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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

The Incorporation of Pump-and-Treat Data in the New Tracking Restoration and Closure (TRAC)Web-Based Mapping Tool - 20339

Groundwater pump-and-treat (P and T) systems are a common remediation strategy for sites with contaminated groundwater within the U.S. Department of Energy (DOE) Office of Environmental Management (EM). Currently, there are six DOE-EM sites with active P and T systems, but integrated information on individual systems is only primarily available in separate annual reports. To provide a broad view of both current and historical P and T operations, data has been collected on all of the P and T systems. This summary information not only includes capital and average annual costs, but also identifies contaminants treated and forecasted P and T closure dates. Principal contaminants of concern treated with P and T systems include trichloroethylene, chromium (VI), strontium-90, technetium-99, and uranium. DOE-EM is continuing to create a web-based mapping tool that compiles information on groundwater contaminants, remediation strategies and, current data related to plumes and cleanup progress within all DOE-EM sites. The information collected on P and T systems will be added to the TRAC (Tracking Restoration and Closure) tool using interactive maps to quickly access and share information. This information sharing supports the transition from active to passive remediation methods and long-term monitoring approaches. TRAC can ensure that managers, stakeholders, regulators, and contractors remain up to date on cleanup progress for all sites within the complex. The information on P and T systems expands the use of TRAC and can assist in project monitoring and budget planning and support potential shifts in management plans based on plume data. (authors)

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

Immobilization of Radioactive Materials Using Porous Iron Composite Media - 18582

High levels of nitrate (NO{sub 3}) derived from agriculture and naturally occurring radioactive materials (NORM), primarily uranium (U) and natural decay chain products that have been mobilized by the influx of oxygenated, high-alkalinity water, are a threat to shallow groundwater resources in California. Porous Iron Composite (PIC) materials were evaluated as a water treatment option in a series of batch experiments. While the use of Zero Valent Iron (ZVI) to address groundwater contaminated with NO{sub 3} or U has been demonstrated, the ability of PICs to chemically reduce and immobilize U and other associated contaminants in oxygenated alkaline groundwater where soluble uranyl-carbonate species dominate and high NO{sub 3} levels may compete for reductive capacity remain a concern. Additionally, ZVI materials have been used to immobilize other redox sensitive radionuclides (e.g., technetium-99 ({sup 99}Tc)) with limited success. Batch tests were conducted using an artificial groundwater surrogate (i.e., CAGW for California Groundwater) containing ∼ 100 mg L{sup -1} NO{sub 3}{sup -}, 200 mg L{sup -1} alkalinity, 100 μg L{sup -1} U, 100 μg L{sup -1} rhenium (Re) as a non-radioactive surrogate for the nuclear fission product technetium-99 ({sup 99}Tc), 100 μg L{sup -1} stable strontium (Sr) as a surrogate for radium (Ra) and the fission product strontium-90 ({sup 90}Sr), and 50 μg L{sup -1} arsenic (As), a common groundwater contaminant. Five gram samples of the PIC material were reacted with 200 mL of CAGW under two different atmospheric conditions to evaluate the impact of O{sub 2}. Batch tests were also conducted in the presence and absence of another oxidized species, NO{sub 3}. At specific time intervals, a small aliquot was collected from each batch treatment for chemical analysis (i.e., U, Re, As, Fe, NO{sub 3}, etc.) and the pH and ORP of the remaining suspension was measured. The PIC was effective at fully reducing NO{sub 3}{sup -} levels over the extended reaction time, even though the reactor pH quickly increased above 8.0 during equilibration. Soluble levels of U and Re decreased in a similar fashion to NO{sub 3}{sup -} in the presence and absence of O{sub 2}. Soluble As levels decreased more rapidly than other contaminants, but remained at ∼ 1 μg L{sup -1}. Very little nitrite (NO{sub 2}{sup -}) was detected in the batch tests, with NH{sub 4}{sup +} levels consistent with the amount of NO{sub 3}{sup -} reduced. Similar rates of immobilization for U, Re, and As were observed in the presence and absence of NO{sub 3}{sup -}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Zero Valent Iron for Reductive Removal of Technetium-99 from Aqueous Sulfate Solutions - 20347

This research investigates the reductive removal of technetium-99 ({sup 99}Tc) by zero valent iron (ZVI) from low activity waste (LAW) off-gas condensate simulant as a secondary treatment after recovery from vitrification planned at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Due to its high volatility, only a fraction of Tc{sup (VII)} will be incorporated into glass waste forms. Volatilized Tc will be captured by an off-gas treatment system with current plans to recycle off-gas condensate back to the vitrification facility. The scheme with off-gas recycling will increase Tc loading in the glass waste, but will also increase the concentrations of sulfate, halides, and other problematic constituents impeding overall LAW processing and increasing volume of the glass product. This study focuses on Tc{sup (VII)} removal by ZVI as a feasible pathway to minimize off-gas condensate recycling, which may result in a reduction in the volumes of LAW waste to be immobilized and subsequent cost savings. ZVI is an established treatment agent for redox-active contaminants such as trichloroethylene, nitrate, arsenic, chromium, phenol, and others. It is a commercially available and cost-effective material. Our previous experiments showed that ZVI is very efficient for the reductive removal of {sup 99}Tc. In this work we studied ZVI oxidation under aerobic conditions at pH 7 with 0.1 M Na{sub 2}SO{sub 4} solution (ionic strength, IS, 0.3 M) in the presence and absence of {sup 99}Tc. The concentration of {sup 99}Tc and the changes in pH, dissolved oxygen (DO), and oxidation-reduction potential (ORP) of the simulated solutions were monitored over 8 days. The formation of iron oxide phases was probed by X-ray diffraction (XRD). Obtained results suggest that ZVI contact time with {sup 99}Tc containing solutions for 6 hours resulted in nearly complete removal of {sup 99}Tc from the aqueous phase. XRD analysis showed that the oxidation of ZVI is rapid with formation of magnetite (Fe{sub 3}O{sub 4}) with minor percentage of goethite and maghemite. This work is a part of a larger set of studies investigating the feasibility of {sup 99}Tc reductive removal from the LAW off-gas condensate. (authors)

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

Selection and Validation of Alternate Slag Sources for Saltstone Production - 16145

To stabilize low activity salt solution produced at the Savannah River Site (SRS) the salt waste is mixed with a combination of fly ash (FA), ground granulated blast furnace slag (GGBFS), and ordinary Portland cement (OPC) to form a grout termed saltstone. Upon curing, the low activity waste is encapsulated within the hardened cementitious matrix. While the sources of the dry feed components have remained unchanged for several years, SRS was recently notified that its domestic source of GGBFS would no longer be available due to the shutdown of the steel making plant that produced the ferrous slag as a by-product. This shut-down follows a trend in the reduction of domestic ferrous slag sources. Alternate GGBFS vendors were sought and a program developed to validate the alternate sources for use in saltstone. GGBFS is primarily utilized in the construction industry as a replacement for OPC and the requirements of a GGBFS source, based on an American Society for Testing and Materials (ASTM) standard, are predominantly limited to demonstrating that the GGBFS replacement will provide equivalent strength in comparison to the baseline OPC-based concrete. This affords the construction industry significant flexibility with respect to the GGBFS source that can be used. In contrast, GGBFS has a role of significantly more importance when utilized in saltstone. In particular, the reducing blast furnace environment under which GGBFS is produced yields reducing species, such as sulfide (S{sup 2-}) and ferrous iron (Fe{sup 2+}), which are believed to facilitate the long-term immobilization of redox sensitive contaminants (e.g. technetium-99 ({sup 99}Tc)). In addition, a change in GGBFS source has the potential to impact the rheological properties and the reactive heat generation of the grout, which may adversely affect the ability to pump the grout to a Saltstone Disposal Unit (SDU) and the temperatures attained within an SDU, respectively. As such, alternate GGBFS sources for the processing of saltstone require significantly more scrutiny than is prescribed by the ASTM standard. Four alternate GGBFS sources were evaluated. All four sources met the ASTM requirements, only one source was considered viable for use in saltstone based on preliminary evaluation. The other three sources will require a more intensive assessment to determine their potential applicability. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Technetium and iodine getters to improve Cast Stone performance - 15420

Technetium-99 ({sup 99}Tc) and iodine-129 ({sup 129}I) have been determined to be tank waste components contributing most to the environmental impacts associated with the cleanup of the Hanford site. Cast Stone, which has the ability to immobilize {sup 99}Tc, {sup 129}I, and other contaminants of concern, is being considered as a low temperature waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) An environmental assessment of Cast Stone performance, using measured diffusivities of the radionuclides from the Cast Stone product, calculates groundwater at the 100 m down-gradient in excess of the allowable maximum permissible concentrations for both radionuclides. Therefore, an opportunity exists to improve the performance of Cast Stone. One method for improvement is through the addition of immobilizing agents known as 'getters' that selectively sequester Tc and I. In the present paper, we report on seven getter materials that have been tested for Tc and five materials that have been tested for I to determine their effectiveness for radionuclide removal in a series of batch sorption tests in 18.2 MΩ DI H{sub 2}O and a 7.8 M Na Ave LAW waste simulant. Contact experiments have been carried out in anoxic conditions at room temperature, with varying solid-to-solution ratios, and in the presence and absence of competing species. Results indicate that most materials perform better in the 18.2 MΩ DI H{sub 2}O compared to the 7.8 M Na LAW waste simulant for both Tc and I getter. Radionuclide sequestration may be affected by the presence of the redox sensitive Cr(VI) in the simulant. The top performing Tc getter materials, Sn(II)-apatite, BFS-2, KMS-2 and Sn-hydroxyapatite, were examined through various solid-state characterization techniques such as SEM/XEDS, XANES and EXAFS. Tc removal is theorized to occur through reduction of the highly soluble Tc(VII) species to the less soluble Tc(IV). Solution results from the series of I tests show that iodide is effectively removed from solution via, most likely, a precipitation reaction of AgI. A wide range of I getter materials and the silver exchanged zeolite getter performed well in both DI water and the simulant. (authors)

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

Diffusion and Leaching Behavior of Radionuclides in Category 3 Waste Encasement Concrete and Soil Fill Material – Summary Report

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Such concrete encasement would contain and isolate the waste packages from the hydrologic environment and would act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed, and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. The mobilized radionuclides may escape from the encased concrete by mass flow and/or diffusion and move into the surrounding subsurface environment. Therefore, it is necessary to assess the performance of the concrete encasement structure and the ability of the surrounding soil to retard radionuclide migration. The retardation factors for radionuclides contained in the waste packages can be determined from measurements of diffusion coefficients for these contaminants through concrete and fill material. Some of the mobilization scenarios include (1) potential leaching of waste form before permanent closure cover is installed; (2) after the cover installation, long-term diffusion of radionuclides from concrete waste form into surrounding fill material; (3) diffusion of radionuclides from contaminated soils into adjoining concrete encasement and clean fill material. Additionally, the rate of diffusion of radionuclides may be affected by the formation of structural cracks in concrete, the carbonation of the buried waste form, and any potential effect of metallic iron (in the form of rebars) on the mobility of radionuclides. The radionuclides iodine-129 ({sup 129}I), technetium-99 ({sup 99}Tc), and uranium-238 ({sup 238}U) are identified as long-term dose contributors in Category 3 waste (Mann et al. 2001; Wood et al. 1995). Because of their anionic nature in aqueous solutions, {sup 129}I, {sup 99}Tc, and carbonate-complexed {sup 238}U may readily leach into the subsurface environment (Serne et al. 1989, 1992a, b, 1993, and 1995). The leachability and/or diffusion of radionuclide species must be measured to assess the long-term performance of waste grouts when contacted with vadose-zone pore water or groundwater. Although significant research has been conducted on the design and performance of cementitious waste forms, the current protocol conducted to assess radionuclide stability within these waste forms has been limited to the Toxicity Characteristic Leaching Procedure, Method 1311 Federal Registry (EPA 1992) and ANSI/ANS-16.1 leach test (ANSI 1986). These tests evaluate the performance under water-saturated conditions and do not evaluate the performance of cementitious waste forms within the context of waste repositories which are located within water-deficient vadose zones. Moreover, these tests assess only the diffusion of radionuclides from concrete waste forms and neglect evaluating the mechanisms of retention, stability of the waste form, and formation of secondary phases during weathering, which may serve as long-term secondary hosts for immobilization of radionuclides. The results of recent investigations conducted under arid and semi-arid conditions (Al-Khayat et al. 2002; Garrabrants et al. 2002; Garrabrants and Kosson 2003; Garrabrants et al. 2004; Gervais et al. 2004; Sanchez et al. 2002; Sanchez et al. 2003) provide valuable information suggesting structural and chemical changes to concrete waste forms which may affect contaminant containment and waste form performance. However, continued research is necessitated by the need to understand: the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties, and the associated impact on contaminant release. Recent reviews conducted by the National Academies of Science recognized the efficacy of cementitious materials for waste isolation, but further noted the significant shortcomings in our current understanding and testing protocol for evaluating the performance of various formulations.

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

Cementitious Waste Form Development for Treatment of EMF Evaporator Bottoms - 18611

Direct Feed Low-Activity Waste operations at the Hanford Site (Washington State, USA) involve concentrating the Hanford low-activity waste (LAW) melter off-gas condensate waste stream by evaporation at the Effluent Management Facility (EMF) followed by recycle back to the melter feed. The EMF evaporator bottoms waste stream is expected to contain variable, high levels of halides and sulfate that could strain LAW melter performance and co-mingled radionuclides, technetium-99 ({sup 99}Tc) and iodine-129 ({sup 129}I), that readily volatilize in the LAW melter and are expected to accumulate during recycle. To this end, there is a need for alternative disposition paths for the EMF evaporator bottoms waste stream that bypass recycling to the LAW melter. One such alternative is solidification in low-temperature cementitious waste forms (CWFs) that meet disposal facility acceptance criteria. The work herein addresses recent waste form development and testing performed to identify CWF candidates that meet off-site disposal acceptance criteria and/or on-site Hanford Integrated Disposal Facility waste acceptance criteria. Phase one testing activities included the production and evaporation of simulated EMF evaporator feed from the Hanford Waste Treatment and Immobilization Plant melter offgas effluent liquid secondary waste stream, which was then analyzed to demonstrate the behavior and fate of contaminants of concern (COCs). Based on analyses from these preliminary tests, a simulant for the EMF evaporator bottoms waste stream was formulated and used to evaluate several dry material compositions proposed for CWF development and testing. In particular, dry ingredient recipes used in early testing efforts contained a combination of ordinary Portland cement (OPC), Class F fly ash (FA), blast furnace slag (BFS), Aquaset{sup R} II-H, and/or Aquaset{sup R} II-GH to generate CWFs at various water-to-dry material ratios. CWFs were then tested according to the Toxicity Characteristic Leaching Procedure (TCLP) for COCs, such as Cr, Zn, As, and Se, to determine which specimens comply with land disposal restrictions (40 CFR 268). The results of TCLP testing showed that chromium leaching was a persistent challenge, except when using the original Cast Stone dry ingredient recipe (8% OPC, 47% BFS, 45% FA) to immobilize the EMF evaporator bottoms simulant. Improving Cr immobilization and evaluating leaching behavior for COC Hg were identified as objectives for the second phase of testing. During second phase testing efforts, a screening test was performed to determine whether the bulk EMF waste composition is a factor in producing an acceptable solidified waste form. A series of eight simulants with varying concentrations of major salts of boron, chloride, nitrite, and sulfate were prepared and immobilized using three dry ingredient recipes. Each dry ingredient recipe contained two to three of the following: BFS, OPC, FA, and Aquaset{sup R} II-GH. All eight simulants were spiked with Zn, Cr, As, Se, and Hg prior to production of the CWFs, similar to previous tests, and used to test for residual free liquids and TCLP analysis. The TCLP results indicate that all phase two formulated CWFs meet land disposal requirements when compared to Universal Treatment Standards (40 CFR 268) for hazardous wastes, regardless of simulant composition. Furthermore, re-absorption of residual free liquids is most quickly achieved for nearly all simulant compositions using the original Cast Stone recipe; however, replicas analysis is required to corroborate these findings. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Technetium Complexation with Multidentate Carboxylate-Containing Ligands: Trends in Redox and Solubility Phenomena

The chemistry of technetium (t 1/2 ( 99 Tc) = 2.11 × 10 5 years) is of particular importance in the context of nuclear waste disposal and historic contaminated sites. Polycarboxylate ligands may be present in some sites and are potentially capable of strong complexing interactions, thus increasing the solubility and mobility of 99 Tc under environmentally relevant conditions. This work aimed to determine the impact of five organic complexing ligands [L = oxalate, phthalate, citrate, nitrilotriacetate (NTA), and ethylenediaminetetraacetate (EDTA)] under anoxic, alkaline conditions (pH ≈ 9–13) on the solubility of technetium. X-ray absorption spectroscopy confirmed that TcO 2 (am,hyd) remained the solubility-controlling solid phase in undersaturation solubility experiments. Ligands with maximum coordination numbers (CN) ≥ 3 (EDTA, NTA, and citrate) exhibited an increase in solubility from pH 9 to 11, while ligands with CN ≤ 2 (oxalate and phthalate) at all investigated pH and CN ≥ 3 at pH ≈ 13 were outcompeted by hydrolysis reactions. Though most available thermodynamic values were determined under acidic conditions, these models satisfactorily explained high-pH undersaturation solubility of technetium for citrate and NTA, whereas experimental data for Tc(IV)–EDTA were highly overestimated. Finally, this work illustrates the predominance of hydrolysis under hyperalkaline conditions and provides experimental support for existing thermodynamic models of Tc–L except Tc–EDTA, which requires further research regarding aqueous speciation and solubility.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Theoretical Modeling of 99 Tc NMR Chemical Shifts

Technetium (Tc) displays a rich chemistry due to the wide range of oxidation states (from -I to +VII) and ability to form coordination compounds. Determination of Tc speciation in complex mixtures is a major challenge, and 99Tc NMR spec-troscopy is widely used to probe chemical environments of Tc in odd oxidation states. However interpretation of the 99Tc NMR data is hindered by the lack of reference compounds. DFT computations can help fill this gap, but to date few com-putational studies have focused on 99Tc NMR of compounds and complexes. This work systematically evaluates the inclu-sion small percentages of Hartree-Fock exchange correlation and relativistic effects in DFT computations to support in-terpretation of the 99Tc NMR spectra. Hybrid functionals are found to perform better than their pure GGA counterparts, and non-relativistic calculations have been found to generally show a lower mean absolute deviation from experiment. Overall non-relativistic PBE0 and B3PW91 calculations are found to most accurately predict 99Tc NMR chemical shifts.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2013

Selective Internal Radiotherapy (SIRT) of Hepatic Tumors: How to Deal with the Cystic Artery

PurposeSelective internal radiotherapy (SIRT) with the beta emitter yttrium-90 (Y90) is a rapidly developing therapy option for unresectable liver malignancies. Nontarget irradiation of the gallbladder is a complication of SIRT. Thus, we aimed to assess different strategies to avoid infusion of Y90 into the cystic artery (CA).MethodsAfter hepatic digital subtraction angiography and administration of technetium-99m-labeled human serum albumin ({sup 99}mTc-HSA), 295 patients with primary or secondary liver tumors underwent single-photon emission computed tomography/computed tomography (SPECT/CT). Different measures were taken before repeated Y90 mapping and SIRT to avoid unintended influx into the CA where necessary. Clinical symptoms, including pain, fever, or a positive Murphy sign, were assessed during patient follow-up.ResultsA significant {sup 99}mTc-HSA accumulation in the gallbladder wall (higher {sup 99}mTc-HSA uptake than in normal liver tissue) was seen in 20 patients. The following measures were taken to avoid unintended influx into the CA: temporary/permanent occlusion of the CA with gelfoam (n = 5)/microcoil (n = 1), induction of vasospasm with a microwire (n = 4), or altering catheter position (n = 10). Clinical signs of cholecystitis were observed in only one patient after temporary CA occlusion with gelfoam and were successfully treated by antibiotics. Cholecystectomy was not required for any patient.ConclusionIt is important to identify possible nontarget irradiation of the gallbladder. The risk for radiation-induced cholecystitis can be easily minimized by temporary or permanent CA embolization, vasospasm induction, or altering the catheter position.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2019

Characterization of Tc-99 Speciation and Chemical Composition of 241-AN-102 Tank Waste Supernatant - 19201

Technetium (Tc) in its isotopic form {sup 99}Tc is generated in high yields from the fission of {sup 235}U and {sup 239}Pu. A significant inventory of {sup 99}Tc remains at the U.S. Department of Energy's Hanford Site, and it is one of the most difficult contaminants to address. In the strongly alkaline solutions that are typical of Hanford tank waste, Tc exists as both pertechnetate (TcO{sub 4}{sup -} in oxidation state VII) and in reduced forms (oxidation state <VII) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized anionic Tc species that can be removed from low-activity waste by anion exchange or other methods. There is no definitive information on the origin or a comprehensive description of the non-pertechnetate species in Hanford tanks. Designing strategies for effective management of Tc, including separation and immobilization, requires an understanding of the molecular structure of the non-pertechnetate species and their identification in actual tank waste samples. Developing this advanced understanding would facilitate development of new technologies for treating dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO){sub 3}]{sup +} and Tc(IV, VI, VII) compounds was generated using a range of techniques and applied to the characterization of actual 241-AN-102 tank waste supernatant. A sample of the 241-AN-102 tank waste (collected in 2013) was processed to remove {sup 137}Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin after the sodium concentration was adjusted to 5.6 M. The Cs-loaded sRF column was eluted with 0.45 M HNO{sub 3}. As-received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using {sup 99}Tc nuclear magnetic resonance (NMR) spectroscopy and X-ray absorption spectroscopy. The results of these analyses demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO){sub 3}L{sub 3}]{sup +}, Tc(IV), and possibly Tc(VI) compounds. This is the second experimental observation of the [fac-Tc(CO){sub 3}L{sub 3}]{sup +} species in the Hanford tank waste that highlights their importance for waste processing. In addition, the chemical composition of the 241-AN-102 tank waste was characterized with the focus on small chelators. (authors)

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

Scoping Tests of Technetium and Iodine Removal from Tank Waste Using SuperLig® 639 Resin

The primary chemical form of 99 Tc found in Hanford Low Activity Waste (LAW) is pertechnetate anion (TcO4 -), which is highly soluble in water, and is mobile if released to the environment. Pertechnetate will not be removed from the aqueous waste in the Hanford waste treatment plant, and the primary disposition path is immobilization in the LAW glass waste form, which will be disposed in the Integrated Disposal Facility (IDF). Due to the soluble properties of pertechnetate, and the potential for impact to the Performance Assessment (PA), effective management of 99 Tc is important to the overall success of the River Protection Project mission. Washington River Protection Solutions (WRPS) is developing some conceptual flow-sheets for LAW treatment and disposal that could benefit from technetium removal. While 99 Tc is the primary radionuclide of interest, 129 I also contributes to the calculated future dose of disposed LAW, and it would be of interest to examine if removal is possible.

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

Structural Properties of Ditechnetium Heptoxide, Tc{sub 2}O{sub 7}

Group 7 elements form unique binary oxides with the stoichiometry M{sub 2}O{sub 7}. The heptoxides of manganese and technetium crystallize as molecular solids while rhenium forms an extended solid with mixed valances, 4 and 6. Unlike manganese and rhenium heptoxides which are used in the chemical industry technetium heptoxide has found very few applications due to its radioactive nature; it has been primarily use as a chemical precursor. While both of the molecular solids, Mn{sub 2}O{sub 7}{sup 3} and Tc{sub 2}O{sub 7}{sup 4} can be described by two corner-sharing MO{sub 4} tetrahedron. The most distinctive difference is that the manganese's compound has a metal-oxygen-metal bond angle of 120.7 deg. while the technetium species is 180 deg.. Additionally, Tc{sub 2}O{sub 7} has the largest molar volume of the heptoxides which suggest that it will be more susceptible to external stimuli of pressure and temperature than the others heptoxides. A pressure induced twisting has been predicted by Fang et al. Anomalous behavior is seen in the c-axis where below 2 GPa the Van der Waals interactions between the Tc{sub 2}O{sub 7} molecules is over come and compression is observed but at higher pressures the twisted distortion forces the lattice to expand. The terminology diagonal-double tetrahedrons (DDTs) is introduced to discuss the molecules of Tc{sub 2}O{sub 7}. The rigid nature of the DDTs has been attributed to the cause of this twisted distortion mechanism. Recently, the solid state and gas phase behavior of technetium heptoxide was revisited by Childs et al. in light of {sup 99}Tc being an important dose contributor in the Hanford waste and vitrification being proposed for immobilization of nuclear waste. Technetium heptoxide is one of the proposed volatilized species of technetium found in the vitrification process. In this study the crystal structure of Tc{sub 2}O{sub 7} was reported at 100 K and compared to the structure first determined by Krebs at room temperature. Inspection of the metal-oxygen distances revealed the 100 K structure had longer bond lengths for the bridging oxygen than at high temperature. It is suggested that transverse thermal vibrational modes can be used to explain this negative thermal expansion behavior. This describes the average bond length of a central atom (i.e. the bridging oxygen atom) being measured as shorter at higher temperatures due to the larger thermal motion. Diffraction methods that measure the long range average structure are susceptible to observing this shorter average length as opposed to the true bond length, atom position at oscillation node that a method like Pair-Distribution Functional PDF Analysis would yield. In the case of Tc{sub 2}O{sub 7}, the magnitude of the transverse thermal vibrational mode is at the limit of being detectable by SCXRD and likely not observable by PDF analysis. Here we report the preliminary results of our study on the nature of the thermal expansion behavior of Tc{sub 2}O{sub 7}. The transverse thermal vibration mode compression mechanism is considered and compared to the pressure induced twisted distortion mechanism predicted at high pressure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Irradiation of 20L LEU Uranyl Sulfate Solution for Production of Mo-99

Argonne National Laboratory, with support from the National Nuclear Security Administration’s Office of Material Management and Minimization, is assisting SHINE Medical Technologies in the development of accelerator-driven production of 99 Mo using low-enriched uranium (LEU). Technetium-99m ( 99m Tc), a daughter of 99 Mo, is a workhorse of nuclear medicine, used in approximately 40,000 medical testing procedures each day just in the U.S.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Recovering high-purity uranyl nitrate from simulated used nuclear fuel dissolver solutions by crystallization: rejecting technetium

The separation of U from Tc and other problematic fission product elements like Mo and Ru, along with Sr, Zr, Cs, and Nd, has been achieved via the crystallization of uranyl nitrate hexahydrate (UNH). Rejection of technetium as pertechnetate anion ( 99 TcO 4 – ) is an especially important feature of this system, as it otherwise tends to follow U (VI) in extractive separations. It also raises the salient question regarding why this oxoanion cannot replace nitrate within the crystalline lattice of UNH. Results showed high-yield (>90 %), high-purity (>99 %) recovery of U as UNH from solutions containing 99 TcO 4 – by simple reduction of temperature from 60°C to 20°C. There was no observable interaction of 99 TcO 4 – with UO 2 2+ . The addition of other cations like, Sr 2+ , Zr 4+ , Cs + , and Nd 3+ , also did not form secondary, contaminant solid phases, leaving the > 99 % of the fission product elements in the mother liquor, while the U was recovered at > 90 %. Similarly, Mo and Ru, when added to the mixture, were shown to behave as the other fission-product elements, remaining in the mother liquor during crystallization. As a result, DFT calculations showed that, despite the higher binding strength of TcO 4 – , HMoO 4 – , and BiO 3 – with the UO 2 2+ cation compared to NO 3 – , the hydrogen-bonding network of the two coordinated ions and four waters of hydration in the UNH crystal structure is the driving force for the high specificity of this separation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Deep liquid-chromatographic purification of uranium extract from technetium

The recycling of uranium in the nuclear fuel cycle requires the removal of a number of radioactive and stable impurities like {sup 99}Tc from spent fuels. In order to improve the grade of uranium extract purification from technetium the method of liquid chromatography and the apparatus for its performance have been developed. Process of technetium extraction and concentrating in aqueous solution containing reducing agent has been studied on simulated solutions (U-Tc-HNO{sub 3}-30% TBP-isoparM). The dynamic tests of the method have been carried out on the laboratory unit. Solution of diformyl-hydrazine in nitric acid was used as a stationary phase. Silica gel with specific surface of 186 m{sup 2}/g was used as a carrier of the stationary phase. It is shown that the volume of purified extract increases as the solution temperature increases, concentration of reducing agent increases and extract flow rate decreases. It is established that the technetium content in uranium by this method could achieve a value below 0.3 ppm. Some variants of overload and composition of the stationary phase containing the extracted technetium have been offered and tested. It is defined that the method provides reduction of processing medium-active wastes by more than 10 times during finish refining process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Laboratory-Scale Melter Systems for Simulated and Real AP-107 Tank Waste - 20483

The Hanford Waste Treatment and Immobilization Plant (WTP) will process and stabilize waste that is stored in underground tanks on the Hanford Site. At the WTP, the tank waste will be combined with glass-forming and modifying additives to form an aqueous and solid mixture called melter feed, which will then be vitrified in joule-heated melters. A small-scale melter system, the continuous laboratory-scale melter (CLSM), is designed to operate with a continuous feeding process while periodically pouring glass product and collecting off-gas. Two CLSMs were constructed: one in a non-radioactive environment for processing waste simulants and another in a fume hood capable of handling radioactive material for processing actual Hanford tank waste. The waste sample received for vitrification in the CLSM was from Hanford tank 241-AP-107. Approximately 12 liters of melter-feed slurry was vitrified into 7 kilograms of glass product over the course of 10 hours of charging to the CLSM. The partitioning of technetium-99 in the product streams was used to calculate the technetium-99 recovery and help determine when the vitrification process had reached a steady state. The average single-pass retention of technetium-99 in the glass product was then calculated for the processing of AP-107 waste in the CLSM system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗