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

Moly 99 Reactor Case Study

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

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

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

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

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

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

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

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

Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

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

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

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

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2017

Experimental Observations of Nuclear Activity in Deuterated Materials Subjected to a Low-Energy Photon Beam

Exposure of highly deuterated materials to a low-energy (nom. 2 MeV) photon beam resulted in nuclear activity of both the parent metals of hafnium and erbium and a witness material (molybdenum) mixed with the reactants. Gamma spectral analysis of all deuterated materials, ErD2.8+C36D74+Mo and HfD2+C36D74+Mo, showed that nuclear processes had occurred as shown by unique gamma signatures. For the deuterated erbium specimens, posttest gamma spectra showed evidence of radioisotopes of erbium ((163)Er and (171)Er) and of molybdenum ((99)Mo and (101)Mo) and by beta decay, technetium (99mTc and 101Tc). For the deuterated hafnium specimens, posttest gamma spectra showed evidence of radioisotopes of hafnium (180mHf and 181Hf) and molybdenum ((99)Mo and (101)Mo), and by beta decay, technetium ((99m)Tc and (101)Tc). In contrast, when either the hydrogenated or non-gas-loaded erbium or hafnium materials were exposed to the gamma flux, the gamma spectra revealed no new isotopes. Neutron activation materials showed evidence of thermal and epithermal neutrons. CR-39 solid-state nuclear track detectors showed evidence of fast neutrons with energies between 1.4 and 2.5 MeV and several instances of triple tracks, indicating (is) greater than 10 MeV neutrons. Further study is required to determine the mechanism causing the nuclear activity.

Steinetz, Bruce M.↗
DOE OSTI2013

Ion Exchange Column Tests Supporting Technetium Removal Resin Maturation

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

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

Testing of New Technetium Selective Media TcTreat - 16497

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

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

Design of a Commercial Scale Accelerator Driven Subcritical Aqueous Assembly

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

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Fabrication and Properties of Advanced Ceramic Technetium Waste Forms - 17311

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

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