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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Investigating the Effect of Sorbed Humic Acid on the Mobility of Uranium - 19461

The Savannah River Site (SRS) was one of the major nuclear facilities for the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, the F-Area consisted of three unlined seepage basins that collected approximately 6.8 billion liters of low-level waste that contained nitric acid, radionuclides, and dissolved metals due to plutonium separation operations. Sufficient quantities of uranium isotopes, I-129, Tc-99, and tritium migrated into the groundwater, creating an acidic plume with a pH between 3 and 5.5. Although the basins functioned as designed, the acidic nature of the waste created a source of contamination for the groundwater and soil below the basins. With radionuclides above their maximum contaminant level (MCL) remediation efforts took place. Humic acid as a remediation technology has shown potential for controlling radionuclide mobility. Humic acid is a fraction of humic substance that are ubiquitous in the environment. Humic substances (HS) are complex heterogeneous mixtures of polydispersed materials formed by biochemical and chemical reactions during the decay and transformation of plant and microbial remains. Humic substances account for 50-80% of the organic carbon in the soil or sediment and are known for their excellent binding capacity for metals. Humic acid can be extracted with a potassium hydroxide base solution and then drying it. The high pH solubilizes the humic acid molecules and generates a dark-brown highly-concentrated solution, rich in humic acid, which can be diluted for use. Humic acid's strong sorption capacity and its ability to develop a bond with uranium at slightly acid pH constitute it as remediation technology to reduce uranium (VI) mobility. Column experiments were performed to identify the sorption and desorption of HS in flow-through porous media representative of the acidic aquifer at the F-Area Seepage Basins. Two commercially available HA were selected for comparison, Huma-K and KW15 modified humics (modified humic acid, mod-HA). By conducting a nonreactive rhenium tracer test, on columns filled with soil obtained from SRS's FAW-1 at a depth interval of 18 m -21 m, column parameters such as pore volume (PV), variance, dispersion and peclet number were determined. Columns were then used to estimate the influence of HA on uranium mobility. Data obtained from this study shows that using HA as an amendment for uranium remediation improves uranium removal significantly. (authors)

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

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

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

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

Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Development of an accelerator-based neutron source to prototype Mo-99 production, part I: A liquid LBE windowless target

In this article, Molybdenum-99 (Mo-99)’s decay product, technetium-99 (Tc-99 m), is one of the most critical isotopes for medical diagnostics. To provide U.S. domestic supply of Mo-99 without using high-enriched uranium (HEU), a subcritical uranium target assembly (UTA) is irradiated by an accelerator-based neutron source to create Mo-99 through fission. This study discusses the development of the accelerator-based neutron source. The high-energy electrons from the accelerator irradiate a liquid lead-bismuth eutectic (LBE) target to produce neutrons. Part I of this work focuses on numerical and experimental analysis towards the development of a liquid LBE windowless target. Unlike the existing windowless targets in literature, the current design creates a vertical free surface for a beam to irradiate. First, a hydrodynamic analysis of the LBE windowless target is performed. Simplified analytical calculations are assisted by 2D computational fluid dynamics (CFD) simulations to design the target, with the focus on eliminating recirculation zones and avoiding cavitation. With the optimized geometry, the experimental study is performed to investigate the flow hydrodynamics using liquid LBE. The experiments (1) compare pressure drop in the system to correlation predictions; (2) visualize the free surface liquid LBE flow from the beam view; (3) validate the LBE flow profile using temperature sensitive paint from the side view; and (4) validate the liquid LBE film thickness using gamma densitometer measurements. Second, the power handling capability of the designed windowless target is investigated. The divider plate in the current design is susceptible to overheating due to the thin LBE film in front. As LBE erosion and corrosion is likely to occur at an LBE velocity of 2.0 m/s and temperature above 500 °C, a power limit of 10 kW of beam power was established to prevent this corrosion from occurring, which is calculated by a Nusselt number correlation. The divider plate surface temperature at 10 kW agrees well with the 3D CFD simulation results. Part I demonstrates the fundamental physics in liquid LBE windowless target design and associated testing. A companion paper, Part II will demonstrate how to couple this windowless target into the Mo-99 production system, including an accelerator system operating under an ultra-high vacuum and the UTA cooled by water at room temperature.

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

Producing radiometals in liquid targets: Proof of feasibility with {sup 94m}Tc

{sup 94m}Tc was produced in a liquid target loaded with a molybdenum-salt solution. This novel technique allows for the irradiation of metals dissolved in a liquid solution, normally only available in metal powder or foil form. By using this approach, the existing liquid targets and transfer infrastructure of many PET cyclotrons can be used to produce radiometals, avoiding the need, expense and challenges of operating solid targets. Such an approach allows for rapid testing of new isotopes for proof of feasibility studies. Different concentrations of Mo solution and their effect on the target performance were tested. Sufficient quantities to allow for preclinical studies were produced.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

ELECTRON-CAPTURE SUPERNOVAE AS SOURCES OF {sup 60}Fe

We investigate the nucleosynthesis of the radionuclide {sup 60}Fe in electron-capture supernovae (ECSNe). The nucleosynthetic results are based on a self-consistent, two-dimensional simulation of an ECSN as well as models in which the densities are systematically increased by some factors (low-entropy models). {sup 60}Fe is found to be appreciably made in neutron-rich ejecta during the nuclear quasi-equilibrium phase with greater amounts being produced in the lower-entropy models. Our results, combining them with the yields of core-collapse supernovae in the literature, suggest that ECSNe account for at least 4%-30% of live {sup 60}Fe in the Milky Way. ECSNe co-produce neutron-rich isotopes, {sup 48}Ca, {sup 50}Ti, {sup 54}Cr, some light trans-iron elements, and possibly weak r-process elements including some radionuclides such as {sup 93}Zr, {sup 99}Tc, and {sup 107}Pd, whose association with {sup 60}Fe might have been imprinted in primitive meteorites or in the deep ocean crust on the Earth.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2013

Radiation Stability of Benzyl Tributyl Ammonium Chloride Towards Technetium-99 Extraction

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning process, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, low nitric acid concentration in the aqueous phase of a UREX based process will co-extract U(VI) and Tc(VII) by tri-n-butyl phosphate (TBP). Technetium (Tc-99) is recognized to be one of the most abundant, long-lived radiotoxic isotopes in UNF (half-life, t1/2 = 2.13 × 105 years), and as such, it is targeted in UNF separation strategies for isolation and encapsulation in solid waste forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flow sheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste forms for ultimate disposal. Furthermore, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macrocompounds such as crown-ethers, aza-crown ethers, and resorcinarenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO4- by benzyltributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligand’s matrix conditions and concentration, as well as varying the organic phase composition (i.e., diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using external Co-60 source. Post-irradiation solvent extraction measurements will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Large-Batch Reduction of Molybdenum Trioxide

Unconverted, isotopically-enriched molybdenum metal must be recovered from the spent radiopharmaceutical solution used in NorthStar’s Technetium-99m generator and reused. The recycle process begins by recovering the metal from the aqueous potassium molybdate (K 2 MoO 4 ) solutions as molybdenum trioxide (MoO 3 ) employing a process developed at Argonne National Laboratory. The MoO 3 powder is subsequently reduced to molybdenum metal powder which can be blended with new powder and further processed into a flowable form to be used to produce target disks for irradiation. The molybdenum oxide reduction process has been examined and scaled to produce kilogram quantities of metal powder suitable for processing into a useable form employing spray drying or similar technique and ultimately used for target fabrication.

36 MATERIALS SCIENCE↗
NASA NTRS2004

Lunar and Planetary Science XXXV: Presolar Grains

Topics discussed include: Presolar Silicate Grains; Presolar Silicates from Primitivr Chondrites; Oxygen and Magnesium Isotopic Ratios of Presolar Spinel Grains; Study of Two New Presolar Grains from Bishunpur Ordinary Chondrite; Extinct Technetium in Presolar Grains; etc.

Source record↗
DOE OSTI2021

Radiological Handling and Containment Considerations in Support of an American Medical Isotope Producer

This work will discuss evaluation and design of new and modified facility level radiological containment, transfer, and handling systems in support of the United States’ American Medical Isotope Production Act. In the United States alone, Molybdenum-99 (Mo-99) is used as a precursor to locally generate its decay product, technetium-99m (Tc-99m), for use in approximately 40,000 medical procedures every day to detect cancer and heart disease. Currently, however, Mo-99 is primarily produced overseas and typically requires highly enriched uranium (HEU) which is classified as weapon -usable. This poses a serious security risk to the United States and, in 2012, Congress passed the American Medical Isotope ACT which aimed to cease all Mo-99 production that requires HEU and support American based companies who have found safer production methods. Through this effort, a system was designed and developed to access irradiated material in processing hot cells for ease of material introduction and removal in the confined area. The proposed solution uses a double-door design that allows for easy access to material while also providing shielding through a mechanically manipulated carousel. A system was also designed to transport irradiated materials from target reactors to a target receipt box using a rail guide system, pneumatic motor, and lead screw drive.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Preparation of Radiopharmaceuticals Labeled with Metal Radionuclides

The overall goal of this project was to develop methods for the production of metal-based radionuclides, to develop metal-based radiopharmaceuticals and in a limited number of cases, to translate these agents to the clinical situation. Initial work concentrated on the application of the radionuclides of Cu, Cu-60, Cu-61 and Cu-64, as well as application of Ga-68 radiopharmaceuticals. Initially Cu-64 was produced at the Missouri University Research Reactor and experiments carried out at Washington University. A limited number of studies were carried out utilizing Cu-62, a generator produced radionuclide produced by Mallinckrodt Inc. (now Covidien). In these studies, copper-62-labeled pyruvaldehyde Bis(N{sup 4}-methylthiosemicarbazonato)-copper(II) was studied as an agent for cerebral myocardial perfusion. A remote system for the production of this radiopharmaceutical was developed and a limited number of patient studies carried out with this agent. Various other copper radiopharmaceuticals were investigated, these included copper labeled blood imaging agents as well as Cu-64 labeled antibodies. Cu-64 labeled antibodies targeting colon cancer were translated to the human situation. Cu-64 was also used to label peptides (Cu-64 octriatide) and this is one of the first applications of a peptide radiolabeled with a positron emitting metal radionuclide. Investigations were then pursued on the preparation of the copper radionuclides on a small biomedical cyclotron. A system for the production of high specific activity Cu-64 was developed and initially the Cu-64 was utilized to study the hypoxic imaging agent Cu-64 ATSM. Utilizing the same target system, other positron emitting metal radionuclides were produced, these were Y-86 and Ga-66. Radiopharmaceuticals were labeled utilizing both of these radionuclides. Many studies were carried out in animal models on the uptake of Cu-ATSM in hypoxic tissue. The hypothesis is that Cu-ATSM retention in vivo is dependent upon the oxygen retention of the tissue and the significantly greater retention amounting in hypoxic tissue. This hypothesis was confirmed in a series of animal studies. Cu-64 can be used both as an imaging radionuclide and a therapeutic radionuclide. The therapeutic efficacy of Cu-64 ATSM was proven in hamsters bearing the CW39 human colorectal tumors. The administration of Cu-64 ATSM significantly increased the survival time of tumor-bearing animals with no acute toxicity. This copper agent therefore shows promise for radiotherapy. The flow tracer Cu-64 PTSM also demonstrates therapeutic potential by inhibiting cancer cells implanted in animal models. Again, this inhibition occurred at doses which showed no sign of toxicity to the animals. Cu-ATSM was translated to humans, under other support a series of tumors were investigated; these included head and neck cancer, non-small cell lung cancer, cervical cancer and renal cancer. Another radionuclide that was investigated was titanium 45. This radionuclide was successfully produced by radiation of a scandium foil with 15 MeV protons. The titanium 45 was processed and separated from residual scandium by high exchange chomotrophy. Titanium titanocene has been utilized as a therapeutic agent; this compound was prepared and studied in vitro and in vivo. Another project was the preparation of cyclodextrin dimers as a new pre-targeting approach for tumor uptake. Beta-cyclodextradin and two other dimers were synthesized. These dimers were studied for the in vivo application. Work continued on the application of the radionuclide already discussed. Technetium 94m, a positron emitting radionuclide of the widely used 99m Tc nuclide was also prepared. This allows the quantification of the uptake of technetium radiopharmaceuticals. In collaboration with Professor David Piwnica-Worms, technetium 94m, sestamibi was studied in animal models and in a limited number of human subjects.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2013

Recovery of Mo for Accelerator Production of Mo-99 Using (y,n) Reaction on Mo-100

Technetium-99m is a widely used radiopharmaceutical. Its parent, Mo-99, is produced worldwide to supply this important isotope. One means to produce Mo-99 is by bombarding a Mo-100 target with an electron beam from a linear accelerator; the γ/n reaction on Mo-100 produces Mo-99. After dissolving Mo-100 enriched disks in hydrogen peroxide, the solution is converted to potassium molybdate (0.2 g-Mo/mL) in 5 M KOH. After milking the Tc-99m in the TechneGen generator over a period of 7-10 days, the molybdenum solution needs to be treated to recover valuable Mo-100 for production of sintered Mo disks. However, during the production of Mo-99 by (γ, n) reaction on the Mo-100 target, several byproducts are formed. Therefore, recycling Mo will require the conversion of K 2 MoO 4 in 5 M KOH solution to MoO 3 powder, and purification from other metals present in the Mo solution. The starting Mo-100 enriched material contains less than 20 mg of potassium in 1 kg of molybdenum (<20 ppm). However, after dissolving the irradiated Mo-100 target in hydrogen peroxide and converting it to K 2 MoO 4 in 5 M KOH (0.2 g-Mo/mL), the solution contains about 1.8 kg of potassium per kilogram of molybdenum. The most challenging separation for this recovery step is purifying molybdenum from potassium. One requirement to facilitate the acceptance of the recycled material by the U.S. Food and Drug Administration (FDA) is that the impurities in the recycled material need to be at or below the levels present in the starting material. Therefore, the amount of potassium (K) in purified MoO 3 powder should be below 20 ppm; this will require a decontamination factor for removal of K to be ~1 × 10 5 . Such a low K-contamination level will also prevent the production of large amounts of K-42 during irradiation of Mo-100. Based on economic concerns (due to the significant cost of enriched Mo-100) recycling Mo requires the conversion of K 2 MoO 4 in a 5 M KOH solution to MoO 3 powder with high Mo recovery yields (>98%).

Tkac, Peter↗
DOE OSTI2016

Van de Graaff Irradiation of Materials

Through irradiations using our 3 MeV Van de Graaf accelerator, Argonne is testing the radiation stability of components of equipment that are being used to dispense molybdenum solutions for use as feeds to 99m Tc generators and in the 99m Tc generators themselves. Components have been irradiated by both a direct electron beam and photons generated from a tungsten convertor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

The use of carbohydrazide for plutonium concentration stripping in separator with inert packing

For the purpose of removing plutonium from uranium- plutonium extract it is proposed to employ concentration stripping process with the use of separator and a new reducing reagent: Carbohydrazide CO(N{sub 2}H{sub 3}){sub 2}. Using plutonium stripping from solution simulating the composition of extract of spent nuclear fuel from VVER-1000 reactor (without γ-emitting isotopes), with O: A ratio of = 28, a product solution was obtained containing 17.8 g/l of plutonium, 29.2 g/l of uranium and more than 1 g/l of technetium. The experiment on real spent fuel from VVER-1000 with burn-up of more than 50 GW*d/t of uranium after 17 year exposure, performed in the shielded box of FSUE 'MCP', confirmed the effectiveness and feasibility of the proposed process. Through concentration stripping (O:A = 20), a plutonium product solution was obtained with a part of uranium with the following composition: [U] = 150 g/l; [Pu] = 23,5 g/l; [Np] = 1,7 g/l, [Tc] = 1.5 g/l; gamma exposure rate - 0,022 mR/s*l. Direct extraction of plutonium in this operation was 95.3%, the rest of plutonium is refluxing to the preceding stage of the extraction cycle. A process flow diagram with organization of plutonium recycling is proposed, allowing for its complete removal into a single stream. Carbohydrazide is an effective reducing agent of plutonium (IV), ensuring the stability of uranium-plutonium separation process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Modeling and Analysis Support for Acceleratro-based Production of Mo-99

NorthStar Medical Radioisotopes, LLC, is in the process of commissioning a medical isotope production facility, which will use high-power electron accelerators to produce molybdenum-99 (Mo-99), the parent of technetium-99m (Tc-99m), through photonuclear reactions in molybdenum-100 (Mo-100). In this approach, a target comprising multiple thin disks of enriched molybdenum metal is bombarded with a 40-MeV electron beam. Electrons that impinge on the molybdenum target produce bremsstrahlung x-rays that cause the nuclear reaction. Because enriched Mo-100 is expensive, there is a desire to use as much beam power as possible to achieve maximum production yield and minimize the size of the target. This requirement leads to very high beam power density (heat deposition in the target), which creates challenging requirements for target cooling. In the latest concept developed by NorthStar, a stack of target (sintered molybdenum) disks is irradiated from two sides by a 40-MeV electron beam, with a total power of 250 kW (125 kW from each side). The target disks are cooled by pressurized helium gas flowing through channels between the disks. Inconel windows in the target housing form a pressure boundary between the helium within the housing and the evacuated beam tube.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

A New Process for Small-Batch Purification of the Medical Isotope Molybdenum-99: Non-Technical Overview

The U.S. medical community depends on a reliable supply of the radioisotope molybdenum-99 (Mo-99) for nuclear medical diagnostic procedures. Mo-99's decay product, technetium-99m (Tc-99m), is used in over 40,000 medical procedures in the United States each day to diagnose heart disease and cancer, to study organ structure and function, and to perform other important medical applications. For example, patients undergoing a common procedure—the cardiac “stress test”—likely have benefited from Tc-99m. Historically, Mo-99 was primarily produced through the fission of uranium-235, in the form of highly enriched uranium (HEU) targets irradiated in research and test reactors. HEU is a proliferation-sensitive material that, if diverted or stolen, could be used as a component of a nuclear weapon. NNSA’s Office of Material Management and Minimization (M3) manages the Molybdenum-99 (Mo-99) Program as part of its mission to minimize the use of HEU in civilian applications. The Mo-99 Program assists global Mo-99 production facilities in converting to non-HEU processes and supports the establishment of domestic supplies of Mo-99 without the use of proliferation-sensitive HEU. As part of this program, M3 funds U.S. national laboratories to provide non-proprietary technical support to U.S. companies working to establish non-HEU-based Mo-99 production capabilities. The results of this research are published on OSTI.gov for the benefit of the Mo-99 community and the public. However, it can be difficult for readers without a scientific background to understand and interpret these publications. In order to increase public understanding of the work being done in M3’s Mo-99 Program, this paper aims to provide an overview of a key, recent national laboratory technical publication in terms that can be understood by readers without a technical background. To accomplish this, the paper first explains key scientific concepts—primarily related to chemistry—that provide a foundation for understanding research in this area. This includes chromatography, absorption vs. adsorption, dissolution and precipitation, and liquid-liquid extraction. Drawing on these concepts, the paper then provides an explanation for non-technical audiences of the Argonne National Laboratory publication entitled Recovery of High Specific Activity Molybdenum-99 from Accelerator-Induced Fission on Low-Enriched Uranium for Technetium-99m Generators (Brown, M.A. et al., 2021) and related article Separation and Purification of Mo-99 Produced from Natural U3O8 Targets via Photo-Fission (Brown, M.A. et al., 2021).

Stamler, Bradley↗
DOE OSTI2013

Radiation Stability of Benzyl Tributyl Ammonium Chloride towards Technetium-99 Extraction - 13016

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning processes, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, the UREX process has been developed in the United States to co-extract hexavalent uranium (U) and hepta-valent technetium (Tc) by tri-n-butyl phosphate (TBP). Tc-99 is recognized to be one of the most abundant, long-lived radio-toxic isotopes in UNF (half-life, t{sub 1/2} = 2.13 x 10{sup 5} years), and as such, is targeted in UNF separation strategies for isolation and encapsulation in solid waste-forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste-form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flowsheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste-forms for ultimate disposal. In addition, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste-forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macro-compounds such as crown-ethers, aza-crown ethers, quaternary ammonium salts, and resorcin-arenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO{sub 4}{sup -} by benzyl tributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligand's matrix conditions and concentration, as well as varying the organic phase composition (i.e. diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using an external Co-60 source. Post-irradiation solvent extraction measurements will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to create samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)

07 ISOTOPE AND RADIATION SOURCES↗