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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Literature review of the potential impact of glycolic acid on the technetium chemistry of srs tank waste

This document presents a literature study of the impact of glycolate on technetium chemistry in the Savannah River Site (SRS) waste system and specifically Saltstone. A predominant portion of the Tc at SRS will be sent to the Saltstone Facility where it will be immobilized. The Tc in the tank waste is in the highly soluble chemical form of pertechnetate ion (TcO 4 - ) which is reduced by blast furnace slag (BFS) in Saltstone, rendering it highly insoluble and resistant to leaching.

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

Accumulation mechanisms for contaminants on weak-base hybrid ion exchange resins

Mechanism of hexavalent chromium removal (Cr(VI) as CrO 4 2- ) by the weak-base ion exchange (IX) resin ResinTech® SIR-700-HP (SIR-700) from simulated groundwater is assessed in the presence of radioactive contaminants iodine-129 (as IO 3 - ), uranium (U as uranyl UO 2 2+ ), and technetium-99 (as TcO 4 - ), and common environmental anions sulfate (SO 4 2- ) and chloride (Cl-). Batch tests using the acid sulfate form of SIR-700 demonstrated Cr(VI) and U(VI) removal exceeded 97%, except in the presence of high SO 4 2- concentrations (536 mg/L) where Cr(VI) and U(VI) removal decreased to ≥ 80%. However, Cr(VI) removal notably improved with co-mingled U(VI) that complexes with SO 4 2- at the protonated amine sites. These U–SO 4 2- complexes are integral to U(VI) removal, as confirmed by the decrease in U(VI) removal (<40%) when the acid chloride form of SIR-700 was used instead. Solid phase characterization revealed that CrO 4 2- is removed by IX with SO 4 2- complexes and/or reduced to amorphous Cr(III)(OH) 3 at secondary alcohol sites. Tc(VII)O 4 - and I(V)O 3 - also undergo chemical reduction, following a similar removal mechanism. Oxyanion removal preference is determined by the anion reduction potential (CrO 4 2 - >TcO 4 - >IO 3 - ), geometry, and charge density. For these reasons, 39% and 69% of TcO 4 - and 17% and 39% of IO 3 - are removed in the presence and absence of Cr(VI), respectively.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Tank Closure Cesium Removal at Savannah River Site - 19604

Savannah River Remediation (SRR) has developed a supplemental at-tank ion exchange capability for the removal of cesium from liquid salt waste. Building on the development of previous ion exchange initiatives and the experience of commercial nuclear plant decontamination services, and following the disaster response associated with Fukushima, the technology has matured to deploy a modular at-tank treatment system for the selective removal of cesium from bulk salt waste. If feasible, this Tank Closure Cesium Removal (TCCR) initiative would provide a supplemental treatment capability to existing and future salt processing and improve confidence in supporting the desired acceleration of waste retrieval and tank closure efforts. In addition to accelerating waste retrieval and tank closure activities, demonstration of this capability has several other benefits. Deployment of commercially available technologies with minor modification can expand Department of Energy (DOE) utilization of commercially supplied treatment processes. This 'at-tank' treatment system could be modified to evaluate and demonstrate removal of other radioactive constituents (strontium, technetium, iodine, etc.) of bulk waste. The modular aspect of this capability can allow for additional units to be delivered and deployed at SRS, as well as, provide supplemental treatment capabilities for waste treatment at other DOE sites. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Raman Analysis of Perrhenate and Pertechnetate in Alkali Salts and Borosilicate Glasses

Sodium borosilicate glasses containing various concentrations of rhenium or technetium were fabricated, and their vibrational spectra studied using a Raman microscope. Spectra were interpreted with reference to new high resolution measurements of alkali pertechnetates and perrhenates NaReO4, KReO4, NaTcO4, and KTcO4. At low concentrations of ReO4- or TcO4-, glass spectra show weak peaks superimposed on a dominant spectrum of glass characteristic of silicate and borate network vibrations. At high concentrations, sharp peaks characteristic of crystal field splitting and C4h symmetry dominate the spectra of glasses, indicating alkali nearby tetrahedral Re or Tc. Often peaks indicative of both the K and Na pertechnetates/ perrhenates are evident in the Raman spectrum, with the latter being favored at high additions of the source chemical, since Na is more prevalent in the glass and ion exchange takes place. These results have significance to immobilization of nuclear waste containing radioactive 99Tc in glass for ultimate disposal.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2014

Perrhenate incorporation into binary mixed sodalites: The role of anion size and implications for technetium-99 sequestration

Perrhenate (ReO 4 - ), as a TcO 4 - analogue, was incorporated into mixed-anion sodalites from binary solutions containing ReO 4 - and a competing anion X n- (Cl - , CO 3 2- , SO 4 2- , MnO 4 - , or WO 4 2- ). For this study, our objective was to determine the extent of solid solution formation and the dependence of competing ion selectivity on ion size. Using equivalent aqueous concentrations of the anions (ReO 4 - /X n- molar ratio = 1:1), we synthesized mixed-anion sodalites from zeolite and NaOH at 90 °C for 96 h. The resulting solids were characterized by bulk chemical analysis, powder X-ray diffraction, scanning electron microscopy, and X-ray absorption near edge structure (XANES) spectroscopy to determine crystal structure, chemical composition, morphology, and rhenium (Re) oxidation state. Rhenium in the solid phase occurred predominately as Re(VII)O 4 - in the sodalites, which have a primitive cubic pattern in the space group P43n. The refined unit-cell parameters of the mixed sodalites ranged from 8.88 to 9.15 Å and showed a linear dependence on the size and mole fraction of the incorporated anion(s). The ReO 4 - selectivity, represented by its distribution coefficient (K d ), increased in the following order: Cl - < NO 3 - < MnO 4 - and CO 3 2- < SO 4 2- < WO 4 2- for the monovalent and divalent anions, respectively. The relationship between the ReO 4 - distribution coefficient and competing anion size was nonlinear. When the difference in ionic radius (DIR) between ReO 4 - and X n - (n = 1 or 2) was greater than ~ 12%, then ReO 4 - incorporation into sodalite was insignificant. The results imply that anion size is the major factor that determines sodalite anion compositions. Given the similarity in chemical behavior and anion size, ReO 4 - serves as a suitable analogue for TcO 4 - under oxidizing conditions where both elements are expected to remain as oxyanions in the + 7 oxidation state.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Functionalized Magnetic Nanoparticles for Technetium Sequestration from Groundwater

Technetium 99 (Tc) is among the most common environmental contaminants at DOE sites and one of the most common risk drivers in low- and high-level waste disposal sites. The majority of Tc is generated from anthropogenic sources, such as nuclear power plants, global weapons, nuclear storage facilities and medical applications. Through these sources, Tc contamination has been unintentionally introduced in to the environment. The most common chemical form of Tc is Tc(VII)O{sub 4}{sup -}. Due to its high solubility and mobility, Tc can enter the food chain and cause adverse health effects to humans. Currently, ion exchange resins and reduction processes are the most common approaches for Tc immobilization. Although these techniques have shown to be effective, they also possess major drawbacks, such as high cost, low adsorption capacity, and complex creation and maintenance. Therefore, development of more efficient and simple technologies for the remediation of Tc-contaminated systems are needed. Functionalized magnetic nanoparticles have been used to remove organic and inorganic contaminants from water resources. These nanoparticles have attracted extensive attention as an adsorbent material due to their large surface area, high efficiency, low-cost, easy functionalization and separation with a magnet. This study seek to develop functionalized magnetic iron oxide nanoparticles for the efficient removal of Tc and other heavy metal contaminants from water resources under ambient conditions. Objectives: Synthesize magnetic iron oxide nanoparticles and functionalize their surface with Cetyltrimethylammonium Bromide (CTAB) and tetraethyl-orthosilicate (TEOS). Characterize the synthesized nanoparticles using scanning electron microscopy (SEM) coupled with energy dispersive X-ray spectroscopy (EDS), Dynamic Light Scattering (DLS) and Zeta PALS. Perform adsorption studies to evaluate their adsorption behavior and capacity for (a) Technetium using Rhenium (ReO{sub 4}{sup -}) as a surrogate and (b) heavy metals, e.g. Cu{sup 2+}. Conclusions: Magnetic iron oxide nanoparticles were successfully functionalized with CTAB and TEOS. The functionalization of the iron oxide nanoparticles affects their surface charge and their hydrodynamic diameter. The addition of CTAB or TEOS decreased the hydrodynamic diameter of the nanoparticles due to repulsive and steric forces. The SEM micrographs show spherical nanoparticles of different sizes. The EDX analysis shows the presence of iron and oxygen from the iron oxide crystalline structure, and the different constituents of the CTAB and TEOS molecules. Proof-of-concept shows the successful adsorption of rhenium (ReO{sub 4}{sup -}) and copper Cu{sup 2+}) onto CTAB-Fe{sub 2}O{sub 3} and TEOS-Fe{sub 2}O{sub 3} respectively.

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

Migration and Distribution of Natural Organic Matter Injected into Subsurface Systems - 16523

The Savannah River Site (SRS) was one of the major U.S. Department of Energy facilities that produced large amounts of radioactive and hazardous waste during the Cold War era, a legacy of material production for defense purposes. The unlined F-Area seepage basins received approximately 1.8 billion gallons of acidic waste solutions containing low-level radioactivity from numerous isotopes generated by uranium slug and irradiated fuel processing. Sufficient quantities of uranium isotopes, I-129, Tc-99, and tritium migrated into the groundwater, creating an acidic plume with a pH between 3 and 5.5. Humic substances (HS) are complex heterogeneous mixtures of poly-dispersed materials formed by biochemical and chemical reactions during the decay and transformation of plant and microbial remains. They are known for their excellent binding capacity for metals, making them a strong candidate for remediation efforts to reduce the mobility of uranium (VI) in the subsurface. Studies showed that HS are an important ion exchange and metal-complexing ligand, carrying a large number of functional groups with high complexing capacity that can greatly affect the mobility behavior of actinides in natural systems. Hence, it is a potential in situ amendment for treating groundwater contaminated with these constituents. Column experiments were conducted to address the complexities of HS in flow-through porous media representative of the acidic aquifer at the SRS F-Area Seepage Basins. The humate source used, Huma-K, is high in humic and fulvic compounds. The column studies were done using the soil obtained from SRS's FAW-1 core-hole at a depth interval of 60'-70'. A bromide tracer test was performed to obtain the pore volume (PV), variance and Peclet number. The columns were saturated with an artificial groundwater solution prepared to mimic the groundwater at SRS that was pH adjusted to 3.5 and 5 until the effluent pH reached equilibrium. Approximately one (1) PV of 5000 ppm of Huma-K was injected into the columns at 2 ml/min to study the sorption of Huma-K on the sediment. Desorption of Huma-K was studied by injecting approximately four (4) PVs of the pH adjusted artificial groundwater solution. A faster increase in pH was observed in column 1 (saturated with pH 3.5 AGW) compared to column 2 (saturated with pH 5 AGW). After injection of 2 PV of AGW, the pH of the columns reached 6.5 and 7, respectively, for columns 1 and 2. After injecting an additional PV of AGW, the pH remained relatively stable. With an increase in pH from 3.5 to 5.0, the overall retention of HA increased from 461 mg/kg for the column with pH 3.5 to 642 mg/kg for the column with pH 5. This suggests that retention is not due to adsorption alone, but adsorption coupled with precipitation and re-dissolution of Huma-K as it flows through the columns is a possible explanation for greater retention at pH 5. This negates the use of a simple Langmuir isotherm to explain humate behavior in the acidic environment of column 1. (authors)

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

Vitrification of Hanford Tank 241-AP-107 with Recycled Condensate

During the vitrification of nuclear waste at the Hanford Waste Treatment and Immobilization Plant (WTP) – the primary mission of the U.S. Department of Energy Office of River Protection – the offgas condensate generated from the waste-to-glass conversion is currently planned to be concentrated by evaporation in the Effluent Management Facility (EMF). This concentrated condensate can then be recycled back to the incoming waste and vitrified. To test the recycle process, a test apparatus was designed to mimic the EMF evaporator and used to concentrate a volume of condensate that had been previously produced during the vitrification of Hanford tank 241-AP-107 (referred to herein as AP-107) waste in a continuous laboratory-scale melter (CLSM). The test apparatus successfully concentrated the AP-107 condensate by a factor of 10 while retaining over 90 % of the technetium-99 ( 99 Tc), Cs, and I inventory. A second portion of AP-107 waste was retrieved by Washington River Protection Solutions, LLC, given to Pacific Northwest National Laboratory, and combined with the AP-107 condensate concentrate after undergoing solids filtration and cesium removal by ion exchange. This combination served to approximate the recycling action to be performed at the WTP. After the addition of glass-forming chemicals (GFCs), the combined AP-107 waste and AP-107 condensate concentrate were processed in the CLSM to produce a glass, called AP-107-1R, that was designed to satisfy the WTP baseline requirements (Kim et al. 2012). During the 8.87 hours of processing, 7.27 kg of AP-107-1R glass were produced for an average glass production rate of 1739 kg m 2 d -1 . Compared to the previous run in the CLSM without recycled condensate, the run with the recycle had a greater average glass production rate, but the rate was within the potential range of variability when processing melter feeds with similar composition in the CLSM. The glass produced from the AP-107 recycle run in the CLSM was within 10 % of the target AP-107-1R glass composition with respect to the primary glass components. Analysis of the minor component impurities revealed that their content in the glass product had approached their nominal target after 2 turnovers of the glass inventory in the CLSM while the activity of the minor radionuclides was retained in the glass product. The 99 Tc and total cesium content in the combined AP-107 waste and recycled condensate were maintained at concentrations expected to be experienced at the WTP. During processing in the CLSM, at discrete sampling time periods, the target 99 Tc/Cs mass ratio in the glass formulation varied from 0.9 to 62.9. Across this range, the Cs retention in the glass ranged from 53 to 60 %, while the retention from the entire runtime totaled 68 %, values which align with Cs retention in other scaled melter systems while processing LAW melter feeds at 99 Tc/Cs mass ratios ranging from 1 to 100. The 99 Tc retention in the glass ranged from 22 to 32 %, primarily due to the cold-cap coverage on the glass melt surface, the area covered by reacting melter feed, varying from ~80 % to ~95 % during processing, demonstrating greater volatility of 99 Tc from the glass while more surface was exposed, as expected based on previous 99 Tc volatility studies.

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

Nuclear Data Sheets for A = 87

The evaluated experimental data are presented for 14 known nuclides of mass 87 (Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru). New datasets have been added for {sup 87}Ga, {sup 87}Se, {sup 87}Kr, {sup 87}Rb, {sup 87}Sr, {sup 87}Nb, and {sup 87}Tc. The observation of and new lifetime measurements for {sup 87}As at the NSCL Coupled Cyclotron Facility and the Oak Ridge Isochronous Cyclotron at the HRIBF–ORNL Facility are included in the adopted levels for this isotope. The new evaluation also includes an isomeric level at unknown energy for {sup 87}Tc, changes to the adopted levels for {sup 87}Nb based on newly published data including an extension of the high spin levels and new lifetime measurements, and new heavy ion data for {sup 87}Kr. New high spin levels for {sup 87}Rb were observed, indicating possible neutron core excitations. In addition, the ground state lifetime of {sup 87}Rb has been revised and a new lifetime measurement for an excited level in {sup 87}Rb indicates a possible proton spin flip. New excitation levels have been observed for the first time in {sup 87}Se. New direct and precise measurement of atomic masses of {sup 87}Tc, {sup 87}Mo, {sup 87}Nb, {sup 87}Zr, {sup 87}Mo and {sup 87}Tc have greatly improved the Q value landscape in this mass region. This work supersedes the previous A=87 evaluation published in 2002He09, and Q values for all isotopes were updated to incorporate the mass evaluation in 2012Wa38.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Influence of Metal Ion Complexation on the Radiolytic Longevity of Butyramide Extractants under Direct Dissolution Process Conditions

The direct dissolution of voloxidized used nuclear fuel (UNF) into an organic solution–comprised of diluent and specialized extractants–poses a promising alternative to the traditional liquid–liquid solvent extraction approach to reprocessing UNF. However, moving to direct dissolution removes the presence of a concentrated nitric acid aqueous phase, which has been shown to significantly influence the radiolytic longevity of extractants in conventional extraction flowsheets. Given the limited knowledge of radiation effects under direct dissolution conditions, here we present a time-resolved and dose-accumulation study on the impact of direct dissolution conditions on the radiolytic longevity of two candidate butyramide extractants, N,N-di(2-ethylhexyl) butyramide (DEHBA) and N,N-di(2-ethylhexyl)isobutyramide (DEHiBA), in pre-equilibrated n-dodecane solvent in the presence and absence of process-relevant metal ions, specifically, uranium and rhenium. Loss G(DEHBA) and G(DEHiBA) values were found to be comparable to each other, with an average of 0.37 ± 0.02 μmol J –1 , and to previous data from the γ irradiation of DEHBA and DEHiBA under conventional solvent extraction conditions. Rhenium, and by extension technetium, extraction had a modest decrease (~10%) in the overall radiolytic stability of DEHiBA only, despite >2× observed increases in chemical kinetic reactivity of the corresponding complexes with the n-dodecane radical cation. Uranium loading, on the other hand, significantly improved the lifetime of both ligands (>30%) under γ irradiation, with a greater stabilization observed for DEHBA over DEHiBA. The observed radioprotective effect afforded by uranium loading is fortuitous for the longevity of direct dissolution solvent.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2014

LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in greatest abundance in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are low but are also expected to be in measurable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, {sup 241}Pu, and {sup 241}Am. These are present due to their partial volatility and some entrainment in the off-gas system. This report discusses results of optimized {sup 99}Tc decontamination testing of the simulant. Testing examined use of inorganic reducing agents for {sup 99}Tc. Testing focused on minimizing the quantity of sorbents/reactants added, and minimizing mixing time to reach the decontamination targets in this simulant formulation. Stannous chloride and ferrous sulfate were tested as reducing agents to determine the minimum needed to convert soluble pertechnetate to the insoluble technetium dioxide. The reducing agents were tried with and without sorbents. The sorbents, hydroxyapatite and sodium oxalate, were expected to sorb the precipitated technetium dioxide and facilitate removal. The Phase 1 tests examined a broad range of conditions and used the initial baseline simulant. The Phase 2 tests narrowed the conditions based on Phase 1 results, and used a slightly modified simulant. Test results indicate that excellent removal of {sup 99}Tc was achieved using SnCl{sub 2} as a reductant, and was effective with or without sorption onto hydroxyapatite. This reaction worked even in the presence of air (which could oxidize the stannous ion) and at room temperature. This process was very effective at neutral pH, with a Decontamination Factor (DF) >199 in one hour with only 1 g/L of SnCl{sub 2}. Prior work had shown that it was much less effective at alkaline pH. The only deleterious effect observed was that the chromium co-precipitates with the {sup 99}c during the SnCl{sub 2} reduction. This effect was anticipated, and would have to be considered when managing disposition paths of this stream. Reduction using FeSO{sub 4} was not effective at removing {sup 99}Tc, but did remove the Cr. Chromium is present due to partial volatility and entrainment in the off-gas, and is highly oxidizing, so would be expected to react with reducing agents more quickly than pertechnetate. Testing showed that sufficient reducing agent must be added to completely reduce the chromium before the technetium is reduced and removed. Other radionuclides are also present in this off-gas condensate stream. To enable sending this stream to the Hanford ETF, and thereby divert it from the recycle where it impacts the LAW glass volume, several of these also need to be removed. Samples from optimized conditions were also measured for actinide removal in order to examine the effect of the Tc-removal process on the actinides. Plutonium was also removed by the SnCl{sub 2} precipitation process. Results of this separation testing indicate that sorption/precipitation is a viable concept and has the potential to decontaminate the {sup 99}Tc from the stream, allowing it to be diverted away from WTP and thus eliminating the impact of the recycled halides and sulfate on the LAW glass volume. Based on the results, a possible treatment scenario could involve the use of a reductive precipitation agent (SnCl{sub 2}) with or without sorbent at neutral pH to remove the Tc. Although hydroxyapatite was not necessary to effect the {sup 99}Tc removal, it may be beneficial in solid-liquid separations. Other testing will examine removal of the other radionuclides. This testing was the second phase of testing, which aimed at optimizing the process by examining the minimum amount of reductant needed and the minimum reaction time. Although results indicated that SnCl{sub 2} was effective, further work on a pH-adjusted Fe(SO{sub 4}) mixture are needed. Additional tasks are needed to examine removal of the other radionuclides, solid-liquid separation technologies, slurry rheology measurements, composition variability impacts, corrosion and erosion, and slurry storage and immobilization.

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

Sulfur-Modified Zero-Valent Iron for Remediation Applications at DOE Sites - 13600

Many DOE remediation sites have chemicals of concern that are compounds in higher oxidation states, which make them both more mobile and more toxic. The chemical reduction of these compounds both prevents the migration of these chemicals and in some cases reduces the toxicity. It has also been shown that zero-valent iron is a very effective substance to use in reducing oxygenated compounds in various treatment processes. These have included the treatment of halogenated hydrocarbons in the form volatile organic compounds used as solvents and pesticides. Zero-valent iron has also been used to reduce various oxidized metals such as chromium, arsenic, and mercury in order to immobilize them, decrease their toxicity, and prevent further transport. In addition, it has been used to immobilize or break down other non-metallic species such as selenium compounds and nitrates. Of particular interest at several DOE remediation sites is the fact that zero-valent iron is very effective in immobilizing several radioactive metals which are mobile in their oxidized states. These include both technetium and uranium. The main difficulty in using zero-valent iron has been its tendency to become inactive after relatively short periods of time. While it is advantageous to have the zero-valent iron particles as porous as possible in order to provide maximum surface area for reactions to take place, these pores can become clogged when the iron is oxidized. This is due to the fact that ferric oxide has a greater volume for a given mass than metallic iron. When the surfaces of the iron particles oxidize to ferric oxide, the pores become narrower and will eventually shut. In order to minimize the degradation of the chemical activity of the iron due to this process, a modification of zero-valent iron has been developed which prevents or slows this process, which decreases its effectiveness. It is called sulfur-modified iron, and it has been produced in high purity for applications in municipal water treatment applications. Sulfur-modified iron has been found to not only be an extremely economical treatment technology for municipal water supplies, where very large quantities of water must be treated economically, but it has also been demonstrated to immobilize technetium. It has the added benefit of eliminating several other harmful chemicals in water supplies. These include arsenic and selenium. In one large-scale evaluation study an integrated system implemented chemical reduction of nitrate with sulfur-modified iron followed by filtration for arsenic removal. The sulfur-modified iron that was used was an iron-based granular medium that has been commercially developed for the removal of nitrate, co-contaminants including uranium, vanadium and chromium, and other compounds from water. The independent study concluded that 'It is foreseen that the greatest benefit of this technology (sulfur-modified iron) is that it does not produce a costly brine stream as do the currently accepted nitrate removal technologies of ion exchange and reverse osmosis. This investigation confirmed that nitrate reduction via sulfur-modified iron is independent of the hydraulic loading rate. Future sulfur-modified iron treatment systems can be designed without restriction of the reactor vessel dimensions. Future vessels can be adapted to existing site constraints without being limited to height-to-width ratios that would exist if nitrate reduction were to depend on hydraulic loading rate'. Sulfur-modified iron was studied by the Pacific Northwest National Laboratory (PNNL) for its effectiveness in the reduction and permanent sequestration of technetium. The testing was done using Hanford Site groundwater together with sediment. The report stated, 'Under reducing conditions, TcO{sub 4} is readily reduced to TcIV, which forms highly insoluble oxides such at TcO{sub 2}.nH{sub 2}O. However, (re)oxidation of TcIV oxides can lead to remobilization. Under sulfidogenic conditions, most TcIV will be reduced and immobilized as Tc{sub 2}S{sub 7}, which is less readily re-mobilized, even under oxic conditions. This process should be favored by stimulation of sulfidogenic conditions'. The sulfur-modified iron provides the sulfur, together with the iron, to maintain this stable sequestration of technetium. As a result of these and other studies demonstrating the cost-effectiveness of sulfur-modified iron in treating technetium and other hazardous compounds in Hanford Site groundwater and its cost-effectiveness in reducing nitrate, the Richland Operations Office of the Department of Energy issued a change order to the Central Plateau Contractor providing for the testing of sulfur-modified iron in a mobile pilot unit at the Hanford Site. Further testing is anticipated to produce refinements in operating conditions and further optimization of the existing process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Radiolytic evaluation of a new technetium redox control reagent for advanced used nuclear fuel separations

Technetium is a problematic radioisotope for used nuclear fuel (UNF) and subsequent waste management owing to its high environmental mobility and coextraction in reprocessing technologies as the pertechnetate anion (TcO 4 - ). Consequently, several strategies are under development to control the transport of this radioisotope. A proposed approach is to use diaminoguanidine (DAG) for TcO 4 - and transuranic ion redox control. Although the initial DAG molecule is ultimately consumed in the redox process, its susceptibility to radiolysis is currently unknown under envisioned UNF reprocessing conditions, which is a critical knowledge gap for evaluating its overall suitability for this role. To this end, we report the impacts of steady-state gamma irradiation on the rate of DAG radiolysis in water, aqueous 2.0 M nitric acid (HNO 3 ), and in a biphasic solvent system composed of aqueous 2.0 M HNO 3 in contact with 1.5 M N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) dissolved in n-dodecane. In addition, we report chemical kinetics for the reaction of DAG with key transients arising from electron pulse radiolysis, specifically the hydrated electron (e aq - ), hydrogen atom (H ˙ ), and hydroxyl ( ˙ OH) and nitrate (NO 3 ˙ ) radicals. The DAG molecule exhibited significant reactivity with the ˙ OH and NO 3 ˙ radicals, indicating that oxidation would be the predominant degradation pathway in radiation environments. This is consistent with its role as a reducing agent. Steady-state gamma irradiations demonstrated that DAG is readily degraded within a few hundred kilogray, the rate of which was found to increase upon going from water to HNO 3 containing solutions and solvents systems. This was attributed to a thermal reaction between DAG and the predominant HNO 3 radiolysis product, nitrous acid (HNO 2 ), k(DAG + HNO 2 ) = 5480 ± 85 M -1 s -1 . Although no evidence was found for the radiolysis of DAG altering the radiation chemistry of the contacted DEHiBA/n-dodecane phase in the investigated biphasic system, the utility of DAG as a redox control reagent will likely be limited by significant competition with its degradation by HNO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Spectroscopic Properties of Tc(I) Tricarbonyl Species Relevant to the Hanford Tank Waste

Technetium-99 (Tc) exists predominately in soluble forms in the liquid supernatant and salt cake fractions of the nuclear tank waste stored at the U.S. DOE Hanford Site. In the strongly alkaline environments prevalent in the tank waste, its dominant chemical form is pertechnetate (TcO4-, oxidation state +7). However, attempts to remove Tc from the Hanford tank waste using ion-exchange processes specific to TcO 4 - only met with limited success, particularly processing tank waste samples containing elevated concentrations of organic complexants. This suggests that a significant fraction of the soluble Tc can be present as non-pertechnetate low-valent Tc (oxidation state < +7) (non-pertechnetate). The chemical identities of these non-pertechnetate species are poorly understood. Previous analysis of the SY-101 and SY-103 tank waste samples provided strong evidence that non-pertechnetate can be comprised of [Tc(CO) 3 ] + complexes containing Tc in oxidation state +1 (Lukens et al. 2004). During the last two years, our team has expanded this work and demonstrated that high-ionic-strength solutions typifying tank waste supernatants promote oxidative stability of the [Tc(CO) 3 ] + species (Rapko et al. 2013; Levitskaia et al. 2014). It also was observed that high-ionic-strength alkaline matrices stabilize Tc(VI) and potentially Tc(IV) oxidation states, particularly in presence organic chelators, suggesting that the relevant Tc compounds can serve as important redox intermediates facilitating the reduction of Tc(VII) to Tc(I). Designing strategies for effective Tc processing, including separation and immobilization, necessitates understanding the molecular structure of these non-pertechnetate species and their identification in the actual tank waste samples. To-date, only limited information exists regarding the nature and characterization of the Tc(I), Tc(IV), and Tc(VI) species. One objective of this project is to identify the form of non-pertechnetate in the Hanford waste. To do this, we are developing a spectral library of reference non-pertechnetate compounds that can be compared against actual waste samples. The emphasis of the fiscal year 2015 work was Tc(I) tricarbonyl [Tc(CO) 3 ] + compounds. The key findings are summarized below.

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

Metallic Technetium Sequestration in Nickel Core/Shell Microstructure during Fe(OH)2 Transformation with Ni doping

This study investigates the impacts of Ni doping on technetium-99 (Tc) sequestration in aqueous solutions through transformation of Fe(OH)2(s) to iron spinel (magnetite) under alkaline conditions. Extensive solid characterization was performed for the mineral phases produced, as well as the Tc/Ni speciation and distribution within these phases. X-ray diffraction results show that iron spinel was the dominant mineral product without detectable Ni incorporation. The doped Ni ions mainly precipitated as fine Fe/Ni oxide/hydroxide particles, including strongly reduced nanometer?sized spheroidal Ni-rich and metallic Ni phases. High-resolution analytical scanning transmission electron microscopy using energy dispersive X-ray spectroscopy and electron energy loss spectroscopy on the produced solid samples (focused ion beam-prepared specimens) revealed three Tc distribution domains dominated by nanocrystals and, especially, a Tc-rich metallic phase. Instances of metallic Tc were specifically found in spheroidal, Ni-rich and metallic nanoparticles exhibiting a core/shell microstructure that suggests strong reduction and sequential precipitation of Ni-Tc-Ni. Mass balance analysis showed nearly 100% Tc removal from the 4.8 × 10-4 M Tc solutions. The finding of the metallic Tc encapsulation indicates that Tc sequestration through Ni-doped Fe(OH)2(s)?to?iron spinel transformation process likely provides an alternative treatment pathway for Tc removal and could be combined into further waste treatment approaches.

Wang, Guohui↗
DOE OSTI2024

Partitioning of Fe(II) in reduced nontronite (NAu-2) to reactive sites: Reactivity in terms of Tc(VII) reduction

Abstract Clay minerals impart important chemical properties to soils, in part, by virtue of changes in the redox state of Fe in their crystal structures. Therefore, measurement of Fe(III)/Fe(II) and partitioning of Fe(II) in different reactive sites in clay minerals (during biological and chemical Fe(III) reduction) is essential to understand their role and their relative reactivity in terms of reduction and immobilization of heavy metal contaminants such as technetium. This study had three objectives: (1) to understand the degree of dissolution of nontronite (Fe-rich smectite) as a result of chemical and biological reduction of Fe(III) in the structure; (2) to quantify partitioning of chemically and biologically produced Fe(II) into different reactive sites in reduced nontronite, including aqueous Fe 2+ , ammonium chloride-extractable Fe(II) (mainly from the ion-exchangeable sites, denoted as Fe ( II ) NH 4 Cl ${\rm{Fe}}{\left( {{\rm{II}}} \right)_{{\rm{N}}{{\rm{H}}_4}{\rm{Cl}}}}$ ), sodium acetate-extractable Fe(II) (mainly from the surface complexation sites, denoted as Fe(II) acetate ), and structural Fe(II) (denoted as Fe(II) str ); and (3) to evaluate the reactivity of these Fe(II) species in terms of Tc(VII) reduction. Chemical and biological reduction of Fe(III) in nontronite (NAu-2) was performed, and reduced nontronite samples with different extents of Fe(III) reduction (1.2–71%) were prepared. The extent of reductive dissolution was measured as a function of the extent of Fe(III) reduction. Our results demonstrated that chemically and biologically produced Fe(II) in NAu-2 may be accommodated in the NAu-2 structure if the extent of Fe(III) reduction is small (< ∼30%). When the extent of reduction was >∼30%, dissolution of nontronite occurred with a corresponding decrease in crystallinity of residual nontronite. The Fe(II) produced was available for partitioning into four species: Fe ( ab ) 2 + ${\rm{Fe}}_{\left( {{\rm{ab}}} \right)}^{2 + }$ , Fe(II) acetate , Fe ( II ) NH 4 Cl ${\rm{Fe}}{\left( {{\rm{II}}} \right)_{{\rm{N}}{{\rm{H}}_4}{\rm{Cl}}}}$ , and Fe(II) str . The increase in Fe(II) acetate during the early stages of Fe(III) reduction indicated that the Fe(II) released had the greatest affinity for the surface-complexation sites, but this site had a limited capacity (∼60 µmol of Fe(II)/g of NAu-2). The subsequent increase in Fe ( II ) NH 4 Cl ${\rm{Fe}}{\left( {{\rm{II}}} \right)_{{\rm{N}}{{\rm{H}}_4}{\rm{Cl}}}}$ indicated that the released Fe(II) partitioned into the exchangeable sites once the amount of Fe at the surface-complexation sites reached half of its maximum site capacity. The fraction of Fe(II) str decreased concomitantly, as a result of Fe(II) release from the NAu-2 structure, from 100% when the extent of Fe(III) reduction was <30% to nearly 65% when the extent of Fe(III) reduction reached 71%. The Fe(II) acetate and Fe(II) str exhibited greater reactivity in terms of Tc(VII) reduction than the Fe ( II ) NH 4 Cl ${\rm{Fe}}{\left( {{\rm{II}}} \right)_{{\rm{N}}{{\rm{H}}_4}{\rm{Cl}}}}$ . Clearly, the surface-complexed and structural Fe(II) are the desirable species when reduced clay minerals are used to reduce and immobilize soluble heavy metals in contaminated groundwater and soils. These results have important implications for understanding microbe—clay mineral interactions and heavy metal immobilization in clay-rich natural environments.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Computational Investigation of Technetium(IV) Incorporation into Inverse Spinels: Magnetite (Fe 3 O 4 ) and Trevorite (NiFe 2 O 4 )

Iron oxides and oxyhydroxides play an important role in minimizing the mobility of redox-sensitive elements in engineered and natural environments. For the radionuclide technetium-99 (Tc), these phases hold promise as primary hosts for increasing Tc loading into glass waste form matrices, or as secondary sinks during the long-term storage of nuclear materials. Recent experiments show that the inverse spinel, magnetite [Fe(II)Fe(III)2O4], can incorporate Tc(IV) into its octahedral sub-lattice, and in that same class of materials, trevorite [Ni(II)Fe(III)2O4] is also being investigated for its ability to host Tc(IV). However, questions remain regarding the most energetically favorable charge-compensation mechanism for Tc(IV) incorporation in each structure, which will affect Tc behavior under changing waste processing or storage conditions. Here, quantum-mechanical methods were used to evaluate incorporation energies and optimized lattice bonding environments for three different, charge-balanced Tc(IV) incorporation mechanisms in magnetite and trevorite. In both cases, the removal of two octahedral Fe(II) or Ni(II) ions upon the addition of Tc(IV) to an octahedral site is the most stable mechanism, relative to the creation of octahedral Fe(III) defects or increasing octahedral Fe(II) content. Following hydration-energy corrections, Tc(IV) incorporation into magnetite is energetically favorable while an energy barrier exists for trevorite.

Environmental Molecular Sciences Laboratory↗