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All sourcesNASA NTRSDOE OSTI
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↗
DOE OSTI2012

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Properties of Pertechnic Acid

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

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

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

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

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

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

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

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

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

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

Chemistry↗
DOE OSTI2019

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

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

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

Technetium Removal Using Tc-Goethite Coprecipitation

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

Goethite↗
DOE OSTI2014

Technetium Inventory, Distribution, and Speciation in Hanford Tanks

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

storage tanks↗
DOE OSTI2015

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

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

Cocktail↗