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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Actinide Separation Inspired by Self-Assembled Metal–Polyphenolic Nanocages

The separation of actinides has a vital place in nuclear fuel reprocessing, recovery of radionuclides, and remediation of environmental contamination. Here we propose a new paradigm of nanocluster-based actinide separation, namely, nanoextraction, that can achieve efficient sequestration of uranium in an unprecedented form of giant coordination nanocages using a cone-shaped macrocyclic pyrogallol[4]arene as the extractant. The U 24 -based hexameric pyrogallol[4]arene nanocages with distinctive [U 2 (PG) 2 ] binuclear units (PG = pyrogallol) that rapidly assembled in situ in monophasic solvent were identified by single-crystal X-ray diffraction, MALDI-TOF mass spectrometry, NMR spectroscopy, and small-angle X-ray and neutron scattering. Furthermore, comprehensive biphasic extraction studies showed that this novel separation strategy has enticing advantages such as fast kinetics, high efficiency, and good selectivity over lanthanides, thereby demonstrating its potential for efficient separation of actinide ions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Functionalized metal organic frameworks for effective capture of radioactive organic iodides

Highly efficient capture of radioactive organic iodides (ROIs) from off-gas mixtures remains a substantial challenge for nuclear waste treatment. Current materials utilized for ROI sequestration suffer from low capacity, high cost ( e.g. use of noble metals), and poor recyclability. Recently, we have developed a new strategy to tackle this challenge by functionalizing MOF materials with tertiary amines to create molecular traps for the effective capture and removal of ROIs ( e.g. radioactive methyl iodide) from nuclear wastes. To further enhance the uptake capacity and performance of CH 3 I capture by ROI molecular traps, herein, we carry out a systematic study to investigate the effect of different amine molecules on ROI capture. The results demonstrate a record-high CH 3 I saturation uptake capacity of 80% for MIL-101–Cr–DMEDA at 150 °C, which is 5.3 times that of Ag 0 @MOR (15 wt%), a leading adsorbent material for capturing ROIs during nuclear fuel reprocessing. Furthermore, the CH 3 I decontamination factors (DFs) for MIL-101–Cr–DMEDA are as high as 5000 under simulated reprocessing conditions, largely exceeding that of facility regulatory requirements (DF = 3000). In addition, MIL-101–Cr–DMEDA can be recycled without loss of capacity, illustrating yet another advantage compared to known industrial adsorbents, which are typically of a “single-use” nature. Our analysis also shows that both physisorption and chemisorption of CH 3 I occur at the three amine-grafted MOFs. While chemisorption takes place at the amine functionalized sites, the amount of physisorption correlates with the MOF porosity. A possible binding site of amine–CH 3 I interaction has been identified via an in situ IR spectroscopic study. The results suggest that CH 3 I interacts strongly and directly with the tertiary nitrogen of the amine molecules. The CH 3 I uptake amount decreases as the amine chain length increases, in trend with the decreasing pore space of the corresponding framework. The strategy to build MOF-based molecular traps developed in this work not only leads to a new record-high performance for ROI capture, but also offers an effective way of systematically tuning the porosity by varying the length of functionalized amine molecules. This study also demonstrates that MOFs represent a promising new platform for selective capture and removal of radioactive nuclear waste.

Chemistry↗
DOE OSTI2021

Curium(iii) radiation-induced reaction kinetics in aqueous media

Insight into the effects of radiolytic processes on the actinides is critical for advancing our understanding of their solution chemistry because the behaviour of these elements cannot be easily separated from the influence of their inherent radiation field. However, minimal information exists on the radiation-induced redox behaviour of curium (Cm), a key trivalent transuranic element present in used nuclear fuel and frequently used as an alpha radiation source. Here we present a kinetic study on the aqueous redox reactions of Cm(III) with radicals generated through the radiolysis of aqueous media. In particular, we probe reaction kinetics in nitric acid solutions that are used as the aqueous phase component of used nuclear fuel reprocessing solvent systems. Second-order rate coefficients (k) were measured for the reaction of Cm(III) with the hydrated electron (e aq - , k = (1.25 ± 0.03) × 10 10 M -1 s -1 ), hydrogen atom (H˙, k = (5.16 ± 0.37) × 10 8 M -1 s -1 ), hydroxyl radical (˙OH, k = (1.69 ± 0.24) × 10 9 M -1 s -1 ), and nitrate radical (NO 3 ˙, k = (4.83 ± 0.09) × 10 7 M -1 s -1 ). Furthermore, the first-ever Cm(II) absorption spectrum (300–700 nm) is also reported. These kinetic data dispel the status quo notion of Cm(III) possessing little to no redox chemistry in aqueous solution, and suggest that the resulting Cm(II) and Cm(IV) transients could exist in irradiated aqueous solutions and be available to undergo subsequent redox chemistry with other solutes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Elemental analysis of air-sensitive frozen molten salt samples using an inert transfer chamber for LIBS/LA-ICP-TOF-MS analysis

A novel inert sample transfer system was developed and employed to enable, for the first time, the analysis of air-sensitive salt samples in a two-volume ablation cell using simultaneous laser-induced breakdown spectroscopy (LIBS) and laser ablation (LA)-inductively coupled plasma (ICP)-time-of-flight (TOF)-mass spectrometry (MS) analysis. Molten salts are of growing interest as a medium for advanced nuclear reactors and nuclear fuel reprocessing technologies continue to be developed around their use. However, compositional analysis of molten salt samples can be challenging because of their air-sensitive nature and varying solubilities leading to inaccurate measurements when digested. LA-based analysis provides an alternate method to digestion and can provide rapid elemental information with little sample preparation. In this study, LIBS and LA-ICP-TOF-MS were used to analyze the Ce content in frozen salt samples taken from a series of electrochemical experiments. Calibrations were built for each technique, and the resulting limits of detection for Ce were estimated to be 107 and 58 µg g −1 for LIBS and LA-ICP-TOF-MS, respectively. Test samples from the electrochemical experiments were analyzed using these calibrations. The results matched bulk digestion-based ICP-optical emission spectroscopy values, and daily trends in Ce concentration changes were identified. Additionally, the LIBS and LA-ICP-TOF-MS analysis was demonstrated for identifying microgram per gram levels of components and detecting trace contaminants. The impurities detected by LIBS included Al, Mg, Ca, and Na. The impurities detected by LA-ICP-TOF-MS included W, Ag, Al, Fe, Ni, Mo, Nd, Sm, Th, and U.

Andrews, Hunter B. [Oak Ridge National Laboratory ↗
DOE OSTI2019

Dissolution of used nuclear fuel using a tributyl phosphate/n-paraffin solvent

The complete recycle of used nuclear fuel (UNF) is necessary to recover the energy value from the remaining U and higher actinides; yet, construction of a large-scale facility is extremely expensive. To address this issue, a hybrid process was evaluated in which UNF pretreated for tritium removal is dissolved in a tributyl phosphate (TBP) solvent. The distribution of the actinides and selected fission products was measured between the solvent and residual solids. Dissolution of these elements in the surrogate fuels was very effective. Most fission products from fuel irradiation except for the lanthanide elements would remain in the undissolved solids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

In Situ Monitoring and Kinetic Analysis of the Extraction of Nitric Acid by Tributyl Phosphate in n-Dodecane using Raman Spectroscopy

Understanding the fundamental rates of transfer and complex formation is crucial in maintaining optimal efficiency and effectivity in solvent extraction. Methodologies to effectively analyze solvent extraction systems are lacking in that they are commonly off-line. Thus, a method to effectively investigate solvent extraction systems in a Lewis cell using in situ Raman spectroscopy paired with chemometric modelling has been developed to allow for on-line analysis and monitoring. Kinetic parameters of interfacial mass transfer for the extraction of nitric acid by tributyl phosphate, a process crucial to used nuclear fuel reprocessing, were measured in by spectroscopic monitoring of both phases simultaneously in a two-phase solvent extraction system. Resulting concentration data, derived from a chemometric model using Raman spectra as the base set, allowed for the calculation of kinetic parameters in the system when paired with extraction time. The forward and reverse interfacial mass transfer coefficients, kf and kr, were determined for solutions that were process relevant and of pro-cess interest, where kf was calculated as 5.3 (± 0.3) × 10-3 cm min 1 and kr was calculated as 2.9 (± 0.2) × 10-2 cm min-1. The information obtained regarding solvent extraction parameters and transfer can be applied in the further study of systems of higher complexity involving complex formation and their transfer across a liquid:liquid interface.

extraction, interfacial transport, Lewis Cell, Liq↗
DOE OSTI2022

Radiation effects on the structure and alteration behavior of an SiO 2 –Al 2 O 3 –B 2 O 3 –Na 2 O glass

As borosilicate glasses are used in many countries to immobilize fission products and minor actinides after spent fuel reprocessing before storage in a deep geological repository, assessing that their chemical durability is of paramount importance. Here, pristine and preirradiated (952 MeV, 136Xe) SiO 2 –B2O 3 –Al2O 3 –Na 2 O glasses with the same molar ratios as in the French SON68 and ISG glasses have been subjected to aqueous corrosion in deionized water and in silica-saturated solution to measure the initial and longer term alteration rates. Pristine and preirradiated glasses corrode following the same mechanisms, but the preirradiation has a strong impact on the initial dissolution rate (increase by a factor of 5.6), and on the alteration layer depth in silica-saturated conditions (by two- to threefolds). The later result is related to the formation of a more porous, less passivating gel on the preirradiated glass specimen. Using both experimental spectroscopies (NMR, IR, and SFG) and classical molecular dynamics, the radiation effects on the glass structure and water diffusion have been assessed. After preirradiation, the density and the polymerization degree of the glass decrease, whereas the topological disorder increases. In consequence, water diffusion accelerates. These observations allow to correlate the radiation impact on the alteration behavior to the structural changes.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Solution-Derived, Chloride-Containing Minerals as a Waste Form for Alkali Chlorides

Sodalite [Na8(AlSiO4)6Cl2] and cancrinite [(Na,K)6Ca2Al6Si6O24Cl4] are environmentally stable, chloride-containing minerals and are a logical waste form option for the mixed alkali chloride salt waste stream that is generated from a proposed electrochemical separations process during nuclear fuel reprocessing. Due to the volatility of chloride salts at moderate temperatures, the ideal processing route for these salts is a low-temperature approach such as the sol-gel process. The sodalite structure can be easily synthesized by the sol-gel process; however, it is produced in the form of a fine powder with particle sizes on the order of 1–10 µm. Due to the small particle size, these powders require additional treatment to form a monolith. In this study, the sol-gel powders were pressed into pellets and fired to achieve > 90% of theoretical density. The cancrinite structure, identified as the best candidate mineral form in terms of waste loading capacity, was only produced on a limited basis following the sol-gel process and converted to sodalite upon firing. Here we discuss the sol-gel process specifics, chemical durability of select waste forms, and the steps taken to maximize chloride-containing phases, decrease chloride loss during pellet firing, and increase pellet densities.

cancrinite↗
DOE OSTI2015

Development and Durability Testing of a Low‐Temperature Sintering Bi–Si–Zn Oxide Glass Composite Material (GCM) 129 I Waste Form

The capture and safe storage of radiological iodine ( 129 I) from nuclear fuel reprocessing is of concern due to its long half‐life and potential mobility in the environment. The development of durable waste forms in which to store captured iodine requires materials that are both compatible with the iodine capture phases and durable to repository environments. To that end, Sandia is developing a low‐temperature sintering Bi–Si–Zn oxide glass composite material (GCM) waste form and herein presents results of durability testing. Furthermore, durability studies were extended to both the occluded iodine capture material Ag‐Zeolite (Mordenite, MOR) as well as the GCM, synthesized with compositional variations including: amount of Ag flake added, AgI‐MOR particle size in the GCM, and mass loading of I within the AgI‐MOR. Product consistency test (PCT‐B), chemical durability (MCC‐1) tests, and single‐pass flow‐through (SPFT) tests, were performed on both the individual components of the GCM and the completed GCMs. Durability tests indicate low GCM dissolution rates (<10 −3 g/m 2 ·d) across wide variable ranges including: pH, AgI‐MOR loading, I loading, and AgI‐MOR particle size. Results indicate that the Bi–Si–Zn oxide glass matrix sharply limits the release of iodine from the otherwise relatively fast degrading AgI‐MOR getter material. Furthermore, the formation of an amorphous AgI phase during sintering of the GCM results in the limitation of iodine release during waste form degradation. Durability of GCM and release rates approximate those of established nuclear waste glasses, or analogues such as basaltic glass. This suggests that the Bi–Si–Zn GCM is a viable candidate as a repository iodine waste form.

Mowry, Curtis D.↗
DOE OSTI2015

Process and equipment development for hot isostatic pressing treatability study

Battelle Energy Alliance (BEA), LLC, has developed processes and equipment for a pilot-scale hot isostatic pressing (HIP) treatability study to stabilize and volume reduce radioactive calcine stored at Idaho National Laboratory (INL). In 2009, the U. S. Department of Energy signed a Record of Decision with the state of Idaho selecting HIP technology as the method to treat 5,800 yd^3 (4,400 m^3) of granular zirconia and alumina calcine produced between 1953 and 1992 as a waste byproduct of spent nuclear fuel reprocessing. Since the 1990s, a variety of radioactive and hazardous waste forms have been remotely treated using HIP within INL hot cells. To execute the remote process at INL, waste is loaded into a stainless-steel or aluminum can, which is evacuated, sealed, and placed into a HIP furnace. The HIP simultaneously heats and pressurizes the waste, reducing its volume and increasing its durability. Two 1 gal cans of calcine waste currently stored in a shielded cask were identified as candidate materials for a treatability study involving the HIP process. Equipment and materials for cask-handling and calcine transfer into INL hot cells, as well as remotely operated equipment for waste can opening, particle sizing, material blending, and HIP can loading have been designed and successfully tested. These results demonstrate BEA’s readiness for treatment of INL calcine.

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

Electrochemical behavior of uranyl in anhydrous polar organic media

Abstract Weak complexes between pentavalent and hexavalent actinyl cations have been reported to exist in acidic, non-complexing high ionic strength aqueous media. Such “cation-cation complexes” were first identified in the context of actinide-actinide redox reactions in acidic aqueous media relevant to solvent extraction-based separation systems, hence their characterization is of potential interest for advanced nuclear fuel reprocessing. This chemistry could be relevant to efforts to develop advanced actinide separations based on the upper oxidation states of americium, which are of current interest. In the present study, the chemical behavior of pentavalent uranyl was examined in non-aqueous, aprotic polar organic solvents (propylene carbonate and acetonitrile) to determine whether UO

Chemistry↗
DOE OSTI2022

Process optimization of caustic scrubber and iodine-129 immobilization in sodalite-based waste forms

Caustic scrubbers are proposed to remove vaporized iodine from nuclear fuel reprocessing plants. Resulting caustic slurries of radioactive iodine ( 129 I) and other anions can be captured and immobilized into the cage structures of sodalite-type minerals. Here, mixed anion sodalites were hydrothermally synthesized from a simulated caustic scrubber solution containing hydroxide/water, carbonate, chlorine, bromine, iodine, nitrate, and nitrite ions with added kaolinite. Experiments were conducted with twenty-five different sets of process conditions. Resulting powders were characterized by quantitative X-ray diffraction (XRD) and multi-functional thermal analysis, including gravimetry, calorimetry, and mass spectrometry. Using statistical analysis, process variables were tailored to maximize conversion of kaolinite into cage silicates, immobilize caustic scrubber anions, and minimize amorphous content. Further, chemical analysis showed the incorporation of all the targeted ions into the solids. Detailed investigation of a batch characterized after washing (creating secondary waste) and one without washing but subsequent calcination is given. This work contributes to defining the constraints required for subsequent consolidation process steps of the waste form to avoid volatilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Why natural monazite never becomes amorphous: Experimental evidence for alpha self-healing

Monazite, a common accessory rare-earth orthophosphate mineral in the continental crust widely used in U-Pb geochronology, holds promise for (U-Th)/He thermochronology and for the immobilization of Pu and minor actinides (MA) coming from spent nuclear fuel reprocessing. Previous results obtained on natural and plutonium-doped monazite have demonstrated the ability of this structure to maintain a crystalline state despite high radiation damage levels. However, the low critical temperature (180 °C), above which amorphization cannot be achieved in natural monazite under ion irradiation, does not explain this old and unsolved paradox: why do natural monazites, independent of their geological history, remain crystalline even when they did not experience any thermal event that could heal the defects? This is what the present study aims to address. Synthetic polycrystals of LaPO 4 -monazite were irradiated sequentially and simultaneously with α particles (He) and gold (Au) ions. Here, our results demonstrate experimentally for the first time in monazite, the existence of the defect recovery mechanism, called α-healing, acting in this structure due to electronic energy loss of α particles, which explains the absence of amorphization in natural monazite samples. This mechanism is critically important for monazite geo- and thermochronology and to design and predictively model the long-term behavior of ceramic matrices for nuclear waste conditioning.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Generalized Test Plan for the Vitrification of Simulated High-Level-Waste Calcine in the Idaho National Laboratory‘s Bench-Scale Cold Crucible Induction Melter

This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine's make-up is HLW containing relatively high concentrations of zirconium and aluminum, representative of the cladding material of the reprocessed fuel that generated the calcine. A separate study to define the CCIM testing needs of these other calcine classifications in currently being prepared under a separate work package (WP-0) and will be provided as a milestone report at the end of this fiscal year.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO 2 -based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Deep Bed Adsorption Testing using Silver-Functionalized Aerogel

Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing and evolve in gaseous species into the reprocessing facility off-gas systems. Analyses have shown that I129, due to its radioactivity, high potential mobility in the environment, and high longevity (half life of 15.7 million years), can require control efficiencies of up to 1,000x or higher to meet regulatory emission limits. Two Aerogel sorption tests that have been performed this fiscal year. The maximum iodine decontamination factor (DF) was measured to be over 10,000, above the 1,000-10,000 target DF range. The mass transfer zone may be as short as 0.5 inches under the sorption conditions of the first test. Only a small fraction of the iodine sorbed on Bed 1 was desorbed during the purge periods. The silver-functionalized Aerogel appears to have potential to be a very effective and efficient iodine sorbent.

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

Functionalized ultra-porous titania nanofiber membranes as nuclear waste separation and sequestration scaffolds for nuclear fuels recycle.

Advanced nuclear fuel cycle concept is interested in reducing separations to a simplified, one-step process if possible. This will benefit from the development of a one-step universal getter and sequestration material so as a simplified, universal waste form was proposed in this project. We have developed a technique combining a modified sol-gel chemistry and electrospinning for producing ultra-porous ceramic nanofiber membranes with controllable diameters and porous structures as the separation/sequestration materials. These ceramic nanofiber materials have been determined to have high porosity, permeability, loading capacity, and stability in extreme conditions. These porous fiber membranes were functionalized with silver nanoparticles and nanocrystal metal organic frameworks (MOFs) to introduce specific sites to capture gas species that are released during spent nuclear fuel reprocessing. Encapsulation into a durable waste form of ceramic composition was also demonstrated.

Liu, Haiqing↗
DOE OSTI2013

Assessment of Methods to Consolidate Iodine-Loaded Silver-Functionalized Silica Aerogel

The U.S. Department of Energy is currently investigating alternative sorbents for the removal and immobilization of radioiodine from the gas streams in a nuclear fuel reprocessing plant. One of these new sorbents, Ag 0 -functionalized silica aerogels, shows great promise as a potential replacement for Ag-bearing mordenites because of its high selectivity and sorption capacity for iodine. Moreover, a feasible consolidation of iodine-loaded Ag0-functionalized silica aerogels to a durable SiO 2 -based waste form makes this aerogel an attractive choice for sequestering radioiodine. This report provides a preliminary assessment of the methods that can be used to consolidate iodine-loaded Ag 0 -functionalized silica aerogels into a final waste form. In particular, it focuses on experimental investigation of densification of as prepared Ag 0 - functionalized silica aerogels powders, with or without organic moiety and with or without sintering additive (colloidal silica), with three commercially available techniques: 1) hot uniaxial pressing (HUP), 2) hot isostatic pressing (HIP), and 3) spark plasma sintering (SPS). The densified products were evaluated with helium gas pycnometer for apparent density, with the Archimedes method for apparent density and open porosity, and with high-resolution scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS) for the extent of densification and distribution of individual elements. The preliminary investigation of HUP, HIP, and SPS showed that these sintering methods can effectively consolidate powders of Ag 0 -functionalized silica aerogel into products of near-theoretical density. Also, removal of organic moiety and adding 5.6 mass% of colloidal silica to Ag0-functionalized silica aerogel powders before processing provided denser products. Furthermore, the ram travel data for SPS indicated that rapid consolidation of powders can be performed at temperatures below 950°C.

Ag-Functionalized Silica Aerogel↗