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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

[{sup 13}N] Ammonia Cardiac Program At West Virginia University Health Sciences

Due to the shortage of the more traditional cardiac imagining isotopes, specifically, Technicium-99, the Cardiologists at WVU have had to look to alternative imagining techniques such as PET. This has led to a dramatic increase in the use of [{sup 13}N] Ammonia PET scans at the Health Sciences Center. The patient load has gone from one to two patients one day a week to typically two to three patients, two days a week, with occasional add-on in-house patients; each patient typically requiring two target irradiations. In this paper, we will discuss the process that is being used to meet this increased demand from the production of the isotope through the final result.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2020

Seasonal to Decadal Variability in Focused Groundwater and Contaminant Discharge along a Channelized Stream

Abstract Focused groundwater discharge to streams is problematic at contaminated sites because high fluxes can limit natural attenuation in the hyporheic zone. However, information on location, spatial evolution, and temporal persistence of springs in unlithified sediments over multiyear time scales is limited. We examine discharge at point (~1‐m) to reach (~300‐m) scales along a stream that intercepts trichloroethene and technetium‐99 plumes from a Superfund site. During 2011 to 2012, we seasonally monitored stream and spring flow and contaminant concentrations, along with probing streambed temperatures on a grid in winter and summer, building on prior monitoring during 1999 to 2002. Baseflow measured by both gauging and dye dilution generally increased with distance downstream, and stream and spring discharge varied seasonally, from minima in October to January to maxima in February to June. Thermal anomalies identified by probing occupied approximately 3% to 6% of the reach and typically coincided with visible springs or seeps. Locations of anomalies were similar to those identified in summer 2002, although some orifices disappeared and others emerged. Vertical groundwater fluxes calculated from probing tended to be less than net fluxes calculated from stream discharge, perhaps in part because the assumption of one‐dimensional, steady‐state flow in calculating point fluxes was simplistic. Maximum contaminant concentrations and fluxes decreased between 1999 to 2001 and 2011 to 2012 as a result of partial capture by an upgradient pump‐and‐treat system. Our findings confirm that springs in unlithified sediments can remain stationary within a few meters over decadal time scales, and seasonal variability in discharge can be greater than decadal variability.

Tripathi, Ganesh N.↗
DOE OSTI2019

Leach Testing of Laboratory-Scale Melter Evaporator Condensate Cementitious Waste Forms

Cementitious waste forms immobilizing evaporated AP-105 off-gas condensate were used to determine radionuclide and contaminant leachability in deionized water over a 63-day period. The off-gas condensate was collected during the actual tank AP-105 low activity waste continuous laboratory-scale melter (LSM)) run performed in fiscal year 2018 and was then solidified as a cementitious waste form using the Cast Stone formulation (8 weight percent [wt%] ordinary Portland cement, 45 wt% class F fly ash, 47 wt% blast furnace slag). Here, using the U.S. Environmental Protection Agency Method 1315 leach testing procedure, the release rate of technetium 99 (Tc), iodine-129 (I), nitrate (NO 3 - ), and sodium (Na + ) from two Cast Stone specimens was measured. From these measurements, observed diffusivity values for each constituent were calculated and can be used in future iterations and maintenance of the Integrated Disposal Facility performance assessments. Furthermore, by testing Cast Stone specimens immobilizing off-gas produced while vitrifying actual tank waste, the results of this effort improve the technical defensibility of long-term waste form performance estimates for the Cast Stone formulation. Finally, 99 Tc distribution within a representative waste form at the end of the 63-day leach period as well as bulk mineralogy were determined using solid phase characterization methods.

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

Systems-based approach to understanding Technetium immobilization in the subsurface - 15326

The purpose of this work is to identify the chemical and biological reactions that occur and potentially influence the mobility of Technetium (Tc-99) in the subsurface. Tc-99 is a by-product of nuclear fission processes and has a relatively long half-life (t{sub 1/2}=212,000 years); additionally, it is highly soluble and mobile in groundwater and is most often remediated through pump and treat systems. The oxidized form of Tc-99 [Pertechnetate, Tc(VII)O{sub 4}-] is highly soluble and very mobile, while the reduced form of Tc-99 [Technetium-oxide, Tc(IV)O{sub 2}·nH{sub 2}O] is insoluble. Historically, complex interactions within the subsurface have made clean-up efforts for Tc-99 challenging. It is thought that environmental bacteria play a role in reducing contaminants to less soluble forms but it is difficult to predict or estimate the interactions and impacts of the complex, mixed microbial communities found in the environment. We have set up controlled batch experiments to examine a selection of organisms (Anaeromyxobacter dehalogenans strain CP-C, Geobacter sulfurreducens strain PAC, Geobacter daltonii strain FRC-32, and Shewanella oneidensis strain MR1) that have known metal reducing capabilities, for their ability to reduce Tc-99 to less soluble forms. An additional organism, Cellulomonas strain ES6 was chosen based on its origin of isolation from the Hanford site. Organisms were evaluated under anaerobic conditions both as isolates and in mixed consortia with and without the presence of iron-bearing mineral substrate. Experiments were monitored using spectroscopy, qPCR, volatile organic acid (VOA) analysis, and inductively coupled plasma-optical emission spectroscopy (ICP-OES) and mass spectrometer techniques (ICP-MS). Initial data collected has shown that the presence and type of substrate may increase the reduction of Tc-99 to less mobile forms. Data collected from these experiments will allow for (a) a better understanding of the mobility of Tc-99 in the subsurface, (b) better predictions of contaminant fate and transport over time, and (c) the development of long term remediation strategies for Tc-99. Additionally, information collected will be used to develop a systems-based metabolic model with predictive capabilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Exploiting Knowledge of Technetium in the Subsurface to Develop Remediation Strategies - 18223

Technetium-99 (Tc) contamination remains a major environmental problem at legacy nuclear reprocessing sites, such as the Hanford Site in south-central Washington State, where the total inventory of released Tc is estimated at 700 curies (Ci). The environmental risk associated with Tc is due to its long half-life (213,000 years) and the high mobility of the oxidized anionic pertechnetate species (Tc(VII)O{sub 4}{sup -}). Thus, there is a critical need to resolve scientific and technical issues underpinning the ability to safely, cost-effectively, and efficiently remediate Tc-contaminated vadose zone sediments and groundwater. Developing enhanced attenuation strategies for released Tc that reduce contaminant migration, while mitigating risk to human health and the environment, requires characterization of any retardation processes and Tc mass flux. The current conceptual model for Tc transport in the vadose zone assumes that Tc is predominantly present in its oxidized form as TcO{sub 4}{sup -}, which is highly soluble and mobile. However, under certain conditions, Tc can become entrained in solid phases. Tc-containing mineral phases from sediment cores extracted from various Hanford waste disposal sites were characterized to determine Tc present in vadose zone sediments as (i) TcO{sub 4} - that remains in water films associated with fine-grained sediments; (ii) Tc co-precipitated with other lanthanides in phosphate mineral phases; and (iii) Tc present as TcO{sub 2}.nH{sub 2}O in localized reducing zones. These results will be used to develop improved long-term Tc remediation strategies that are most likely to succeed during field application, through stimulation of conditions that enhance Tc attenuation. (authors)

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

LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in greatest abundance in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are low but are also expected to be in measurable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, {sup 241}Pu, and {sup 241}Am. These are present due to their partial volatility and some entrainment in the off-gas system. This report discusses results of optimized {sup 99}Tc decontamination testing of the simulant. Testing examined use of inorganic reducing agents for {sup 99}Tc. Testing focused on minimizing the quantity of sorbents/reactants added, and minimizing mixing time to reach the decontamination targets in this simulant formulation. Stannous chloride and ferrous sulfate were tested as reducing agents to determine the minimum needed to convert soluble pertechnetate to the insoluble technetium dioxide. The reducing agents were tried with and without sorbents. The sorbents, hydroxyapatite and sodium oxalate, were expected to sorb the precipitated technetium dioxide and facilitate removal. The Phase 1 tests examined a broad range of conditions and used the initial baseline simulant. The Phase 2 tests narrowed the conditions based on Phase 1 results, and used a slightly modified simulant. Test results indicate that excellent removal of {sup 99}Tc was achieved using SnCl{sub 2} as a reductant, and was effective with or without sorption onto hydroxyapatite. This reaction worked even in the presence of air (which could oxidize the stannous ion) and at room temperature. This process was very effective at neutral pH, with a Decontamination Factor (DF) >199 in one hour with only 1 g/L of SnCl{sub 2}. Prior work had shown that it was much less effective at alkaline pH. The only deleterious effect observed was that the chromium co-precipitates with the {sup 99}c during the SnCl{sub 2} reduction. This effect was anticipated, and would have to be considered when managing disposition paths of this stream. Reduction using FeSO{sub 4} was not effective at removing {sup 99}Tc, but did remove the Cr. Chromium is present due to partial volatility and entrainment in the off-gas, and is highly oxidizing, so would be expected to react with reducing agents more quickly than pertechnetate. Testing showed that sufficient reducing agent must be added to completely reduce the chromium before the technetium is reduced and removed. Other radionuclides are also present in this off-gas condensate stream. To enable sending this stream to the Hanford ETF, and thereby divert it from the recycle where it impacts the LAW glass volume, several of these also need to be removed. Samples from optimized conditions were also measured for actinide removal in order to examine the effect of the Tc-removal process on the actinides. Plutonium was also removed by the SnCl{sub 2} precipitation process. Results of this separation testing indicate that sorption/precipitation is a viable concept and has the potential to decontaminate the {sup 99}Tc from the stream, allowing it to be diverted away from WTP and thus eliminating the impact of the recycled halides and sulfate on the LAW glass volume. Based on the results, a possible treatment scenario could involve the use of a reductive precipitation agent (SnCl{sub 2}) with or without sorbent at neutral pH to remove the Tc. Although hydroxyapatite was not necessary to effect the {sup 99}Tc removal, it may be beneficial in solid-liquid separations. Other testing will examine removal of the other radionuclides. This testing was the second phase of testing, which aimed at optimizing the process by examining the minimum amount of reductant needed and the minimum reaction time. Although results indicated that SnCl{sub 2} was effective, further work on a pH-adjusted Fe(SO{sub 4}) mixture are needed. Additional tasks are needed to examine removal of the other radionuclides, solid-liquid separation technologies, slurry rheology measurements, composition variability impacts, corrosion and erosion, and slurry storage and immobilization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Iron oxide waste form for stabilizing 99 Tc

Crystals of goethite were synthesized with reduced technetium [{sup 99}Tc(IV)] incorporated within the solid lattice. The presence of {sup 99}Tc(IV) as a substituting cation in the matrix and 'armoring' by an additional layer of precipitated goethite isolated the reduced {sup 99}Tc(IV) from oxidizing agents. These products were used to make monolithic pellets to quantify an effective diffusion coefficient for {sup 99}Tc from goethite waste form contacted with a synthetic Hanford IDF (integrated disposal facility) pore water solution (pH = 7.2, I = 0.05 M) at room temperature for up to 120 days in static reactors. XANES analysis of the goethite solids recovered post-run demonstrated that the {sup 99}Tc in the goethite crystals remains in the reduced {sup 99}Tc(IV) state. The slow release of pertechnetate concentration with time in the static experiments with the monolith followed a square root of time dependence, consistent with diffusion control for {sup 99}Tc release. An apparent diffusion coefficient of 6.15 x 10{sup -11} cm{sup 2}/s was calculated for the {sup 99}Tc-goethite pellet sample and the corresponding leaching index (LI) was 10.2. The results of this study indicate that technetium can be immobilized in a stable, low-cost Fe oxide matrix that is easy to fabricate and these findings can be useful in designing long-term solutions for nuclear waste disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2017

Synthesis and Characterization of 5- and 6- Coordinated Alkali Pertechnetates

The local chemistry of technetium-99 ( 99 Tc) in oxide glasses is important for understanding the incorporation and long-term release of Tc from nuclear waste glasses, both those for legacy defense wastes and fuel reprocessing wastes. Tc preferably forms Tc(VII), Tc(IV), or Tc(0) in glass, depending on the level of reduction of the melt. Tc(VII) in oxide glasses is normally assumed to be isolated pertechnetate TcO 4 - anions surrounded by alkali, but can occasionally precipitate as alkali pertechnetate salts such as KTcO 4 and NaTcO 4 when Tc concentration is high. In these cases, Tc(VII) is 4-coordinated by oxygen. A reinvestigation of the chemistry of alkali-technetium-oxides formed under oxidizing conditions and at temperatures used to prepare nuclear waste glasses showed that higher coordinated alkali Tc(VII) oxide species had been reported, including those with the TcO 5 - and TcO 6 - anions. The chemistry of alkali Tc(VII) and other alkali-Tc-oxides is reviewed, along with relevant synthesis conditions. Additionally, we report attempts to make 5- and 6-coordinate pertechnetate compounds of K, Na, and Li, i.e. TcO 5 - and TcO 6 -. It was found that higher coordinated species are very sensitive to water, and easily decompose into their respective pertechnetates. It was difficult to obtain pure compounds, but mixtures of the pertechnetate and other phase(s) were frequently found, as evidenced by x-ray absorption spectroscopy (XAS), neutron diffraction (ND), and Raman spectroscopy. Low temperature electron paramagnetic resonance (EPR) measurements showed the possibility of Tc(IV) and Tc(VI) in Na 3 TcO 5 and Na 5 TcO 6 compounds. Here, it was hypothesized that the smaller counter cation would result in more stable pertechnetates. To confirm the synthesis method, LiReO 4 and Li 5 ReO 6 were prepared, and their Raman spectra match those in the literature. Subsequently, the Tc versions LiTcO 4 and Li 5 TcO 6 were synthesized and characterized by ND, Raman spectroscopy, XANES, and EXAFS. The Li 5 TcO 6 was a marginally stable compound that appears to have the same structure as that known for Li 5 ReO 6 . Implications of the experimental work on stability of alkali technetate compounds and possible role in the volatilization of Tc are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Results of four one-day electron-accelerator irradiations of enriched Mo-100 targets for the production of Mo-99

A series of four one-day irradiations was conducted with 100 Mo-enriched disk targets. After irradiation, the enriched disks were removed from the target and dissolved. The resulting solution was processed using a NorthStar RadioGenix™ 99m Tc generator either at Argonne National Laboratory or at the NorthStar Medical Radioisotopes facility. Runs on the RadioGenix system produced inconsistent analytical results for 99m Tc in the Tc/Mo solution. These inconsistencies were attributed to the impurities in the solution or improper column packing. During the irradiations, the performance of the optic transitional radiation (OTR) and infrared cameras was tested in high radiation field. The OTR cameras survived all irradiations, while the IR cameras failed every time. The addition of X-ray and neutron shielding improved camera survivability and decreased the number of upsets.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2011

Photoreduction of 99 Tc Pertechnetate by Nanometer-Sized Metal Oxides: New Strategies for Formation and Sequestration of Low-Valent Technetium

Technetium-99 ( 99 Tc)(β - max : 293.7 keV; t 1/2 : 2.1 x 10 5 years) is a byproduct of uranium-235 fission and comprises a large component of radioactive waste. Under aerobic conditions and in a neutral- basic environment, the pertechnetate anion ( 99 TcO 4 - ) is stable. 99 TcO 4 - is very soluble, migrates easily through the environment and does not sorb well onto mineral surfaces, soils or sediments. This study moves forward a new strategy for the reduction of TcO 4 - and chemical incorporation of the reduced 99 Tc into a metal oxide material. This strategy employs a single material, a polyoxometalate (POM), α 2- [P 2 W 17 O 61 ] 10- , that can be photoactivated in the presence of 2-propanol to transfer electrons to TcO 4 - and incorporate the reduced 99 Tc covalently into the α 2 - framework to form the Tc V O species, Tc V O(α 2- P 2 W 17 O 61 ) 7- . This occurs via the formation of an intermediate species that slowly converts to Tc V O(α 2- P 2 W 17 O 61 ) 7- . EXAFS and XANES analysis and preliminary EPR analysis, suggests that the intermediate consists of a Tc(IV) α 2 - species where the 99 Tc is likely bound to only 2 of the 4 W-O oxygen atoms in the α 2- [P 2 W 17 O 61 ] 10- defect. This intermediate then oxidizes and converts to the 99 Tc V O(α 2- P 2 W 17 O 61 ) 7- product. The reduction and incorporation of 99 TcO 4- was accomplished in a ''one pot'' reaction using both sunlight and UV irradiation, and monitored as a function of time using multinuclear NMR and radio TLC. The process was further probed by the ''step-wise'' generation of reduced α 2- P 2 W 17 O 61 12- through bulk electrolysis followed by the addition of TcO 4 - . The reduction and incorporation of ReO 4 - , as a non-radioactive surrogate for 99 Tc, does not proceed through the intermediate species, and Re V O is incorporated quickly into the α 2- [P 2 W 17 O 61 ] 10- defect. These observations are consistent with the periodic trends of 99 Tc and Re. Specifically, 99 Tc is more easily reduced compared to Re. In addition to serving as models for metal oxides, POMs may also provide a suitable platform to study the molecular level dynamics and mechanisms of the reduction and incorporation of Tc into a material.

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

Evaluation of Technetium Getters to Improve the Performance of Cast Stone

Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. One of the major radionuclides that Cast Stone has the potential to immobilize is technetium (Tc). The mechanism for immobilization is through the reduction of the highly mobile Tc(VII) species to the less mobile Tc(IV) species by the blast furnace slag (BFS) used in the Cast Stone formulation. Technetium immobilization through this method would be beneficial because Tc is one of the most difficult contaminants to address at the U.S. Department of Energy (DOE) Hanford Site due to its complex chemical behavior in tank waste, limited incorporation in mid- to high-temperature immobilization processes (vitrification, steam reformation, etc.), and high mobility in subsurface environments. In fact, the Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington (TC&WM EIS) identifies technetium-99 ( 99 Tc) as one of the radioactive tank waste components contributing the most to the environmental impact associated with the cleanup of the Hanford Site. The TC&WM EIS, along with an earlier supplemental waste-form risk assessment, used a diffusion-limited release model to estimate the release of different contaminants from the WTP process waste forms. In both of these predictive modeling exercises, where effective diffusivities based on grout performance data available at the time, groundwater at the 100-m down-gradient well exceeded the allowable maximum permissible concentrations for 99 Tc. (900 pCi/L). Recent relatively short-term (63 day) leach tests conducted on both LAW and secondary waste Cast Stone monoliths indicated that 99 Tc diffusivities were at or near diffusivities where the groundwater at the 100-m down-gradient well would exceed the allowable maximum permissible 99 Tc concentrations. There is, therefore, a need and an opportunity to improve the retention of Tc in the Cast Stone waste form. One method to improve the performance of the Cast Stone waste form is through the addition of “getters” that selectively sequester Tc inside Cast Stone.

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

Isotope Specific Remediation Media and Systems - 13614

On March 11, 2011, now two years ago, the magnitude 9.0 Great East Japan earthquake, Tohoku, hit off the Fukushima coast of Japan. While, of course, most of the outcome of this unprecedented natural and manmade disaster was a negative, both in Japan and worldwide, there have been some extremely invaluable lessons learned and new emergency recovery technologies and systems developed. As always, the mother of invention is necessity. Among these developments has been the development and full-scale implementation of proven isotope specific media (ISMs) with the intent of surgically removing specific hazardous isotopes for the purpose of minimizing dose to workers and the environment. The first such ISMs to be deployed at the Fukushima site were those removing cesium (Cs-137) and iodine (I-129). Since deployment on June 17, 2011, along with treated cooling water recycle, some 70% of the curies in the building liquid wastes have been removed by the Kurion system alone. The current levels of cesium are now only 2% of the original levels. Such an unprecedented, 'external cooling system' not only allowed the eventual cold shut down of the reactors in mid-December, 2011, but has allowed workers to concentrate on the cleanup of other areas of the site. Water treatment will continue for quite some time due to continued leakage into the buildings and the eventual goal of cleaning up the reactors and fuel pools themselves. With the cesium removal now in routine operation, other isotopes of concern are likely to become priorities. One such isotope is that of strontium, and yttrium (Sr-90 and Y-90), which is still at original levels causing further dose issues as well as impediments to discharge of the treated waste waters. For over a year now, a new synthetic strontium specific media has been under development and testing both in our licensed facility in Oak Ridge, Tennessee, but also in confirmatory tests by the Japan Atomic Energy Agency (JAEA) in Japan for Tokyo Electric Power Company (TEPCO). The tests have proven quite successful, even in high salt conditions, and, with loading and dose calculations being completed, will be proposed to add to the existing cesium system. There is no doubt, as high gamma isotopes are removed, other recalcitrant isotopes such as this will require innovative removal media, systems and techniques. Also coming out of this international effort are other ISM media and systems that can be applied more broadly to both Commercial Nuclear Power Plants (NPPs) as well as in Department of Energy (DOE) applications. This cesium and strontium specific media has further been successfully tested in 2012 at a Magnox station in the UK. The resulting proposed mitigation systems for pond and vault cleanup look quite promising. An extremely unusual ISM for carbon 14 (C-14), nickel (Ni-63) and cesium (Cs-137) has been developed for Diablo Canyon NPP for dose reduction testing in their fuel pool. These media will be deployed in Submersible Media Filter (SMF) and Submersible Columns (SC) systems adapted to standard Tri-Nuclear{sup R} housings common in the U.S. and UK. External Vessel Systems (mini-Fukushima) have also been developed as a second mitigation system for D and D and outages. Finally, technetium (Tc- 99) specific media developed for the Waste Treatment Plant (WTP) recycle or condensate (secondary) waste streams (WM 2011) are being further perfected and tested for At-Tank Tc-99 removal, as well as At Tank Cs media. In addition to the on-going media development, systems for deploying such media have developed over the last year and are in laboratory- and full-scale testing. These systems include the fore mentioned Submersible Media Filters (SMF), Submersible Columns (SC) and external pilot- and full-scale, lead-lag, canister systems. This paper will include the media development and testing, as well as that of the deployment systems themselves. (authors)

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

Interim glycol flowsheet reduction/oxidation (redox) model for the Defense Waste Processing Facility (DWPF)

Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe +2 /ΣFe ratios of between 0.09 and 0.33, a range which is not overly oxidizing or overly reducing, helps retain radionuclides in the melt, i.e. long-lived radioactive 99 Tc species in the less volatile reduced Tc 4+ state, 104 Ru in the melt as reduced Ru +4 state as insoluble RuO 2 , and hazardous volatile Cr 6+ in the less soluble and less volatile Cr +3 state in the glass. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam. Currently, the Defense Waste Processing Facility (DWPF) is running a formic acid-nitric acid (FN) flowsheet where formic acid is the main reductant and nitric acid is the main oxidant. During decomposition formate and formic acid releases H 2 gas which requires close control of the melter vapor space flammability. A switch to a nitric acid-glycolic acid (GN) flowsheet is desired as the glycolic acid flowsheet releases considerably less H 2 gas upon decomposition. This would greatly simplify DWPF processing. Development of an EE term for glycolic acid in the GN flowsheet is documented in this study.

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

Development of Tc(IV)-incorporated Fe minerals to enhance {sup 99}Tc retention in glass waste form - 15239

Iron minerals have been considered to be good hosts for Tc immobilization because the Tc(IV) ion substitutes for Fe(III) in the crystal structure of the Fe oxide due to similarities in (1) cation size [Tc(IV) = 78.5 pm; Fe(III) = 69 or 78.5 pm], (2) metal-oxygen interatomic distance (Tc-O = 0.199 nm, Fe-O = 0.203 nm), (3) number of coordinating oxygen atoms (both 6-fold coordinated), and (4) the redox potential (Eh=ca. +20 mV at pH = 7) for a redox couple between Tc(VII)/Tc(IV) and Fe(III)/Fe(II). Magnetite, maghemite, and trevorite are iron oxide minerals and all belong to spinel mineral group. Laboratory testing shows that Tc can be removed from aqueous waste solutions by a process of Tc reduction from Tc(VII) to Tc(IV) followed by co-precipitation with iron oxide minerals during recrystallization of Fe(OH){sub 2}(s) used as an initial solid precursor. X-ray absorption near edge structure (XANES) spectroscopy confirmed that Tc was in the +4 oxidation state in final Tc-Fe minerals. The Tc-incorporated Fe minerals (Tc-goethite and Tc-magnetite/maghemite) were also tested for Tc retention in glass melts at different temperatures between 600 - 1,000 deg.C in a furnace. After being cooled in air, the solid glass specimens collected at different temperatures were analyzed for Tc oxidation state using XANES and Tc retention using liquid scintillation counting (LSC). Although Tc(IV) started to re-oxidize at 600 deg.C, Tc retention in the final glass specimen prepared with Tc-incorporated Fe mineral at high temperatures was at least two times higher than glass prepared with KTcO{sub 4} salt. Higher Tc retention in glass is considered to result from limited and delayed Tc volatilization process due to Fe mineral encapsulation for Tc. Therefore, the results showing the presence of Tc(IV) in the Fe mineral structure indicate strong possibility to enhance Tc retention in borosilicate glass as well as to reduce the remediation costs at the Hanford Site. (authors)

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

Structure and Thermochemistry of Perrhenate Sodalite and Mixed Guest Perrhenate/Pertechnetate Sodalite

Treatment and immobilization of technetium-99 ( 99 Tc) contained in reprocessed nuclear waste and present in contaminated subsurface systems represents a major environmental challenge. One potential approach to managing this highly mobile and long-lived radionuclide is immobilization into micro- and meso-porous crystalline solids, specifically sodalite. We synthesized and characterized the structure of perrhenate sodalite, Na 8 [AlSiO 4 ]6(ReO 4 ) 2 , and the structure of a mixed guest perrhenate/pertechnetate sodalite, Na 8 [AlSiO 4 ] 6 (ReO 4 ) 2-x (TcO 4 ) x . Perrhenate was used as a chemical analogue for pertechnetate. Bulk analyses of each solid confirm a cubic sodalite-type structure (P$\overline{43}$n, No. 218 space group) with rhenium and technetium in the 7+ oxidation state. High-resolution nanometer scale characterization measurements provide first-of-a-kind evidence that the ReO 4 – anions are distributed in a periodic array in the sample, nanoscale clustering is not observed, and the ReO 4 – anion occupies the center of the sodalite β-cage in Na 8 [AlSiO 4 ] 6 (ReO 4 ) 2 . We also demonstrate, for the first time, that the TcO 4 – anion can be incorporated into the sodalite structure. Lastly, thermochemistry measurements for the perrhenate sodalite were used to estimate the thermochemistry of pertechnetate sodalite based on a relationship between ionic potential and the enthalpy and Gibbs free energy of formation for previously measured oxyanion-bearing feldspathoid phases. The results collected in this study suggest that micro- and mesoporous crystalline solids maybe viable candidates for the treatment and immobilization of 99 Tc present in reprocessed nuclear waste streams and contaminated subsurface environments.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Los Alamos National Laboratory Support for Commercial U.S. Production of 99 Mo without the Use of Highly Enriched Uranium

There is currently a serious shortage of 99 Mo, from which to generate the medically significant isotope 99m Tc. Most of the world's supply comes from the fission of highly enriched uranium targets--this is a proliferation concern. This document focuses on the technology involved in two alternative methods: electron accelerator production of 99 Mo from the 100 Mo(γ,n) 99 Mo reaction and production of 99 Mo as a fission product in a subcritical, DT accelerator-driven low enriched uranium salt solution.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

The Texas A&M Radioisotope Production and Radiochemistry Program

The main motivation of the project at Texas A&M University was to carry out the production of critically needed radioisotopes used in medicine for diagnostics and therapy, and to establish an academic program in radionuclide production and separation methods. After a lengthy battle with the Texas A&M University Radiation Safety Office, the Texas Department of State Health Services granted us a license for the production of radionuclides in July 2015, allowing us to work in earnest in our project objectives. Experiments began immediately after licensing, and we started the assembly and testing of our target systems. There were four analytical/theoretical projects and two experimental target systems. These were for At-211 production and for Zn-62/Cu-62 production. The theoretical projects were related to the production of Mo-99/Tc-99m using a) a subcritical aqueous target system and b) production of Tc-99m from accelerator-generated Mo-99 utilizing a photon-neutron interaction with enriched Mo-100 targets. The two experimental projects were the development of targetry systems and production of At-211 and Zn-62/Cu-62 generator. The targetry system for At-211 has been tested and production of At-211 is chronic depending of availability of beam time at the cyclotron. The installation and testing of the targetry system for the production of Zn-62/Cu-62 has not been finalized. A description of the systems is described. The academic program in radionuclide production and separation methods was initiated in the fall of 2011 and, due to the lack of a radiochemistry laboratory, it was suspended. We expect to re-start the academic program at the Texas A&M Institute for Preclinical Studies under the Molecular Imaging Program.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗