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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Design and Application of Materials for Sequestration and Immobilization of 99 Tc

99 Technetium ( 99 Tc) is a hazardous radionuclide generated by the nuclear industry that poses a serious environmental threat. The wide variation and complex chemistries of nuclear waste streams containing 99 Tc often create unique, site specific challenges when sequestering and immobilizing the waste in a matrix suitable for long-term storage and disposal. Therefore, an effective management plan for 99 Tc containing radioactive waste will likely require a variety of suitable materials/matrices capable of adapting to and addressing these challenges. In this review, we discuss and highlight the key developments for effective removal and immobilization of 99 Tc in inorganic waste forms. Specifically, we review the synthesis, characterization, and application of materials for targeted removal of 99 Tc from (simulated) waste solutions under various experimental conditions. These materials include: i) layered double hydroxides (LDHs), ii) metal-organic frameworks (MOFs), iii) ion-exchange resins (IERs) as well as cationic organic polymers (COPs), and iv) surface modified natural clay materials (SMCMs). Secondly, we discuss some of the major and recent developments towards 99 Tc immobilization in i) glass, ii) cement, and iii) iron mineral waste forms. Finally, we present future challenges that need to be addressed for the design, synthesis, and selection of suitable matrices for the efficient sequestration and immobilization of 99 Tc from targeted wastes. Finally, the purpose of this review is to inspire the researches on the design and application of various suitable materials/matrices for selective removal of 99 Tc present globally in different radioactive wastes and its immobilization in stable/durable waste forms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Functionalized Magnetic Nanoparticles for Technetium Sequestration from Groundwater

Technetium 99 (Tc) is among the most common environmental contaminants at DOE sites and one of the most common risk drivers in low- and high-level waste disposal sites. The majority of Tc is generated from anthropogenic sources, such as nuclear power plants, global weapons, nuclear storage facilities and medical applications. Through these sources, Tc contamination has been unintentionally introduced in to the environment. The most common chemical form of Tc is Tc(VII)O{sub 4}{sup -}. Due to its high solubility and mobility, Tc can enter the food chain and cause adverse health effects to humans. Currently, ion exchange resins and reduction processes are the most common approaches for Tc immobilization. Although these techniques have shown to be effective, they also possess major drawbacks, such as high cost, low adsorption capacity, and complex creation and maintenance. Therefore, development of more efficient and simple technologies for the remediation of Tc-contaminated systems are needed. Functionalized magnetic nanoparticles have been used to remove organic and inorganic contaminants from water resources. These nanoparticles have attracted extensive attention as an adsorbent material due to their large surface area, high efficiency, low-cost, easy functionalization and separation with a magnet. This study seek to develop functionalized magnetic iron oxide nanoparticles for the efficient removal of Tc and other heavy metal contaminants from water resources under ambient conditions. Objectives: Synthesize magnetic iron oxide nanoparticles and functionalize their surface with Cetyltrimethylammonium Bromide (CTAB) and tetraethyl-orthosilicate (TEOS). Characterize the synthesized nanoparticles using scanning electron microscopy (SEM) coupled with energy dispersive X-ray spectroscopy (EDS), Dynamic Light Scattering (DLS) and Zeta PALS. Perform adsorption studies to evaluate their adsorption behavior and capacity for (a) Technetium using Rhenium (ReO{sub 4}{sup -}) as a surrogate and (b) heavy metals, e.g. Cu{sup 2+}. Conclusions: Magnetic iron oxide nanoparticles were successfully functionalized with CTAB and TEOS. The functionalization of the iron oxide nanoparticles affects their surface charge and their hydrodynamic diameter. The addition of CTAB or TEOS decreased the hydrodynamic diameter of the nanoparticles due to repulsive and steric forces. The SEM micrographs show spherical nanoparticles of different sizes. The EDX analysis shows the presence of iron and oxygen from the iron oxide crystalline structure, and the different constituents of the CTAB and TEOS molecules. Proof-of-concept shows the successful adsorption of rhenium (ReO{sub 4}{sup -}) and copper Cu{sup 2+}) onto CTAB-Fe{sub 2}O{sub 3} and TEOS-Fe{sub 2}O{sub 3} respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Interpretation of large‐scale, long‐term electrical geophysical monitoring guided by a process simulation

Abstract Surface electrical resistivity tomography (ERT) was used at a waste site to monitor vadose zone changes in electrical properties as a proxy for contaminant flux over a span of 17 years. The BC Cribs and Trenches (BCCT) site at the Hanford site contains 20 disposal trenches and six disposal cribs. Wastes include a large inventory of technetium‐99 and large masses of nitrate and uranium‐238. ERT data were collected along 41 profiles in 2005 to characterize regions of elevated bulk electrical conductivity (BEC) associated with past liquid waste discharges. Previous analyses performed on samples from four boreholes showed a high correlation between nitrate concentration and BEC. In 2022, ERT data were re‐collected along the same profiles and six additional profiles in an area not previously surveyed. Compared to background uncontaminated areas, BEC was higher in contaminated areas at the waste sites. Given the correlation between nitrate concentration and BEC previously found at this site, ERT images show the spatial distribution and relative ionic concentration of vadose zone contaminants at BCCT. Between 2005 and 2022, ERT difference images showed a decrease in BEC surrounding most waste sites, with exceptions where there were known anthropogenic surface changes. An evaluation of recharge‐driven nitrate migration using synthetic flow and transport simulations showed that downward migration causes a decrease in BEC from the decrease in ionic strength at the trailing end of the plume where contaminants migrated downward. From this, we interpret ERT difference images as showing the predominant regions of downward ion flux.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2020

Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Low Count and Background Radionuclides Analysis - 20488

The US Department of Energy (DOE) is often faced with the need to evaluate radionuclides at low concentrations. When site sample data are likely to be close to threshold activity concentrations of interest, then the means by which the radiochemical analysis is performed and reported is critical. This situation can occur when differentiating from zero (presence/absence) for radionuclides that do not occur naturally, close comparison with environmental background for naturally occurring radionuclides, close comparison with a risk- or dose-based threshold concentrations of interest, or even comparisons across studies. There are several analytical issues that are of concern, but the two that appear to cause incorrect decisions to be made most often involve establishing detection limits and subtracting ambient background conditions in the laboratory. These issues are not critical when radionuclide activity concentrations are large relative to thresholds of concern, but they seem to be poorly understood when it matters. When the comparisons are important and are likely to be close to a threshold of interest, then the general contract with the analytical laboratories needs to be changed so that the right or appropriate data are obtained. The concern is that important decisions are made incorrectly more often as greater scrutiny is placed on DoE's radionuclide cleanup or monitoring decisions by the public and other stakeholders. Examples are presented of problems that have been observed for different projects, both within and outside the realm of DOE and NRC remediation and radioactive waste disposal problems, and solutions are offered that should lead to better data from which important decisions need to be made. The first example is from Los Alamos National Laboratory (LANL) and involves radionuclide concentrations in soil and rock beneath LANL's Material Disposal Area (MDA) G. An initial review of the data led to a conclusion that americium and plutonium are a long way present beneath MDA G. A more thorough review of the data that accounted properly for ambient background and the detection limits that had been established led to the opposite conclusion. Another example is from the Nevada National Security Site where tritium results from one of the wells were unexpectedly high. Proper understanding and analysis of ambient background led to the conclusion that the increased concentrations were not so obvious, and that a different contract with the analytical laboratory was needed to provide more appropriate data to support a better determination. Other examples are used from regulatory review of projects in Nevada, where background levels and secular equilibrium for naturally occurring radionuclides are not established correctly because of analytical issues. The same basic issues have also been found to create difficulties analyzing historical data from the West Valley Demonstration Project. There is evidence in the data that the apparent lack of secular equilibrium where it is expected to exist is related to ambient background subtraction or other analytical issues. A final example is presented for analysis of Tc-99 in samples of depleted uranium. In this case, two different studies that were performed only three months apart provide quite different results. The US Environmental Protection Agency (EPA) established the data quality objectives (DQO) process in the mid-1980's to establish decision performance criteria for data collection. EPA guidance (EPA G-4, for example) clearly distinguishes between DQOs and measurement performance objectives (MQOs) that should be addressed for laboratory analysis of samples. The language of DQOs and MQOs has become confused over time it seems, and the subsequent effects seem to include a lack of attention to decision performance and a routine approach to measurement quality. In order to better address radionuclide sample analysis when the concentrations are close to thresholds of concern, which might be zero for some radionuclides, background for others, and risk-based thresholds for yet others, it is important that routine laboratory analysis methods are adjusted, and that the project team and the laboratory work closely together to ensure that the data meets the MQO requirements of laboratory analysis and reporting of results, and that the MQOs effectively support project-specific DQOs. This basic approach will be applied in Los Alamos in the coming year to the collection of moisture data from underneath MDA T that will be analyzed for americium and neptunium isotopes. Proper understanding of the radiochemistry methods and reporting, and of appropriate statistical methods is critical to the success of such projects, ensuring that the right decisions are made. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

129 I, 99 Tc, and U Distribution Coefficients of Subsurface Sediments Collected from the Proposed Site of the Environmental Management Disposal Facility

Performance Assessment calculations were completed in 2020 to evaluate the Environmental Management Disposal Facility (EMDF), a proposed new low-level radioactive waste (LLW) disposal facility on the U.S. Department of Energy’s Oak Ridge Reservation (ORR). Among the large number of input parameters needed for such calculations, are distribution coefficients (K d values; radionuclide concentration solid: liquid ratio) that provide a measure of the tendency of radionuclides to bind to sediments. The objective of this study was to measure K d values of three radionuclides that may pose a disproportionately large amount of risk, U, iodine-129 ( 129 I) and technetium-99 ( 99 Tc). The average 129 I K d value for the 14 geological materials recovered from the proposed EMDF site was 37.8 mL/g and ranged from 0.45 to 140.9 mL/g. These values were consistent, but somewhat larger than previous measurements made with ORR sediments and were about an order of magnitude greater than those used in previous EMDF PA calculations. The median 99 Tc K d value was 365.7 mL/g, much greater than previously reported using ORR geological materials. Five of the 14 tested geological materials sorbed large quantities of 99 Tc, suggesting that the weakly sorbing 99 Tc(VII) species had been reduced to the sparingly soluble 99 Tc(IV) species. The five strongly sorbing sediments had apparent 99 Tc solubility values of approximately <10 -8 mol/L. The median U K d value was 5,726 mL/g. All of the tested geological materials had large K d values, ranging from 625 to >10,208 mL/g. Among the sediment samples that exhibited strong U binding, the apparent solubility value was approximately <10 -9 mol/L. Based on sediment properties and general ORR geological considerations, it was proposed that much of the 129 I and 99 Tc retention could be attributed to the site materials exhibiting low pH (average pH = 4.94), and/or the elevated levels of iron oxides, manganese oxides, and natural organic matter. Similarly, the extremely high U binding measured in these sediments may also be attributed to the low conditions of carbonates, which can complex and therefore solubilize uranyl in these tests due to the low pH, and also the relatively high concentrations of iron and organic coatings on these samples. An implication of this study is that the areas of the EMDF subsurface environment may have natural properties for attenuating 129 I, 99 Tc, and U movement, and potentially other radionuclides, thereby possibly reducing risk posed by burial of LLW at this site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Sorption of Anionic Iodine and Molybdenum to Multi-sorbing Barrier (MSB) Materials - 20030

A multi-sorbing barrier (MSB) material containing sodium bentonite (NaB), organoclay, and anion resin is being evaluated for containment of anionic radionuclides. Batch tests on the MSB were conducted using two non-radioactive analog isotopes ({sup 127}I, {sup 95}Mo) as surrogates for {sup 129}I and {sup 99}Tc due to their similar valence and molecular weight relative to the radionuclides. Tests were conducted on the organoclay, anion resin, and NaB individually and on NaB-resin blends with resin contents of 1% and 5% by mass. The anion resin exhibits very strong sorption of {sup 127}I and {sup 95}Mo relative to the other materials. The organoclay exhibited modest sorption of {sup 95}Mo and {sup 127}I, and NaB little sorption. The sorption isotherms follow a linear model for NaB, a Freundlich model for the organoclay, and a Langmuir model for the anion resin and NaB-resin mixtures. Combining resin with NaB significantly enhances sorption relative to NaB alone, with a larger resin fraction yielding greater sorption. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

The Nature and State of Groundwater Contamination at the Nevada National Security Site: What Have We Learned from Decades of Groundwater Analysis? - 20337

The regulatory framework for remediating radionuclide contamination from underground nuclear testing at the Nevada National Security Site (NNSS) is based on a combination of characterization and modeling studies, monitoring, and institutional controls [1]. Currently, tritium is the largest contributor (∼90%) to the estimated 44.6 million-curie radionuclide inventory resulting from underground testing [2]. Because of its short half-life (12.32 years), its relative contribution reduces below 10% of the total radiologic inventory over the next 120 years as a result of radioactive decay. Although tritium levels are observed well above the Safe Drinking Water Act (SDWA) maximum contaminant levels (MCLs) in groundwater, other radionuclides are well below their MCLs except within the nuclear test near-field (nuclear test cavity and chimney) environment. In fact, most device-derived radionuclides are below their MCL in groundwater even in samples collected from this near-field environment. The distribution of radionuclides following the nuclear detonation greatly influences the availability of potential contaminants for groundwater transport. Tritium is initially distributed in the gas phase, later as tritiated water in steam, and finally as liquid water, and is available to groundwater transport away from the near-field environment. Other radionuclides that are mobile in groundwater are {sup 14}C, {sup 36}Cl, {sup 99}Tc, and {sup 129}I though their radiologic inventory is small when compared to tritium. Many radionuclides (e.g. U, Pu, Am) are incorporated to a significant extent into the melt glass at the bottom of the cavity and are accessible to groundwater primarily through the slow process of glass dissolution. These radionuclides are also adsorbed to the surfaces of the crushed rock within the cavity and chimney which limits their migration in groundwater. Although colloid facilitated transport of radionuclides at the NNSS has been observed [3], radionuclide concentrations decrease with time and migration distances due to desorption and colloid filtration processes. Current studies indicate that radionuclides associated with colloids are unlikely to migrate downgradient from NNSS underground nuclear tests at concentrations above the SDWA MCL [4][5]. The results of over 50 years of sampling, along with an understanding of these post detonation processes, indicate that tritium is the only contaminant of concern downgradient of testing and that even tritium will not exceed its MCL in groundwater after ∼120 years. Although other longer-lived radionuclides may continue to be released slowly from the near-field environment they will likely never reach levels exceeding their MCLs in groundwater downgradient of the NNSS. Groundwater monitoring will continue to verify these observations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Energetics of Salt-Bearing Sodalites, Na 8 Al 6 Si 6 O 24 X 2 (X = SO 4 , ReO 4 , Cl, I): A Treatment Option for Pertechnetate-Enriched Nuclear Waste Streams

An alternative option for treating anion-enriched reprocessed nuclear waste streams is to immobilize technetium-99 ( 99 Tc, β = 293.7 keV, t 1/2 = 2.1 × 10 5 years) and other anions in micro- and mesoporous materials. Here we determine the thermodynamic stability of anion bearing sodalites, Na 8 Al 6 Si 6 O 24 X 2 (X = SO 4 , ReO 4 , Cl, I), to improve our understanding of the driving forces that control framework assembly using high temperature oxide melt solution calorimetry. Raman and FTIR spectroscopy illustrate a strong dependence for vibrational features on anion size and enabled the development of a linear model that predicted the vibrational features for numerous anion bearing sodalites to within ±20 cm –1 (i.e., OH, F, Br, ClO 4 , NO 3 , and MnO 4 ). The largest negative enthalpy of formation from elements and the lack of structural water demonstrate that the perrhenate sodalite (Na 8 Al 6 Si 6 O 24 [ReO 4 ] 2 ), a chemical analogue for pertechnetate sodalite (Na 8 Al 6 Si 6 O 24 [TcO 4 ] 2 ), is more thermodynamically stable than all other anion bearing sodalites evaluated. The enthalpies of the reaction between nepheline and the sodium salt, which provides the guest anion species, was negative only for the ReO 4 and NO 3 bearing sodalites. Overall, we report for the first time the enthalpy of the ion exchange reactions for different anion bearing sodalites relative to the perrhenate sodalite, which is a key step in gaining the ability to tune sodalite material properties and structure during treatment and the immobilization of 99 Tc in the presence of competing anions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

A Focused Ion Beam-Scanning Transmission Electron Microscopy with Energy-Dispersive X-ray Spectroscopy Study on Technetium Incorporation within Iron Oxides through Fe(OH) 2 (s) Mineral Transformation

Incorporation of Tc(IV) into iron oxide/hydroxide minerals has been explored and proposed as a promising pathway to preventing Tc(IV) reoxidation to environmentally mobile pertechnetate (TcO 4 - ) and improving long-term immobilization of radioactive technetium-99 (Tc). However, visual evidence evaluating the distribution of Tc(IV) incorporated within iron oxide/hydroxide phases has not been available, until now, despite potential implications on Tc(IV) stability within the host phase. For the purpose of this study Tc(IV) incorporation into iron oxide/hydroxide phases was facilitated via Fe(OH) 2 (s) oxidation and mineral transformation to magnetite (Fe 3 O 4 ). Focused ion beam - scanning transmission electron microscopy equipped with energy-dispersive X-ray spectroscopy (FIB/STEM-EDS) methods were then combined with X-ray diffraction and absorption spectroscopy techniques to characterize and visually demonstrate that, for the first time, Tc(IV) is heterogeneously incorporated into different iron oxide/hydroxide phases as Tc(IV)-incorporated magnetite and/or TcO 2 ·2H 2 O(s) via different incorporation mechanisms. Heterogeneous distribution of Tc(IV) in magnetite suggests either (i) TcO4- is reduced quickly at the magnetite surface and then encapsulated into magnetite during continued octahedral crystal growth, or (ii) Tc(IV) alternatively partitioned into multiple layers of a blocky, plate-like morphological magnetite structure showing stratified Tc. With limited Tc-hematite (Fe 2 O 3 ) incorporation, the results suggest that TcO 2 ·2H 2 O(s) is formed and mainly associated/embedded in fibrous nanometer-sized polycrystalline hematite. This work highlights the power of modern state-of-the-art FIB/STEM-EDS approach to provide essential visual insights of the Tc-iron oxide/hydroxide incorporation and generate reliable mechanism-informed designs for waste forms relying on Tc mineral incorporation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Advanced Porous Materials as Designer Platforms for Sequestering Radionuclide Pertechnetate

Technetium-99 ( 99 Tc), predominantly present as pertechnetate ( 99 TcO 4 – ), is a challenging contaminant in nuclear waste from artificial nuclear fission. The selective removal of 99 TcO 4 – from nuclear waste and contaminated groundwater is complex due to (i) the acidic and intricate nature of high-level liquid wastes; (ii) the highly alkaline environment in low-activity level tank wastes, such as those at Hanford, and in high-level wastes at locations like Savannah River; and (iii) the potential for 99 TcO 4 – to leak into groundwater, risking severe water pollution due to its high mobility. This Review focuses on recent developments in advanced porous materials, including metal–organic frameworks (MOFs), covalent organic frameworks (COFs), and their amorphous counterparts, porous organic polymers (POPs). These materials have demonstrated exceptional effectiveness in adsorbing 99 TcO 4 – and similar oxyanions. We comprehensively review the adsorption mechanisms of these anions with the adsorbents, employing macroscopic batch/column experiments, microscopic spectroscopic analyses, and theoretical calculations. In conclusion, we present our perspectives on potential future research directions, aiming to overcome current challenges and explore new opportunities in this area. Our goal is to encourage further research into the development of advanced porous materials for efficient 99 TcO 4 – management.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Vitrification of Hanford Tank 241-AP-107 with Recycled Condensate

During the vitrification of nuclear waste at the Hanford Waste Treatment and Immobilization Plant (WTP) – the primary mission of the U.S. Department of Energy Office of River Protection – the offgas condensate generated from the waste-to-glass conversion is currently planned to be concentrated by evaporation in the Effluent Management Facility (EMF). This concentrated condensate can then be recycled back to the incoming waste and vitrified. To test the recycle process, a test apparatus was designed to mimic the EMF evaporator and used to concentrate a volume of condensate that had been previously produced during the vitrification of Hanford tank 241-AP-107 (referred to herein as AP-107) waste in a continuous laboratory-scale melter (CLSM). The test apparatus successfully concentrated the AP-107 condensate by a factor of 10 while retaining over 90 % of the technetium-99 ( 99 Tc), Cs, and I inventory. A second portion of AP-107 waste was retrieved by Washington River Protection Solutions, LLC, given to Pacific Northwest National Laboratory, and combined with the AP-107 condensate concentrate after undergoing solids filtration and cesium removal by ion exchange. This combination served to approximate the recycling action to be performed at the WTP. After the addition of glass-forming chemicals (GFCs), the combined AP-107 waste and AP-107 condensate concentrate were processed in the CLSM to produce a glass, called AP-107-1R, that was designed to satisfy the WTP baseline requirements (Kim et al. 2012). During the 8.87 hours of processing, 7.27 kg of AP-107-1R glass were produced for an average glass production rate of 1739 kg m 2 d -1 . Compared to the previous run in the CLSM without recycled condensate, the run with the recycle had a greater average glass production rate, but the rate was within the potential range of variability when processing melter feeds with similar composition in the CLSM. The glass produced from the AP-107 recycle run in the CLSM was within 10 % of the target AP-107-1R glass composition with respect to the primary glass components. Analysis of the minor component impurities revealed that their content in the glass product had approached their nominal target after 2 turnovers of the glass inventory in the CLSM while the activity of the minor radionuclides was retained in the glass product. The 99 Tc and total cesium content in the combined AP-107 waste and recycled condensate were maintained at concentrations expected to be experienced at the WTP. During processing in the CLSM, at discrete sampling time periods, the target 99 Tc/Cs mass ratio in the glass formulation varied from 0.9 to 62.9. Across this range, the Cs retention in the glass ranged from 53 to 60 %, while the retention from the entire runtime totaled 68 %, values which align with Cs retention in other scaled melter systems while processing LAW melter feeds at 99 Tc/Cs mass ratios ranging from 1 to 100. The 99 Tc retention in the glass ranged from 22 to 32 %, primarily due to the cold-cap coverage on the glass melt surface, the area covered by reacting melter feed, varying from ~80 % to ~95 % during processing, demonstrating greater volatility of 99 Tc from the glass while more surface was exposed, as expected based on previous 99 Tc volatility studies.

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

Vitrification of Hanford Tank Wastes for Condensate Recycle and Feed Composition Changeover Testing

During the vitrification of Hanford Site nuclear waste at the Waste Treatment and Immobilization Plant (WTP), which is a part of the safe and efficient retrieval, treatment, and disposal mission of the U.S. Department of Energy Office of River Protection, the offgas condensate generated from the waste-to-glass conversion is currently planned to be concentrated by evaporation in the Effluent Management Facility (EMF). This concentrated condensate can then be recycled back to the incoming waste and vitrified. To test the recycle process, an apparatus was designed and built to mimic the EMF evaporator and was then used to concentrate a volume of condensate produced during the vitrification of a sample of Hanford tank 241-AP-107 (referred to herein as AP-107) waste in a continuous laboratory-scale melter (CLSM). The concentrated condensate was added to an additional sample of AP-107 waste, to mimic one round of the recycle process, and the combined solution was vitrified, producing a second round of recycle condensate. In the current study, the EMF test apparatus was used to concentrate the second-round recycle condensate under evaporation conditions (at 45 °C and 1.4 psia) designed to emulate EMF operation. The condensate was successfully concentrated by a factor of ~10 while retaining over 95 % of the technetium-99 ( 99 Tc), Cs, and I inventories in the concentrate. Another portion of AP-107 waste was retrieved by Washington River Protection Solutions, LLC (WRPS) and transferred to Pacific Northwest National Laboratory (PNNL), where it was pretreated and then combined with the second-round recycle AP-107 condensate concentrate and glass-forming chemicals (GFCs) to form the two-time recycle AP-107 melter feed, approximating a second round to the recycling action to be performed at the WTP. A portion of AP-105 waste was also retrieved by WRPS and provided to PNNL for pretreatment and combining with GFCs to form AP-105 melter feed. The two-time recycle AP-107 and AP-105 melter feeds were processed consecutively in the CLSM. The CLSM run proceeded for 13.63 hours, producing 9.70 kg of glass for an average glass production rate of 1464 kg m 2 d -1 during the two-time recycle AP-107 feed charging and 1568 kg m 2 d -1 during the AP-105 feed charging. The rate during AP-107 charging was essentially equivalent to the rate when processing no-recycle AP-107 feed and lower than that achieved when processing one-time recycle AP-107 feed. However, all rates were within the potential range of variability when processing melter feeds with similar composition in the CLSM. Likewise, the rate during AP-105 charging was higher than the previous rate processing AP-105, but within the potential CLSM range. The cold-cap characteristics changed from the typically thin AP-107 cold cap to a foamy-edged cold cap as previously seen with AP-105 shortly after transitioning to the AP-105 melter feed. The glass produced during the CLSM run was within 10 % of its target composition for the primary glass components. The CaO and Li 2 O targets varied by more than 1 wt% between the two-time recycle AP-107 and AP-105 glass targets and it took about 2 turnovers of the CLSM glass inventory to reach a relative chemical steady state in the glass for CaO and Li 2 O after the melter feed inputs were switched. The 99 Tc and total cesium content in the melter feeds were maintained at concentrations expected to be experienced at the WTP. During the CLSM run, while processing the two-time recycle AP-107 melter feed at a relative chemical steady state, the 99 Tc/Cs ratio was 10, and 34% of 99 Tc and 74% of Cs were retained in the glass. These values were higher than those measured in the CLSM run with one-time recycle AP-107 melter feed. After the transition to processing the AP-105 melter feed, when the production reached a relative chemical steady state, the 99 Tc/Cs ratio was 77 while 44% of 99 Tc was retained in the glass. The C's retention during this time frame reached 200% due to the excess C's in the glass after the target content decreased to 15% of its initial level in the two-time recycle AP-107 melter feed to the lower target in the Ap-105 melter feed. While iodine was below inductively coupled plasma mass spectrometry analytical reporting limits in the melter feed and glass samples, it was detected in quantities above the analytical reporting limits in the liquid and filter samples collected from the CLSM offgas treatment system. The behavior of iodine in the CLSM offgas treatment system followed a similar pattern to those of 99 Tc and Cs.

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DOE OSTI2020

Pump-and-Treat Impacts on Aquifer Conditions and the Development of System-Wide Performance Strategies - 20377

The pump-and-treat (P and T) system on the Hanford Site's Central Plateau currently treats multiple contaminants within the 200-ZP-1 operable unit, including nitrate through its biological treatment processes. Although it was initially designed to support the 200-ZP-1 Remedial Action Objectives, the P and T system has evolved in use to address contaminants from other regions of the aquifer as well. The P and T influents are biogeochemically distinct, treating groundwaters with elevated concentrations of cyanide, uranium, technetium-99, and nitrate. Given the varying geochemical signatures from current and potential future sources, operational conditions, and amendments considered for treatment processes, there is a need to identify the impacts that the P and T effluent has on the aquifer, and how the aquifer affects P and T performance. To this end, laboratory efforts have focused on characterizing P and T effluent and its biogeochemical effects on the sediments through a series repacked column experiments. The P and T effluent is rich in particulate matter. A systematic inspection of this particulate matter in the P and T effluent showed significant levels of manganese/iron oxides, calcite crystals, relatively large organic matter molecules, and biomass. Therefore, the column experiments were designed and conducted for four different treatment systems: artificial groundwater, filtered P and T effluent, non-filtered P and T effluent, and chlorinated P and T effluent. Different levels of microbiologically- and geochemically-induced changes in the columns occurred and modified the flow systems. For example, physical filtration occurred in the sediments where the particulate material had accumulated in the top portion of the column. Reactive transport modeling of the laboratory experiments is being carried out so that the reaction network can be extrapolated to the field scale. These simulations will provide predictive evaluations of system performance based on P and T operational changes and aquifer mass loading and capacity estimates. This information can be used to address combined subsurface and operational challenges, such as the need for frequent well redevelopment to maintain injection capacity. This information also provides the technical basis for decisions related to the P and T system operations that support remedy optimization efforts and short- and long-term remedy decisions related to multiple operable units on the Hanford Site Central Plateau. (authors)

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DOE OSTI2020

Evaluating Remediation Options for a Perched-Water Unit at the Hanford Site - 20378

At the U.S. Department of Energy's Hanford Site, liquid-phase waste was discharged using surface disposal features such as ponds, trenches, and cribs. These practices distributed contaminants in the subsurface, some of which are currently in the vadose zone above the groundwater. At the 200-DV-1 operable unit (OU) at the Hanford Site, a key feature is a perched-water zone. This perched-water zone contains several million liters of water contaminated with high concentrations of uranium, technetium-99 (Tc-99), and nitrate that serves as a continuing source of contamination to the underlying aquifer. Currently, a pumping system is used to remove contaminant mass from the perched-water zone. However, water extraction will become more challenging as the perched unit dewaters. As a result, residual contaminants remaining after pumping is no longer effective needs to be addressed with post-extraction mitigation approaches, appropriate monitoring, and selection of suitable remediation endpoints. Thus, a remedy evaluation needs to include the unique conditions of this perched-water zone. As part of the initial evaluation, a set of technologies have been preliminarily identified as potentially applicable to the perched-water zone, including in situ remediation technologies to sequester uranium and Tc-99 and to degrade nitrate. While there is sufficient information available for direct consideration of some of the reviewed technologies in a feasibility study, five in situ remediation technologies were identified as needing additional information on their effectiveness under the geochemical and contaminant conditions of the perched-water zone. (authors)

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DOE OSTI2020

Field-Scale Lysimeter Studies of Glass and Cementitious Waste Forms at the Hanford Site - 20392

The Hanford site Integrated Disposal Facility (IDF) will receive waste forms from vitrification activities at the Hanford Waste Treatment and Immobilization Plant (WTP). The waste form inventory to be disposed of at the IDF will consist of vitrified low-activity waste (LAW) in glass forms and solidified (or encapsulated) secondary wastes in cementitious forms. The IDF is a near-surface burial facility located near Hanford's Waste Treatment and Immobilization Plant in the 200 East Area of the site's Central Plateau. An extensive set of laboratory experimental data has been collected to support IDF performance assessment calculations of the eventual degradation of waste forms and mobility of contaminants. This paper describes the start of a field experiment to generate waste-form performance data on a larger scale (tens of centimeters), over a longer duration (five years or more), and under field conditions representative of the IDF. Results from this study are expected to improve model descriptions of contaminant mobility, reduce uncertainties about the representativeness of laboratory results in the IDF performance assessment, improve stakeholder confidence in the safe disposal of treated waste at the IDF, and facilitate the adoption of informed facility designs with the potential to reduce operational costs. A lysimeter test facility near Hanford's 200 West Area has been repurposed to carry out long-term assessments of waste forms buried in sediments excavated from the IDF. The waste forms are buried in large drainage lysimeters, open caissons that are two meters in diameter and three meters deep. Waste forms are buried in up to six locations in each lysimeter. Samplers collect pore water and drainage water beneath the waste forms and air from the pore space adjacent to waste forms. Sensors measure temperature, soil water content, soil water tension, and drainage flux. One lysimeter containing cementitious waste forms and one containing glass waste forms have been completed, along with a control lysimeter (without waste forms). The cementitious waste forms are Cast Stone grout formulations using a liquid secondary waste simulant and a generic Hanford high-salt simulant, as well as a hydrated lime-based grout formulation using a liquid secondary waste simulant. Technetium-99 and iodine-127 were added as tracers to monitor contaminant mobility. Glass waste forms were fabricated in two formulations (LAWA44 and ORLEC28), with rhenium and molybdenum tracers to monitor waste form degradation. The predicted performance of the waste forms in the lysimeters was modeled prior to installation to determine waste form size, tracer concentration/activity, and locations of sensors and samplers. (authors)

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