Search NASA⌕ Search

SEARCH · Search NASA

Results for “RESINS”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
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)↗
NASA NTRS1992

Comparison of resin film infusion, resin transfer molding, and consolidation of textile preforms for primary aircraft structure

Innovative design concepts and cost effective fabrication processes were developed for damage tolerant primary structures that can perform at a design ultimate strain level of 6000 micro inch/inch. Attention focused on the use of textile high performance fiber reinforcement concepts that provide improved damage tolerance and out-of-plane load capability, low cost resin film infusion (RFI) and resin transfer molding (RTM) processes, and thermoplastic forming concepts. The fabrication of wing 'Y' spars by four different materials and/or processes methods is described: fabricated using IM7 angle interlock 0 to 90 deg woven preforms with + or - 45 deg plies stitched with Toray high strength graphite thread and processed using RFI and 3501-6 epoxy; fabricated using G40-800 knitted/stitched preforms and processed using RFI and 3501-6 epoxy; fabricated using G40-800 knitted/stitched preforms using RTM and Tactix 123/H41 epoxy; and fabricated preforms using AS4(6K)/PEEK 150 g commingled angle interlock 0 to 90 deg woven preforms with + or - 45 deg commingled plies stitched using high strength graphite thread and processed by consolidation. Structural efficiency, processability, and acquisition cost are compared.

Suarez, J.↗
NASA NTRS1992

Comparison of resin film infusion, resin transfer molding, and consolidation of textile preforms for primary aircraft structure

Under NASA's Novel Composites for Wing and Fuselage Applications (NCWFA) Program, Grumman is developing innovative design concepts and cost-effective fabrication processes for damage-tolerant primary structures that can perform at a design ultimate strain level of 6000 micro-inch/inch. Attention has focused on the use of textile high-performance fiber-reinforcement concepts that provide improved damage tolerance and out-of-plane load capability, low-cost resin film infusion (RFI) and resin transfer molding (RTM) processes, and thermoplastic forming concepts. The fabrication of wing 'Y' spars by four different materials/processes methods is described: 'Y' spars fabricated using IM7 angle interlock 0/90 deg woven preforms with +/- 45 deg plies stitched with Toray high-strength graphite thread and processed using RFI and 3501-6 epoxy; 'Y' spars fabricated using G40-800 knitted/stitched preforms and processed using RFI and 3501-6 epoxy; 'Y' spars fabricated using G40-800 knitted/stitched preforms and processed using RTM and Tactix 123/H41 epoxy; and 'Y' spars fabricated using AS4(6k)/PEEK 150-g commingled angle interlock 0/90 deg woven preforms with +/- 45 deg commingled plies stitched using high-strength graphite thread and processed by consolidation. A comparison of the structural efficiency, processability, and projected acquisition cost of these representative spars is presented.

Suarez, J.↗
DOE OSTI2012

Nitrate Conversion of HB-Line Reillex™ HPQ Resin

Reillex™ HPQ ion exchange resin is used by HB Line to remove plutonium from aqueous streams. Reillex™ HPQ resin currently available from Vertellus Specialties LLC is a chloride ionic form, which can cause stress corrosion cracking in stainless steels. Therefore, HB Line Engineering requested that Savannah River National Laboratory (SRNL) convert resin from chloride form to nitrate form in the Engineering Development Laboratory (EDL). To perform this task, SRNL treated two batches of resin in 2012. The first batch of resin from Reilly Industries Batch 80302MA was initially treated at SRNL in 2001 to remove chloride. This batch of resin, nominally 30 liters, has been stored wet in carboys since that time until being retreated in 2012. The second batch of resin from Batch 23408 consisted of 50 kg of new resin purchased from Vertellus Specialties in 2012. Both batches were treated in a column designed to convert resin using downflow of 1.0 M sodium nitrate solution through the resin bed followed by rinsing with deionized water. Both batches were analyzed for chloride concentration, before and after treatment, using Neutron Activation Analysis (NAA). The resin specification [Werling, 2003] states the total chlorine and chloride concentration shall be less than 250 ppm. The resin condition for measuring this concentration is not specified; however, in service the resin would always be fully wet. Measurements in SRNL showed that changing from oven dry resin to fully wet resin, with liquid in the particle interstices but no supernatant, increases the total weight by a factor of at least three. Therefore, concentration of chlorine or chloride expressed as parts per million (ppm) decreases by a factor of three. Therefore, SRNL recommends measuring chlorine concentration on an oven dry basis, then dividing by three to estimate chloride concentration in the fully wet condition. Chloride concentration in the first batch (No.80302MA) was nearly the same before the current treatment (759 ppm dry) and after treatment (745 ppm dry or ~248 ppm wet). Treatment of the second batch of resin (No.23408) was very successful. Chloride concentration decreased from 120,000 ppm dry to an average of 44 ppm dry or ~15ppm wet, which easily passes the 250 ppm wet criterion. Per guidance from HB Line Engineering, SRNL blended Batch 80302 resin with Batch P9059 resin which had been treated previously by ResinTech to remove chloride. The chloride concentrations for the two drums of Batch P9059 were 248 ppm dry (~83 ppm wet) ±22.8% and 583 ppm dry (~194 ppm wet) ± 11.8%. The blended resin was packaged in five gallon buckets.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
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 OSTI2010

Uranium Adsorption on Ion-Exchange Resins - Batch Testing

The uranium adsorption performance of five resins (Dowex 1, Dowex 21K 16-30 [fresh], Dowex 21K 16-30 [regenerated], Purofine PFA600/4740, and ResinTech SIR-1200) were tested using unspiked, nitrate-spiked, and nitrate-spiked/pH adjusted source water from well 299-W19-36. These batch tests were conducted in support of a resin selection process in which the best resin to use for uranium treatment in the 200-West Area groundwater pump-and-treat system will be identified. The results from these tests are as follows: • The data from the high-nitrate (1331 mg/L) tests indicated that Dowex 1, Dowex 21K 16-30 (fresh), Purofine PFA600/4740, and ResinTech SIR-1200 all adsorbed uranium similarly well with Kd values ranging from ~15,000 to 95,000 ml/g. All four resins would be considered suitable for use in the treatment system based on uranium adsorption characteristics. • Lowering the pH of the high nitrate test conditions from 8.2 to 7.5 did not significantly change the uranium adsorption isotherms for the four tested resins. The Kd values for these four resins under high nitrate (1338 mg/L), lower pH (7.5) ranged from ~15,000 to 80,000 ml/g. • Higher nitrate concentrations greatly reduced the uranium adsorption on all four resins. Tests conducted with unspiked (no amendments; nitrate at 337 mg/L and pH at 8.2) source water yielded Kd values for Dowex 1, Dowex 21K 16-30 (fresh), Purofine PFA600/4740, and ResinTech SIR-1200 resins ranging from ~800,000 to >3,000,000 ml/g. These values are about two orders of magnitude higher than the Kd values noted from tests conducted using amended source water. • Compared to the fresh resin, the regenerated Dowex 21K 16-30 resin exhibited significantly lower uranium-adsorption performance under all test conditions. The calculated Kd values for the regenerated resin were typically an order of magnitude lower than the values calculated for the fresh resin. • Additional testing using laboratory columns is recommended to better resolve differences between the adsorption abilities of the resins and to develop estimates of uranium loading on the resins. By determining the quantity of uranium that each resin can adsorb and the time required to reach various levels of loading, resin lifetime in the treatment system can be estimated.

36 MATERIALS SCIENCE↗
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 OSTI2012

Ion Exchange Temperature Testing with SRF Resin - 12088

Ion exchange using the Spherical Resorcinol-Formaldehyde (SRF) resin has been selected by the U.S. Department of Energy's Office of River Protection for use in the Pretreatment Facility of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) and for potential application in an at-tank deployment for removing Cs-137. Recent proposed changes to the WTP ion exchange process baseline indicate that higher temperatures (50 deg. C) to alleviate post-filtration precipitation issues prior to reaching the ion exchange columns may be required. Therefore, it is important to understand the behavior of SRF resin performance under the conditions expected with the new equipment and process changes. This research examined the impact of elevated temperature on resin loading and resin degradation during extended solution flow at elevated temperature (45 deg., 50 deg., 55 deg., 60 deg., 65 deg., 75 deg. C). Testing for extended times at elevated temperatures showed that the resin does degrade and loading capacity is reduced at and above 45 deg. C. Above 60 deg. C the resin appears to not load at all. It was observed that the resin disintegrated at 75 deg. C until not much was left and partially disintegrated at 65 deg. C, which caused the column to plug in both tests after ∼336 hours. The results indicate that WTP will lose resin loading capacity if the ion exchange process is performed above 25 deg. C, and the resin will disintegrate above 65 deg. C. Therefore, WTP will have a restricted operating range of temperatures to perform the ion exchange process with this resin. PNNL and WTP are currently evaluating the operating limits of the resin in further detail. Aging in 0.5 M HNO{sub 3} also caused the resin to lose capacity above 25 deg. C and to completely dissolve at 55 deg. C. Again, WTP will have a restricted operating range of temperatures when eluting the resin with nitric acid in order to maintain resin loading capacity and avoid disintegration of the resin. (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↗
DOE OSTI2013

Westinghouse Modular Grinding Process - Enhancement of Volume Reduction for Hot Resin Supercompaction - 13491

In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. Spent resins can contain a significant amount of contaminates which makes treatment for disposal of spent resins mandatory. Several treatment processes are available such as direct immobilization with technologies like cementation, bitumisation, polymer solidification or usage of a high integrity container (HIC). These technologies usually come with a significant increase in final waste volume. The Hot Resin Supercompaction (HRSC) is a thermal treatment process which reduces the resin waste volume significantly. For a mixture of powdered and bead resins the HRSC process has demonstrated a volume reduction of up to 75 % [1]. For bead resins only the HRSC process is challenging because the bead resins compaction properties are unfavorable. The bead resin material does not form a solid block after compaction and shows a high spring back effect. The volume reduction of bead resins is not as good as for the mixture described in [1]. The compaction properties of bead resin waste can be significantly improved by grinding the beads to powder. The grinding also eliminates the need for a powder additive.Westinghouse has developed a modular grinding process to grind the bead resin to powder. The developed process requires no circulation of resins and enables a selective adjustment of particle size and distribution to achieve optimal results in the HRSC or in any other following process. A special grinding tool setup is use to minimize maintenance and radiation exposure to personnel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Radioactive Spent Ion-Exchange Resins Conditioning by the Hot Supercompaction Process at Tihange NPP - Early Experience - 12200

Spent ion-exchange resins are considered to be problematic waste that, in many cases, requires special approaches and precautions during their conditioning to meet the acceptance criteria for disposal. In Belgium, for economical reasons, the Volume Reduction Factor is a key criterion. After Tractebel Engineering performed a technical and economical comparison of the industrially available systems, Tihange NPP decided to install a spent ion-exchange resins hot supercompaction unit with Tractebel Engineering in the role of architect-engineer. The treatment and conditioning unit processes the spent ion-exchange resins through the following steps: dewatering of the resins, drying the resins under deep vacuum, discharging the dried resins into compactable drums, super-compacting the drums to generate pellets, grouting the pellets into standard 400 litres waste drums (overpacks) licensed for final disposal in the near-surface repository in Belgium. Several developments were required to adapt the reference process and equipment to PWR spent ion-exchange bead resins and Belgian radioactive waste acceptance criteria. In order to avoid cracks on the compacted drum, and external surface contamination from resin leaks, some improvements were achieved to minimize spring-back as well as the risk of cracking the drum wall. Placing the compactable drum inside a second, slightly larger drum, guarantees clean and reproducible pellets. Currently the commissioning phase is on-going. Numerous process validation tests have been completed. An acceptance file was transmitted to the Belgian Waste Management Authority recently. This will be followed by demonstration tests necessary to obtain their final acceptance of the installation. More than 3 800 drums of mixed powdered and bead resins have been processed by the reference Hot Compaction process, achieving a Volume Reduction Factor (VRF) of 2.5. The equipment has been proven to be a reliable technology with low operation and maintenance costs. Tractebel Engineering has managed the construction of a new application of this process in Belgium at Tihange NPP. Several developments were required to adapt the reference process and equipment to PWR spent ion-exchange bead resins and Belgian radioactive waste acceptance criteria. The chosen method of conditioning (draining, drying and compaction of the spent resins followed by grouting of the pellets in a 400 litres drum) immobilizes the spent resins under the form of a solid, compact, stable, and non dispersible block free of interstitial water. The various series of inactive tests which were conducted at Tihange NPP, helped among others to determine the best design of the compactable drum and lid and to set the value of critical parameters such as vapour temperature at the end of drying, speed, force and duration of compaction. In an environment of very limited space for interim storage and in the absence of an operational final repository site, or in the case of high final disposal costs, the process exhibits the following key advantages: - Achieving a Volume Reduction Factor (VRF) close to 1 (overpack included) for the interim storage instead of increased volumes observed with other currently available processes; - Achieving a water free end product; - Creating a flexible waste product for interim storage (pellet), which can be retrieved and routed into alternative types of package later, if not initially grouted; - Using well proven standard technologies like drying and compaction; - Flexible use of the system components for the supercompaction of other operational solid waste streams when not conducting resins conditioning campaigns. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗