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All sourcesNASA NTRSDOE OSTI
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↗
DOE OSTI2011

GLASSES CONTAINING IRON (II III) OXIDES FOR IMMOBILIZATION OF RADIOACTIVE TECHNETIUM

Technetium-99 (Tc-99) has posed serious environmental threats as US Department of Energy's high-level waste. This work reports the vitrification of Re, as surrogate for Tc-99, by iron-borosilicate and iron-phosphate glasses, respectively. Iron-phosphate glasses can dissolve Re as high as {approx} 1.2 wt. %, which can become candidate waste forms for Tc-99 disposal, while borosilicate glasses can retain less than 0.1 wt. % of Re due to high melting temperature and long melting duration. Vitrification of Re as Tc-99's mimic was investigated using iron-borosilicate and iron-phosphate glasses. The retention of Re in borosilicate glasses was less than 0.1 wt. % and more than 99 wt. % of Re were volatilized due to high melting temperature and long melting duration. Because the retention of Re in iron-phosphate glasses is as high as 1.2 wt. % and the volatilization is reduced down to {approx}50 wt. %, iron-phosphate glasses can be one of the glass waste form candidates for Tc (or Re) disposal. The investigations of chemical durability and leaching test of iron-phosphate glasses containing Re are now underway to test the performance of the waste form.

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

Reduction and Simultaneous Removal of 99 Tc and Cr by Fe(OH) 2 (s) Mineral Transformation

Technetium (Tc) remains a priority remediation concern due to persistent challenges, including rapid re-oxidation of immobilized Tc, and competing contaminants, e.g. Cr(VI), that inhibit targeted Tc reduction and incorporation into stable mineral phases. Here Fe(OH) 2 (s) is investigated as a comprehensive solution for overcoming these challenges, by serving as both the reductant, (Fe(II)), and immobilization agent to form Tc-incorporated magnetite (Fe 3 O 4 ). Trace metal analysis suggests removal of Tc(VII) and Cr(VI) from solution occurs simultaneously; however, complete removal and reduction of Cr(VI) is achieved earlier than the removal/reduction of co-mingled Tc(VII). Bulk oxidation state analysis of the magnetite solid phase by XANES confirms that the majority of Tc is Tc(IV), which is corroborated by XPS. Furthermore, EXAFS results show successful Tc(IV) incorporation into magnetite octahedral sites without additional substitution of Cr or Tc into neighboring Fe octahedral sites. XPS analysis of Cr confirms reduction to Cr(III) and the formation of a Cr-incorporated spinel, Cr2O 3 , and Cr(OH)3 phases. Spinel (modeled as Fe 3 O 4 ), goethite, and feroxyhyte are detected in all reacted solid phase samples analyzed by XRD, where Tc(IV) incorporation has little effect on the spinel lattice structure. In the presence of Cr(III) a spinel phase along the magnetite-chromite (Fe 3 O 4 -FeCr 2 O 4 ) solid-solution line is formed.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2023

Mo-99 Concentration and processing by Solvent Extraction and Ion Exchange

Normally derived from its transient-equilibrium parent molybdenum-99 ( 99 Mo) (t½ = 66 h), the short-lived isotope technetium-99m ( 99m Tc) (t ½ = 6.0 h) is the most widely used medical isotope in diagnostic imaging today. The National Nuclear Security Administration’s (NNSA’s) Material Management and Minimization (M3) program—established under the auspices of the American Medical Isotope Production Act—has been tasked with facilitating the work of domestic 99 Mo suppliers that do not use highly enriched uranium (HEU). Superconducting electron linear accelerators that employ high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that can induce photonuclear reactions and uranium (U) fission. Argonne, in collaboration with industrial partners, has been developing the process chemistry for superconducting linear accelerator (LINAC)-irradiated triuranium octoxide (U 3 O 8 ) targets to produce 99 Mo. In short, the process involves dissolving the irradiated targets in nitric acid, which simultaneously releases valuable fission products (xenon and iodine). After adjusting the acid concentration, producers process the feed using tri-n-butyl phosphate (TBP), then mix the raffinate derived from this extraction cycle with a phosphinic acid such as Cyanex 272 or di(2-ethylhexyl) phosphoric acid (HDEHP), which is selective for Mo (molybdenum liquid liquid extraction [MoLLE]). In the next step, a concentration column comprising an anion exchange platform is used to decontaminate the remaining fission products, generating a pure 99 Mo stream in sodium hydroxide (NaOH)/sodium chloride (NaCl).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Evaluate opportunities for monitoring the oxidation states of technetium in both organic and aqueous phases using Raman spectroscopy

An important isotope in the nuclear fuel cycle, 99 Tc is characterized by its long half-life of 2.1 × 10 5 years and its beta emission. It is generated through the fission of 235 U. As the burnup levels of nuclear waste continue to rise, the concentration of technetium in spent fuel increases, leading to a heightened interest in its quantification during the reprocessing of spent nuclear fuel. In HNO3 environments, technetium predominantly exists as TcO 4 − ; however, it tends to interfere with reducing agents, decreasing the separation efficiency for actinides. Therefore, accurate quantitative determination of technetium is critical to controlling its distribution and mitigating its negative effect on the reprocessing of spent fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Development of Alternative Technetium Waste Forms

The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ( 99 Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior of a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of 99 Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.

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

Non-Pertechnetate Technetium Sensor Research and Development

There remain several significant uncertainties in the understanding and modeling of the fate and speciation of 99 Tc in Hanford waste tanks, glass, and low-temperature waste forms. A significant (2% to 25%) fraction of the 99 Tc in the water-soluble portion of the tank waste may be present as a non-pertechnetate species that has not been identified and, based on experimentation to date, cannot be effectively separated from the wastes. This task will provide a sensor specifically tuned to detect the Tc(I)-carbonyl species believed to constitute the main fraction of the non-pertechnetate form of technetium. By direct measurement of the non-pertechnetate species, such a sensor will help reduce the uncertainties in the modeling of the fate and speciation of 99 Tc in Hanford tanks and waste forms. This report summarizes work done in FY 2014 exploring the chemistry of a low-valence technetium species, [Tc(CO) 3 (H 2 O) 3 ] + , a compound of interest due to its implication in the speciation of alkaline-soluble technetium in several Hanford tank waste supernatants. Progress made in FY 2014 was sponsored by the Department of Energy’s Office of Environmental Management and is summarized in this report.

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

Ion Exchange Distribution Coefficient Tests and Computer Modeling at High Ionic Strength 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 (WTP) that is 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 poured into canisters for disposition. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium, so its disposition path is the LAW glass. Due to the water solubility properties of pertechnetate and long half-life of 99 Tc, effective management of 99 Tc is important to the overall success of the Hanford River Protection Project mission. To achieve the full target WTP throughput, additional LAW immobilization capacity is needed, and options are being explored to immobilize the supplemental LAW portion of the tank waste. Removal of 99 Tc, followed by off-site disposal, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for supplemental LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing 99 Tc from the LAW feed stream to supplemental immobilization. To enable an informed decision regarding the viability of technetium removal, further maturation of available technologies is being performed. This report contains results of experimental ion exchange distribution coefficient testing and computer modeling using the resin SuperLig ® 639 a to selectively remove perrhenate from high ionic strength simulated LAW. It is advantageous to operate at higher concentration in order to treat the waste stream without dilution and to minimize the volume of the final wasteform. This work examined the impact of high ionic strength, high density, and high viscosity if higher concentration LAW feed solution is used. Perrhenate (ReO 4 - ) has been shown to be a good nonradioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin, and the performance bias is well established. Equilibrium contact testing with 7.8 M [Na + ] average simulant concentrations indicated that the SuperLig ® 639 resin average perrhenate distribution coefficient was 368 mL/g at a 100:1 phase ratio. Although this indicates good performance at high ionic strength, an equilibrium test cannot examine the impact of liquid viscosity, which impacts the diffusivity of ions and therefore the loading kinetics. To get an understanding of the effect of diffusivity, modeling was performed, which will be followed up with column tests in the future.

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

Technetium incorporation in glass for the Hanford Tank Waste treatment and immobilization plant - 15066

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of {sup 99}Tc (∼9 x 10 E2 TBq or ∼2.5 x 10{sup 4} Ci or ∼1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with {sup 99}Tc is its high mobility in addition to a long half-life (2.1 x 10{sup 5} yrs). The pertechnetate ion (TcO{sup 4-}) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed. (authors)

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

Niowave Neutron Source Converter: Lead-Bismuth-Eutectic (LBE) Windowless Target Design and Evaluation

Los Alamos National Laboratory (LANL) is working with Niowave on the design and evaluation of their lead-bismuth-eutectic (LBE) windowless target (i.e., neutron source converter). Niowave plans to use 200 kW electron beam at 40 MeV beam energy to produce neutrons by photonuclear reaction with LBE. Then, the neutrons undergo fission at the surrounding uranium target assembly (UTA) to produce Molybdenum 99 (Mo-99) as a fission product, which eventually decays to Technetium-99 (Tc-99m). Tc-99m is one of important radioisotopes that is used for medical diagnostics. LANL conducted 3D multiphysics analysis for the Niowave neutron converter design and provided design assessment in thermohydraulic aspects. LANL conducted radiation transport calculations using Monte-Carlo N-Particle (MCNP) code with unstructured meshing scheme. The 3D volumetric heating profiles in the LBE and Stainless-Steel (SS) housing were imported into multiphase computational fluid dynamics (CFD) to obtain 3D temperature profiles of LBE and SS through conjugate heat transfer (CHT) analysis. The key findings are: LBE film thickness at the center of the beam is approximately 1.6 cm with a maximum LBE velocity of approximately 1.8 m/s, which is below a 2 m/s limit to avoid erosion issues on supporting structures; Heat deposition in the LBE peaks at ~1 cm depth from the LBE free-surface because of the forward interactions of electron, photon, and neutron with LBE; LBE maximum temperature is ~360 °C, which is below LBE evaporation initiative temperature, ~450 °C; LBE-SS interface temperature is ~350 °C, which is below the safety thermal limit to prevent severe corrosion on SS. The results indicate that Niowave’s neutron converter design satisfies both hydraulic and thermal criteria for safe operation. By virtue of such computational analysis, Niowave can move toward establishing an experimental setup to experimentally test their LBE neutron converter. The following sections describe the detailed work done by LANL.

43 PARTICLE ACCELERATORS↗
DOE OSTI2011

Goethite Bench-scale and Large-scale Preparation Tests

The Hanford Waste Treatment and Immobilization Plant (WTP) is the keystone for cleanup of high-level radioactive waste from our nation's nuclear defense program. The WTP will process high-level waste from the Hanford tanks and produce immobilized high-level waste glass for disposal at a national repository, low activity waste (LAW) glass, and liquid effluent from the vitrification off-gas scrubbers. The liquid effluent will be stabilized into a secondary waste form (e.g. grout-like material) and disposed on the Hanford site in the Integrated Disposal Facility (IDF) along with the low-activity waste glass. The major long-term environmental impact at Hanford results from technetium that volatilizes from the WTP melters and finally resides in the secondary waste. Laboratory studies have indicated that pertechnetate ({sup 99}TcO{sub 4}{sup -}) can be reduced and captured into a solid solution of {alpha}-FeOOH, goethite (Um 2010). Goethite is a stable mineral and can significantly retard the release of technetium to the environment from the IDF. The laboratory studies were conducted using reaction times of many days, which is typical of environmental subsurface reactions that were the genesis of this new process. This study was the first step in considering adaptation of the slow laboratory steps to a larger-scale and faster process that could be conducted either within the WTP or within the effluent treatment facility (ETF). Two levels of scale-up tests were conducted (25x and 400x). The largest scale-up produced slurries of Fe-rich precipitates that contained rhenium as a nonradioactive surrogate for {sup 99}Tc. The slurries were used in melter tests at Vitreous State Laboratory (VSL) to determine whether captured rhenium was less volatile in the vitrification process than rhenium in an unmodified feed. A critical step in the technetium immobilization process is to chemically reduce Tc(VII) in the pertechnetate (TcO{sub 4}{sup -}) to Tc(Iv)by reaction with the ferrous ion, Fe{sup 2+}-Fe{sup 2+} is oxidized to Fe{sup 3+} - in the presence of goethite seed particles. Rhenium does not mimic that process; it is not a strong enough reducing agent to duplicate the TcO{sub 4}{sup -}/Fe{sup 2+} redox reactions. Laboratory tests conducted in parallel with these scaled tests identified modifications to the liquid chemistry necessary to reduce ReO{sub 4}{sup -} and capture rhenium in the solids at levels similar to those achieved by Um (2010) for inclusion of Tc into goethite. By implementing these changes, Re was incorporated into Fe-rich solids for testing at VSL. The changes also changed the phase of iron that was in the slurry product: rather than forming goethite ({alpha}-FeOOH), the process produced magnetite (Fe{sub 3}O{sub 4}). Magnetite was considered by Pacific Northwest National Laboratory (PNNL) and VSL to probably be a better product to improve Re retention in the melter because it decomposes at a higher temperature than goethite (1538 C vs. 136 C). The feasibility tests at VSL were conducted using Re-rich magnetite. The tests did not indicate an improved retention of Re in the glass during vitrification, but they did indicate an improved melting rate (+60%), which could have significant impact on HLW processing. It is still to be shown whether the Re is a solid solution in the magnetite as {sup 99}Tc was determined to be in goethite.

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

Non-pertechnetate Technetium Sensor Research and Development

Several significant uncertainties remain regarding the understanding and modeling of the fate and speciation of technicium-99 ( 99 Tc) in Hanford waste tanks, glass, and low-temperature waste forms. A significant (2% to 25%) fraction of the 99Tc in the water-soluble portion of the tank waste may be present as one or more non pertechnetate species that have not been identified and to date, cannot be effectively separated from the wastes. This task will provide a sensor specifically tuned to detect the Tc(I)-carbonyl species believed to constitute the main fraction of the non-pertechnetate form of technetium. By direct measurement of the non-pertechnetate species, such a sensor will help reduce the uncertainties in the modeling of the fate and speciation of 99 Tc in Hanford tanks and waste forms. This report summarizes work performed in FY2016 that was sponsored by the Department of Energy’s Office of Environmental Management and demonstrates the protocol for using fluorescent Tc(I)-tricarbonyl complex as a means to detect the non-pertechnetate species within tank waste solutions. The protocol was optimized with respect to ligand concentration, solvent choice, reaction temperature and time. This work culminated in the quantitation of Tc(I)-tricarbonyl within a waste simulant, using a standard addition method for measurement. This report also summarizes the synthesis and high-yield preparation of the low-valence technetium species, [Tc(CO) 3 (H 2 O) 3 ] + , which will be used as the technetium standard material for the demonstration of the non-pertechnetate species in actual wastes.

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

Recovery of high specific activity molybdenum-99 from accelerator-induced fission on low-enriched uranium for technetium-99m generators

Abstract A new process was developed to recover high specific activity (no carrier added) 99 Mo from electron-accelerator irradiated U 3 O 8 or uranyl sulfate targets. The process leverages a novel solvent extraction scheme to recover Mo using di(2-ethylhexyl) phosphoric acid following uranium and transuranics removal with tri- n -butyl phosphate. An anion-exchange concentration column step provides a final purification, generating pure 99 Mo intended for making 99 Mo/ 99m Tc generators. The process was demonstrated with irradiated uranium targets resulting in more than 95% 99 Mo recovery and without presence of fission products or actinides in the product.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

A biokinetic model for systemic technetium in adult humans

The International Commission on Radiological Protection (ICRP) currently is updating its biokinetic and dosimetric models for internally deposited radionuclides. Technetium (Tc), the lightest element that exists only in radioactive form, has two important isotopes from the standpoint of potential risk to humans: the long-lived isotope 99 Tm(T 1/2 =2.1x10 5 y) is present in high concentration in nuclear waste, and the short-lived isotope 99m Tc (T 1/2 =6.02 h) is the most commonly used radionuclide in diagnostic nuclear medicine. This paper reviews data on the biological behavior of technetium and proposes a biokinetic model for systemic technetium in the adult human body for use in radiation protection. Compared with the ICRP s current occupational model for systemic technetium, the proposed model provides a more realistic description of the paths of movement of technetium in the body; provides greater consistency with experimental and medical data; and, for most radiosensitive organs, yields substantially different estimates of cumulative activity (total radioactive decays within the organ) following uptake of 99 Tm or 99 mTc to blood.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2014

Low Activity Waste Feed Solids Characterization and Filterability Tests

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so its disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of 99 Tc from the LAW Feed, followed by off-site disposal of the 99 Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flow-sheets will specifically examine removing 99 Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the 99 Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process. Although a Supplemental LAW feed simulant has previously been prepared, this feed composition differs from that simulant because those tests examined only the fully soluble aqueous solution at room temperature, not the composition formed after evaporation, including the insoluble solids that precipitate after it cools. The conceptual flow sheet for Supplemental LAW immobilization has an option for removal of 99 Tc from the feed stream, if needed. Elutable ion exchange has been selected for that process. If implemented, the stream would need filtration to remove the insoluble solids prior to processing in an ion exchange column. The characteristics, chemical speciation, physical properties, and filterability of the solids are important to judge the feasibility of the concept, and to estimate the size and cost of a facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Chelators and metal complex stability for radiopharmaceutical applications

Abstract Diagnostic and therapeutic nuclear medicine relies heavily on radiometal nuclides. The most widely used and well-known radionuclide is technetium-99m ( 99m Tc), which has dominated diagnostic nuclear medicine since the advent of the 99 Mo/ 99m Tc generator in the 1960s. Since that time, many more radiometals have been developed and incorporated into potential radiopharmaceuticals. One critical aspect of radiometal-containing radiopharmaceuticals is their stability under in vivo conditions. The chelator that is coordinated to the radiometal is a key factor in determining radiometal complex stability. The chelators that have shown the most promise and are under investigation in the development of diagnostic and therapeutic radiopharmaceuticals over the last 5 years are discussed in this review.

Chemistry↗