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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Technetium Chemistry in the Fuel Cycle: Combining Basic and Applied Studies

Technetium is intimately linked with nuclear reactions. The ultraminute natural levels in the environment are due to the spontaneous fission of uranium isotopes. The discovery of technetium was born from accelerator reactions, and its use and presence in the modern world are directly due to nuclear reactors. While occupying a central location in the periodic table, the chemistry of technetium is poorly explored, especially when compared to its neighboring elements, i.e., molybdenum, ruthenium, and rhenium. This state of affairs, which is tied to the small number of laboratories equipped to work with the long-lived (99)Tc isotope, provides a remarkable opportunity to combine basic studies with applications for the nuclear fuel cycle. An example is given through examination of the technetium halide compounds. Binary metal halides represent some of the most fundamental of inorganic compounds. The synthesis of new technetium halides demonstrates trends with structure, coordination number, and speciation that can be utilized in the nuclear fuel cycle. Examples are provided for technetium-zirconium alloys as waste forms and the formation of reduced technetium species in separations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Properties of Pertechnic Acid

Dilute aqueous pertechnic acid has long been known as a strong monoprotic acid which behaves as simple pertechnetate ion in aqueous solution. As pertechnic acid concentrates by evaporation, it becomes yellow and then dark red, and dark red crystalline material may ultimately be obtained. We show that as pertechnic acid concentrates, at least three compounds are formed – a yellow viscous liquid, a colorless (not red) crystalline solid, and a small amount of an intensely colored red-purple compound. The colorless crystalline compound melts at 118° C and can be melted and recrystallized several times with little decomposition. The red-purple compound is apparently not stable at room temperature and quickly decomposes if it is isolated. UV-Vis spectra show that Beer’s law does not hold as pertechnic acid concentrates by evaporation. We report densities, 99Tc nuclear magnetic resonance spectra, and ultraviolet-visible absorption spectra for highly pure aqueous pertechnic acid (accompanied by the other technetium compounds that form) ranging from 1 to 14 M in technetium concentration.

99Tc, pertechnic acid, technetium, pertechnetate↗
DOE OSTI2014

Protocol for Identifying the Presence of and Understanding the Nature of Soluble, Non-pertechnetate Technetium in Hanford Tank Supernatants

The objective of this report is to propose a method to evaluate the presence and extent of soluble, non-pertechnetate Tc in Hanford tank supernatants as well as methods that might be used to gain insight as to the nature of the specie(s) that make up this fraction. This study will then provide a recommendation as to the preferred approach for identifying and quantifying the presence of Hanford tank supernatant-soluble, non-pertechnetate, technetium. The recommendation will also describe an approach to address the issue of whether inductively coupled plasma mass spectrometry (ICP-MS) analysis, which is useful as a monitoring tool for Tc, may be confounded by the presence of other mass 99 species.

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

Design of a Fission 99 Mo Recovery Process and Implications toward Mo Adsorption Mechanism on Titania and Alumina Sorbents

Molybdenum-99 is a parent of the most widely used medical isotope technetium-99m. Proliferation concerns have prompted development of alternative Mo production methods utilizing low enriched uranium. Alumina and titania sorbents were evaluated for separation of Mo from concentrated uranyl nitrate solutions. System, mass transfer, and isotherm parameters were determined to enable design of Mo separation processes under a wide range of conditions. A model-based approach was utilized to design representative commercial-scale column processes. The designs and parameters were verified with bench-scale experiments. The results are essential for design of Mo separation processes from irradiated uranium solutions, selection of support material and process optimization. Mo uptake studies show that adsorption decreases with increasing concentration of uranyl nitrate; howeveL, examination of Mo adsorption as a function of nitrate ion concentration shows no dependency, indicating that uranium competes with Mo for adsorption sites. These results are consistent with reports indicating that Mo forms inner-sphere complexes with titania and alumina surface groups.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Uranyl-Tc( vii )/Tc( v ) hybrid clusters

Pertechnetate ( 99 Tc VII O 4 − ) and autoreduced 99 Tc(v), species coordinate to UO 2 2+ to form hybrid clusters. These clusters have important implications to understanding technetium chemistry in spent nuclear fuel and legacy wastes.

Chemistry↗
DOE OSTI2011

Laboratory and Modeling Evaluations in Support of Field Testing for Desiccation at the Hanford Site

The Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau includes testing of the desiccation technology as a potential technology to be used in conjunction with surface infiltration control to limit the flux of technetium and other contaminants in the vadose zone to the groundwater. Laboratory and modeling efforts were conducted to investigate technical uncertainties related to the desiccation process and its impact on contaminant transport. This information is intended to support planning, operation, and interpretation of a field test for desiccation in the Hanford Central Plateau.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Kinetics of Co-Mingled 99 Tc and Cr Removal during Mineral Transformation of Ferrous Hydroxide

In this work, four simulated waste streams relevant to the vitrification of Hanford nuclear waste were studied to evaluate the removal kinetics of technetium-99 (Tc) and co-mingled Cr(VI) during treatment with solid ferrous hydroxide (Fe(OH) 2 (s)). Simulants treated with Fe(OH) 2 (s) were reacted for 24 h and sub-sampled periodically to monitor Tc and Cr removal. Solution sample analysis during the reaction was coupled with solid phase characterization, for example, X-ray absorption spectroscopy (XAS) and X-ray diffraction (XRD), to establish influence of the solid phase product on Tc and Cr removal rates. Based on these results, the majority of Tc and Cr removal occurs within the first 5 min of simulant contact with Fe(OH) 2 (s). However, the order in which Tc and Cr are completely removed from each simulant depends on the simulant chemistry, the preferred reduction pathway, and the solid phase product, for example, magnetite (Fe 3 O 4 ) versus goethite (α-FeOOH). Low pH and low Cr concentrations favor rapid Tc removal, with XRD and XAS confirming that Fe 3 O 4 readily incorporates reduced Tc(IV) into its structure. High pH, high Cr concentrations, and the presence of other co-mingled constituents favor heterogeneous removal of Tc early in the reaction (<1 h) with removal rates often faster than those determined for Cr. At reaction times of >1 h, Tc removal slows as homogeneous removal of Cr begins to dominate, concurrent with an increase in the formation of FeOOH as the solid phase reaction product. These results suggest that for complex, high pH waste streams, Tc removal within the first hour after Fe(OH) 2 (s) treatment is necessary for complete Tc reduction and improved mineral immobilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Technetium Removal Using Tc-Goethite Coprecipitation

This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory for the U.S. Department of Energy (DOE) EM-31 Support Program (EMSP) subtask, “Low temperature waste forms coupled with technetium removal using an alternative immobilization process such as Fe(II) treated-goethite precipitation” to increase our understanding of 99Tc long-term stability in goethite mineral form and the process that controls the 99Tc(VII) reduction and removal by the final Fe (oxy)hydroxide forms. The overall objectives of this task were to 1) evaluate the transformation process of Fe (oxy)hydroxide solids to the more crystalline goethite (α-FeOOH) mineral for 99 Tc removal and 2) determine the mechanism that limits 99 Tc(IV) reoxidation in Fe(II)-treated 99 Tc-goethite mineral and 3) evaluate whether there is a long-term 99 Tcoxidation state change for Tc sequestered in the iron solids.

Goethite↗
DOE OSTI2014

Technetium Inventory, Distribution, and Speciation in Hanford Tanks

The purpose of this report is three fold: 1) assemble the available information regarding technetium (Tc) inventory, distribution between phases, and speciation in Hanford’s 177 storage tanks into a single, detailed, comprehensive assessment; 2) discuss the fate (distribution/speciation) of Tc once retrieved from the storage tanks and processed into a final waste form; and 3) discuss/document in less detail the available data on the inventory of Tc in other "pools" such as the vadose zone below inactive cribs and trenches, below single-shell tanks (SSTs) that have leaked, and in the groundwater below the Hanford Site. A thorough understanding of the inventory for mobile contaminants is key to any performance or risk assessment for Hanford Site facilities because potential groundwater and river contamination levels are proportional to the amount of contaminants disposed at the Hanford Site. Because the majority of the total 99 Tc produced at Hanford (~32,600 Ci) is currently stored in Hanford’s 177 tanks (~26,500 Ci), there is a critical need for knowledge of the fate of this 99 Tc as it is removed from the tanks and processed into a final solid waste form. Current flow sheets for the Hanford Waste Treatment and Immobilization Plant process show most of the 99 Tc will be immobilized as low-activity waste glass that will remain on the Hanford Site and disposed at the Integrated Disposal Facility (IDF); only a small fraction will be shipped to a geologic repository with the immobilized high-level waste. Past performance assessment studies, which focused on groundwater protection, have shown that 99 Tc would be the primary dose contributor to the IDF performance.

storage tanks↗
DOE OSTI2015

Liquid scintillation counting methodology for 99 Tc analysis. A remedy for radiopharmaceutical waste

This paper presents a new approach for liquid scintillation counting (LSC) analysis of single-radionuclide samples containing appreciable organic or inorganic quench. This work offers better analytical results than existing LSC methods for technetium-99 ( 99g Tc) analysis with significant savings in analysis cost and time. The method was developed to quantify 99g Tc in environmental liquid and urine samples using LSC. Method efficiency was measured in the presence of 1.9 to 11,900 ppm total dissolved solids. The quench curve was proved to be effective in the case of spiked 99g Tc activity calculation for deionized water, tap water, groundwater, seawater, and urine samples. Counting efficiency was found to be 91.66% for Ultima Gold LLT (ULG-LLT) and Ultima Gold (ULG). Relative error in spiked 99g Tc samples was ±3.98% in ULG and ULG-LLT cocktails. Minimum detectable activity was determined to be 25.3 mBq and 22.7 mBq for ULG-LLT and ULG cocktails, respectively. A pre-concentration factor of 1000 was achieved at 100°C for 100% chemical recovery.

Cocktail↗
DOE OSTI2019

Synthesis and Characterization of Non-Aqueous [Tc X M-PW11O39] n – with M = O, N: Comparing Tc V and Tc VI in Metal Oxide Matrices

Technetium-99, a prevalent by-product of the nuclear fuel cycle, is a transition metal with 9 accessible oxidation states and a half-life of 2.1 × 10 5 years. 99 Tc VII has been found in the environment surrounding National Lab sites and fuel reprocessing centers, and its separation and long-term storage is complicated by its extensive redox chemistry. In our prior work we have reported the synthesis and reduction of Tc V O to Tc IV O in α 1- and α 2 -Wells–Dawson P 2 W 17 O 61 10- polyoxometalates and reported on the polyoxometalate features that promote reduction and stabilization. In this work we report on the oxidation state stabilization of Tc V O, Tc VI O and Tc VI N within the α-Keggin PW 11 O 39 7- polyoxometalate. Both Density Functional Theory and experimental results show that Tc in TBA 4 [Tc VI N-PW 11 O 39 ] can be reduced to TcV, but any subsequent electrons reduce the tungsten framework of the polyoxometalate. TBA 4 [Tc V O-PW 11 O 39 ] however, can be theoretically reduced down to Tc(I) before reduction of the polyoxometalate occurs.

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

Thermodynamic parameters for the complexation of technetium(IV) with EDTA

Technetium-99 is a high yield (~6% fission yield) fission product and long-lived (2.13×10 5 year half-life) component of nuclear waste that will be disposed of in a geological repository. Some 99 Tc has been released into the environment due to nuclear fuel and weapon production activities at sites such as Hanford, WA. Strongly complexing ligands such as ethylenediamine- N , N , N ', N '-tetraacetic acid (EDTA) are known to increase Tc(IV) solubility and mobility in environmental systems and an accurate quantification of the complexation of Tc(IV) with EDTA is important for predicting its behavior in a geological repository. A liquid–liquid extraction system utilizing 0.2 M TOPO in dodecane was used to measure the stability constants of Tc(IV)-EDTA in 0.50 m NaNO 3 at variable temperatures (14.0±0.1, 25.0±0.1, and 32.0±0.1°C). The acid dependence of the apparent stability constants in the p C H range of 2.00–2.70 indicated the formation of TcO(EDTA) 2- (log β 101 =17.9±0.3, 25.0±0.1°C) and a protonated complex TcO(H)(EDTA) - (log β 111 =20.5±0.1, 25.0±0.1°C). The associated thermodynamic parameters Δ r G 101 =-101.7±0.4 kJ·mol -1 , Δ r H 101 =-47±9 kJ·mol -1 , Δ r S 101 =179±36 J·mol -1 ·K -1 , Δ r G 111 =-117.2±0.3 kJ·mol -1 , Δ r H 111 =-23±5 kJ·mol -1 , and Δ r S 111 =315±63 J·mol -1 ·K -1 (0.50 m NaNO 3 , 25.0±0.1°C) were determined by van’t Hoff analysis. The formation of each Tc(IV)–EDTA complex is exothermic and present favorable entropy terms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Technetium Sorption by Cementitious Materials Under Reducing Conditions

The objective of this study was to measure Tc sorption to cementitious materials under reducing conditions to simulate Saltstone Disposal Facility conditions. Earlier studies were conducted and the experimental conditions were found not to simulate those of the facility. Through a five month subcontract with Clemson University, sorption of 99 Tc to four cementitious materials was examined within an anaerobic glovebag targeting a 0.1% H 2 (g)/ 99.9% N 2 (g) atmosphere. Early experiments based on Tc sorption and Eh indicated that 0.1% H 2 (g) (a reductant) was necessary to preclude experimental impacts from O 2 (g) diffusion into the glovebag. Preliminary data to date (up to 56 days) indicates that sorption of 99 Tc to cementitious materials increased with increasing slag content for simulated saltstone samples. This is consistent with the conceptual model that redox active sulfide groups within the reducing slag facilitate reduction of Tc(VII) to Tc(IV). These experiments differ from previous experiments where a 2% H 2 (g) atmosphere was maintained (Kaplan et al., 2011 (SRNL-STI-2010-00668)). The impact of the 2% H 2 (g) reducing atmosphere on this data was examined and determined to cause the reduction of Tc in experimental samples without slag. In the present ongoing study, after 56 days, Tc sorption by the 50-year old cement samples (no slag) was undetectable, whereas Tc sorption in the cementitious materials containing slag continues to increase with contact time (measured after 1, 4, 8, 19 and 56 days). Sorption was not consistent with spike concentrations and steady state has not been demonstrated after 56 days. The average conditional K d value for the Vault 2 cementitious material was 6,362 mL/g (17% slag), for the TR547 Saltstone (45% slag) the conditional K d was 1258 mL/g, and for TR545 (90% slag) the conditional K d was 12,112 mL/g. It is anticipated that additional samples will be collected until steady state conditions are established to permit measuring more representative K d and solubility values under these experimental conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Molybdenum-99 from Molten Salt Reactor as a Source of Technetium-99m for Nuclear Medicine: Past, Current, and Future of Molybdenum-99

Technitium-99m ( 99m Tc), a widely used radioisotope, is used in tens of millions of medical diagnostic procedures annually. However, it is hard to store and must be immediately used upon production due to its short half-life (i.e., 6 h); thus, it is currently produced from 99 Mo, which itself is a result of 235 U fission. The majority of 99 Mo supplies to U.S. patients are currently provided by foreign producers and produced using highly enriched uranium (HEU). In order to minimize the proliferation risks of HEU-based medical isotope production, the U.S. Department of Energy’s National Nuclear Security Administration has funded a program to accelerate the development of technologies to produce 99 Mo without the use of HEU. Today, the global supply of 99 Mo depends on a limited number of nuclear reactors, and production has been interrupted unexpectedly since 2009 due to the fleet’s advanced age. Herein, alternative options for 99 Mo production are discussed, and one potential option is to obtain 99m Tc from molten salt reactors (MSRs). A MSR is a nuclear fission reactor that can operate at or close to atmospheric pressure with liquid fuel, which allows for producing isotopes in a timely manner. In this paper, the past and current production of 99 Mo via nuclear reactors is described, and the future of 99 Mo production by MSRs is discussed. The behavior and chemical properties of molybdenum in fluoride salts in MSRs and the possible extraction methods are also examined in addition to the limitation of current studies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Technetium Sorption By Cementitious Materials Under Reducing Conditions

The objective of this study was to measure Tc sorption to cementitious materials under reducing conditions to simulate Saltstone Disposal Facility conditions. Earlier studies were conducted and the experimental conditions were found not to simulate those of the facility. Through a five month subcontract with Clemson University, sorption of 99Tc to four cementitious materials was examined within an anaerobic glovebag targeting a 0.1% H 2 (g)/ 99.9% N 2 (g) atmosphere. Early experiments based on Tc sorption and Eh indicated that 0.1% H 2 (g) (a reductant) was necessary to preclude experimental impacts from O 2 (g) diffusion into the glovebag. Preliminary data to date (up to 56 days) indicates that sorption of 99 Tc to cementitious materials increased with increasing slag content for simulated saltstone samples. This is consistent with the conceptual model that redox active sulfide groups within the reducing slag facilitate reduction of Tc(VII) to Tc(IV).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Chalcogen-Based Aerogels as Sorbents for Radionuclide Remediation

The efficient capture of radionuclides having long half-lives such as technetium-99 (99Tc), uranium-238 (238U), and iodine-129 (129I) is pivotal to prevent their transport into groundwater and/or release into the atmosphere. While different sorbents have been considered for capturing each of them, in the current work, a new nanostructured chalcogen-based aerogel, called a chalcogel, is shown to be very effective to capture ionic forms of 99Tc and 238U, as well as nonradioactive gaseous iodine (i.e., a surrogate for 129I), irrespective of the sorbent polarity. Some of the chalcogels performed better than others but the PtGeS sorbent performed the best with capture efficiencies of 98% and 99.4% for 99Tc and 238U, respectively. All sorbents showed >99% capture efficiency for iodine over the test duration. This unified sorbent would be an attractive option in environmental remediation for various radionuclides associated with legacy wastes from nuclear weapons production, wastes from nuclear power production, or potential future nuclear fuel reprocessing.

chalcogel↗
DOE OSTI2013

Investigations Into the Nature of Alkaline Soluble, Non-Pertechnetate Technetium

This report summarizes work accomplished in fiscal year (FY) 2013, exploring the chemistry of a low-valence technetium(I) species, [Tc(CO) 3 (H 2 O) 3 ] + , a compound of interest due to its implication in the speciation of alkaline-soluble technetium in several Hanford tank waste supernatants. Various aspects of FY 2013’s work were sponsored both by Washington River Protection Solutions and the U.S. Department of Energy’s Office of River Protection; because of this commonality, both sponsors’ work is summarized in this report. There were three tasks in this FY 2013 study. The first task involved examining the speciation of [(CO) 3 Tc(H 2 O) 3 ] + in alkaline solution by 99 Tc nuclear magnetic resonance spectroscopy. The second task involved the purchase and installation of a microcalorimeter suitable to study the binding affinity of [(CO) 3 Tc(H 2 O) 3 ] + with various inorganic and organic compounds relevant to Hanford tank wastes, although the actual measure of such binding affinities is scheduled to occur in future FYs. The third task involved examining the chemical reactivity of [(CO) 3 Tc(H 2 O) 3 ] + as relevant to the development of a [(CO) 3 Tc(H 2 O) 3 ] + spectroelectrochemical sensor based on fluorescence spectroscopy.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗