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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
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↗
DOE OSTI2020

Moly 99 Reactor Case Study

Molybdenum-99 (Moly 99) is a critical raw material for Technetium (Tc) 99m, a radioactive isotope most widely used in nuclear medicine procedures. Moly 99 has a short half-life of about six hours, which means it cannot be stockpiled. When Moly 99 decays, it turns into Tc 99m, which has a half-life of 214,000 years. The photon energy emitted from the decay of Moly 99 is used in a variety of nuclear imaging technologies such as gamma cameras. Furthermore, radiopharmaceutical manufacturers use the photon energy emitted from the decay of Moly 99 to produce generators for hospitals, clinics, and radiopharmacies. Once Moly 99 decays to Tc 99m, it is used to make individual patient doses for a variety of diagnostic imaging procedures. The Moly 99 Reactor Design is a conceptual blueprint for Moly 99 production that does not use weapon-grade uranium; instead, the reactor has a target core of low-enriched uranium. The reactor design is small, reaching a foot-and-a-half in height and diameter and consumes less than two megawatts of power. The reactor's only purpose is for medical isotope production and with every fission, Moly 99 is produced.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Photo-Transmutation of {sup 100}Mo to {sup 99m}Tc with Laser-Compton Scattering Gamma-Ray

Technetium-99m ({sup 99m}Tc) is an important medical radioisotope which is largely used in hospitals for diagnosis of cancer. It is produced in the form of {sup 99}Mo which decays to {sup 99m}Tc with a half-life of 66 hours. More than 90% of the world's demand of {sup 99}Mo are sourced from five nuclear reactors. Two of these reactors have already been closed and the rest are more than 45 years old. Relatively short half-life of the parent {sup 99}Mo requires continuous re-supply to meet the medical industry's requirement. Therefore there is an urgent need to produce the {sup 99}Mo and {sup 99m}Tc isotopes by alternative ways. This paper presents one such alternative via the use of laser-Compton scattering (LCS) gamma beam. In this paper, production rate (reaction rate) of {sup 99}Mo using the photonuclear (γ, n) reaction is evaluated. It is expected that the LCS technique can be used to generate the gamma ray required for the aforementioned photonuclear reaction. Rigorous optimization of the LCS spectrum is also performed to maximize production of the {sup 99}Mo. The optimal maximum energy of the LCS photon is about 17 MeV for the production of {sup 99}Mo. The analysis shows that production of {sup 99}Mo using LCS gamma beam offers a promising alternative for the production of {sup 99m}Tc. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Initial Testing of Off-Gas Recycle Condensate Liquid from the WTP Low Activity Waste Vitrification Process - 16432

The Waste Treatment and Immobilization Plant (WTP) under construction at the Hanford site is designed to treat and immobilize the inventory of High-Level Waste stored in the 177 underground storage tanks. The WTP process flow was designed to pre-treat this waste, separating it into High Level Waste (HLW) and Low Activity Waste (LAW) fractions for vitrification. Vitrification of the waste generates a condensate stream from the off-gas processes originating from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP). The baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processed through the LAW melter again. If the Pretreatment facility is unavailable, this baseline disposition pathway is not viable. Additionally, chloride, fluoride, and sulfate are partially volatile in the melter, accumulating to high concentrations while recycling and impacting the glass waste loading and facility throughput. The primary radionuclide that volatizes and accumulates is {sup 99}Tc. The long half-life and environmental mobility of {sup 99}Tc complicates disposal options for this secondary aqueous waste stream. This task is investigating {sup 99}Tc removal to examine the potential for diverting this stream to an alternate disposition path and avoiding impacting the glass waste loading. Separation methods for {sup 99}Tc have focused on using inorganic precipitation agents that can remove it by reducing it from soluble pertechnetate ion, with a +7 oxidation state, to insoluble technetium dioxide in the +4 oxidation state. Stannous chloride (SnCl{sub 2}) has been found highly effective at causing the reduction and precipitating {sup 99}Tc for removal, but it also precipitates other components. The primary competitor that precipitates is chromium, which consumes the redox activity of the Sn(II), causing addition of more of the SnCl{sub 2}. Research has examined the optimum quantity of SnCl{sub 2}, mixing time, and the impact of other redox-active species. It is also important to know the stability of the precipitated {sup 99}Tc in process conditions, since it is expected it will eventually be re-oxidized by oxygen and dissolve. This work is expected to determine if a {sup 99}Tc removal step could be used to decontaminate the bulk of this stream, and make it suitable for other disposal paths so that the chloride, fluoride, and sulfate can be diverted away from the glass melter. (authors)

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

Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

The isotope 99 Tc is an important fission product generated from nuclear power production. Because of its long half-life (t 1/2 = 2.13 ∙ 10 5 years) and beta-radiotoxicity (β⁻ = 292 keV), it is a major concern in the long-term management of spent nuclear fuel. In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up. In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form. Technetium waste forms under consideration include metallic alloys, oxide ceramics and borosilicate glass. In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository. Metallic alloys under consideration include Tc–Zr alloys, Tc–stainless steel alloys and Tc–Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long-term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed, but efforts to model the behavior of Tc metallic alloys are limited. One parameter that should also be considered in predicting the long-term behavior of the Tc waste form is the ingrowth of stable Ru that occurs from the radioactive decay of 99 Tc ( 99 Tc → 99 Ru + β⁻). After a geological period of time, significant amounts of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance. In this context, we studied the electrochemical behavior of Tc metal, Tc-Ni alloys (to model Tc-Inconel alloy) and Tc-Ru alloys in acidic media. The study of Tc-U alloys has also been performed in order to better understand the nature of Tc in metallic spent fuel. Computational modeling and simulations were performed to shed light on experimental results and explain structural and kinetics trends.

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

Reactive amendment saltstone (RAS). A novel approach for improved sorption/retention of radionuclides such as technetium and iodine

This study examined the use of reactive amendments (hydroxyapatite, activated carbon, and two types of organoclays) that prior research suggests may improve retention of 99 Tc and 129 I. Tests were conducted using surrogates for 99 Tc (NaReO 4 ) and 129 I (NaI). Results showed that adding up to 10% of organoclay improved the retention of Re without adversely impacting hydraulic properties. To a lesser extent, iodine retention was also improved by adding up to 10% organoclay. Numerical modeling showed that using organoclay as a reactive barrier may significantly retard 99 Tc release from saltstone disposal units.

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

Ion Exchange Column Tests Supporting Technetium Removal Resin Maturation

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant, currently under construction. The baseline plan for this facility is to treat the waste, splitting it into High Level Waste (HLW) and Low Activity Waste (LAW). Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed on site. There are currently no plans to treat the waste to remove technetium, so its disposition path is the LAW glass. Due to the soluble properties of pertechnetate and long half-life of 99 Tc, effective management of 99 Tc is important. Options are being explored to immobilize the supplemental LAW portion of the tank waste, as well as to examine the volatility of 99 Tc during the vitrification process. Removal of 99 Tc, followed by off-site disposal has potential to reduce treatment and disposal costs. A conceptual flow sheets for supplemental LAW treatment and disposal that could benefit from technetium removal will specifically examine removing 99 Tc from the LAW feed stream to supplemental immobilization. SuperLig® 639 is an elutable ion exchange resin. In the tank waste, 99 Tc is predominantly found in the tank supernate as pertechnetate (TcO 4 - ). Perrhenate (ReO 4 - ) has been shown to be a good non-radioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin. This report contains results of experimental ion exchange distribution coefficient and column resin maturation kinetics testing using the resin SuperLig® 639a to selectively remove perrhenate from simulated LAW. This revision includes results from testing to determine effective resin operating temperature range. Loading tests were performed at 45°C, and the computer modeling was updated to include the temperature effects. Equilibrium contact testing indicated that this batch of SuperLig® 639 resin has good performance, with an average perrhenate distribution coefficient of 291 mL/g at a 100:1 phase ratio. This slightly exceeds the computer-modeled equilibrium distribution. The modeling agreed well with the experimental data for perrhenate removal with minor adjustments. Predicted breakthrough performance was on average within about 20% of measured values.

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

Testing of New Technetium Selective Media TcTreat - 16497

A new zirconium oxide based material TcTreat is under development and testing for the removal of radioactive oxo-anionic species (Tc-99 pertechnetate, Ru-106 ruthenate) from solution. TcTreat exhibited a very high {sup 99}Tc uptake over a broad pH-range of 4-10 and the uptake was reasonably high also in mildly alkaline and acidic ranges. Nitrate and sulphate had a decreasing effect on {sup 99}Tc uptake but is was still strong (k{sub d} = 100,000 ml/g and above) even at 1 molar concentrations.

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

Design of a Commercial Scale Accelerator Driven Subcritical Aqueous Assembly

SHINE Medical Technologies is working towards producing a reliable, domestic supply of medical isotopes, including molybdenum-99 (Mo-99), xenon-133, and iodine 131, without using highly enriched uranium. Approximately 88% of Mo-99 decays to technetium-99m, which is the most widely used medical isotope in the world, accounting for approximately 80% of all nuclear medicine procedures. A novel aqueous, subcritical nuclear assembly has been designed by SHINE Medical Technologies. The assembly can produce medical radioisotopes in significant quantities for U.S. and foreign needs. The assembly produces isotopes without using a reactor, instead using a fusion-based neutron source coupled to a fission-based aqueous assembly. The subcritical nature of the aqueous assembly has safety and licensing benefits, including the elimination of the reactor control system and elimination of high level waste. SHINE plans to construct eight subcritical assemblies with a plant capacity of greater than 3000 6-day Curies, which is more than one-quarter of the world's Mo-99 requirements. From a nuclear design standpoint, the aqueous subcritical assembly needs to be able to generate sufficient fission power to meet the production capacity requirements while remaining subcritical at all times with an acceptable margin. To address these conflicting design goals, a simple nuclear design and reactivity control philosophy has been developed that allows for high power operation with a high degree of safety. A fusion-driven subcritical fission assembly has been designed by SHINE Medical Technologies to provide a reliable, domestic supply of medical isotopes. Nuclear models have been created and show that the assembly can be started up, operated, and shutdown in a safe manner without active reactivity control systems. A unique reactivity control strategy allows SHINE to operate the aqueous subcritical assembly at much higher power levels than would otherwise be practical, while still retaining the benefits of subcritical operation and avoiding the need for complex subcritical measurement techniques. Based on the current design, the subcritical target is capable of generating significant fission power (on the order of 100 kW) when being driven by a simple, low-cost deuterium-tritium accelerator, enabling the SHINE facility to supply over one-fourth of the worldwide demand of Mo- 99. SHINE has recently been issued a construction permit by the U.S. Nuclear Regulatory Commission to build the commercial facility in Janesville, WI, with eight subcritical assemblies and the associated production equipment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Fabrication and Properties of Advanced Ceramic Technetium Waste Forms - 17311

Technetium-99 (Tc, t{sub 1/2} = 2.13 x 10{sup 5} years) is a challenge from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes found in used nuclear fuel (UNF). Within the Hanford Tank Waste Treatment and Immobilization Plant, technetium volatilises at typical glass melting temperature, is captured in the off-gas treatment system and recycled back into the feed to eventually increase Tc-loadings of the glass. In this research we are providing a technical solution for continuously depleting the Tc-content of the glass melter feed by precipitating Tc from the off-gas effluents and effectively immobilizing {sup 99}Tc as durable ceramic waste form. We have developed a wet-chemical coprecipitation-based synthesis route to fabricate gram-size Tc-based pyrochlores with the model composition Sm{sub 2}(Tc{sub 0.5-x}Ru{sub x}Ti{sub 0.5}){sub 2}O{sub 7}. These ceramic Tc waste forms exhibit better technetium retention and corrosion resistance in ASTM C1220-10 based testing compared with technetium-containing LAWE4-type borosilicate glass, combined with optimal waste loading. (authors)

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

Extraction and reductive stripping of pertechnetate from spent nuclear fuel waste streams.

An approach directed at rapid sequestration and disposal of technetium-99 from UREX (uranium extraction) liquid waste streams is presented. This stream is generated during reprocessing of light-water-reactor spent fuel to recycle the actinides and separate fission products for waste disposal. U and Tc are co-extracted from a nitric acid solution using tri-n-butylphosphate in dodecane, so that Tc(VII) is present in the strip solution after the actinide separations. The goal is to separate uranyl from the pertechnetate in this U-Tc stream and then sequester Tc in the metallic form. Our approach is based on reductive stripping of pertechnetate either from aqueous solution (for column extractions) or organic solvents (for liquid-liquid extractions). In both of these methods, metallic zinc in the presence of formic acid serves as a reducing agent, and {sup 99}Tc is recovered as a co-precipitate of Zn(II) hydroxide and hydrous Tc(IV) oxide, with a Zn:Tc ratio between 1:1 and 2:1 mol/mol. This solid residue can be reduced to a Zn-Tc alloy by high temperature (500-700 C) hydrogenation, and the resulting heterophase alloy can be added to a metallic Fe-Zr-Mo waste form that is processed at 1600 C, with subsequent loss of Zn by evaporation. Alternatively, Zn and Tc can be separated and {sup 99}Tc sequestered as NH{sub 4}TcO{sub 4} for further reduction to Tc(0) metal. The aqueous Zn reduction process removes {approx}90% of {sup 99}Tc per cycle. The nonaqueous Zn reduction in 1:1 methanol-formic acid removes 60-70% of {sup 99}Tc per cycle, depending on the extracting agent (such as a tetraalkylammonium nitrate). The extracting agent is recycled in the process. The pertechnetate is extracted from the aqueous phase into 1,2-dichloroethane, which is removed by evaporation and reused. The residue is either calcined and steam reformed to Tc(0) or processed by the nonaqueous Zn reduction method. These methods can be used not only to remove the pertechnetate from the U-Tc product stream, but also to sequester the pertechnetate from aqueous waste streams generated through the processes described in this paper, thereby closing the cycle. The same approaches can be used to close the {sup 99}Tc cycle for other methods that are currently being developed at Los Alamos and Argonne National Laboratories.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Impact of nitrate on the redox chemistry and solubility of Tc(IV) in alkaline, dilute to concentrated aqueous NaCl solutions

Here, the impact of nitrate on the redox behaviour and solubility of 99 Tc(IV) was investigated in 0.1–5.0 M NaCl–NaNO 3 –NaOH solutions with 0.1 M ≤ [NO 3 – ] ≤ 1.0 M and 8.5 ≤ pH m ≤ 14.5 (with pH m = –log [H + ]) in the presence or absence of reducing agents (as Fe powder, Na 2 S 2 O 4 , Sn(II)). The evolution of the redox conditions as a function of time (for t ≤ 450 days) was investigated: (i) in the absence of technetium (inactive samples), (ii) in the presence of 99 Tc(IV) (added as 1–2 mg of TcO 2 ∙0.6H 2 O(am)), and (iii) in the presence of 137 Cs(I) (and absence of 99 Tc). All experiments were conducted in an Ar-glovebox (O 2 < 1 ppm) at T = (22 ± 2) °C. An improved understanding of Tc redox processes in nitrate-rich alkaline solutions is relevant to predict Tc behaviour in specific nuclear waste disposal scenarios and a topic of fundamental scientific interest.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗