-15.201- 2.1).

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DOE OSTI2022

Removal of Tc-99, Zr-95, and Nb-95, From Solutions Obtained After Dissolution of Irradiated Mo Targets

Decontamination of 99 Tc, Nb, and Zr impurities in dissolved irradiated Mo disks can be accomplished using Fe(II) precipitation. Here, we report bench-scale experiments and large-scale demonstrations on removal of 99 Tc (0.1 mM), 95 Nb, and 95 Zr from dissolved Mo disk simulant solutions. Tc removal is accomplished through the reduction and simultaneous immobilization of Tc(VII) by Fe(II), producing an insoluble, Tcincorporated Fe(II)-Fe(III) solid that is removed by filtration. Large-scale testing showed that Tc (97.6% ± 1.7%) is removed without affecting 99 Mo yields. Other important side reaction product impurities in irradiated Mo targets, namely various isotopes of Nb and Zr, are also removed (>99%) in this process. The effects of metal (Fe, Tc, Mo) concentrations, Fe(II)/Fe(III) ratio, pH, temperature, Fe addition method, and hydrogen peroxide addition were tested. The minimum amount of Fe(II) needed to remove up to 99% Tc is 10 mM and the optimal pH value for simultaneous removal of Tc, Nb, and Zr impurities is pH 13.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Low Pressure Oxygen-Assisted Acidic Dissolution of U Metal Foils

Traditional methods of dissolving uranium in nitric acid involve the generation of nitrogen oxide gases (NO x ), which not only are hazardous but also contribute to increased pressure within the dissolver. When uranium metal foils are dissolved in nitric acid, a series of redox reactions occurs, leading to the formation of NO x , such as nitrogen dioxide (NO 2 ) and nitric oxide (NO).

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