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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Reuse of Ion Exchange Resins for Reduction of Spent Radioactive Resins - 17125

A process is proposed to reuse ion exchange resins. The process consists of ion exchange beds for the ion exchange and regeneration, regeneration solution (regenerant) make-up tanks, an evaporator for concentration of spent regenerant, a dryer for the drying of concentrate, a pulverizer for grinding dried salt and a mixer for polymer solidification of the salt. A spent resins storage tank and a spent regenerant storage tank are also involved in the process. The role of each component in the process for reuse of the resins is explained. In addition, the effect of the reuse of ion exchange resins by regeneration was estimated in view of the reduction of spent resins. As the number of regenerations increases, the volume of spent resins is obviously inversely proportional to the number of regenerations. Although a polymer waste form is generated instead of spent resins when the ion exchange resins are reused, its volume is only about one-fifth of that of spent resins. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Infusible thermoplastic resin based sandwich structures for wind blade applications and the influence of scrim on facesheet to core interface debonding

Infusible thermoplastic Elium® family of resins from Arkema have garnered much attention in recent years as a possible replacement for thermoset resins in laminate and sandwich composite manufacturing for wind blade applications due to its ease of recyclability and the ability to utilize existing manufacturing processes without imposing complicated variations. However, physical and mechanical properties of the proposed Elium® based thermoplastic composites must be comparable to existing epoxy (thermoset) based composites using manufacturing processes relevant for large wind turbine blades. A 13-meter-long demonstration blade was manufactured for that purpose and sandwich samples were obtained from that project for a detailed study. This paper details three-point flexural properties of unidirectional E-glass fiber reinforced acrylic and epoxy based sandwich panels with identical balsa wood core materials. In addition, to evaluate the relative merit considering debond failure mode, the interfacial critical strain energy release rate, predominantly in mode-1, was compared via single cantilever beam testing. In sandwich composites constructed with balsa wood core material, resin uptake by the balsa core is traditionally impeded via the insertion of a scrim material at the facesheet to core interface. Results revealed that inclusion of scrim mesh layer at the facesheet to core interface reduced flexural properties and strain energy release rates in panels infused with acrylic resin but did not significantly reduce these properties in epoxy infused facesheets.

Engineering↗
DOE OSTI2018

Tandem column isolation of zirconium-89 from cyclotron bombarded yttrium targets using an automated fluidic platform: Anion exchange to hydroxamate resin columns

The development of a tandem column purification method for the production of high-purity 89Zr is presented. The primary column was a macroporous strongly basic anion exchange resin on styrene divinylbenzene co-polymer. The secondary column, with an internal volume of ~27 µL, was packed with hydroxamate resin. A condition of inverted selectivity was developed, whereby the 89Zr eluent from the primary column is synonymous with the load condition on the secondary column. The ability to transfer 89Zr from one column to the next allows two sequential column clean-up methods to be performed prior to the final elution of the 89Zr. This approach assures delivery of high purity 89Zr product and assures a 89Zr product that is eluted in a substantially smaller volume than is possible when using the traditionally-employed single hydroxamate resin column method. The tandem column purification process has been implemented into a prototype automated fluidic system. The system is configured with on-line gamma detection so column effluents can be monitored in near-real time. The automated method was tested using seven cyclotron bombarded Y foil targets. It was found that 95.1±1.3% of the 89Zr present in the foils was recovered in the secondary column elution fraction. Furthermore, elution peak analysis of several 89Zr elution profile radiochromatograms made possible the determination of 89Zr recovery as a function of volume; a 89Zr product volume that contains 90% of the mean secondary column elution peak can be obtained in 0.29±0.06 mL (representing 86±5% of the 89Zr activity in the target). This product volume represents a significant increase in isotope product concentration over the predominant method used in the field. In addition to the reduced 89Zr product elution volume, titrations of the 89Zr product with deferoxamine mesylate salt across two preparatory methods resulted in mean effective specific activity (ESA) values of 279 and 340 TBq·mmole-1 and mean bindable metals concentrations ([MB]) of 13.5 and 16.7 nmole·g-1. These ESA and [MB] values infer that the 89Zr product resulting from this tandem column isolation method has the highest purity reported to date.

immunoPET, zirconium-89, laboratory automation, de↗
DOE OSTI2020

Characterization of Spent Purolite A530E Resin with Implications for Long-Term Radioactive Contaminant Removal

Direct removal of pertechnetate anion (TcO4-) from the subsurface contaminated plumes at legacy nuclear weapons production sites, is effective remediation strategy to prevent their spread to adjacent aquifers but is also challenging because TcO4- is a trace component of the contaminated groundwater typically containing large access of other anionic constituents, such nitrate, sulfate, chloride and others. It can be achieved through ion exchange treatment using Purolite A530E resin which to date however has only been evaluated under laboratory conditions and many questions regarding its long-term operational performance remain. To address this need, here we report comprehensive characterization of the spent Purolite A530E resin which processed over 5.38x109 L of contaminated groundwater and successfully removed about 3.78 Ci of Tc-99 during four years of operation at the 200 West Pump & Treat facility at the US DOE Hanford site. This Tc-99 loading constitutes however only about 0.034% of the theoretical capacity of the resin which retained significant amounts of sulfate. Among other radioactive contaminants, small quantities of U-238, Co-60, and I-129 were retained by the resin. The total loading of iodine (combined 1-127 and I-129) on the spent resin exceeded that of Tc-99. To elucidate the mechanism of iodine retention, ion exchange behavior of iodide and iodate was investigated. Purolite A530E resin exhibited highly efficient uptake of iodide and only moderate affinity for iodate in accord with their Gibbs energy of hydration. Sorption isotherms for both anions obeyed Freundlich model.

Levitskaia, Tatiana G.↗
DOE OSTI2023

Iodine Removal from Carbonate-Containing Alkaline Liquids Using Strong Base Resins, Hybrid Resins, and Silver Precipitation

The ability of several material types to remove aqueous iodine from a mildly alkaline, carbonate-rich nuclear waste stream was evaluated: strong base anion exchange resins (SBAR), hybrid resins, Ag-containing materials and Bi-containing hybrid resins. A combination of batch testing and flow through column testing were used in the evaluation. In batch testing, hybrid resins CHM-20, SIR-110-CE and RTBI were shown to have high efficiency for the removal of both iodide and iodate simultaneously, while Ag-containing materials and SBAR demonstrated high capacity for iodide removal. One example of each material type (CHM-20, A532E and Ionex 400) were further evaluated for their sorption isotherms and column performance. The Langmuir isotherm, or a Langmuir-Freundlich hybrid isotherm, best described the sorption of iodide to the CHM-20 hybrid resin and Purolite A532E. The Freundlich isotherm best described the uptake of iodate to CHM-20 and A532E and for both iodide and iodate to the Ag-containing Ionex-400. In column testing the Purolite A532E had exceptional performance for overall iodide removal. With the capacity demonstrated the A532E resin would exceed Class C waste classification before breakthrough initiated, and column change outs in processing would be dictated by eventual waste classification, not breakthrough. The Ionex 400, a Ag-zeolite, was observed to degrade over time in the column in the mild alkaline conditions whereas the hybrid CHM-20 was limited in the single pass through design and would be best suited for applications where iodide and iodate are present and recirculation of the column effluent is feasible. Here this work highlights the feasibility of commercially available materials to separate radioiodine from liquid environments.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Applications of MP-1 anion exchange resin and Eichrom LN resin in berkelium-249 purification

Purification of berkelium-249 ( 249 Bk) at Oak Ridge National Laboratory requires a multi-step process with Bk Finishing as the final step. The current multi-process takes 6–7 months, including ~ 3 months spent on Bk Finishing. Recent trials of using columns of either LN resin or MP-1 anion exchange resin adsorbing the tetravalent 249 Bk4+ in 8M HNO 3 -0.5M NaBrO 3 , followed by eluting impurities (e.g. Ce 4+ , Cf 3+ ) and stripping 249 Bk resulted in a new dual column method of selectively separating 249 Bk from other trans-curium elements and fission product impurities, which is proposed to replace the current Bk Finishing with higher Bk purification efficiency and shorter processing period. This paper displays the experimental results and describes the dual column method with relevant chemistry issues.

Du, Miting (ORCID:0000000336741567)↗
DOE OSTI2025

Loading Capacity and Dilute Nitric Acid Rinse of Diglycolamide Resin for the Recovery of Transplutonium and Rare Earth Elements from Mark-18A Targets

N,N,N′,N′-tetraoctyldiglycolamide (TODGA) as a resin (“DGA resin”) produced by Eichrom Technologies will be used by Savannah River National Laboratory for the indiscriminate extraction of trivalent actinides and rare earth elements with the intent of recovering Cm and Am from dissolved irradiated 242 Pu Mark-18A targets in 7 to 9 M nitric acid. The extracted constituents will be recovered as an oxide by direct thermal decomposition of the loaded resin followed by calcination of the resultant residue. The characteristics of DGA resin with non-radiological feed simulant representative of the anticipated feed in the Mark-18A process, with Sm and Nd as surrogates for Cm and Am, respectively, including breakthrough point and saturation capacity were evaluated in this work. Additionally, this work examined the losses from the loaded resin by rinsing the resin bed with dilute acid to reduce the nitrate concentration in the resin bed prior to thermal decomposition operations to improve the safety posture of the process. A resin loading profile was developed, and the resin was determined to have a trivalent metal saturation capacity of 74 μmol/mL resin under the experimental conditions evaluated. Following a wash of the loaded resin bed with fresh 8 M HNO 3 , the trivalent metals retained was reduced to 68 μmol/mL resin, which represents the practical capacity of the resin for the Mark-18A process. Lighter lanthanides breakthrough the resin well before the saturation capacity is reached. Rinsing the saturated resin bed with 0.26 M HNO 3 was found to result in a rapid reduction in retention of rare earth elements by the resin. After 2.8 bed volumes of dilute acid rinse, the mean resin bed free acid concentration was reduced to 0.28 M and 3.1 bed volumes of dilute acid rinse resulted in a reduction of the cumulative rare earth element retention to 52 μmol/mL of resin.

Transplutonium separations↗
DOE OSTI2019

Analysis of Tc and U Spent Ion Exchange Resins from Hanford 200 West P and T Facility - 19202

The 200 West Area Pump-and-Treat (P and T) facility at the Hanford Site is being used to remove radionuclides such as uranium (U) and technetium-99 (Tc-99) from the contaminated groundwater. To support evaluation of U and Tc ion exchange (IX) treatment operational conditions and performance with various influent conditions, knowledge of resin capacity and contaminant retention is needed. To meet these objectives, samples of the spent Dowex{sup R} 21K and Purolite{sup R} A530E IX resins from the 200 West P and T facility used for the treatment of U and Tc, respectively, have been obtained and analyzed for the target analytes and other retained contaminants. The objectives of this work included 1) determination of the effective U and Tc loading on the resins, 2) evaluation of redox and chemical speciation of U and Tc retained by the resins and impact of the feed composition on the retention of U by the resin, and 3) quantification of other contaminants and groundwater constituents retained on the resins. Speciation of U and Tc-99 on the resin was determined by a combination of X-ray absorption fine structure, nuclear magnetic resonance, X-ray photoelectron spectroscopy (XAS), and optical spectroscopic techniques. Nondestructive analytical techniques including gamma energy analysis and low-energy photon spectroscopy were incorporated to quantify radionuclides retained on the resin. Scanning electron microscopy and energy-dispersive X-ray spectroscopy were used to observe the IX beads and quantify chemical constituents on the surface and penetration through the beads. Destructive analysis of the resin was explored through wet oxidation to identify and quantify co-contaminants sorbed to the resins. Inductively coupled plasma-optical emission spectroscopy (ICP-OES) and inductively coupled plasma-mass spectrometry (ICP-MS) were used to analyze for metals, iodine, and specific isotopes retained on the resin, including Tc-99 and U-238. A single pertechnetate Tc-99 species was identified to be retained on the resin. For U, the obtained structural information was correlated with the laboratory findings using pristine resin as reference material to identify the nature of the U complexes and predict U retention by the resin as a function of variable influent conditions. Both resins demonstrated high loading with sulfate, silica, calcium, iron, zinc, and other groundwater constituents, suggesting that the competing anions, fouling, or a combination thereof significantly reduced their effective capacity for U and Tc-99. A significant amount of nonradioactive iodine-127 (I-127) and long-lived radioactive iodine-129 (I-129) isotope was also retained on the resins. Obtained quantitative information provided insights on the impact of water chemistry and the presence of common mineral constituents on the retention of contaminants of concern, including U, Tc-99, and I-129. (authors)

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

Method development for thermal analyses testing on Reillex HPQ resin using the advanced reactive system screening tool (ARSST)

Reillex™ HPQ resin was developed by Los Alamos Laboratory and Reilly Industries Inc. in an effort to increase safety and process efficiency during the recovery and purification of plutonium. Ionac™ A-641, another strong base macroporous anion exchange resin used in the nuclear industry, was known to undergo a runaway reaction in hot nitric acid solutions. Because of this, an extensive amount of thermal analyses testing on the Reillex™ HPQ resin in SRNL was performed in 1999-2001 prior to use. A report on the thermal stability qualification of the Reillex™ HPQ resin in 8M (35%) and 12M (53%) HNO 3 was reported in 2000. In 2001, the reactivity of Reillex™ HPQ resin in 14.4M (64%) HNO 3 was evaluated. In January of 2001, thermal stability scoping tests were performed on irradiated Reillex™ HPQ resin in 14.4M (64%) HNO 3 (as a worst case scenario) and the results sent to Fauske and Associates to calculate a rupture disk size for the HB-Line resin column. A technical report by Fauske and Associates was issued in February 2001 recommending a 2.0” vent line with a rupture disk set pressure of 60 psig. This calculation was based on ARSST thermal analyses scoping tests at SRNL in which 4 grams of dried resin and 6.0 grams of 64% nitric acid in a 10 gram test cell, produced a maximum pressure rate (dP/dt) of 720 psi/min (12 psi/sec) and a maximum temperature of 250 °C. In 2015, a new batch of Reillex™ HPQ resin was manufactured by Vertellus Industries. A test sample of the resin was sent to SRNL to perform acceptance and qualification thermal stability testing using the ARSST. During these tests, method development was performed to ensure that a representative resin to acid ratios were used while running the tests in the ARSST. Fauske and Associates recommended to either use a full test cell representative of the HB-Line column or a 10 gram sample in the test cell that was representative of the ratios of resin to nitric acid in the actual HB-Line column. An observation was made during method development testing that the 4 grams of resin to 6 grams of 64% acid ratio used in 2001 to determine the rupture disk size did not entirely wet the resin and was not representative of the actual HB-Line column. The results of this study raised questions as to whether the ratio of 4 grams of resin to 6 grams of 64% nitric acid recommended by Fauske and Associates in 2001 was the correct ratio.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Evaluating ion exchange resin efficiency and oxidative capacity for the separation of uranium(IV) and uranium(VI)

Background: Previously described methods to separate dissolved U(IV) from dissolved U(VI) under acidic anoxic conditions prior to laboratory analysis were ineffective with materials currently available commercially. Three strong anion exchange resins were examined for their efficiency in separating, recovering, and preserving both redox states during separation. Results: Under oxic conditions, recovery of U(VI) from three exchange resins (Bio-Rad AG® 1x8 Poly-Prep® prefilled columns, Bio-Rad AGW 1x8 powder, and Dowex® 1x8 powder) ranged from 72% to 100% depending on the dosed mass, eluent volume, and resin selected. Dowex® 1x8 resin was the only resin found to provide 100% recovery of U (VI) with fewer than 5 bed volumes of eluent. Under anoxic conditions, all three resins oxidized U(IV) in aqueous solutions with relatively low U(IV) concentrations (<3x10-6 M). Resin-induced oxidation was observed visually using a leuco dye, safranin-o. Oxidants associated with the resin were irreversibly reduced by the addition of Ti(III). After anoxic resin pre-treatment, a series of U(IV)/U(VI) mixtures at micro-molar levels were prepared and separated using the Dowex® 1x8 resin with 100% recovery of both U(IV) and U(VI) with no resin-induced changes in oxidation state. Conclusions: Currently available anion exchange resins with apparently identical physical properties were found to have significantly different recoveries for hexavalent uranium at micro-molar concentrations. A novel qualitative technique was developed to visually assess oxidative capacities of anion exchange resins under acidic anoxic conditions. A protocol was developed for pre-treatment and use of currently available anion exchange resins to achieve quantitative separation of U(IV) and U(VI) in aqueous solutions with low U(IV) concentrations. This method can be applied to future work to quantitatively assess dissolved U(IV) and U(VI) concentrations in both laboratory and field samples.

54 ENVIRONMENTAL SCIENCES↗